ML20083A759

Safety Evaluation Summary Rept, Rev 7

ML20083A759
Person / Time
Site:	Fermi
Issue date:	05/08/1995
From:	DETROIT EDISON CO.
To:
Shared Package
ML20083A748	List: ML20083A759 NRC-95-0043, Forwards Rev 7 to Fermi 2 Updated FSAR & Fermi 2 Safety Evaluation Summary Rept
References
NUDOCS 9505110127
Download: ML20083A759 (207)

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Safety Evaluation No: 92-0071 UFSAR Revision No. NA Reference Document: ABN 13204-1 Section(s) NA Table (s) NA Figure Change Yes X No Title of Change: Replacement of UFSAR Figure 11.4-6, Post Accident Sampling System (PASS)

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SUMMARY

This evaluation justifies replacing Figure 11.4-6 with the Detroit Edison drawings 61721-2400-I 25 (Figure 11.4-6, sheet 1) and 61721-2400-26 (Figure 11.4-6, Sheet 2). ABN 13204-1 documents the as-built configuration of PASS. It does not make any physical changes to the system.

I Replacement of Figure 11.4-6, a GE sketch, with Edison drawings more accurately represents 1

the installed PASS and enhances Fermi's capability to respond to a malfunction of system equipment. Figure changes made by this A BN were submitted to the NRC in Revision 6 of the UFSAR submittal.

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Safety Evaluation No: 92-0076, Rev 1 UFSAR Revision No. 7 Reference Document: ABN 12834-1 Section(s) 3.12 I Table (s) NA I Figure Change Yes X No Title of Change: Emergency Diesel Generator (EDG) Circuit Protection, Standby Motor, and Standby Heater Qualification Change

SUMMARY

This evaluation justified qualifying the existing isolation devices between the EDG balance-of-I plant (BOP) motor or heater loads and their Class 1E power supplies as Class 1E MCC to non-Q load isolation devices and, through design calculations, took exception to Regulatory Guide 1.75 requirements for fault sensing breakers. The classification of the EDG standby motors ,

and heaters has been changed from QA level 1, Seismic I to non-Q, Seismic II/I because there  ;

I is no seismic analysis to support the previous classification and the previously unclassified EDG generator standby heater is now classified as non-Q, Seismic II/l. This required upgrading the subject fused disconnects and circuit breakers to QA level I, Seismic l. Previous I Regulatory Guide 1.75 exceptions were identified in the UFSAR and were reviewed and approved by the NRC in NUREG 0798. As a result of this change, UFSAR Subsection 3.12.3 was revised to state that each essential /non-essential load separation case that is an )

exception to Regulatory Guide 1.75 is analyzed in design calculations and that future changes  :

are evaluated using the same criteria as the original cases. In addition, UFSAR Subsection 3.12.4 was revised to describe the use of double fuses or a fuse and a molded case circuit breaker as an attemate means of meeting the Regulatory Guide 1.75 objectives.

The component reclassification is supported by seismic and failure effect analyses. It I maintains the EDG and AC power electrical design described in UFSAR 3.12.8. No credit is taken for the EDG BOP motors or heaters during EDG operation under accident conditions and there is no change to EDG performance during a design basis accident. The design I calculations show that there is no adverse effect to the safety related Class 1E MCCs or loads for all fault and open circuit combinations possible in the BOP design.

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Safety Evaluation No: 92-0091 UFSAR Revision No. 7 Reference Document: ABN 12554-1 Section(s) NA Table (s) NA Figure Change l X l Yes (_j No Title of Change: Assigning of PIS Numbers to Drywell Cooler Discharge Dampers

SUMMARY

This evaluation justifies assigning plant identification system (PIS) numbers T4700F001A I through T4700F014 to drywell fan cooler units T4700C001 through T4700C014 air discharge volume dampers. In addition, air flows as shown on the system P&lD (UFSAR Figure 9.4-08) and isometric have been revised to agree with air flows determined during preoperational testing. Assigning PIS numbers to volume dampers will aid plant personnel in locating, I verifying and documenting the alignment of the volume dampers. This should also help to minimize stay times in the drywell and therefore lower dose received in verifying drywell i

cooling system volume damper alignment during plant shutdown conditions. Revising air flows I I to agree with preoperational test data will provide the accurate indication of expected fan cooler unit discharge air flows.

f The changes made by ABN-12554-1 do not physically affect the drywell cooling system or its design function. Proper alignment of volume dampers will help to ensure that temperature limits of Technical Specifications are met.

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Safety Evaluation No: 93-0007 UFSAR Revision No. 7 l

Reference Document: ABN 13765-1 Section(s) NA Table (s) NA f Figure Change ] Yes No Title of Change: Removal of Shield Plugs Above Turbine Reheaters

SUMMARY

This evaluation justifies removing the twelve shield plugs located above the turbine reheaters.

The shield plugs are located in the low roofs that run the length of the turbine bay. The plugs

.I allow access to the turbine intercept and stop valves. The roofs were installed solely to reduce i skyshine. Skyshine is the radiation dose rate around the site environs resulting from various l g turbine bay N-16 sources which travel upward through the turbine building roof and then scatter off air molecules and return to earth. Skyshine dose rates result from normal power operation. Shield plug removal limits the localized heatup in the vicinity of the turbine valve l3 g unitized actuators. It allows greater radiative heat transfer from the unitized actuators resulting g in lower actuator operating temperatures. At the present time, one shield plug is removed.

This safety evaluation assumed that all twelve plugs were removed to provide a bounding evaluation for the possible future removal of additional low roof shield plugs.

The reheater low roofs do not protect equipment important to safety and do not provide an

. accident mitigation safety function. They are not part of any other plant systems or components. The openings created by the removal of the low roof shield plugs will not cause or initiate a major accident involving the reheaters, steam piping, valves, or unitized actuators.

Skyshine calculations indicate that with full plug removal, the skyshine dose could increase by 2.5%. However, this increase is considered insignificant because the measured skyshine values are much lower than the values originally estimated in the UFSAR. There is no impact on the local ventilation system because the existing air flow pattems are not affected. There is no significant increase in radiation exposure rates. Therefore, there is no affect on radiation access control. The low roof floor loading capability is not affected by the removal of the plugs. There is no affect on the fire protection system because the low roof is not a rated fire bamor and the safety margins implied in the Offsite Dose Calculation Manual are not affected bect use skyshine is not considered an effluer t.

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[L Safety Evaluation No: 93-0015 UFSAR Revision No. 7 b

Reference Document: ABN 13861-1 ,_, Section(s) NA

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h. Figure Change x Yes No Title of Change: Residual Heat Removal (RHR) System Keep Fill Lineup Change

SUMMARY

( This evaluation justifies isolating the RHR keep fill condensate makeup control valve E11-F100 and opening the keep fill pressure condensate makeup valve bypass valve E11-F088 to maintain condensate makeup to the RHR discharge piping. The 3/4" makeup control valve could not maintain adequate makeup flow due to accumulation of material in th valve inlet and cage. The quantity of foreign material was minimal to have any impact on the RHR System funcbon (RHR System inventory is >20,000 gallons); however it was large enough to impede the makeup flow through 3/4" valve causing pressure to drop downstream of E11-F100 and activate the keep fill alarm set at 50 psig decreasing. Using the 4" bypass valve maintains adequate makeup flow to the RHR keep fill system. The new valve lineup improves RHR keep fill system reliability because the passive bypass valve is less prone to failure than the active

{ makeup control valve. The increased keep fill pressure (the keep fill system pressure is equal to the condesate system pressure) applies additional seating pressure to the RHR pump discharge check valves resulting in a greater potential to reduce any leakage to the torus.

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The RHR keep fill system is not part of any system that contributes to any accident evaluated in the UFSAR. The 100 psig condensate supply pressure is well below the 300 psig RHR

[ piping design pressure and any over-pressure transients would be mitigated by the RHR relief valves.

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Safety Evaluation No: 93-0019 UFSAR Revision No. 7 Reference Document: ABN 13791-1 Section(s) 9A.2; 9A.4; 9A.5 LCR 93-0080-UFS Table (s) NA I Figure Change Yes X No Title of Change: Drawing and UFSAR Fire Door Rating Changes

SUMMARY

This safety evaluation justifies the following changes to plant drawings and the UFSAR:

1. Fire protection evaluation drawings A-2413 through A-2415 and A-2419 through A-2423 have been revised to show the correct fire rating on certain doors whose fire ratings were incorrectly identified. Certain Class B and C doors were incorrectly identified as Class A fire doors.

2 Turbine building door schedule drawing A-2072 and radwaste building door schedule drawing A-2073 have been revised to show the correct fire rating. Certain Class A and Class C doors were incorrectly identified as Class B fire doors.

3. A note has been added to drawings A-2072 and A-2073 to state that Class B fire doors supplied without windows are constructed in the same manner as Class A doors and I are, therefore, considered equivalent to Class A fire doors.

4 UFSAR Subsection 9A.2.3.1.5 has been revised to state that " doors in rated barriers are either listed or labelled by a nationally recognized laboratory or have been evaluated by a specific fire hazards analysis". Previously, only doors that were either I labelled or evaluated by a specific fire hazards analysis were permitted. This change brings the UFSAR text into conformance with NFPA (National) Codes and standards by permitting the use of listed as well as labelled doors.

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5. UFSAR Subsections 9A.4.4.1, 9A.4.4.2, and 9A.4.5.1 have been revised to document the presence of both Class B and C fire doors in the three hour rated fire barrier separating the turbine and radwaste buildings. These changes bring the UFSAR into conformance with the fire protection evaluation and door schedule drawing changes above.

6. UFSAR Subsection 9A.S.d.1.] and 9A.S.F.14 have been revised to refer back to the fire hazards analysis portion of UFSAR Section 9A.4 for a detailed discussion of specific fire door ratings. This change is required because of the presence of non-Class A fire doors in the turbine / radwaste building interface as described above.

These changes resolve inconsistencies between the fire protection evaluation drawings, the turbine and radwaste building door schedule drawings, and the UFSAR.

SAFETY EVALUATIONS AS-BUILT NOTICES I PAGE 7 Safety Evaluation 93-0019 (Continued)

These changes reflect actual field conditions and no physical changes were made to the plant.

They do not affect the operation or reliability of any plant systems or equipment. The reduction in the fire resistance ratings for the subject doors does not reduce the level of fire barrier protection because: (1) the Class B doors covered by this safety evaluation are constructed in I the same manner as Class A doors and can be considered equivalent to Class A doors and; (2) the Class C doors covered by this safety evaluation are provided with automatic fire detection on one side and an insufficient combustible loading on both sides of these doors to sustain a fire of sufficient intensity and duration to breach a 3/4 hour rated fire barrier. The use of labelled and listed doors is acceptable because, in either case, the door is rated and approved as a fire door for a specific fire duration. The UFSAR changes that reference other ,

UFSAR sections are editorial in nature.

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Safety Evaluation No: 93-0050 UFSAR Revision No. 7 Reference Document: ABN 14009-1 Section(s) NA Table (s) NA 4!I Figure Change X Yes No I Title of Change: Revision to UFSAR and Other Drawings on EDG System

SUMMARY

This evaluation justifies revising drawings 6M721-5734 and 6M721-5827 to:

• Add pipe cap or plug symbols to the end caps for certain drain valves.

I 3 = Add the PIS number, valve symbol, and end caps for the drain valves on the EDG air jE compressor pressure switch sensing lines. l

• Revise the gate valve symbol to ball valve for the ball valves provided by the EDG vendor. I These valves meet the requirements.

in addition, this evaluation justifies adding a pipe plug symbol to UFSAR Figure 1.7-1 and adding a

.I ball valve symbol in lieu of gate valve symbol in UFSAR Figures 9.5-7 and 9.5-11 pertaining to Emergency Diesel Generator System. The change to the UFSAR Figure 1.7-1 is an editorial

. change to add a new symbol to the list of acceptable symbols and legends. The changes to the UFSAR Figures 9.5-7 and 9.5-11 are merely to reflect plant as-built conditions.

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Safety Evaluation No: 93-0052 UFSAR Revision No. 7 Reference Document: ABN 14040-1 Section(s) NA Table (s) NA I Figure Change l X l Yes U No Title of Change: Reactor Water Cleanup (RWCU) System Valve Configuration Change

SUMMARY

This evaluation changed the RWCU system maintenance isolation valves G33-F027A(B) from normally closed to normally open in order to prevent overpressurization of the 150 psig rated piping associated with the vents and drains of the RWCU demineralizers. Changing the maintenance valves from normally closed to normally open does not prevent the RWCU I system from isolation on the appropriate signal and does not increase the likelihood of a pipe break because the valve configuration change prevents the overpressurization of the 150 psig piping involved.

The only system functional change made by having the G33-F027A(B) valve open is the continuous vent of high pressure leakage from the low pressure piping to the RWCU phase separators. The venting does not affect the safety-related function of isolating the RWCU I system.

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Safety Evaluation No: 93-0060 UFSAR Revision No. 7 Reference Document: ABN-26020-1 Section(s) NA

{ Table (s) NA Figure Change X Yes [_] No Title of Change: Updating the UFSAR and Other Design Documents on Post Accident Sampling System (PASS) Refrigeration Unit

SUMMARY

PDC-13854 evaluated and installed a like-for-like replacement refrigeration unit for an obsolete unit in the PASS. However, the new vendor supplied unit did not provide a bypass valve (P3400F062) around the chilled water pump. Discussions with the vendor confirmed that the bypass valve is not needed for the system operation. The bypass valve on the old refrigeration unit was to reduce flow of the pump that was oversized to handle a variety of applications. The newer pump, now installed in the system, is much smaller and a low head pump and, thus, does not require a bypass valve. This evaluation justifies that the missing valve is not needed and also updates the UFSAR and other design documents on the PASS.

The refrigeration unit installed under PDC-13854 is the vendor recommended replacement which enhances the system by having increased cooling capacity. The PASS is to provide ,

r remote sampling capabilities under post-accident conditions when radiological conditions L which make routine samples unavailable. The new refrigeration unit performs the same function as the original unit.

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Safety Evaluation No: 94-0098 UFSAR Revision No. 7 Re'erence Document: ABN 27063-1 Section(s) None Table (s) 8.3-2, 8.3-3, 8.3-4, 8.3-6, 8.3-7 Figure Change Yes X No Title of Change: Revision to UFSAR EDG Loading Tables to incorporate Plant As-Built Conditions

SUMMARY

I Licensee Event Report (LER)94-003 corrective actions resulted in a review of all Loss of Power (LOP) and Loss of Power / Loss of Coolant Accident (LOP /LOCA) procedures, schematics, load diagrams and design calculations overlaps. This review resulted in an I issuance of As-Built Notice (ABN) 27063-1, which in tum required the following changes in the UFSAR Tables referenced above:

Added load for PIS number P44F606A (0.6 kW) to EDG-12 since the valve changes I

position with LOCA signal Added load for PIS number R1700S016B (27kW), C4101S002 (40kW) and T4901F602 I

(0.3kW) to EDG-13. l Added load for PIS number G1154F018 (0.4kW) to EDG-14 I

i The above changes in the UFSAR tables result in the highest total load of 3123 kW on EDG- i 14 with loss of off-site power and loss of coolant accident at 0-10 minutes with all diesel I generators available. Similarly, EDG-12 and EDG-13 highest totalloads were calculated to be 3104 kW and 2878 kW respectively.

It was determined that the highest load (EDG-14) is still within the short time rating of the EDG in accordance with paragraph C.2 of the Regulatory Guide 1.9, Revision 2 and item 3.7.2 of IEEE Standard 387-1977.

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Safety Evaluation No: 94-0107 UFSAR Revision No. 7 Reference Document: ABN-27142-1 Section(s) NA Table (s) 8.3-6 Figure Change Yes X No Title of Change: Revising UFSAR Table 8.3-6

SUMMARY

ABN 27142-1 added 1 KW load to EDG-14 for the following reasons:

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1. EDP 11938 was issued on September 24,1991, to replace the MOV mo^or (P4400F6018) with a smaller size, which reduced the load by 1KW on EDG-14. The EDP did not correct table numbers in DC 5003 and Table 8.3-6 in the UFSAR since the charige was in the conservative direction.

2. EDP 27064 was issued on September 27,1994, to replace the MOV (P4400F6018) motor with a larger size which increased the load by 1KW on EDG-14. The EDP corrected table number 5 in DC 5003 for the addition of 1KW and, at the same time, subtracted 1KW to correct the problem of item 1 above.

3. ABN 27063-1 was issued on October 6,1994, to add 0.4 KW (G1154F018 MOV) on EDG-

14. The ABN corrected table number 5 in DC 5003 for the addition of 0.4 KW and at the same time subtracted 1KW to correct the problem of item 1 above.

In conclusion, the problem in EDP 11938 was corrected twice (in EDP 27064 and ABN 27063-

1) and that caused subtracting 1 KW load twice, which resulted in 1 KW shortage.

It has been determined that the total load on EDG-14 for all conditions is within the short time rating of the diesel generator in accordance with paragraph C.2 of Regulatory Guide 1.9 Revision 2 and item 3.7.2 of IEEE Standard 387-1977.

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Safety Evaluation No: 90-0061, Rev 2 UFSAR Revision No. 7 Reference Document: EDP 7808 Section(s) NA Table (s) NA Figure Change Yes No Title of Change: Replacement of the Sound Powered Headset Communication System with a Telephone System

SUMMARY

This modification converted the existing sound powered headset system to a telephone system; provided additional telephone jacks, telephones, and sound proof telephone booths; provided additional Hi-Com public address handsets, amplifiers, and loud speakers; and improved the radio and Hi-Com communications in the control room. This modification provides expanded communications coverage and enables the plant operators and l&C technicians to conduct their activities more efficiently.

This change does not affect the operation of safety related equipment or systems and does not impact existing accident analyses. This modification pertains to non-Q systems and does not impact the fire protection or Appendix R criteria. The direct communication system between the control room and the refueling platform required by the Technical Specifications is not affected by this modification.

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Safety Evaluation No: 91-0009 UFSAR Revision No. 7 Reference Document: EDP 11890 Section(s) NA Table (s) NA h

Figure Change Yes No Title of Change: Transferring Oil from the Turbine Building Oil / Water Separator Directly to the Waste Oil Reprocessing Tank (WORT)

SUMMARY

This evaluate justifies transferring the oil output of the turbine building oil / water separator to the WORT instead of transferring it to the turbine building waste oil tank. This will reduce turtune oil purchases by allowing the turbine waste oil in the turbine building waste oil tank to be recycled. The waste oil from the turbine building oil / water separator contains solvents and is not compatible with turbine waste oil.

The routing of oil through the turbine building waste oil tank is not a design requirement for oil processing. There is no impact on the function of the radwaste system or on any accident analyses in the UFSAR.

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Safety Evaluation No: 910017, Rev 1 UFSAR Revision No. 7 Reference Document: EDP 11583 Section(s) NA

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[ Figure Change Yes No Title of Change: Residual Heat Removal (RHR) Service Water Piping Relief Valve Replacement

SUMMARY

This modification replaces RHR service water relief valves E1100F056A and E1100F0568 with lower pressure relief setpoints. The original relief valve setpoint was 450 psig. The replacement relief valve setpoint is 175 psig, the design pressure of the service water piping.

This modification will protect the service water piping from exceeding its design pressure.

The replacement valves comply with the requirements of ASME Code Section Vill. There are

{ no accidents evaluated in the UFSAR that apply to this section of piping. The new valves will lessen the chance of an accident due to overpressurization. The relief valve setpoint is above

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the shutoff head of the RHR service water pumps (approximately 140 psig). Therefore, inadvertent valve lift is unlikely during normal operation.

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Safety Evaluation No: 91-0084 UFSAR Revision No. 7 Reference Documen': EDP 11783 Section(s) NA Table (s) NA

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( Figure Change x Yes No Title of Change: Installation of a Platform at the Residual Heat Removal (RHR) Complex

SUMMARY

( This change installs a metal platform and reinforced concrete slab at the existing north center access point of the RHR complex. It provides additional space in which to load, unload, and 7

move equipment and material to the first floor entrance door with forklifts or trucks. This larger working area enhances personnel safety when handling equipment and material.

This modification does not change the function or operation of any system because it is l

located outside of the RHR complex and is not part of any boundary, barrier or system that is required for plant operation, safe shutdown, or maintaining the plant in a safe shutdown condition. The new platform does not create any conditions different from existing conditions because two similar platforms exist at the north and south RHR complex access points.

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Safety Evaluation No: 91-0111 UFSAR Revision No. 7 Reference Document: EDP-12778 Section(s) NA l

Table (s) NA

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l Figure Change [Xl Yes l l No Title of Change: Moving of Vent Connection From Pump Nozzle to Pipe f

SUMMARY

This modification moved the 1-inch vent from the 16-inch heater drain pump nozzle to the 16-inch

! pipe. The threaded vent connechon to the pump nozzle had failed at all three heater drain pumps.

l The vent connechon was located on a highly stressed nozzle, with the threaded connection i conbibuting to the stress increase. To reduce this stress, and reduce the stress intensificabon, the

[ vent connecbon was moved to the 16-inch pipe. The new tap connects to the 16-inch pipe with a welded sockolet. The stress intensification of a welded connecbon is less than that of the original threaded connecbon The funcbon of the 1-inch vent will not change as a result of the slight locabon change. The vent will still act to vent any air from the heater drain pump and its associated piping durin0 fill and start up of the system. Relocating the vent downstream of the instrument tap will have no impact on the funcbon of the vent, the instrumentabon, the pump, nor any other Heater Drain Pump System components i

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Safety Evaluation No: 924023 UFSAR Revision No. 7 Reference Document: EDP 10129 Section(s) 7.6 Table (s) NA

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( Figure Change Yes No Title of Change: Installation of the Emergency Response Data System (ERDS)

SUMMARY

j This modification installed the ERDS at Fermi 2. ERDS is an electronic data transmissum system that was developed to provide the NRC Operations Center with accurate and timely data during emergencies on the following plant parameters:

o Reactor core and reactor coolant system conditions.

o Containment conditions.

o Radiosctivity releases.

o Meteorological conditions.

E Using a modem provided by the NRC, the Fermi 2 ERDS data link will operate as an integral part of the Emergency Response information System (ERIS) in conjunebon with. the Emergency Nclifier. tion System (ENS). Fifty-nine (59) ERIS data points transmit the r it' formation for em above plant parameters and the ENS red phone b used for voice h transmission of essential data not available on the system. The ERDS will be used only during emergencies and will be activated by Detroit Edison during declared emergencies classified at r the alert level or higher. The system is activated with the ERIS safety parameter display L system (SPDS). This change was made in accordance with the NRC requirements in NUREG 1394, Revision 1,

• Emergency Recoonse Data System (ERDS) Implementation" and satisfies

- the NRC commitment made in response to the final rule on ERDS.

L The plant specific parameters transmitted to the NRC have been rev'ewed and approved by the NRC. This modification does not change the operation or function of any system, equipment, or component important to safety. The ERIS computer is a non-safety related system and the only function of the modem is to establish the data link with the NRC.

Therefore, failure of this equipment will not affect any system, equipment, or component i important to safety.

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Safety Evaluation No: 92-0024 UFSAR Revision No. 7 Reference Document: EDP 12825 Section(s) NA Table (s) NA Figure Change Yes No Title of Change: Replacement of Off-Gas System Condensste Receiver Tank Float Switches with LevelTransmitters

SUMMARY

This modificahon upgraded the level detechon instrumentation on the off-gas system condensate receiver tank. The original Magnetrol float switches were replaced with level f transmitters with sealed impulse lines. The transmMer outputs go to bistables that perform the h

same funcbon as the original float switches. This modification eliminates spurious condensate receiver tank low level alarms caused by steam flashing in the condensate tank instrument

f. lines.

This modification does not change the function of the alarm design. The bistables have the l

same setpoints as the original float switches. The new instrumentation does not affect equipment important to safety. If the new instrumentation were to fail, the off-gas condensate l receiver tank could flood causing a loss of steam jet air ejector funcbon, causing a subsequent loss of condenser vacuum, and resultant turbine trip. This scenario is bounded by the UFSAR turbine trip analyses. I r i t

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Safety Evaluation No: 92-0025, Rev 1 UFSAR Revision No. 7 h

Reference Document: EDP 12914 Rev A Section(s) NA Table (s) -NA Figure Change Yes No p Title of Change: Relocation of the Lawn Sprinkler System Supply Tap L

SUMMARY

( - This modification moved the lawn sprinkler system supply tap from the bottom of the 24" general sennce water (GSW) header to the 48" GSW header vertical pipe run located in the southeast comer of the turbine building basement. Valve P4100F810 was installed as part of

(. the " hot tap" process used to make the new connection. This valve remained in place and became the sprinkler system header isolation valve. The original connechon trapped sitt and sediment which plugged the 2" lawn sprinkler header during the winter season when the line

' was not in use. The sprinkler header could not be cleaned because therc Jre no cleanout plugs. The header could not be isolated and disassembled because sprinkler header isolation valve P4100F043 leaks by. It was undesirable to repair P4100F043 because it would require a GSW system outage. The new sprinkler system tap location eliminates the depostbon of silt and sediment and the resultant sprinkler header pluggage.

The potential for malfunction due to the relocation of the lawn sprinkler tap is no greater than it ;

was for the original tap location. There is no impact on plant system cooling because the amount of GSW water used by the sprinkler system is insignificant. The GSW system does not support any equipment required for safe reactor shutdown. This modification does not change ,

s the funcbon of the GSW system and there is no effect on the analyses contained in the  !

UFSAR or the Fermi 2 Safety Evaluation Report.

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Safety Evaluation No: 92 0056 Rev 1 UFSAR Revision No. NA Reference Document: EDP 13210 Section(s) NA NA

{ Table (s)

[ Figure Change Yes X No Title of Change: High Pressure Turbine Govemor Valve Modifications

SUMMARY

This modification replaced the original drilled sleeves on high pressure turbine govemor valves N3021F004A, B, C, and D with shorter sleeves. The purpose of the drilled sleeves was to provide additional valve head support at low lift positions to minimize mechanical looseness.

The new sleeve extends farther into the valve cover to compensate for the loss of valve head support near the seat. The additional flow area between the valve seat and the now, shorter sleeve reduces the pressure drop across the valve and provides adequate throttle margin for power uprate conditions. To ensure that the Cycle 4 Reload Licensing Analysis (RLA) was

{ bounding for govemor valve operating positions between 75% and 88% open, the operating limit minimum critical power ratio (OLMCPR) was increased by 0.01.

The modified valves are functionally equivalent to the original valves. There are no turbine govemor control system hardware modifications made by this modification and the govemor valve stroke and stroke time are not affected. The OLMCPR originally established by the Cycle 4 RLA ensured that the safety limit minimum critical power ratio (MCPR) margin would be i maintained with the turbine govemor valve position is equal to or greater than 88%. Operation of the turt>ine govemor valves between 75% and 88% open reduces the entical power ratio margin by less than 0.01. The OLMCPR was increased 0.01 to maintain the safety limit MCPR.

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SUMMARY

l Safety Evaluation No: 92-0077 UFSAR Revision No. 7 l

Reference Document: EDP 13464 Section(s) NA j Table (s) NA I

l- Figure Change Yes No Title of Change: High Pressure Coolant injection (HPCI) Test Retum Isolation Valve Motor l Replacement

SUMMARY

i This modification changed the HPCI test retum isolation test valve E4150F008 motor from a 60 ft-lb motor to a 80 ft-lb motor and replaced the spring pack to accommodate the new motor.

l The maximum and minimum torque switch settings were reduced and anti-cycling logic was added to the auto-close circuitry. The original motor was damaged and the higher capacity

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replacement motor provides the appropriate additional design margin. The torque switch settings were changed to accommodate the new spring pack. The anti-cycling circuitry was installed to prevent motor cycling during valve closure.

The torque switch settings, maximum allowable thrust values, motor replacement, and spring pack replacement do not impact structures, systems, or components described in the UFSAR.

] The anti-cycling logic does not change the function of the valve and, therefore, has no negative impact on secondary containment isolation or full flow test requirements described in the UFSAR. The torque switch setting calculations satisfy the requirements of the Fermi 2 response to NRC Generic Letter 89-10. The additional motor weight (16 lbs) imposed on the valve and the piping system has been evaluated and is acceptable. The use of a larger motor

- does not adversely impact the DC electrical distribution system. ,

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Safety Evaluation No: 92-0086, Rev 1 UFSAR Revision No. 7 Reference Document: EDP 13368 Section(s) NA NA

[ Table (s)

( Figure Change Yes No Title of Change: Rerouting the Scram Discharge Volume (SDV) Vent Line and Residual Heat

(. Removal (RHR) Pump Drain Unes

SUMMARY

This modification changed the terminus of the SDV vent line from the reactor building ventilation exhaust duct catch pan to the phase separator overflow line. This modification was initially implemented using the original single vacuum breaker to ensure that no vacuum will

( prevent vent path drainage. However, when this modification was fully implemented, two redundant QA level 1M vacuum relief valves were installed. This change also relocated the

[ terminus of the RHR pump A seal leakage and pump cover drain lines from the radwaste overflow line to the G1101D073 sump. The SDV. vent line relocation provides a more direct path for the colleebon of ejected reactor condensate in the radwaste system and eliminates the i

post-scram contamination problems associated with the previous vent line configuration. The redundant vacuum breakers ensure higher reliability and will allow preventive maintenance to be performed in any plant mode. The relocation of the RHR pump drain lines eliminates a r source of contamination from the phase separator tank overflow and prevents the potential for L redirection or relocation of SDV discharge contamination upon scram reset.

r This modification does not alter the function of the SDV vent line or the RHR pump drain lines.

The new SDV vent line runs are in the same general area of the RHR Division 1 heat exchanger room as the original vent line run. The RHR pump drain lines remain outside of the RHR system pressure boundary. The current use of the original SDV vent line vacuum breaker and the eventual use of the redundant vacuum relief valves ensure a free and clear SDV vent path that satisfies the design bases and commitments stated and evaluated in the UFSAR.

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Safety Evaluation No: 92-0088 UFSAR Revision No. 7 Reference Document: EDP 9207 Section(s) NA Table (s) NA I Figure Change Yes No Title of Change: Feedwater Control System (FCS) Changes

SUMMARY

This modification made the following changes to the FCS:

1. A limiter was installed on the feedwater/ steam flow deviation module to eliminate the possibility of a scram due to the failure of either a feedwater flow or steam flow transmitter. This limiter is set at 70% (30% less than steam flow) to provide sufficient I margin to avoid a reactor scram on high water level.

2. The feedwater flow totalizer, C32-K609 was removed to reduce excess components from I the FCS. The totalizer was not used for any process control or monitoring function.

3. The reactor feedwater pump turbine (RFPT) run up motor operated potentiometer (MOP) logic was modified to automatically drive the MOP up to maximum and deactivate it whenever the FCS RFPT controller is placed in automatic. The original logic automatically drove the MOP up to maximum and deactivated it when the FCS RFPT I controller was in automatic and its output was greater than 10 ma. The 10 ma input was deleted because it is unnecessary and to reduce excess components.

J5 4. The loss of heater drains limiter (#1 of dual limiter C32K610) was removed to reduce the chance of a scram due to its failure. The dual limiter was rewired for single limiter function.

5. The system diagram was revised by deleting the QA level of various feedwater control '

components.

6. Components C32N003A through D and C32N009A through D have been identified as QA level 1 components because their process inputs connect directly to other devices  ;

that are safety related.

These changes are designed to improve FCS performance. FCS failure can result in either an increase or decrease in feedwater flow. Both of these transients are analyzed in the UFSAR.

Accidents and transients involving feedwater flow excursions previously evaluated in the UFSAR are made less probable by this modification because it reduces the probability of FCS I

SAFETY EVALUATIONS l ENGINEERING DESIGN PACKAGES

I PAGE 13 Safety Evaluation No. 92-0088 (Continued) failure (through component reduction) and the effects of FCS component failure (through the

'I use of controllimiters). Safety related devices are located in the same cabinet as the balance of plant components changed by this modification. However, any FCS circuit failure will not I impact the other circuits in the cabinet.

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Safety Evaluation No: 92-0089 UFSAR Revision No. 7 Reference Document: EDP 10257 Section(s) 10.4 Table (s) NA Figure Change Yes No I Title of Change: Main Turbine Gland Sealing Steam Pressure Regulating instrumentation Modification

SUMMARY

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I This modification upgraded the original turbine gland sealing steam pressure controller and added a second pressure controller. The first controller regulates the startup gland seal steam pressure regulating valves N3000F430 and N3000F431 and the second controller regulates I the normal gland seal steam regulating valve N3000F433. New tubing was added from the output of the new controller to the normal control valve N000F433 and the tubing between this valve and N000F430 and N000F431 was disconnected. The modification also replaced the l

l gland seal steam pressure transmitter. The turbine gland seal steam system was previously operated in the manual mode because the original control scheme did not work reliably in )

automatic. This modification was intended to provide a control scheme that allows reliable automatic gland seal steam pressure control in both startup and normal operating conditions.

This change does not affect the function of the turbine gland seal steam system because the i system will operate in its originally specified automatic mode. This modification does not l adversely affect any previously evaluated UFSAR accidents or transients because they I assumed automatic control of the turbine gland sealing steam system.

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SUMMARY

Safety Euluation No: 92-0090 UFSAR Revision No. 7 Reference Document: EDP 12636 Section(s) NA Table (s) 7.3-2 I Figure Change Yes X No Title of Change: Automatic Depressurization System (ADS) Timer Replacement

SUMMARY

This modification added two new Agastat ETR time delay relays to the ADS logic as g replacements for the original General Electric GE CR2820 time delay relays. The Agastat time E delay relays were installed in pa.allel with the original relays. The GE CR2820 relays were retained to provide the in.e+antaneous contacts needed to activate control room annunciator ,

" ADS Timers initiated". This modification was recommended in General Electric Service '

I information Letter SIL 230 because, for this application, the original relay timer setpoint accuracy was found to be beyond the manufacturer's specified +/-10%. The replacement relay f timers maintain the required accuracy. The new time delay relays allow plant maintenance to j

i conduct the timer setpoint periodic surveillances on a quarterly basis as specified in the l Technical Specifications. Previously, the surveillances were performed monthly to ensure  !

l setpoint accuracy.

This modification does not adversely affect the performance of the ADS logic circuits. The  ;

ADS timer setpoint and initiation logic remain unchanged. This change does not introduce any i new ADS timer failures and maintains the same level of ADS component redundancy. An ADS  !

timer failure would result in reduced ADS logic initiation circuit redundancy. However, the ADS logic control has two separate circuits (logic A and B). Each independent circuit is capable of automatically opening the five ADS valves to accomplish the ADS function. The Agastat time ,

delay relays have been procured, designed, tested, and installed in accordance with the same I criteria as the original hardware. l l

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SUMMARY

Safety Evaluation No: 92-0092 UFSAR Revision No. l l

Reference Document: EDP 13712 Section(s) NA Table (s) NA j Figure Change Yes C No Title of Change: Offgas Holdup Pipe Collector Tank Low Level inpui Removal

SUMMARY

This modification removed the offgas holdup tank low level switch N62N464 input to common offgas annunciator and sequence-of-events recorder. The original purpose of the alarm was I to wam the operators of a possible offgas holdup pipe collection tank drain valves closing circuit malfunction. Such a malfunction would increase the main condenser piessure resulting in a possible turbine trip. However, the offgas holdup pipe collector tank normally runs dry because the ambient air tc,nperature is hotter than the process temperature; any liquid collected in the tank is quickly evaporated. As a result, this alarm was continuously received in the control room. This modification eliminates the alarm state. l There is no equipment important to safety in the vicinity of the offgas holdup tank, valves, or control / alarm circuits. This alarm circuit does not have a direct input that could cause an accident or transient. The circuit is not addressed in the Technical Specifications, the UFSAR, or the Fermi 2 Safety Evaluation Report (NUREG 0798). Failure of the offgas holdup pipe collection tank drain valves can still be identified by the control room operators through the high main condenser back pressure annunciator alarm.

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[L Safety Evaluation No: 93-0001 UFSAR Revision No. 7

[- Reference Document: EDP 13684 6.3; 9.2 Section(s)

NA

(: Table (s) h Figure Change Yes No Title of Change: Residual Heat Removal (RHR) Keep Fill Water Source Change

SUMMARY

This modification changed the RHR keep fill system water source from the domineralized water system to the condensate storage and transfer (CST) system. This eliminated the build up in CST inventory due to RHR keep fill venting activities and RHR pump discharge check valve leakage and the eventual release of the excess CST inventory to the environment. This change helps Fermi 2 to meet its zero discharge goal.

This modification did not alter the RHR portion of the keep fill system. Divisional separation of each RHR keep fill line is maintained. Condensate conductivity, chlorides, and ph are well within chemistry specifications. The condensate water supply is not a part of any accidents evaluated in the UFSAR.

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Safety Evaluation No: 93-0003 UFSAR Revision No. 7 Reference Document: EDP 12689 Section(s) 9.4 Table (s) NA I Figure Change Yes X No Title of Change: Fan-Coil Cooling Unit Control Modifications

SUMMARY

This modification changed the control scheme of eleven fan-coil cooling units from I temperature-based automatic control to continuous running manual control. This change was made on the Division I and ll standby gas treatment system (SGTS), residual heat removal (RHR), emergency equipment cooling water (EECW), and hydrogen recombiner system I coolers. The Division I core spray system / reactor core isolation cooling (RCIC) system, Division ll core spray system, and Division 11 high pressure coolant injection (HPCI) system coolers were also modified. The c.iginal 3-position (run, auto, and off-reset) switches were I replaced with 2-position (run and off-reset) switches; the motor trip alarm circuit " auto-off" and "high temperature" contacts were eliminated; and the temperature controller function was disabled. The controller temperature indicator will continue to function. The existing safety related power supply is used. However, since both the controller and temperature indicator I have been down graded from Q to non-Q, the fuses are coordinated to provide the required isolation between the temperature indicator and the safety related power supply. This modification was made because, at the temperature control switch setpoint of 75*F, the fan-I coil cooling units were running continuously while in the automatic mode. Therefore, there is no need for automatic temperature control. This modification increases the reliability of the fan-coil cooling units by eliminating unnecessary control components and eliminates periodic calibration and procurement of qualified temperature controllers.

This modification does not change the function or operation of any system, equipment, or component important to safety and does not change the function of the subject fan-coil cooling units. The subject fan-coil cooling units are designed to run continuously. An engineering evaluation of this modification indicates that running the fan-coil cooling units continuously will not lower the room temperatures to the extent that safety related equipment is affected. The I. resulting temperature of the emergency safety features equipment remains in the normal range.

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SUMMARY

l Safety Evaluation No: 93-0005 UFSAR Revision No. 7 l Reference Document: EDP 12741 NA Section(s)

Table (s) NA l

l Figure Change Yes No Title of Change: Standby Feedwater (SBFW) Pumps Discharge Line Vent Modification E

SUMMARY

This modification extended the SBFW pumps discharge line vent valve to the G1101D017

] sump. This change added a new vent valve and sight glass located within easy reach of the operator. The original vent valve N2103F327 is maintained in the open position. This modification eliminated the need to climb 30 feet overhead to vent the line.

This modification has no adverse affect on the operation or reliability of the SBFW system.

This change does not affect the design bases, functions, or operation of any other structures, systems, or components. All previous accident analysis results remain unchanged.

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Safety Evaluation No: 93 0010, Rev 1 UFSAR Revision No. 7

{ Reference Document: EDP 13775 Section(s) NA

{ Table (s) NA

[ Figure Change Yes No Title of Change: Potable Water System Piping Modifications

SUMMARY

This modification made the following changes to the temporary portion of the potable water system:

1. The Building 40 water supply was moved closer to header supply isolation valve P2100F045. This allowed the old Kuhlman temporary warehouse line to be abandoned.

This line was leaking and could not be isolated.

2. A blowdown #iushing hydrant was installed downstream of P2100F045 and the Building 40 suppif line. It is painted yellow to distinguish it from fire protection hydrants. This hydrant provides a means of flushing the temporary potable water system main.

3. The line downstream of the blowdown / flushing hydrant was capped tominimize the amount r of piping that can leak.

UFSAR Figure 9.2-5 was modified to show the above changes to the potable water system.

1 These modifications were performed on the temporary portion of the system and are outside of the protected area. The potable water system is a non-safety related, QA level NQ, and non-seismic system. This system does not interconnect with any system important to safety and is I

not required for the continuous reliable operation or safe shutdown of the reactor. No credit is taken for the potable water system in UFSAR accident analyses.

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Safety Evaluation No: 93-0012 UFSAR Revision No. 7 Reference Document: EDP 11436 Section(s) NA Table (s) NA I Figure Change Yes No Title of Change: Turbine Building East Floor Drain Sump Pump Replacement

SUMMARY

This modification replaced the turbine building east floor drain sump G1101D017 sump pumps l I G1101C004 E and N. The original pumps were designed for a flow of 50 gpm at a total head of 93 feet. The new pumps are designed for a flow of 50 gpm at a total head of 115 feet. The new pump motors have the same power rating as the original pumps. Piping modifications l

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were performed as necessary to align the piping with the replacement pumps and removable I pipe sections and pipe tees were added to the discharge piping to provide cleanout capability.

The new pumps restore the system to its design capacity and maintenance is enhanced because spare parts are easier to obtain and cleanout capability allows the discharge piping to I be cleaned and flushed to reduce system pressure drop and remove obstructions.

This modification does not change the function or operation of the sump or sump pumps. The affected components do not interface with or support safety related equipment nor are they located in the vicinity of safety related equipment. The input to the radwaste collector tanks is still controlled by the pumping capacity of the oil / water separator. The new pumps' capacity is within the range described in UFSAR Subsection 9.3.3. There are no accidents that specifically evaluate the turbine building floor drains or the sump pumps and the changes do not contribute to the seismically induced tank rupture accident described in UFSAR Subsection 15.7.3.

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Safety Evaluation No: 93-0013 UFSAR Revision No. NA Reference Document: EDP 1598 Section(s) NA NA

( Table (s)

Figure Change Yes X No Title of Change: Primary Containment Atmospheric Monitoring System (PCAMS) Design Modification Deletion

SUMMARY

This safety evaluation justified the deletion of the design modification for replacing the existing non-qualified PCAMS drywell thermocouples. EDP 1598 was issued originally to replace 4 torus and 14 drywell non-qualified PCAMS thermocouples with safety related, environmentally and seismically qualified thermocouples. The purpose of this modification was to meet the Regulatory Guide 1.97 requirements for accident condition assessment. During the first refueling outage,4 torus thermocouples were replaced under EDP 1598. The remaining 14 thermocouples were to be replaced at a later date. Subsequently, in correspondence with the NRC, Detroit Edison stated that it was in compliance with Regulatory Guide 1.97 becauso 2 drywell,2 torus air space, and 2 torus water space thermocouples are environmentally and seismicady qualified. The letter also stated that the remaining thermocouples are not qualified.

The NRC accepted this position in their response dated May 2,1990. Therefore, performing the balance of this modification is no longer necessary.

The deleted modification is not required for plant safety. Since this is a deletion of a proposed modification, there is no change to the existing plant design. The existing qualified PCAMS thermocouples provide sufficient information to monitor and assess important containment variables. The drywell and torus temperature measurement functions described in the UFSAR l continue to be performed by the existing PCAMS thermocouples.

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SUMMARY

Safety Evaluation No: 93-0016 UFSAR Revision No. 7 Reference Document: EDP 13727 Section(s) 9.4

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Table (s) NA Figure Change Yes No B Title of Cl ange: Reactor Building Steam Tunnel Pressure Equalizing Line isolation Valve l Removal

SUMMARY

l I This modification removed the reactor building steam tunnel pressure equalizing line isolation valve T4100F208. The purpose of the pressure equalizing line is to provide a flow path between the steam tunnel and the reactor building so the ventilation system can maintain both the reactor building and the steam tunnel at a negative pressure for secondary containment integrity. The normally open valve was designed to minimize the short term environmental impact of steam entering the reactor building due to a serious steam leak or line break within the steam tunnel by automatically closing on high steam tunnel pressure. It was also designed to fail open on a loss of instrument air or a loss of power to its air supply solenoid valve T41F061. However, it was discovered that the as-built design of the isolation valve allowed

' the valve to fait closed upon loss of instrument air. As corrective actions, a temporary modification was initiated to block T4100F208 open and EDP 13727 permanently opened the E equalizing line. This modification assures that the steam tunnel is maintained as part of secor.dary containment.

T4100F208, its actuator, and logic were not electrically interlocked or mechanically linked with any other equipment. The secondary containment and standby gas treatment system continue to perform as designed and evaluated in the UFSAR and the Fermi 2 Safety Evaluation Report NUREG-0798. This modification maintains an open pressure equalization line between the y steam tunnel and the reactor building. The steam line break evaluation does not take credit for manual or automatic closure of T4100F208. Calculations show that the automatic closure 7 function of this valve on high steam tunnel pressure is not required to minimize the consequences of line breaks within the steam tunnel. The environmental consequences in the reactor building of a steam tunnel steam line break are negligible and, therefore, will not impact r the operability of equipment in the reactor building. There is no impact on the LOCA analysis L because the analysis assumes that the equalizing line is open. This modification has no impact on the level of fire protection in the steam tunnel and reactor building as there are insufficient combustibles on either side of the steam tunnel wall in the vicinity of the equalizing line.

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Safety Evaluation No: 93-0018, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-12447 Section(s) NA Table (s) NA

[ Figure Change X Yes No Title of Change: Changes in the Gland Sealing Steam System

SUMMARY

( This modificabon added manual isolabon valves to the sucbon and discharge of each gland steam exhauster. The gland steam exhausters have a common suchon and discharge line with a nonretum damper in each exhauster discharge line. In its original configurabon, it was not paaeiNa to perform mantenance on one exhauster with the system in service. The isolabon valves abow

( exhauster mantenance while the system is in service. In addebon, the old exhausters have been replaced with new exhausters of the same original capacity. The replacement was necessary snce the old exhausters were a high mantenance item, and spare parts were not readily avalable since

[ the ongnal vendor is no longer in bumness The new exhausters have TBCCW suppled to a shaft seal. The TBCCW pipng to the bearing lube oil cooling is not needed for now exhausters and, therefore, has been cut and capped.

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This modificabon does not affect the funcbon of the turbne gland sealog steam system. This system is not requwed for safe shutdown of the plant. The falute of this system wil not affect systems, structures, or components required for safe shutdown of the plant. The consequences of a suchon and/or discharge valve failure is no worse than the gland sealing steam system loss of vacuum acodent discussed in UFSAR Chapter 15. The capping of the TBCCW connechons to the beanng lube oil coolers has an insignificant impact on the operabon of the TBCCW.

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SUMMARY

j Safety Evaluation No: 93-0020 UFSAR Revision No. -7 ,

Reference Document: EDP 13672 Section(s) NA l Table (s) NA l

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l Figure Change Yes No j Title of Change: Primary Containment Pneumatic Supply (PCPS) System isolation Valve Installation and Check Valve Replacement  !

SUMMARY

This modification:

1. Installed twenty-seven one-inch manual isolation valves in the PCPS lines to drywell equipment that previously did not have supply line isolation capability.

2. Installed test connechons between the isolation valve and the accumulator check valve for the main steam isolation valve (MSIV), automatic depressurization system (ADS) relief valve, and the low-low set relief valve accumulators.

3. Replaced the accumulator check valves for the above accumulators with bolted bonnet, soft seat check valves. The original check valves were welded bonnet, hard seat valves.

These changes allow accumulator check valve maintenance and testing without taking a complete division of the PCPS system out of service and depresurrizing its associated headers. The use of accumulator check valves with soft seats reduces the leakage test failures and bolted bonnet check valves reduce the repair effort and ALARA concems associated with welded bonnet check valve repair procedures.

This modification does not change the operation or function of the PCPS system. All safety relateo components that rely on the PCPS system will perform their safety related funcbon as previously analyzed and evaluated in the UFSAR. All valves were purchased as ASME r Sechon lil, Class 3 (Group C) and meet the seismic and EQ requirements for this applicaten Stress and hanger calculations indicate that the valve installation is acceptable. The manual isolation and test valves are in the locked valve program and are posiboned to assure PCPS system function and pressure boundary integrity.

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0021, Rev 1 UFSAR Revision No. 7 Reference Document: EDP 13893 NA  ;

Section(s)

Table (s) . NA l l'

Figure Change Yes No l

l Title of Change: No. 4 Feedwater Heater Extraction Steam Line Modifications I

SUMMARY

i in April 1993, Fermi 2 experienced a #4 feedwater heater extraction line failure. The cause of this event was tentatively identified as a vibration resonant condition created by the piping and support changes made by engineering design package EDP 13692 in December 1992. Piping system pressure pulsations appear to be the source of vibration. The following modifications were made to the #4 feedwater heater extraction steam line as corrective action:

1. The extraction line expansion joint was removed from the extraction steam line located within the condenser. Removal of the expansion joint stiffens the extraction piping between the turbine and the condenser wall resulting in a fundamental frequency shift above the pressure pulsation frequencies.

2. A stiffening ring was added to the extraction line condenser penetration reinforcing gusset to reduce penetration stresses introduced by piping expansion that was originally accommodated by the expansion joint.

3. Five permanent pressure transducers and 1 temporary pressure transducer were installed ,

to measure the steam pressure pulsations and confirm the root cause analysis and i acceptable piping response.

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4. A drain connection was added to the bottom of the extraction line drain pot located outside of the condenser. This ensures that no water can collect in the pot and flash during power reductions.

These modifications do not change the function of any system. Structural design calculations j have been performed to ensure that the piping and condenser changes satisfy the applicable _!

design criteria. The turbine manufacturer approved the additional low pressure #2 turbine  !

extrachon steam nozzle loads that occur as a result of the expansion joint removal. Pressure l transducer pressure integrity failure was evaluated and the worst case failure was found to be a failure of the main steam manifold transducer which shares an instrument line with one of the main steam line low pressure sensors which feed the main steam line isolation logic. The  !

installation of this transducer does not compromise the ability of the main steam line isolation i

logic to close the main steam isolation valves. This logic is not required to mitigate the consequences of analyzed transients or accidents.

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SUMMARY

Safety Evaluation No: 93-0022 UFSAR Revision No. 7 l

Reference Document: EDP 13894 Section(s) NA Table (s) NA l

Figure Change Yes No

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Title of Change: Installation of Off Gas System Hydrogen Purge Valves I

SUMMARY

l This modification installed 1" globe valves downstream of the east and west off-gas recombiner trains inlet valves N6200F617 and N6200F618 and upstream of the east and west off-gas recombiner trains outlet valves N6200F622 and N6200F623. These valves provide the 1 l capability to purge the standby recombiner train prior to swapping trains and prior to l recombiner train maintenance. This modification will prevent backflow when the recombiner trains are swapped and the resultant migration of palladium catalyst into the preheater.

This change does not affect the operation, function, or performance of the operating off-gas E system train or any other interrelated systems. It does not affect or alter any analyses 3

discussed in the UFSAR or the Fermi 2 Safety Evaluation Report NUREG 0798.

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SAFETY EVALUATION

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Safety Evaluation No: 93 0023 UFSAR Revision No. 7

[ Reference Document: EDP 13907 Section(s) NA 7.3-1

{ Table (s)

[ Figure Change Yes x No Title of Change: High Pressure Coolant Injection (HPCI) Pump Suction Pressure Switch

[ Replacement.

SUMMARY

b This evaluation justifies replacing the HPCI pump suction pressure switch E41N010 with a different model pressure switch This switch monitors the HPCI pump suction pressure to

{ protect the pump from cavitation and water hammer. It provides a HPCl turbine trip signal and operator alarm if pump suction pressure falls below 15" Hg. vacuum. The original switch was a Static-O-Ring model 6N-AA21V-X9STT with a range of 30" Hg Vacuum to 75 psig and a 3.6%

span accuracy. The new switch is a Static-O-Ring model 54N6-BB118-NX-C1A-JJTTX13 with

{ a range of 30" Hg Vacuum to 0.5 psig and a 5% span accuracy. This modification also provided a new seismic pressure switch mount and new calibration tables to account for the resultant head correction change.

{

This change does not impact the performance of the HPCI system or any other accident mebgation systems. The replacement switch function has not changed and there is no change to the low pressure setpoint. The replacement switch has electrical, pressure, environmentti, and seismic ratings that meet or exceed those of the original model.

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Safety Evaluation No: 93-0025 UFSAR Revision No. 7 Reference Document: EDP-11002 Section(s) NA Table (s) NA I Figure Change X Yes No Title of Change: Instrument Control Loop Changes for #5 North / South Feedwater Heaters Level Controls

SUMMARY

This modification removed the split range control of the normal level control valves and emergency 1 drain line valves on the #5 N and S feedwater heaters. It installed indisidual controllers for each of these valves. An independent controller for each nonnal level control valve and emergency drain line valve will eliminate the series operation of the control valves. This change will improve the I response time and stability of the heater drains system. It also eliminates the instantaneous latch up of the forced closure cycle for the normal level control valves when the check valve (located in normal feed forward path) bounce occurs. This modification reduces the potential of unnecessary i reactor recirculation M-G set runback and manual operator control due to check valve bounce.

The modification is associated with nonsafety-related equipment. The majority of the equipment affected by this modification is locally mounted in the turbine building. In addition, scale changes have been made on manual loading stations in the main control room, while wiring and relay I

changes are made in the relay room The addition of new controllers and separating the instrumentation for nonna! drain line and emergency drain line into individual loops to maintain level in #5 feedwater heaters, and the addition of relays and wiring changes to remove instantaneous latch up of the forced closure cycle on check valve bounce does not change the original function of the system aM/or create the possibility of an accident of a different type than arry previously evaluated in the UFSAR. The modification made does not directly or indirectly adversely affect any safety related or important to safety function of the plant.

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Safety Evaluation No: 93-0029 Rev 1 UFSAR Revision No. 7 l

Reference Document: EDP-5173 Section(s) 7.7 I Table (s) 15.2.4-1 Figure Change @ Yes No l l

Title of Change: Addition of an Anticipated Trip Wthout Scram (ATWS) Initiated Trip to the Reactor Recirculation Pumps (RRP) Motor Generator (M-G) Set Drive Motor

. Breakers

SUMMARY

4

This modification added ATWS signal initiated trip coils to each RRP M-G set drive motor breaker.

The automatic trip occurs at either high reactor water pressure or reactor vessel low water level (level 2). This modification addressed an issue that was discussed in NRC letter, " Fermi 2 Compliance with the ATWS Rule,10CFR50.62 (TAC No. 59094)", dated October 10, 1989, conceming the reliability of M-G ni field brukers during an ATWS event. This modification

g ensures that both the field breakers and the M-G set drive motor breakers receive trip signals during

g an ATWS event resulting in increased reliability of the recirculation pump trip (RPT) function and diminished potential for reaching an unacceptable suppression pool temperature.

The new trip output relays are safety related, seismic category 1, and contact-to-coil separation is used to separate the safety related ATWS/RPT logic from the nonsafety-related drive motor breaker trip coils. The reactor recirculation system (RRS)is not an engimered safety system and the RRPs are not considered essential to the rafe shutdown of the plant under either normal or abnormal conditions. However, the RRS does provide a safety support function in that an RRP trip is used to mitigate the consequences of a common mode failure to scram when needed during an ATWS event. Successful tripping of either the M-G set field breakers or the M-G set drive motor breakers will mitigate the consequences of an ATWS event. This modif' cation assures that adequate fuel barrier margins are maintained. Analysis performed to examine the effect of the slower coastdown characteristics associated with the drive motor breaker showed only a slight increase in the reactor peak pressure as compared with the results of the field breaker trip. The inadvertent MSIV closure at full power with a failure to scram was used as the most limiting ATWS event for this plant parameter. However, these results are bounded by the RRP seizure analysis in UFSAR Subseebon 15.3.3.

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Safety Evaluation No: 93-0032, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-13543 Section(s) None Table (s) None I Figure Change X Yes No Title of Change: Improvements in the Condenser Waterbox Draindown System

SUMMARY

This modification improved the reliability of the condenser waterbox draindown system. The modifications included the following:

. Replacement of the existing high volume draindown pump simplex strainer N7100D101 with a duplex strainer.

. Installation of the removed simplex strainer into the low volume draindown pump suction line.

. Replacement of an existing section of the PVC draindown suction header with cart >cn steel pipe. ,

. Installation of a second draindewn pump located at the north end of the condenser bay. ]

Included is all new carbon steel discharge piping, discharge valves, duplex strainer, '

pressure gauges and motor starter.

The simultaneous failure of the draindown taps or piping / equipment within the boundary of this I

modification is bounded by the analysis in UFSAR section 10.4.5.3.  !

l Power feed for the new high volume drain down pump is fed from the plant construction feed I and does not affect or connect with any plant bus distribution panel required for safe shutdown of the plant.

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Safety Evaluation No: 93-0035 UFSAR Revision No. 7 Reference Document: EDP 12836 Section(s) 6.5 l

Table (s) 5.5-3 Figure Change x Yes No Title of Change: Control Rod Drive (CRD) Hydraulic System Relief Valve Replacement

SUMMARY

This modification replaced the original socket-welded CRD hydraulic system relief valves C1100F001 A and B with flange mounted relief valves. The use of flange-mounted relief valves reduces the manpower requirements associated with setpoint pressure testing of the relief valves by eliminating the cutting and rewelding of the valves.

This evaluation also justifies revising UFSAR Subsection 5.5.13.1 to exempt relief valves installed in Class D piping from the requirements of ASME Ill, NC-7000. This change makes UFSAR Subsection 5.5.13.1 consistent with the UFSAR piping and instrumentation diagrams and UFSAR Section 3.2.

The replacement relief valves have the same function as the original relief valves. The pressure setpoint is still 250 psig. The relief capacity has increased from 80 gpm to 90 gpm.

The greater relief capacity provides more conservative overpressure protection and is still within the 100 gpm capacity of the sump to which the relief flow is piped. The use of flanged valves does not impact the accident analysis because the flange in analyzed as if it were a pipe.

The UFSAR revision makes UFSAR subsection 5.5.13.1 consistent with the UFSAR figures and sections which govem classification requirements. The Fermi 2 Safety Evaluation Report NUREG-0798 has no specific requirements for the relief valves installed in Class D piping.

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Safety Evaluation No: 93-0036, Rev 3 UFSAR Revision No. 7

[ L Reference Document: EDP-13751 Section(s) 4.5, 6.2, 7.3, 7.4, 7.5 & 7.7 Table (s) None

[-

Figure Change X Yes No Title of Change: Modificaton to Reactor Water Level Instrumentaten System

SUMMARY

[ This modirsuon added a continuous flow of water to two cold reference legs assooated with the RPV water level measurement system. It connects the control rod drive system piping as a source of backfill water to the RPV water level. This backfill modiruuon will prevent the potenbal level emn that could result from the effects of noncondensible gases in the reactor water level

( instrumentation reference legs. The modification ties the CRD system to the reference legs at instrument rack H21P009 for Division I and H21P010 for Division ll instrumentation The interconnection includes manual isolation capability, flow control, flow measurement and isolaton

( between the systems. The flow control panels between CRD system and instrument racks H21P009 and H21P010 are mounted Seismic I whde the flow control panel components are non-QA and mounted Seismic II/l. The tie into instrument racks H21P009 and H21P010 from the

( panels is QAl, Seismic l. This modification resolves issues related to the reactor vessel water level instrumentaten addressed in Generic Letter 92-04 and Bulletin 93-03.

r The existing reactor level instrumentation is isolated from the nonsafety-related CRD system by two safety-related check valves in series. Any failure or break in the added backfill lines upstream of

, the double check valve lineup would be isolated from the reactor coolant pressure boundary by L existing safety-related excess flow check valves. The addition of the backfill modsficaten does not affect the analysis provided in UFSAR 15.6.2 on instrument line pipe break. The possible effects of g loss of flow or leakage through double check valves due to the failure of backfill flow lines or CRD L pump trip was evaluated, and compensatory action as stipulated in Bulletin 9M)3 will be taken until  ;

backflow is reestablished. This modifeuon improves the reliability of the water level instruments r

while maintaining the accuracy of the instruments ce0,06cW h t% Technical Specifsucas.

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Safety Evaluation No: 93-0038, Rev 2 UFSAR Revision No. 7 l

Reference Document: EDP-13965 Section(s) 9A.4 l Table (s) NA Figure Change Yes X No Title of Change: Thermo-Lag Fire Barriers Removal / Reclassification in Auxiliary Building i

SUMMARY

This modification removed three (3) Thermo-lag material fire baniers installed in the auxiliary building and replaced them with UL listed 1-hour and 3-hour rated fire baniers. Additionally, two other Thermo-lag material fire barriers were redassified as smoke and gas seals. The fire baTiers impacted are:

• HVAC chase floor closure (elevation 613'-6") has been redassified as a smoke and gas barrier.

• HVAC chase floor closure (elevation 630'-6") has been reclassified as a smoke and gas F barrier.

• Ventilation equipmer'. area wall (elevation 659' 6") opening around HVAC ducts was replaced with UL design UL 435 which is a 3-hour rated gypsum board assembly.

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• CCHVAC equipment area wall (elevation 677' - 6") opening seals were replaced with one-r hour rated Promot board material rated fire banier.

The openings in the walls (elevation 677' -6") created by the removal of the cable tray

- en&sure for trays 1C-078 and 1P-073 were sealed with 3-hour rated elastomer seals.

This replacement / reclassification of fire barriers has been technically justified in accordance with 4 USNRC Generic Letter 8610 and shown not to have an adverse impact on the fire protection a program on both sides of the subject baniers. Although this reclassification may represent a minor decrease in the level of protection, it does not represent a decrease in the margin of safety since a i

~ fire cannot spread across the smoke and gas banier in a noncombustible HVAC chase. Thus the L chanoes beino made by this modification do not adversery impact the fire protection program, ratner i improves the existing fire protection program by replacing / reclassifying five inoperable fire barrier penetration seals, herv.:e maintaining the level of protection as described in the UFSAR.

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Safety Evaluation No: 93-0040, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-12873 Section(s) 5.5, 7.6 Table (s) 6.2-2 Figure Change @ Yes O No Title of Change: Reactor Water Cleanup (RWCU) System isolation on High Outlet Temperature on Non-Regenerative Heat Exchanger

SUMMARY

This modification relocated the isolation signals for the RWCU nonregenerative heat exchanger (NRHX) outlet temperature high. To maintain protection of the RWCU filter demineralizer resins and to avoid an ESF actuation, the isolation signal for the NRHX outlet temperature high was moved from primary containment isolation valves G33F001 and G33F004 to the downstream pump I supply valve G33F119. Moving the RWCU NRHX outlet temperature high isolation to the G33F119 valve eliminates the resulting RWCU pump trips that previously resulted from closure of G33F001 and G33F004. Also, inputs to the Steam Leak Detection Ambient Temperature High annunciator I from the RWCU NRHX outlet temperature have been removed by this modification. These temperature inputs are no longer needed following NRC approval of Techni::al Specifications Amendment 21 in 1988.  ;

This modification does not impact the performance of RWCU or any plant system used to mitigate '

accidents. Relocating the isolation from the primary containment isolation valves to downstream nonprimary containment isolation will have no effect on the RWCU isolation capability on a valid containment isolation signal. Also, the changes made to the steam leak detection annunciator do not affect the ability of the steam leak detection system to perform its intended function.

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SAFETY EVALUATION

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Safety Evaluation No: 93-0041, Rev 1 UFSAR Revision No. 7

[ Reference Document: EDP-13687 Section(s) NA 6.2-2,6.2-13,6.2-15

{ Table (s)

[ Figure Change X Yes R No Title of Change: Addition and/or Modification of Valves in Emergency Equipment Coohng

[ Water (EECW) System

SUMMARY

This modification added/ modified the following valves in Division 2 of the EECW System:

e Replaced body and intemals, and reoriented 6-inch MOV (P44F606B). Original motor

[- and position indication rod have been reused.

[ e installed new manual 6-inch gate valves (P44F966B, P44F9678) to isolate the drywell portion of the EECW system for maintenance.

e Installed drain valves (P44F968, P44F969, P44F970 and P44F971) between new manual valve and motor operated valve.

1

, Valve (P44F606B) has been replaced with a valve that is as close to like-for-like as possible.  !

The original valve had been reworked several times in the past and took valuable outage time.

Installation of the 6-inch manual valves will allow maintenance on motor operated valve inside the drywell vnthout draining the EECW system.

k The addition of the two manual valves and the replacement of the valve body of P44F6068 r valve does not affect the design bases, functions, or operations of any structures, systems or L components. All previous accident analyses remain unchanged in relation to this modification.

Minor differences in materials between the old and new valve (P44F6068) have been

, evaluated against design specifications and found acceptable.

This evaluation also evaluated a similar modification on Division 1, but it has not yet been installed.

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SUMMARY

Safety Evaluation No: 93-0042 UFSAR Revision No. NA Reference Document: EDP 11493 Section(s) NA Table (s) NA I

Figure Change Yes X No Title of Change: Division 1 Fuel Zone Water Level Indication Instrumentation Loop Upgrade

SUMMARY

This modification upgraded the Division 1 fuel zone level indication instrumentation loop to meet Regulatory Guide 1.97 Category 1 requirements. It fulfills a commitment made to the NRC in response to its Safety Evaluation Report for Compliance to Regulatory Guide 1.97.

The original non-qualified level indicator was replaced with a qualified indicator. The signal cables from testability panel H21P080 to the indicator were replaced with cables that are routed in a qualified raceway system. The indicator power source is inverter-fed from the Division 1 130 VDC batteries. A qualified isolator was installed in qualified panel H11P613 to provide the required electric separation and isolation for the emergency response information system (ERIS) input signal.

The function of this instrument loop remains unchanged. The trip unit still provides an interiock j signal to the containment spray valves and produces an isolated auxiliary analog output signal.  !

This output signal still provides fuel zone level indication in the control room and to ERIS.  ;

There are no common mode failure concems associated with this modification because the Division 1 and Division 2 fuel zone indication loops are separate channels using independent 3 differential pressure transmitters, trip units, power supplies, and indication devices. The new l indicator is qualified in accordance with the guidelines provided in IEEE Standards 344-1987 and 323-1933. The additional load on the Division 1130 VDC batteries has been reviewed and determined to be acceptable.

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Safety Evaluation No: 93 0043- UFSAR Revision No. 7 Reference Document: EDP-12653 Section(s) NA

f. Table (s) NA l

j. Figure Change X Yes No Title of Change: Replacement of Obsolete Temperature Control Loop Components in l Feedwater System

SUMMARY

This modificabon changed the temperature control loop of the north and south reactor feed pump (RFP) and reactor feed pump turtune (RFPT) drive lube oil coolers to eliminate GEMAC con *ders l

and auto manual stabons and its assocated devices. The changes made include the following:

- The obsolete Bailey posdon transmitters, N21N435A and B were replaced with a Fisher Model 4211-2/MTG-2318 electronic posebon transmitter.

l The omsting Fisher series 546 I/P converters, N21K409A and B, which accept an input I signal of 10-50 ma were replaced with a unit that will accept an input signal of +20 me.

! This is because the Moore controller can only have a +20 mali-SV output signal

. The Transmation T/C signal condiboners N21K804A and B and the Transmabon auto inal=hws N21K838A and B were eliminated. Instead, thermocouple wires were installed from the relay room to COP H11P805 in the control room where the new Moore controller was installed The thermocouple board opbon was added in the Moore controller to receive i the milkvolt signal from the existing thermocouplos. -I i

• The GEMAC controllers N21K803A and B and the assooated amplifiers were elmnated.

The GEMAC manual / auto stabon, N21K806A and B, located in the COP panel H11P805, was also eliminated in its place, the Moore controller was mounted.

• The Moore controllers are powered by 120VAC supply available in COP H11P805, the power supply required is a non-Class 1 E power source.

• The exisbng mulb-conductor cable used by the GEMAC manual /suto stabon in COP H11P805 was used to wire the new Moore 352E controller to and from the associated I remote devices in the relay room I

\ The design of the turtune building closed cooling water and north and south RFP and RFPT drive lube oil coolers is not affected by this modificabon All required controls, indicabons, and alarms are funcbonally unchanged. The accident analysis in Chapter 15 of the UFSAR is not affected since it does not cover the system failure of the RFP system lube oil coolers

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SAFETY EVALUATIONS

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SAFETY EVALUATION

SUMMARY

(

Safety Evaluation No: 93-0045 UFSAR Revision No. 7 Reference Document: EDP-13092 Section(s) 8.3 6.2-9

( Table (s)

[ Figure Change U Yes X No Title of Change: Replacement of Power and Control Cable for Nitrogen inerting Drywell

( Inboard Exhaust isolation Valve (T4803F602)

SUMMARY

This modification replaced a portion of power and control cables for MOV T4803F602 with a high temperature cable due to the high temperature (122.7*C) measured in the drywell for a portion of the power and control cables. The old Okonite cables rated for 90*C (194*F) have been replaced with the new silicone rubber cable used in the drywell for valve T4803F602.

The cable is rated at a maximum operating temperature of 125*C (257*F) for 40 years in r addition to its 61 minutes (includes one-hour margin) post DBE requirement and is Class 1E _

L qualified in accordance with the requirements of IEEE 383-1974.

[ The addition of a new high temperature rated cable for valve T4803F602 does not change the function of the component and/or system. The new cable is qualified for higher temperatures and thus enhances the availability of the motor operated valve T4803F602. l l

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SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 93-0046 UFSAR Revision No. 7

[ Reference Document: EDP-13667 Section(s) NA Table (s) 3.9-22

{

[ Figure Change [XJ Yes l lNo Title of Change: Snubber Reduction inside the Drywell

SUMMARY

[ This modification reduced the number of snubbers on the main steam and attached piping inside the drywell either by elimination and/or replacement with a strut. This reduction was accomplished by changing and optimizing the pipe support configurations using current piping i technology and analysis methods which are currently acceptable to the NRC. Minor modifications were also made to pipe supports, structural steel and piping mirror insulation.

Parameters, such as pipe size, routing, flow rate or pressure, which are vital in complying with E the piping system design objectives, have not been changed. The benefits gained by snubber L removal or replacement with a strut include a decrease in costs associated with periodic maintenance, inspection, and testing of snubbers, decreased radiation exposure levels for e plant personnel, increase in reliability of piping systems, and potential reduction in outage I

duration. The snubber reduction analyses utilized the existing piping analysis as documented in the UFSAR, plant unique analysis report, or other design documents.

I The pipe support configuration changes do not impact the function or operation of the main steam or any interfacing systems. Design criteria and design loadings, as provided in the UFSAR, were utilized in the snubber reduction design calculations. The revised pipe stresses y and component loadings do not exceed allowable values as specified in the UFSAR. The changes have no impact on any accident scenarios previously evaluated in the UFSAR, nor is

- the safety-related function of any component required to mitigate the consequences of an l

accident impacted in any way. Insulation modifications do not have a significant effect on the containment ambient temperatures during operation. Altt

Dugh stresses, loads, and

- displacements have increased in some instances, the design requirements to satisfy ,

applicable code or design limits have been fully satisfied.

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SUMMARY

[

Safety Evaluation No: 93-0048, Rev 1 UFSAR Revision No. 7

[ Reference Document: EDP-13677, Rev A Section(s) NA Table (s) NA

[

[ Figure Change X Yes l l No Title of Change: Replacement of Position Switches on Selected Anchor Dariing Check

[- Valves

SUMMARY

f.

This modification replaced the position switches on four Anchor / Darling exercisable and spring assist close check valves B2100F076A&B and E1100F050A&B. The old switches had a poor I operating performance history. The switches had been f>und with misaligned switch actuator arms and suffer 6d ' m switch boot degradation due, to high ambient temperature and  !

radiation expomre. J new switches utilize a noncontact sensor which operates on the principle of magnetic attraction. The new switches are single pole double throw type, and due ,

[ to this hardware constreint, the B2100F076A&B valve status process computer inputs are i deleted. Similarly, the si.are ISO mimic inputs for E1100F050A&B have been removed. The

[ design of the new switches remains as QA level NQ, Seismic il/l. UFSAR figures have been L

updated to remove the symbols for the ISO mimic and computer interface funcbons wtuch are presently not being utilized.

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L The new valve modification package for this modification was evaluated for electncal I characteristics, rating of switches, and extension wire, electncal termination methods / techniques, physical characteristics, environmental effects, effects on the valve (s)

L and system (s) piping / hanger. The evaluation concluded that the new valve indication package qualification envelop the design application's requirements. The new valve indication package

- is non intrusive and does not alter valve and its indicating function.

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SAFETY EVALUATION

SUMMARY

L Safety Evaluation No: 93 0049 UFSAR Revision No. 7 Reference Document: EDP-11324 NA Section(s)

Table (s) NA Figure Change l X l Yes l lNo Title of Change: Instrument Control Loop Changes on Level Control Valves for Number 6 North and South Feedwater Heaters

SUMMARY

This modification removed the split range control of the normal level control valve (LCV) and emergency drain line (EDL) valves on #6 North and South feedwater heaters. The new individual controllers for LCV and EDL provide smoother and stable operation of the LCVs by independently positioning these valves over the entire range of the controller. Adeaally, with a two-controller arrangement, a single loop failure only affects the valve it cony a The remaining healthy valve control loop provides for partial heater level control. The r<eynse time and stability of the heater drains system is also improved by separating the operation of these valves into two controlloops.

This modification is associated with nonsafety-related equipment located in the turbine building. The addition of new controllers and separating the instrumentation for LCV and EDL valves into individual loops to maintain levv 5 #6 feedwater heaters does not change the original function of the system and/or crea's ne possibility of an accident of a different type than any previously evaluated in the UFSA.i. The modification does not directly or indirectly affect any safety related or important to safety function of the plant.

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SUMMARY

b Safety Evaluation No: 93-0051 UFSAR Revision No. 7

{ Reference Document: EDP-13623 Section(s) NA

{ Table (s) ,NA

( Figure Change W Yes No Title of Change: Removal of Throttling Function on HPCI Test Retum Line Valve

SUMMARY

This modification removed the throttling capability for the HPCI test retum line motor operated valve E4150F008. The throttling capability for E4150F008 is not required after replacement of the motor operated valve E4150F011 with an air operated drag type valve in the HPCI test retum line to condensate storage tank. The removal of throttling capability also resolved the

{ potential operational problems created by the anti-pumping circuit in the manual push-button closure circuit of the valve logic when the valve was previously used as a throttle valve. This F change also includes replacement of a valve mimic symbol on control room panel H11P602.

L This modification to the E4150F008 valve does not affect the safety-related function of the r valve to close and isolate the HPCI test retum line during a HPCI system initiation. The HPCI L system ability to inject water to the reactor is not diminished by implementing this design change. Also, the ability of E4150F008 valve to isolate the secondary containment leakage r path is not adversely affected. There is no change to the bypass leakage criteria associated with E4150F008 valve. This modification does not create a new failure mode, common failure mode failure, or violate single failure / separation criteria. {

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SUMMARY

Safety Evaluation No: 93-0053, Rev 4 UFSAR Revision No. 7 Reference Document: EDP-13921 Section(s) NA EDP-13922 Table (s) NA X Yes

~

Figure Change No Title of Change: Modifications to Residual Heat Removal (RHR) and Fuel Pool Cooling and Cleanup (FPCC) System

SUMMARY

During RFO4, the following modifications were implemented in the RHR and FPCC systems:

• Added two 8" gate valves; G4100F407 to line G41-3668 and G4100F406 to line G41-3669.

These valves are normally open and locked, under the locked valve program, to ensure FPCC-assist mode is available should it be required.

• Added a 24" 300# rotary-disc manual block valve, E1100F211, into the 24" RHR cross-tie line E11-3146.

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• Added a new 8" connection between existing valve E1150F010 and new valve E1100F211 for the FPCC-assist retum line G41-3669.

• Added a 10" branch connection line with a 10" rotary-disc manual isolation valve, E6100F011, and a blank flange to the 20" RHR pump "C" discharge line E11-3157.

• Cut the existing connection of line G41-3669 to line E11-3146 and reattached G41-3669 to the new 8" connection.

• Capped the existing weldolet connection for line G41-3669 to line E11-3146.

The above modifications are a portion of the attemate decay heat removal system which was being implemented at Fermi 2. The attemate decay heat removal system design is a subloop of RHR that is installed primarily in the RHR to the fuel pool cooling and cleanup system interfaces. The modifications to lines E11-3146 and G41-3669 provide improved RHR divisional isolation capability I for maintenance, and greater flexibility in the use of RHR/FPCC assist mode. The new E1100F211 valve will be used to isolate RHR Division I during its maintenance outages. Valve E1150F010 will continue as the crosstie header isolation valve for RHR Division 2 maintenance outages and for remote manual operator action. The valves installed are passive and have no design requirement

,I to be operated except to be manually manipulated during refueling outages. These valves do not affect the active functions of any other safety-related component in the same system or located in the adjacent areas. All components are qualified for, and have been installed in accordance with,

'I the system specifications, ASME Code and piping design requirements, including seismic support and rattle space clearance.

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Safety Evaluation No: 93-0059 UFSAR Revision No. 7 Reference Document: EDP-13714 Section(s) NA Table (s) NA I Figure Change l X l Yes l lNo Title of Change: Dryer Separator Drain Line Change

SUMMARY

l This modification added a removable flanged spool piece with a drain connection for the dryer I separator storage pit drain line. The modification allows flushing / cleaning water to be diverted to another drain such that contamination would not be spread to pipes that are in other main traffic aisle ways. The spool piece is equipped with a drain valve to facilitate draining of the i

line prior to removal of the spool piece. Also, a threaded connection in the pit has been added I which allows flushing of drain line with a pressurized water source. A standpipe could be installed to provide an area where contamination could settle prior to entering the drain line

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during storage pool cleaning. Both the standpipe and the pressurized water flushing would I reduce the radiation fields that would develop on the Reactor Building third floor and possibly reduce the need for cleaning this drain line after every refueling.

The fittings, valve, and piping added are designed and constructed in accordance with ASME Section lil Class C requirements. The system structural integrity is ensured by designing this modification as Seismic Category 11/l. The accident that the dryer-separator storage pit drain l line could impact is the drainage of the storage pit, reactor well and the fuel pool that uncovers the fuel stored in the spent fuel pool. This system of pools is designed such that no connection can albw the spent fuel pool to be drained below the bottom of the fuel transfer I canal or make a connection that would drain the spent fuel pool below this point. Also, the flanged spool piece drain valve will be normally capped when not in use.

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SUMMARY

1 Safety Evaluation No: 93-0061, Rev 1 UFSAR Revision No. 7 1

Reference Document: EDP-14099, Section(s) 9A.2, 9A.4 1 ECR-14099-1 Table (s) NA  :

I Figure Change [X_J Yes l lNo Title of Change: Thermo-lag Fire Barriers in Auxiliary Building l

SUMMARY

Five (5) thermo-lag fire barriers installed in the Auxiliary Building have been removed and replaced with UL listed (1 and 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) fire rated assemblies. Additionally, a sixth thermo-lag I barrier will remain in place. However, it is being reclassified as a continuous smoke and gas barrier. Two (2) 3-hour rated doors and an additional barrier have been added to support the reclassification of the sixth thermo-lag barrier. This modification is a result of actions required by NRC issued Bulletin 92-01 and its Supplement 1. All the seals being replaced / reclassified ,

are not located in the control room pressure boundary. The fire barriers impacted include: (i) J cable tray enclosure for cable trays 1C-037 and IP-070 in the CCHVAC area in El 677' -6", (ii) separation barrier between the redundant CCHVAC air handling units in the CCHVAC area on El 677' -6", (iii) electrical blockout enclosure in the cable tray area on El 631' -0", (iv) electrica! l blockout closure in the cable spreading room on El 630' -6", (v) cable tray enclosure for cable tray 1K-034 in the relay room stairwell between elevations 613' -6" and 643' -6" and (vi) relay room stairwell enclosure in northeast comer of the relay room and control center on El 613'-6".

Fire resistance ratings of the five replacement barriers are either the same as the original thermo-lag barrier they are replacing (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) or, in the case of the cable tray enclosure on El 677'-6", are an extension of a one-hour barrier whose lower rating was previously approved by I the NRC based on a very low combustible loading in the area. Reclassification of the sixth thermo-lag assembly as smoke and gas barrier has been technically justified and also meets requirements stipulated in NRC Generic Letter 86-10. Although this reclassification may i

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represent a minor decrease in the level of protection, it does not represent a decrease in the i margin of safety since a fire still cannot spread beyond the relay room fire zone. In addition, two fire doors have been upgraded and additional fireproofing has been added to support this reclassification. The modifications made are material related changes and do not affect the required separation provided between the auxiliary building fire zones. The replacement barriers have been seismically analyzed and found to meet ll/l requirements imposed on the original barriers.

l The changes made by this modification in the plant and the fire hazards analysis (UFSAR Section 9A) do not decrease the level of safety being provided for plant systems. Additionally, a fire in zones 3,7 and 9 (which adjoin the newly modified stairwell fire barrier) would result in the plant still being able to safely shut down using the dedicated shutdown panel.

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SUMMARY

1 Safety Evaluation No: 93-0063 UFSAR Revision No. 7 Referince Document: EDP-13544 Section(s) 9.4 Table (s) NA Figure Change [Xl Yes U No Title of Change: CRD Rebuild Room Ventilation System Change

SUMMARY

This modification changed the ventilation system for the CRD rebuild room on the third floor of the reactor building. The changes included supply duct work modification, replacement of solid doors with louvered doors and replacement of solid metal panels above the doors with expanded metal. These modifications will reduce the possibility of the spread of contamination 3 from the CRD rebuild room into other areas of the reactor building. The alarm setpoint for the E HEPA filter differential pressure switch (T41N053) has been increased (from 2 inches water gauge to 3 inches water gauge) to avoid nuisance alarms and reduce the frequency of HEPA filter changeout. During this modification, some editorial changes were also made in UFSAR Figure 9.4-4.

This modification does not change the design basis that air flows from the areas of low potential for radioactivity to areas of high potential for radioactivity. The air that was originally supplied (via ducting) to the room is now supplied to the general area outside the room, which in turn moves into the CRD rebuild room through the expanded metal and the louvered door or the open door. This modification does not affect the operation of the reactor building ventilation system or initiation and operation of the Standby Gas Treatment System.

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SUMMARY

f Safety Evaluation No: 93-0064, Rev 2 UFSAR Revision No. 7 l

Reference Document: TSR-26049 Section(s) NA Table (s) NA

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I l Figure Change X Yes l lNo Title of Change: Replacement of Heater Drain Pumps' Seal l

SUMMARY

This modification replaced the seals on the north, center, and south heater drain pumps with

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new cartridge mechanical seals and associated equipment for monitoring and control of seals performance. Additionally, the pump base plates were modified to replace the existing gasket with a flexatallic. This modification was made to stop base plate leakage and ensure reliable operation of heater drain pump seals. Also included with the modification is a seal flush filter

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which provides a clean flush to the seal during startup of pumps. Two of the three orifices on the seal pumps'line were removed and replaced with a Fisher valve to allow the variation of seal intet pressure to above stuffing box pressure. Gate valves N2200F030A, B, and C were replaced with an equivalent globe valve to allow additional stuffing box pressure control.

This modification revises UFSAR figures which depict seal flush lines supplied from the heater feed pumps (HFPs) and the heater drain pumps (HDPs) stuffing box vent and seal leakoff lines. Neither the HDPs nor HFPs are required for plant shutdown accident mitigation or long term recovery. The failure of these pumps is encompassed in the loss of feedwater analysis as described in UFSAR chapter 15.2.7.

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SUMMARY

Safety Evaluation No: 93-0065 UFSAR Revision No. 7 Reference Document: EDP-13937 Section(s) 9.1 Table (s) NA Figure Change X Yes No Title of Change: Refueling Floor Changes

SUMMARY

( The modifications on the reactor building fifth floor include removal of jib crane pedestals (F1100E021) located north of the spent fuel pool on the refueling floor and the two northemmost stops of the refueling bridge rails at the north end of the floor. Additionally, grating of the sections of the refueling bridge rails north of the remaining northem stops and

( documentation of the already removed wedge anchors for the uprighting stand and the angle stop are part of this modification. The purpose of this modification was to remove unused plant components from the refueling floor. Each item removed was originally installed with an

{ intended purpose. However, they are no longer used or needed, and their intended function is being performed by an attemate plant component. The intent of this modificatum is to provide cleaner work areas on the refueling floor and also to eliminate unused installations which present tripping hazards. These changes will facilitate the refueling floor activities by increasing the available unobstructed floor space.

The jib crane boom and its bases, their concrete pedestals, the uprighting stand, and angle stop anchors are not considered to be important to safety and are not associated with any equipment important to safety previously evaluated in the UFSAR. The removal of the northemmost rail stops does not impact the operational characteristics of the refueling bridge 1 because the bridge has never traveled past the remaining north stops, and it does not travel farther north than the existing limits. Furtherrnore, the removal of the fuel pool jib crane may t result in safer conditions by eliminating the possibility of the refueling bridge structure interfacing with the jib crane boom when traveling to the north end of the spent fuel pool. l f

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SUMMARY

Safety Evaluation No: 93-0066 UFSAR Revision No. 7 Reference Document: EDP-13934 Section(s) 9.2, 9.5.

Table (s) NA I  :

Figure Change l X l Yes l l No Title of Change: Removal of General Service Water (GSW) System Pressure Control Valves

SUMMARY

This modification removed the GSW system pressure control valves (PCV) P41F401A and B

'I and replaced them with a straight section of pipe with three orifices in series. The pressure control valves have proven to be unreliable due to repeated component mechanical failures which required manual pressure control by the use of a pump test line that is equipped with an ,

orifice in series with globe valve (P4100F012). In order to provide an attemative (in the event I of failure of valve P4100F012), this modification removed the two pressure control valves, P41F401A and B, and replaced them with three in-series orifices. Should P4100F012 fail, system pressure could be controlled using gate valve P4100F134 in series with three orifices I to bypass flow to the pump pit. Valve P4100F134, which used to be normally open, is now normally closed. In addition to removing the PCVs and the associated controllers, this modification also removed the associated control room indication and annunciator and adjusted the control room TBCCW heat exchanger temperature indicator green, or normal, band from 75'F - 85 F to 75 F - 88 F and associated alarm from 85 F to 88 F.

-l Removing two PCVs and replacing them with three onfices, operating the system manually, and changing the green band for TBCCW does not increase the probability or consequences of an accident previously evaluated in the UFSAR. The GSW and TBCCW systems are nonsafety related, nonseismic and non-Q, and as such, are not considered as operating in any of the Chapter 15 accident scenarios. Failure of GSW or any of its components affected by this change has no impact on the Fire Protection System.

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SUMMARY

b Safety Evaluation No: 93-0070 Rev 4 UFSAR Revision No. 7 Reference Document: EDP-11655 Section(s) 5.5, 6.2, 6.3, 7.3, 7.5, 9A.4 Table (s) 6.2-14,8.3-14 Figure Change X Yes No Title of Change: Replacing Existing DC Motor Operated Valve (E4150F011) With an Air Operated Valve (E41F011) in the HPCI System

SUMMARY

This modification replaced the DC motor operated valve E4150F011 with an air operated control valve (E41F011). The air operated valve is a drag type (modified globe) valve manufactured by Control Components, Inc. with a custom trim to achieve the required differential pressure to meet Technical Specification requirements. The new valve has been designed and manufactured to meet ASME lil, Class 2 ISI bypass leakage requirements of the E previous valve. The valve is a spring to close (fait closed) valve using interruptible air supply (IAS) for stroking the valve.

E The trim inside the new valve is custom designed to provide the required differential pressure for meeting Technical Specification operability during HPCl/RCIC high pressure and low pressure testing. The valve trim is designed to have proportional characteristics which closely l

approximate actual field requirements. Required valve response is obtained through the use

' of Moore programmable controller (model 352E). The intent of this is to use the new AOV to generate the required differential pressure instead of using the existing spectacle flange orifice r plates (E4150D010 E4150D015, E4150D016, E5150D006, E5150D013 and E5150D014).

Since there will be no pressure drop required across the orifice plates, the above spectacle flange orifice plates will remain aligned with the large bore orifice in line. Additionally, J restricting orifices E4150D017, E4150D018, E5150D015 and E5150D016 were replaced with spectacle flange orifice plates of similar design as the other spectacle flange orifice plates.

These orifice plates are also aligned such that the large bore orifice is left in line. This eliminates the need to change out spectacle flange orifice plates for HPCI and RCIC testing.

This modification does not change the characteristics of the system, only the method of developing the required system head pressure and is bounded by hydraulic analysis developed for HPCI and RCIC systems.

L Most controls for the E41F011 valve are not required to be safety related because the valve does not serve a safety-related function other than to isolate the test line. The only safety-related control components related to isolation of the valve are the actuator air isolation and air exhaust solenoid valves (E41F001 and E41F002), and a selector switch provided to de-energize the actuator air isolation and air exhaust solenoid valves (E41F001 and E41F002).

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES PAGE52 Safety Evaluation 93-0070, , Rev 4 (Continued)

The selector switch has been classified as QA-1, safety related for circuit integrity only. The programmable digital controller is not required to be safety related because the controller does not perform the isolation function for the system. The isolation capability is ensured by using safety-related components in the actuator and isolation logic. The safety-related solenoid valves E41F001 and E41F002 isolate air supply to the actuator and exhaust air from the actuator to allow the actuator spring to close E41F011 when isolation signal is present.

To address the Appendix R issue, the E41F011 valve utilizes a spring to close (fail closed) air actuator on the drag valve. An air actuator was chosen for this application due to the power limitation from the existing position for E4150F011 (2PB-1, POS 3C). A manual 3-way isolation valve (P5000F154) is provided to isolate the actuator air supply during execution of the Dedicated Shutdown procedure. This will prevent the valve's position controlling components from opening the valve (after closure) should the valve receive an open signal as a result of an electrical short from the postulated fire in a common fire zone. A manual 3-way valve was chosen for this application to depressurize the air supply line and allow the normally open lock-up valve to vent off the actuator, should the valve be in the full open position prior to entering the dedicated shutdown procedure.

UFSAR Section 7.5.2.5.4 and UFSAR figure 7.5-10 specify the maximum times for operator action during the dedicated shutdowr. process for certain critical parameters (Reactor level, Drywell temperature, and Suppression Pool temperature). A timed performance of the Dedicated Shutdown procedure 20.000.18 was performed to determine the impact of the added operator action of isolating P5000F154 on meeting the time requirements of UFSAR Section 7.5.2.5.4 and UFSAR Figure 7.5-10. Procedurally, isolation of P5000F154 is required as soon as it is determined that shutdown of the plant must be performed from the Dedicated Shutdown panel. It was demonstrated that the added requirement to isolate P5000F154 did not impact the ability of SBFW to inlect to the vessel in the time frame as stated in UFSAR section 7.5.2.5.4. Additionally, is)lation of P5000F154 does not impact the other time requirements of UFSAR Section 7.5.2.5.4 (i.e., containment cooling and torus cooling). Based on the above, sufficient controls are in place to ensure SBFW wil' ae able to perform its I function for Dedicated Shutdown.

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 934071 UFSAR Revision No. 7 Reference Document: EDP-12961 Section(s) 7.3 NA

( Table (s)

Figure Change Yes X No Title of Change: Removal of Tamper Guards fnxn TestabMy Catmets i

SUMMARY

This modificabon removed the tamper guards (locking bars) from the trip unit chassis on testabikty catunets H21P080, H21P081, H21P082, H21P083, H21P084, H21P085, H21P008, and H21P087.

! The tamper guards were originally installed by the vendor as a positive control measure to ensure l that only authonzad personnel would have access to the calibration instrumentabon Based upon the administrabve, procedural and engineering controls now in place to preclude unauthonzed acmss, the tamper guards have been removed. The tamper guards were a passive cuirpcsent and did not in any way affect the safety funcbons associated with trip units.

Removal of the tamper guards does not affect the safety-related funcbons or operabonal rebatMhty of the trip units as desenbod in UFSAR 7.3.2.3.1. This modificabon does not ah,w.'i mpact i any radiabon barriers, and it does not create a new failure mode, common failure mode event, or wolale single failure /separabon entena.

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SUMMARY

Safety Evaluation No: 93-0072 UFSAR Revision No. 7 Reference Document: EDP 13977 Section(s) NA Table (s) NA I

Figure Change X Yes No Title of Change: HPCI and RCIC High Point Vent Lines Changes

SUMMARY

HPCl and RCIC high point vent lines were modified by perfomling the following major changes:

. Extend HPCI and RCIC high point vent lines outside the steam tunnel, and locate the vent valves in an easily accessible low radiation area.

. Install pressure indicators on each line, upstream of new inboard vent valves.

. Provide LLRT taps for ISI use with isolation valves between inboard and outboard vent valves on each line.

. Downstream of the new vent valves, the vent lines for HPCI and RCIC lines are merged into one line. A sight glass and a thermometer are installed at the beginning of the common vent line.

. The vent line is routed to the equipment sump D075. The piping diameter is increased from 3/4" to 2" and material changed to stainless steel before entry into the sump.

This modification will improve both HPCI and RCIC systems by eliminating hazards associated with '

venting practices used earlier during the performance of routine surveillance tests. The operational characteristics of HPCI, RCIC, and the radwaste system will not be affected by the modification.

The HPCI and RCIC high point vent lines extension will not affect any accidents described in the p UFSAR. The system separation criteria between HPCI and RCIC has been maintained in the new  ;

configuration. The extended piping added by this modification was analyzed for the design conditions per ASME code and all stresses and displacements were found acceptable for all loading combinations evaluated.

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SUMMARY

Safety Evaluation No: 93-0073 UFSAR Revision No. 7 L Reference Document: EDP-26025 NA Section(s)

NA

( Table (s)

Figure Change X Yes No Title of Change: Addition of New input to Torus Water Level Trouble Alarm Wodow 7D71 1

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SUMMARY

I l This modification adds a new input to existing reflash torus water level trouble alarm window 7D71.

This ERIS generated input ' Torus Water Level High HPCI Suebon Swap Wamng" provides the control room operator an early waming that the torus water level is increasing and that if the trend continues, the HPCI suchon valve will automabcally transfer from the condensate storage tank l (CST) to the torus. Avoiding unnecessary automabc HPCI suchon transfers will prevent actusbon of l the HPCI suchon valve, E4150F042, which is an engneered safety feature (ESF). The changes resulbng from this modification are in nonsafety-related equipment. The Pnmary Containment Monstonng System figures (7.6-11 and 11.4-01) in the UFSAR have been revised to incorporate this modificabon in addebon, a couple of typographical errors have been corrected in these figures.

The failure of the new added alarm will not prevent any equipment hTW=id to safety from funcbonng This modificabon does not alter the control logic of the HPCI suchon valve transfer circuit.

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SUMMARY

Safety Evaluation No: 94 0007 UFSAR Revision No. 7-Reference Document: EDP-14125 Section(s) NA

{ Table (s) NA

[ Figure Change X Yes No Title of Change: Installation of Precision Resistors, Waschler Digital Bargraph Indicators and

[- Test Jacks in Core Spray (CS) and Residual Heat Removal (RHR) Systems

(

SUMMARY

This modification installed precision resistors, Weschler digital bargraph indicators and test Jacks in flow measurement loops of CS and RHR systems. This modification was performed to provide more accurate reading of pump fkw during RHR and CS pump Technical Specification surveillance testing.

The modification replaced the existing 250 ohm +/(-) 5% resistor with 250 ohm +/(-) 0.25%

resistors in each of the RHR and CS flow loops. A Weschler Model BW 1316 e:actronic bargraph/ digital indicator and a pair of banana Jacks were installed across each resistor..The

[ new circuitry has been designed to minimize the impact on the existing flow loops. For example, the meter and the banana Jacks are mounted across the 250 ohm resistors. This will prevent any short circuit or open circuit across these devices from impacting the flow measurement loop. The new Weschler meters are QA1, Seismic 1, and mounted to Seismic 1 requirements. Meters and jacks are installed in the relay room which should prevent any -

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SUMMARY

Safety Evaluation No: 94-0011 UFSAR Revision No. 7 Reference Document: EDP-26281 Section(s) NA Table (s) NA I Figure Change X Yes No Title of Change: Addition of Pipe Taps in Condensate Storage and Transfer System, Reactor Water Cleanup System and Torus Water Management System I This modification added a 3" tap to each of the following systems: Condensate Storage and Transfer System, Reactor Water Cleanup System, and Torus Water Management System.

The taps consisted of a manualisolation valve and blind flange terminating end. The purpose I of this design change is to assist in providing 1) flushing capability to the control rod drive mechanisms, 2) a letdown flow path to maintain reactor vessel water level, and 3) temporary filter / demineralizing capability for reactor water cleanup prior to retuming to the condensate storage and transfer system per Temporary Modification 94-005 (reference Safety Evaluation I 94-0012). This design change was required after the December 25,1993, turbine generator failure event for the purpose of processing contaminated or dirty water from the systems.

This modification does not functionally or operationally change the CST, RWCU or the TWM systems. The modification has no adverse impact on any existing accident scenarios. The new connections do not affect any safety related equipment / component. Pipe break (UFSAR 3.6.2) and flooding analysis in the Turbine Building are not impacted by additions of these

-l taps.

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SUMMARY

Safety Evaluation No: 94-0013 UFSAR Revision No. 7 Reference Document: TSR-26286 Section(s) NA Table (s) NA I

Figure Change X Yes No Title of Change: Remove intemals From West Station Air Compressor Discharge Check Valve (P5000F014A)

SUMMARY

This modification removed intemals from the west station air compressor discharge check I valve P500F014A. The check valve had a history of maintenance poblems. Removal is permissable because the compressor discharge valves act as check valves and prevent reverse flow. The other two station air compressors' discharge check valves will remain.

The Station Air System is a nonsafety-related system. Removal of the check valve's intemals will not affect the system design and operation as described in the UFSAR, section 9.3 and do I not impact the loss of instrument air system as described in the UFSAR, section 15.16.

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0015 UFSAR Revision No. 7 i

Reference Document: EDP-26303 Section(s) NA r

Table (s) NA Figure Change l X l Yes l l No Title of Change: Attemate Processing of Water From the 'Naste Sample Tanks

SUMMARY

This modification added four (4) 2" pipe taps to the Floor Drain and Waste Demineralizer System to allow for an attemate means of water processing from the Waste Sample Tanks.

These taps consist of a short nipple, manual isolation valve, nipple and a threaded cap l

terminating end. This modification is located in the Radwaste Building Basement and is not l

safety related.

These pipe taps are not required to perform any safety-related function. This modification does not change the function or operation of the Radwaste System or any related structures, system, and components. The pipe taps being installed on the Radwaste system are designed to keep the pipe stresses within acceptable levels. The modification components are structurally mounted to prevent interference with safety-related systems and other components under design accident and seismic conditions.

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0021 UFSAR Revision No. 7 Reference Document: EDP-26257 Section(s) 6.2 Table (s) NA

(

( Figure Change X Yes No r

Title of Change: Monitoring of Drywell Shell Corrosion from Outside L

SUMMARY

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This modification removed sand from four 1-1/2" diameter drain lines and the local area between the drywell shell and the drain pipes. This modification was performed in order to establish a viewing path to monitor the corrosion of the drywell shell from the outside. After

' the completion of this modification, the drain path is clear, and moisture monitoring can be performed more efficiently. A viewing path facilitated the drywell inspections at the four drain locations.

The removal of sand from four localized areas does not have any significance considering the drywell shell stresses. The sand layer is a passive component, located outside the drywell

shell. After the removal of the sand, observation of the drywell shell from these four drainage I

pipe locations was possible. The drywell shell is viewed from the drain pipes where the sand layers have been removed. The removed sand in these drain pipes and the area between the drywell and the entry to the drain pipes are not associated with any equipment. The drywell shell at this area is painted,1-1/2" thick steel, and is monitored for corrosion. The removal of -

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sand from the pipes may even facilitate the draining of condensate from the areas, thus helping keep the moisture away from the drywell shell.

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SUMMARY

Safety Evaluation No: 94-0022, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-10810 Section(s) NA Table (s) NA Figure Change X Yes No Title of Change: Changes to Fuel Pool Cooling and Cleanup (FPCCU) System

SUMMARY

This modification made the following three changes to the Fuel Pool Cooling and Cleanup I (FPCCU) system: 1) Added the two relays in the a! arm circuits to the " Fuel Pool Cooling Trouble" alarm 2D13 and local alarm indication on panels G41P001 and G41P002. The affected circuits are the FPCCU skimmer surge tank high and low level inputs to the alarm and local indication; 2) Added a local pressure gauge scaled in feet of water to indicate FPCCU I skimmer surge tank level; 3) Installed test connections and new fittings to the skimmer surge tank level switches G41N004, G41N005, and G41N006.

The change to the alarm involves FPCCU system alarm circuits only. No other control functions are affected. The instrument mounting is Seismic Category ll/l. The pressure rating of the line is significantly above the pressure of the process, and no new taps are being added I to the tank. This modification does not affect the RHR system and thus does not affect its availability. None of the equipment in this modification is required to perform safety-related functions. The level instrument mounting does not affect safety-related equipment in the area.

The power feed to the relays added to the FPCCU system alarm are fused. A failure of the relays will not affect or prevent the operation of safety-related systems. This modification added relays to an existing FPCCU alarm and level indicator and test connections using an existing tank tap. Failure of the alarm circuit will not initiate nor create an equipment malfunction. The changes made to the FPCCU system by this modification will not affect the water level in the fuel nool. If a failure of the instrument line to the skimmer surge tank did I occur, the fuel in the fuel pool can be cooled by the RHR system and the radioactivity released to secondary containment would be bounded by the existing analysis.

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SUMMARY

Safety Evahtation No: 94 0023, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-12002 Section(s) NA l I

Table (s) NA i Figure Change X Yes No l l

Title of Change: Changes to Extraction Steam Line Drain Pot Drain Lines l

SUMMARY

This modification modified the number 3 and 4 extraction steam line drain pot drain piping and improved the le"*l control and sensing lines for the number 3 N & S drain pot drain lines level .;

control valves r /s); resolved the erosion / corrosion failure of the number 3 south extrachon  !

steam drain fin, aproved control of the number 6 heater extraction steam line drain LCV; and  ;

also added two small bore drain lines for the 3N and 4N extraction steam lines to ensure no i moisture collects in these lines. These modifications were implemented to improve the operation and control of existing systems. ,

This modificabon does not impact the loss of feedwater heating on a line break outside of  !

containment. All of the lines and instruments am not safety related and do not affect the l functica of safety-related systems, structures, or components. The modificabons improved the design and control of existing small bore lines except for the addition of the two 1.5" diameter drain lines which are bounded by the existing evaluations and are isolated from safety-related equipment. The existing alarms in the control room remain to alert operators for schon but do not initiate actions. The two new added lines are essentially identical to existing lines in use in the plant for similar flow and design conditions. The new drain pot level controllers are not i required for safety-related functions.  !

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SUMMARY

Safety Evaluation No: 94-0025 UFSAR Revision No. 7 Reference Document: EDP-26315 Section(s) NA NA

{ Table (s)

[ Figure Change X Yes No Title of Change: Replacement of Radwaste Discharge Line Flow Indicator Transmitters and

[ the Recorder

SUMMARY

[ This modification replaced radwaste discharge line flow indicator transmitters (G11N048 and G11N049), and recorder (G11R703) with a newer model which has a built-in power supply for the loop to accommodate the new transmitters. Radwaste discharge line flow

[ indicator / transmitters (G11N048 and N049) were submerged in the radwaste basement after the December 25,1993, turbine damage event. They were not functioning properly and also

' are obsolete. This modification replaced these obsolete instruments with the new instruments, which perform the same function with the same range and accuracy.

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l r There is no safety related equipment located in the Radwaste Building Basement near the L area where the new flow instrumentation is installed per this modification. The new transmitter mounting is the same as the old ones and works on the same principle. The pressure rating of p the new instrumentation is the same as the old instruments. The new flow instrumentatum L loops work and function the same way as the old ones, with the exception of the electrical signal. The electrical signal put out by the new instrumentation is 4-20 made instead of 10-50 r made as put out by the old ones. This low voltage signal is for the indication in the Radwaste Control Room.

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Safety Evaluation No: 94-0028 UFSAR Revision No. 7

.I Reference Document: TSR-26405 Section(s) NA Table (s) NA Figure Change X Yes No Title of Change: Addition of Low Flow Bypass Lines Around the TBCCW and RBCCW j E Temperature Control Valves (TCVs)

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SUMMARY

This modification added two low flow bypass lines with isolation and throttling valves around the TBCCW TCV P43F402 and around the RBCCW TCV P42F400. Because of the low flow instability 3

~E of the GSW TCVs and the inability of the bypasses to control temperatures at low flows, low flow bypass lines with isolation valves and valves designed for low flow throttling were installed around the TBCCW and RBCCW TCVs. These new low flow bypasses will provide reliable manual I temperature control for TBCCW and RBCCW during low GSW flow conditions.

! The function of the GSW System remains unchanged with this modification. The addition of the i bypass lines allows more operational flexibility and improved temperature control of TBCCW and RBCCW during low GSW flow conditions. The GSW system is nonsafety-related, nonseismic, non4, and does not directly interface with any safety or safety-related systems. RBCCW which i supplies cooling water under normal conditions to safety related equipment does not perform a

safety-related function. Safety-related cooling is provided by the two divisions of EECW (which do l not interface with GSW dire ctly) should RBCCW fail. Therefore, GSW does not perform a safety-related cooling function and has no impact should it or any of its components fail on a safe plant shutdown. Radioactive contamination of GSW is avoided by using closed heat exchangers between the service water and the closed cooling water systems. The GSW remains '

i uncontaminated by operating at higher pressure than the cooled system, and any leakage would be from the GSW system to either the TBCCWS or the RBCCWS. Installation of the low flow bypasses around the heat exchanger temperature control valves will not change this design. The majority of GSW flow to equipment being cooled is controlled by temperature control valves. These valves can be controlled in auto or manual or the flow can be manually bypassed around the TCV via the normal (full) flow bypass lines. Manually bypassing the TCV using the low flow bypasses will

E improve control during low GSW flow conditions. The GSW System pressure range,135160 psig, E is not impacted by this modification.

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h SAFETY EVALUATION

SUMMARY

b Safety Evaluation No: 94 0031 UFSAR Revision No. 7 Reference Document: EDP-13298 Section(s) NA Table (s) NA

{

[ Figure Change l X l Yes No Title of Change: Replacement of the RWCU Filter /Demin Holding Pump Isolation Valves

SUMMARY

[ This modification removed the old RWCU filter /demineralizer holding pump discharge valves and replaced them with new valves, actuators, and limit switches. The old valves were 1-1/2" angled globe valves with air operated diaphragm actuators. They fully opened upon I application of air and fail closed with a spring retum when the air is removed. The new valves k are 1-1/2" ball valves with air operated piston actuators. They function identically to the old valves in that they operate fully open upon application of air and will fail closed with a spring retum when the air is removed. The existing solenoid valves supply the air to the new valves

{ so automatic and manual operation will be identical to the old valves.

F The RWCU filter /demins are nonsafety-related, Seismic Class 11/1 and Non-Q. The holding L pumps with their isolation valves are a support system for the filter /demins and are not considered in any safety-related activities. The RWCU System contains two motor-operated 7 isolatum valves which automatically close to mitigate the loss of coolant and release of L radioactive material, or the loss of boron in the event of a standby liquid control actuation. i Changing the holding pump isolation valves from a globe style to the ball valves does not L prevent the motor-operated valves from isolating the system. The modifc' ation to the RWCU System affects only nonsafety-related equipment. Changing the RWCU filter /demin holding pump isolation valves from globe valves to ball valves does not impact any equipment or e systems needed to maintain the plant in a safe shutdown condition. The safety features of the L RWCU System that are provided to maintain the primary containment's integrity will continue to be operational and will not be affected by the change. Access to safety-related equipment will c not be affected. There are no safety-related systems in the vicinity of this modifc' ation that L could be impacted. This modification is fully contained inside the Clean-up Holding Pump Room. This modification does not change the function of the RWCU System or the function of

, the isolation valves, and will not challenge the RWCU leakage detection system or the RWCU

[ isolation circuitry and valves.

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Safety Evaluation No: 94-0038 UFSAR Revision No. 7 Reference Document: EDP-26375 Section(s) NA Table (s) 6.2-2 Figure Change f5(lyes l l No Title of Change: Installation of Bypass Line to Residual Heat Removal (RHR) Valve E1150F015A

SUMMARY

g This modification added an insulated equalizing bypass line from the RHR Valve E1150F015A 5 to the Local Leak Rate Test (LLRT) line on the downstream (reactor vessel pressure side) side to equalize the pressure in the body cavity of the valve with the piping to the reactor vessel. I I

This was provided to eliminate the susceptibility of pressure locking of the valve.

The addition of the bypass line does not impact the function or operation of the E1150F015A valve or of the RHR System. Design criteria and quality standards for this line are consistent l with the design of the RHR system. The bypass line does not perform an active function. The line is designed to prevent the valve from becoming pressure locked due to trapped high a pressure in the body cavity and thus reduces the likelihood of a malfunction of E1150F015A.

g Containment isolation capability (leakage) is still maintained. The failure of the valve is addressed as part of the analysis for loss of one ECCS subsystem. The function or operation of system components or any interfacing system components has not been altered in any way that would cause an equipment malfunction.

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SUMMARY

Safety Evaluation No: 94-0039 UFSAR Revision No. 7 Reference Document: EDP-13393 Section(s) None Table (s) None Figure Change X Yes [ No Title of Change: Modification in Physical Installation of Flow Restriction Orifices in the RHR l Test Retum Lines I

SUMMARY

l This modification changed the physicalinstallation of the flow restriction orifices in the residual heat removal test retum lines to the suppression pool. The original restriction orifices (E1150D010A/D0108) were positioned in the test lines by flanged connections that required I periodic LLRT testing during the outages. This modification installed each of the restriction orifices within a welded pipe spool thus eliminating the flanges and need of LLRT testing without any impact on system operation. Prior to this modification, LLRT testing inside the torus required installing scaffolding which is a safety concem due to manpower / time required to work in the radiation area of the torus and location of the work over the water in the torus.

The reduction in weight by elimination of the flanges and botting in favor of welded-in orifice plates and pipe spool has been analyzed for pipe stress calculations. The orifice plate l thickness, weld size, and pipe local stresses were also evaluated. j I

The piping in the area of the modification has been reclassified from "D" to "D+". This corects a typographical error in UFSAR Figure 5.5-13(1) since the pipe was fabricated, installed and inspected to Group D+ requirements. The piping design is Seismic I and conforms to the

,a requirements of Regulatory Guide 1.29, Rev 1 seicmic design classification. The piping has g been evaluated for LOCA and SRV related torus-attached piping loading as described in the plant unique analysis report for torus attached piping and NUREG-0661.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES PAGE 68 b

SAFETY EVALUAT!ON

SUMMARY

{,

Safety Evaluation No: 94 0041 UFSAR Revision No. 7

[- . Reference Document: EDP-26520 Section(s) NA NA

{ Table (s)

[. Figure Change l X l Yes No Title of Change: Removal of Disc From Valve E4100F046

SUMMARY

[' This modification removed the disc from HPCl main pump discharge to suppression pool check valve E4100F046. During performance of the PM event, it was observed that the in-body seat for this lift-check valve was missing nearly all of its stellite hard facing. Due to the size of the ;

valve (4"), replacement of the in-body seat cannot be performed

[ l There is no design basis for check valve E4100F046. Pressure boundary is not modified by l H this modification. Therefore, removal of the valve disc does not negate ASME 111, Class 2 code requirements of the valve or the system. E4100F046 is not a containment isolation boundary i

valve. Containment isolation for the line is assured by containment isolation boundary valve )

' E4150F012 and the water seal in the suppression chamber. Drain down of the suppression chamber is prevented by isolation of the HPCI minimum flow line isolation valve E4150F012.

Since E4150F012 is a containment isolation boundary valve and is powered by Division 11 r batteries, E4100F046 is not required to prevent drain down of the suppression pool.

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SUMMARY

(

Safety Evaluation No: 94-0043 UFSAR Revision No. 7 (L Reference Document: EDP-26452 NA Section(s)

Table (s) NA

[ ' .'

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( Figure Change X Yes No p

Title of Change: Condensate Pumps' Minimum Flow Control Loop Change L

SUMMARY

r l This modificaton changed the existing condensate pumps' minimum flow control loop to provide h

local manual control and mnmize oscillaton on the minimum control valve N20F404. The oscillabons dunng plant startup and shutdown lead to pressure transients that affect the operaten of the startup level control valve. To circumvent this situaten, the operators have, in the past, used motor-operated bypass valve N2000F618 which does not have automate control, and the operator must adjust flow as plant power changes. Also, signs of valve damage due to cavitsbon have been nobced on motor operated bypass valve N2000F618. The major components installed by this modificaten included the following

• An annubar flow element, FE-N070, downstream of the condensate pumps' mnmum flow ,

i retum line. This flow element is now used as input to the flow controller for valve N2000F404. I

• A pneumabc differenbal pressure transmitter (FXE-N071) and a pneumabc square root extractor (FY-K051) for receiving and lineanzing the signal from the flow element for input to ,

the flow controller.

• A pneumate flow controller {FCP-K050) mounted locally to feed the control signal to the existing valve possener on valve N2000F404.

The unused electronic controller and E/P converter on flow transmitter FXE-N433 were removed as

, a part of this modificaton The main control room flow indicaton, low flow alarm, and valve position indcabon remain unchanged The funcbon of the condensate pumps' minimum flow control loop to protect the condensate pumps

, during plant startup and shutdown remains unchanged. All the required controls, indicabon, protecbon and alarm funcbons are maintained. The flow control loop modificabon provides the operator with a new type of controller with both manual and automate controls instead of automate control only. The feedwater system piping break accident analysis in Chapter 15 of the UFSAR l

envelops the failure of the condensate pumps' minimum flow control loop and the affected area of the condensate system.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES PAGE 70 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0044 Rev 1 UFSAR Revision No. 7 Reference Document: EDP-13576 Section(s) NA

{ Table (s) NA

[ Figure Change X Yes No Title of Change: Changes in Limit Switch Settings for Core Spray System Test Bypass Valves

SUMMARY

This modification changed the limit switch settings to 20% in the opening direction for core spray system test bypass valves E2150F015A and B. Recent MOV testing indicated a need to change this setting to prevent any spurious trip. The change is only in the torque switch bypass, limit switch part of the MOV opening coil circuit. The function of the valve to deliver the desired ficw or close is not altered.

The opening function of valve E2150F015A and B is considered nonsafety related, as opposed to its closing function which is safety related. This modification was done in the nonsafety-related function of the valve. The open or close travel length of the valve is not affected by this

{ modification.

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SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 94 0046 UFSAR Revision No. 7 Reference Document: EDP-26397 Section(s) NA Table (s) NA

( Figure Change X Yes No Title of Change: Draining of Heater Drain Pumps for Maintenance

SUMMARY

( This modification installed a 2-inch drain line off the suction of the heater drain pumps and routed it to each pump's respective pit. This modification provides a less time consuming process for draining heater drain pumps during pump isolation for maintenance purposes.

[ New 2-inch globe valves (N2200F859A, B, and C, and N2200F860A, B, and C) have been

- provided on these drain lines for isolation and throttling purposes.

I This modification does not impact equipment performance since the drain is used only when its respective pump is out of service. The modification involves equipment / components that do not perform a safety function. It does not affect the function of any safety-related equipment /.

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SAFETY EVALUATION

SUMMARY

[ I Safety Evaluation No: 94-0047 UFSAR Revision No. 7

{ Reference Document: EDP-26376 Section(s) NA

{; Table (s) 6.2-2 h Figure Change X Yes No Title of Change: Installation of a Bypass Line Around Core Spray Division 1 Inboard Isolatum

( Valve

SUMMARY

This modification installed an insulated equalizing bypass line from the core spray dmsion inboard isolabon valve (E2150F005A) to the local leak rate test (LLRT) line on the downstream (reactor vessel pressure side) side to equalize the pressure in the body cavity of the valve with the piping to the reactor vessel The bypass line consists of small bore pipe and elbows which are insulated This modificabon will eliminate the susceptibility of pressure lociong due to trapped high pressure in the body cavity and thus reduce the likelihood of a malfunchon of the valve. Stresses in the exishng LLRT piping and the new pressure relief bypass line were evaluated for the applicable stabc and dyname loading condibons to ASME Seebon 111 NC requirements Stresses in the Class i 3/4" l sockolet fitbng were evaluated for the appleable fabgue loading condsbons to ASME Sechon 111 NB requirements The matenals used in this modificabon and the system piping that it is interfacmg with are QA-1, Group A and are safety related.

E This modificabon does not affect the operabon of the E2150F005A valve or of the core spray system The valve, after this modificabon, is unidirechonal in its ability to seal against leakage but l

will seal against leakage from the RPV in accordance with its isolation funcbon The funcbon of the L bypass line is passive, and the containment isolabon capability (leakage) is still maintained The funcbon or operation of system components or any interfaang system coirpor, ente has not been r altered in any way that would cause an equipment malfuncbon L

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Safety Evaluation No: 94-0049 UFSAR Revision No. 7 Reference Document: EDP-26523 Section(s) NA Table (s) NA I Figure Change X Yes No Title of Change: Replacement of Reactor Feed Pump Seal Water Temperature Controllers

SUMMARY

This modification replaced the old GEMAC seal water temperature controllers with better performing I Fisher Model 5190 controllers. The new controllers have very accessible controls, which include proportional band, reset, and rate adjustments. They also have an " anti-reset windup" feature and a local auto / manual station. Installation of new controllers will allow automatic control on the reactor i j

I feed pump seal water system. This modification eliminated several modules, and interconnecting cables which are not required with new controllers. In general, the defunct GEMAC modules were retumed to stock; and the Transmation devices were de-powered and left as installed spares. The j

new controllers do not require a separate electrical power supply, but do require pneumatic supply I (20 psig). This additional load on air compressors is essentially offset by elimination of defunct 1/Ps in the old temperature control loop. The new controllers, with anti-reset feature also permitted changing the temperature setpoint from 135*F to 150 F, as originally recommended by the vendor.

I The control room digital recorder N21R819 for the RFP seal water temperature was also reprogrammed from 150 F to 200*F maximum. Additionally, the RFP seal water temperature annunciator (5D50) setpoint was raised from 140*F to 180*F which is the manufacturer's maximum I limit.

Automatic temperature control on the seat water system will not increase the probability or l consequences of any accident described in the UFSAR. The failure of the controller will not cause the reactor feed pump to trip. This modification did not change the original intent of design, i.e.,

control seat water temperature to a nominal setpoint.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES I PAGE 74 j SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0051, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-26356 NA I TSR-26817 Section(s)

Table (s) NA I Figure Change X Yes No  ;

i Title of Change: Condenser Hotwell Level Control Instruments Loop Changes

SUMMARY

This modification relocated and replaced level instrumentation associated with north and south I condenser hotwell level controls. The major changes made include the following:

1. Replace Fisher Controls leve! transmitters (N61N409A&B and N410A&B) and Fisher Controls I level controllers (N61N406A&B) and their associated devices. Fisher Controls instruments were replaced with Foxboro Model 13DMP level transmitters, and a Fisher 4195 level controller, respectively.

2. A common set of new process taps was provided on the condenser wall. The high pressure tap was located on the wall just above the bottom, the low pressure tap was located on the wall, well above the operating water level. The old sensing lines were cut and removed and I the process taps were permanently capped. Relocating the taps to the condenser hotwell wall minimizes the possibility of rust particles entering the sensing lines thereby mitigating the probability of sediment clogging the sensing lines.

I 3. The new differential pressure instruments are equipped with sealed pressure sensors including capillary tubing. The use of sealed pressure sensors including capillary tubing shortens sensing line length which in tum maintains sensing line fluid temperature close to process temperature.

4. The old north and south umdenser hotwelllevel controllers were removed. A single common level controller was installed along with a manual 3-way valve and appropriate tubing reconfiguration. The new controller is equipped with a process indicator. Replacing the two-I blind level controllars with a common indicating controller mitigates process bumping when transferring condenser level control from one hotwell to the other. Also, the process indicator on the controller assists the operator in fine tuning the level setpoint to ensure a nearly bumpless transfer. ,

5. The main control room wide range indication selector switch SC35 was rewired and additional I contacts were provided. The additional selector switch contact blocks, in conjunction with rewiring, allows the operator to visually monitor the selected hotwell wide range level while the computer monitors both wide-range levels.

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I SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES I PAGE 75 Safety Evaluation 94-0051, Rev 1 (Continued)

6. The main control room hotwell control selector switch SC36 was removed and the mounting l hole plugged. The remote transfer switch is not needed since transfening condenser level control from one hotwell to the other now requires manipulation of the local controller.

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7. The new level instrumentation was relocated to an area that has lower radiation levels during plant operation. This will allow minimizing radiation exposure to l&C technicians performing routine surveillances.

This modification does not impact the function of the condenser level instrumentation system or any  ;

associated or support systems. The operation of the level control instrumentation is improved by )

I relocating the level control transfer switch from the control room to a local installation. This should improve level control transfer between north and south hotwells. In addition, continuous simultaneous monitoring of the north and south wide range levels on the process computer i improves the information available to operator. l This modification has no impact on any accident scenarios previously evaluated in the UFSAR nor has any safety-related function.

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I SAFE 7Y EVALUATIONS ENGINEERING DESIGN PACKAGES I PAGE 76 l SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0053 UFSAR Revision No. 7 Reference Document: EDP-13548 Section(s) None Table (s) None I Figure Change X Yes No Title of Change: Replacement of Obsolete Hydrogen Monitor in the Off Gas System

SUMMARY

l This modification replaced the existing offgas hydrogen monitor located in panels H21P275A

' and H21P2758. The replacement was needed because the old monitor was obsolete and getting replacement parts was difficult. The new monitor is contained in a single cabinet of the I same size and mounting contigaration as the existing H21P275A which it directly replaces.

The old analyzer components lo.;,ated in H21P2758 are removed, a coverplate installed and H21P2758 is being used as a termination cabinet to utilize the existing cabling to the control center.

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The new monitor contains dual channels, similar to the primary containment atmospheric monitoring system (T50) Hydrogen / Oxygen analyzers, except it will monitor hydrogen only.

' Dual channels will permit continued operation when one channel is out of service, without having to enter Tech Spec requirement. The replacement monitor is procured as non-Q but r will be maintained as QA Level 1M. The design pressure of the new analyzer is 75 psig, the same as the offgas system.

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SAFETY EVALUATION

SUMMARY

[

Safety Evaluation No: 94 0056 UFSAR Revision No. 7 b Reference Document: EDP-11881 Section(s) NA Table (s) NA Figure Change X Yes No t

Title of Change: Replacement of Degraded Transmitters in Various Systems

(

SUMMARY

This modification removed eleven transmitters in various systems (C32, C11, N21, and N20) and replaced them with new Rosemount, Inc. transmitters that are considered like-for-like in function and performance. The basic function of each instrument loop remains unchanged.

The new transmitters are suitable for their respective applications and are qualified for the environment in which they are expected to function. This modification incorporated the following:

[ 1. Removed existing main steam flow transmitters C32N002A and C32N002B from their L respective instrument racks and replaced them with new Rosemount, Inc.,1153 series F transmitters and mounting hardware installed in the same location.

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2. Removed existing feedwater flow transmitters C32N003A, C32N003B, C32N003C, and C32N003D from their existing instrument rack and replaced them with new Rosemount, Inc.,1153 series F transmitters and mounting hardware installed in the same location.

3. Removed existing locally mounted CRD drive water header and cooling water header r transmitters C11N008 and C11N011 and replaced them with new Rosemount, Inc.,1153 L series F transmitters and mounting bracket installed in the same location.

c 4. Removed existing RFP recirculation line flow transmitter N21N409A and N21N4098 from ,

i their existing instrument racks and replaced them with new Rosemount, Inc.,1151 series transmitters and new mounting hardware installed in the same location.

5. Removed existing condenser pump discharge header and condensate line to polishing demineralizer discharge header transmitter N20N404 from its existing instrument rack and replaced it with a new Rosemount, Inc.,1151 series transmitter and new mounting hardware installed in the same location.

6. Installed an analog signal isolator in the loop between the Rosemount, Inc., transmitters and existing GEMAC square root extractors for the main steam flow, feedwater flow, and reactor feed pump recirculation line flow measurement loops. The analog isolators will be located in existing cabinet space in the relay room racks.

SAFETY EVALUATIONS -

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SUMMARY

b Safety Evaluation No: 94-0057 UFSAR Revision No. 7 b Reference Document: EDP-26593 Section(s) NA NA

{ Table (s)

Figure Change X Yes No Title of Change: Removal of Blades and the Air Actuator From CCHVAC Damper

(- T4100F039A

SUMMARY

The CCHVAC Division 1 Retum Air Fan (T4100C031) had exhibited high vibration due to a damaged fan shaft. In order to replace the fan shaft, an access plate on the fan casing and

[- the removal of the shafts and blades from damper T4100F039A was required. The damper is currently abandoned in place in the full open position and performs no active function within the CCHVAC system.

This modification:

a. Added an access plate on the CCHVAC Division 1 Retum Air Fan T4100C031 to allow access to the fan intemals.

b. Removed the blades / shafts from damper T4100F039A. This damper is located on the L suction side of fan T4100C031 and removal of the blades was required to gain access to the fan intamals.

p c. Removed the air actuators on damper T4100F039A.

d. Disconnected and capped the instrument air tubing to the actuators.

E The components' removed (damper blades and actuators) served no active function in the system and are not required in any mode of operation. Removal of the components will not impact the performance or the operation of the system. The addition of a fan access plate will F~ . also not impact operation of the fan and the system. The CCHVAC and control air systems will operate and perform as described in the UFSAR.

SAFETY EVALUATIONS ENGINEERING DESIGN PACKAGES I PAGE 79 l SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0058, Rev 3 UFSAR Revision No. 7 Reference Document: EDP-26310 Section(s) 3.6, 5.5, 6.2, 7.6 Table (s) 3.2-1,6.2-2,6.2-15, 8.3-6, 5.2-4 Figure Change X Yes No Title of Change: Installation of a New Motor Operated Valve (G3352F220) in the Reactor Water Cleanup System (RWCU)

SUMMARY

This modification installed a new Motor Operated Valve (MOV) G3352F220 as outboard containment isolation valve in the RWCU retum line to the reactor feedwater line "B." The

{ design change involved associated piping change, isolation signal logic similar to G3352F004 isolation valve, and Class IE Div 11 AC power supply. The G3352F220 valve is a 4"O,900#

Powell motor operated gate valve which was designed to meet ASME Section ill Class 1

{ requirements. This MOV is installed upstream of G3300F120 check valve which was earlier performing the outboard containment isolation valve function. The piping class break is relocated upstream of the G3352F220 valve. Together, B2100F0108 and G3352F220 valves

[ now provide the containment isolation function for penetration X-9B.

The pipe, piping components, and valves for the pressure boundary up to valve G3352F220 have been designed to comply with ASME Section til Class 1 (Group A) requirements.

L Installation of pipe and valves up to valve G3352F220 was performed in compliance with l ASME Section XI requirements.

This modification also removed the air lines to the air operator on check valves G3300F120 l and G3300F121. The limit switches on these valves were also removed. The testable ,

7 function of the G3300F121 valve was also removed by this modification. The RWCU retum l L line outboard containment isolation valve G3352F220 control logic is the same as the RWCU suction outboard containment isolation valve, G3352F004. The valve is designed to close y when an isolation signal is received. The RWCU isolation signal is initiated by high RWCU L differential flow, high RWCU area temperature, high RWCU area differential temperature, Reactor Low Water Level 2 and Standby Uquid Control System initiation. The valve also has

~ a remote manual operation capability from the control room. Position indication of the valve is

[ provided in the main control room. A valve open and closed position signal is provided to Group 11 isolation mimic display logic. An alarm is also provided to indicate valve motor thermal overload on alarm window 2D40.

The indicating lights and control switch on control room panel H11-P602 for valve G3300F120 have been removed. The push-button switches for the new MOV G3352F220 are installed at F

ENGINEERING DESIGN PACKAGES PAGE 80 Safety Evaluation 94-0058 Rev 3 (Continued) the location previously used for the valve G3300F120 indicating lights. The hole left due to removal of the control switch for valve G3300F120 has been plugged.

The addition of 1.0 hp (MOV G3352F220) load to MCC72E-5A a id EDG #13 has been I evaluated and found to have no significant impact on existing equipment ratings. The addition of this load has been incorporated into EDG loading design calculation, UFSAR Table 8.3-6 and EDG loading sequence drawing. The valve control circuit is interlocked with the existing reactor isolation signal to close the valve on loss of coolant accident.

The rerouting of the RWCU retum piping to accommodate MOV G3352F220 as outboard I isolation valve would slightly increase system resistance, but not significantly enough to reduce the RWCU system nominal flow of 133,000 lb/hr.

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SUMMARY

Safety Evaluation No: 94-0059, Rev 3 UFSAR Revision No. 7 Reference Document: EDP-26631 Section(s) NA ECR-26631-1 Table (s) NA Figure Change X Yes No Title of Change: Demineralized Water Supply to Offgas System Ring Water Pumps

SUMMARY

This modification installed a demineralized water supply for each ring water pump (N6200C003 and C004). The new line is 5/8" tubing routed from the 1" condensate storage and transfer system piping upstream of the ring water buffer tank inlet air operated valve to existing flush water tubing connections at the pump's casing. Each line is equipped with one isolation valve which will manually be operated to provide the required water to the pump casing prior to pump start. This modification ensures proper pump start by providing a sufficient amount of water and increases the ,

reliability and operability without changing the intended design function of the pump. l The offgas system, including the ringwater pumps, performs no safety function. This modification does not affect any protective barriers such as fuel cladding, reactor pressure boundary, and the containment. Any failure related to this modification is enveloped by the accident analysis described in UFSAR 15.11.

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Safety Evaluation No: 94-0060 UFSAR Revision No. 7 l l

Reference Document: EDP-26510 Section(s) 3.4, 9.3 l

Table (s) None I Figure Change X Yes No Title of Change: Reactor Building Protection Against Intemal Flooding

SUMMARY

This modification upgraded the protection of the reactor building against intemal flooding I resulting from backflow through the floor and equipment sump pump discharge lines that run from the reactor building and drywell sumps to the collector tanks in the radwaste building.

This design change added redundant check valves and manual isolation valve to both the I 6-inch diameter common floor drain sump pumps discharge line and the common equipment drain sump pumps discharge line. In addition, three (3) 3/4-inch at connections were added to each line to facilitate leak testing of the check valves. The new check valves will allow the water collected in the individual sumps to be pumped to the radwaste tanks for processing, I and thus do not change the mode of system operation.

The modified portions of the floor and equipment sump pump discharge lines earlier were I classified as a non-O, Seismic II/l, ASME Section 111, Class 3 (Group C) piping system per Edison Design Specification 3071-519. This design change downgrades the portion where the modification has been installed to the ANSI B31.1 Code (Group D). The seismic II/I classification is still being kept. This downgrading is consistent with other modifications made in the past to these lines and is also consistent with the piping classification specified in L UFSAR Tables 3.2-1 and 11.2-6. The Design Specification 3071-519 has been revised to reflect downgrading of the piping code and classification.

UFSAR Section 3.4.4.4.2, Design Analyses, was revised to eliminate the discussion on backflow flooding scenario during a Probable Maximum Meteorological Event (PMME) resulting g from a single failure of postulated pipe cracks in the sump pump discharge piping. The new L redundant check valves and manual isolation valves provide backflow flood protection and prevent water from entering the reactor building.

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Safety Evaluation No: 94-0061 UFSAR Revision No. 7 1

Reference Document: EDP-14080 Section(s) None Table (s) None l _

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{ Figure Change X Yes No

, Title of Change: Replacement of a Dual Coil Solenoid Valve With Two (2) Single Coil I Solenoid Valves in Control Rod Drive Hydraulic System (C1100)

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SUMMARY

This modification replaced the dual coil solenoid valve, C11F409 with two (2) single coil,

- solenoid valves C11F409A and C11F4098. The use of two single coil valves is functionally

! equivalent to the old dual coil valve which was obsolete and had expired its qualified life of four years. The two replacement single coil valves are pneumatically connected in series with one

- being actuated by RPS A and other by RPS B. Although additionallosses are introduced into the air tubing due to the use of two valves and additional tubing and fittings, these losses are more than offset by the increased flow coefficient. The new valves have a flow coefficient of 1.0; the old valve had a flow coefficient of 0.15. The replacement solenoids are rated for the same nominal supply voltage.

Although there are now two valves in series instead of the one valve, the overall system safety margin is not adversely impacted because:

a. All replacement components are QAl and Seismic Category 1. They are suitable for use in this application and are no more likely to fail than any other safety-related components

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b. The two single coil valves are identical (so that one is no more likely to fail than the

- other). The same number of elec+rical components exist in the new configuration as the old.

_ c. A failure of either (or both) single coil valves has the same effect as a failure of the dual coil valve being replaced.

d. The Scram Discharge Volume (SDV) inboard vent and drain valves C1100F010 and C1100F011 are backed by the outboard vent and drain valves C1100F180 and C1100F181.

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j Safety Evaluation No: 94-0065 UFSAR Revision No. 7 Reference Document: EDP-26409 Section(s) NA )

I Table (s) NA l

lI Figure Change X Yes No I Title of Change: Replacement of Obsolete Instruments in Off-Gas System (N6200)

SUMMARY

This modification changed the temperature control loop of the east and west offgas preheater discharge to recombiner to eliminate the obsolete GEMAC controllers (N62K809A and B) and ,

I auto manual stations (N62K800A and B) and its associated devices. The obsolete GEMAC instruments have been replaced with the Moore Products Model 352E single loop digital controllers. This modification also replaced the position transmitters N62N411A and B, which I- were obsolete instruments and replaced the I/P converters N62K401A and B to match the signal with Moore Products Model 352E. l This modification does not affect the basic design of the offgas system. All the required controls, indicators, and alarms are functionally unchanged. The design of the charcoal adsorber is not modified by this modification; therefore, accident analysis in UFSAR I 15.11.4.1.2 is not affected. j I

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SAFETY EVALUATION

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[4 Safety Evaluation No: 94-0066 UFSAR Revision No. 7 ,

r k Reference Document: EDP-26436 Section(s) NA NA

(- Table (s) h Figure Change X Yes No g Title of Change: Modifications in Primary Containment Radiation Monitoring System L (PCRMS)

SUMMARY

) This modification replaced part of the piping on PCRMS rack H21P284, the moisture J separator, the reheater (T5001B002), the temperature controller (T50K001); deleted the non-Q rack isolation valves T5000F040, T5000F046 and associated pressure switch T50N010 and temperature switch T50N011; and added two T/Cs T50N498A&B. The existing 3/4"T moisture separator was replaced with a new large (3"T) moisture separator, an iso-kinetic probe and the piping will be sloped toward the moisture separator. The sample pipe section after the

) separator was projected into the moisture separator so that any moisture accumulation along j the skin surface of the moisture separator cannot be carried over. Deletion of non-Q rack isolation valves T5000F040, T5000F046 and associated switches will eliminate high maintenance and unnecessary LCOs. The requirement of rack isolation to protect the detectors from high sample pressure is achieved by the safety grade isolation valves T50F450, F451, F455 and F456. The isolation of the rack H21P284 by these valves does not compromise the grab sample capability. The grab sample line with the normally closed valve T5000F047A is located upstream of the isolation valves, and thus is available all the time.

The existing temperature controller and the band heater was replaced with a new dual element temperature controller and a new band heater. The new temperature controller operates based on the difference of temperatures at two different locations on the piping system downstream of the moisture separator. The reheater is interlocked with flow control valve T5000F043 (V5-2515). This eliminates the unnecessary cycling of the heater when the PCRMS is not operating.

These changes will improve the efficiency of moisture separation from the sample which should improve the quality of the gas sample delivered to the radiation detector and thereby improve the accuracy of the gas detector and the particulate / iodine filter readings.

These modifications do not negate any design criteria, licensing commitment, function and principles of operation of PCRMS.

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b Safety Evaluation No: 94-0067 UFSAR Revision No. 7

[ Reference Document: EDP-26312 Section(s) NA

[ Table (s) NA Figure Change X Yes No Title of Change: Abandoning Heat Tracing on Extemal Sample Lines on Primary b Containment Atmospheric Monitoring System (PCAMS)

SUMMARY

This modification abandoned in place the nonsafety-related heat tracing on the extemal sample lines by disconnecting the power supply to the heat tracing and identifying the disabled heat tracing control panel and other equipment as " abandoned in place" with proper labeling.

The design calculation on the condensation study for PCAMS was revised to document the impact on T50-H2/02 monitoring system due to deletion of the non-Q heat tracing on the extemal sample lines. It was determined that there is no significant impact on the measurement of T50-H2/02 gas concentration due to deletion of extemal heat tracing. The air sample reaches the H2/02 monitoring panels at a temperature higher than the dew point temperature, and a such , a nearly dry sample is provided at the H2/02 monitor. Under LOCA conditions with no heat tracing on extemal sample line, but with the intemal heat tracing operating, the hydrogen concentration is expected to be only slightly higher than the actual concentration in the conservative direction.

A design calculation was prepared to determine the setpoint for the intemal heat tracings. The

[ calculated value of the intemal heat tracing is 107.5'F and is documented and controlled. The h design calculation provides the technical basis for the new setpoint.

The modification does not negate any design criteria, technical assumptions, or licensing commitments for the system.

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Safety Evaluation No: 94 0068 UFSAR Revision No. 7 Reference Document: EDP-26182 Section(s) 7.6 Table (s) NA I-Figure Change X Yes No Title of Change: Enhancing Capabilities of the Rod Worth Minimizer (RWM)

SUMMARY

This modification enhanced the capabilities of the Rod Worth Minimizer (RWM) by adding the I following festures: single rod scram data display, shutdown verification display, and multiple rod motion (MRM) rod drive block and display function. This modification did not change the existing RWM rod sequence control, i.e., Banked Position Withdrawal Sequence (BPWS)

I adherence function. The three new features (incorporated by this modification) use existing input signals from Reactor Protection System (RPS) and Rod Position Information System (RPIS) to the RWM. The MRM output actuates the existing RWM rod drive block and settle

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i relays to the Reactor Manual Control System (RMCS) to apply rod drive block and settle I signals. No new input or output signals are needed to implement the new features. The Control Rod Drop Accident (CRDA) and Continuous Rod Withdrawal during startup events that l

l are described in UFSAR section 15.4 are not affected by this modification.

I A failure of the RWM with new features will not affect or prevent the operation of safety-related systems. Shutdown verification display provides the operator with indication of rod position I following scram and allows the operator to make an early determination of the success of the scram. The MRM rod block and display will terminate and limit rod motion of both the selected and any unselected rods to one notch if an MRM were to occur. This feature not only limits rod motion to one notch, but provides the operator with an indication that a hardware failure has  ;

occurred in the Reactor Manual Control System. Single Rod Scram Data display is selected by the operator during single rod scram testing to display actual scram time for the rod under test, l the average Technical Specification scram time and margin of the scram time of the tested rod to the Technical Specification time.

I The RWM operability and surveillance requirements as specified in Technical Specification 3/4.1.4.1 are not impacted by this modification.

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b Safety Evaluation No: 94-0069 UFSAR Revision No. 7-

[L Reference Document: EDP-26367 9.2,9.4,10.2 Section(s)

Table (s) 9.4-5, 12.1-21,

{. 12.1-24,12.1-28

[' Figure Change X Yes No Title of Change: Replacement of the Main Generator Shaft Driven Excitation System with h a Static Excitation System

SUMMARY

During the December 25,1993 turbine generator incident, the main generator shaft driven -

excitation system suffered irreparable damage. This modification replaced the shaft driven

( main and pilot exciters, excitation system water cooled rectifier unit, automatic voltage regulator (AVR) unit, and field suppression unit with a static excitation system from Asea Brown Boveri (ABB) consisting of a 5100 KVA dry type excitation transformer, rectifier cubicle,

. excitation control cubicle, and field suppression cubicle. Water cooled air cc*rs have been added to maintain a nominal ambient temperature of 104*F in the excitation equipment area.

The new excitation transformer, rectifier cubicle, field suppression cubicle, and air coolers are located on the second floor of the Turbine Building. The new excitation cubicle is located in the Relay Room in the Auxiliary Building. The new excitation system provides improved performance, reliability and efficiency with an added benefit of reduced maintenance.

Two 100% capacity, 30 ton, water cooled exciter air coolers are installed to remove the additional heat generated by the excitation transformer and air cooled rectifier cubicle and

f. maintain the excitation equipment area at a nomina! ambient temperature of 104*F . Each h cooler is provided with an isolation damper to isolate the non-operating cooler from the operating cooler. New ductwork has been installed from the air coolers to the excitation equipment area for supply and retum air. Turbine building closed cooling water (TBCCW) system supplies cooling needed for the new air omier.

Balance of plant battery 2PC supplies the DC power feed for excitation system indicating lights. The Turbine Building ventilation system drawings have been revised to incorporate the addition of new excitation equipment area coolers. Air coolers and the cooling water lines do not affect any of the analyzed accidents or transients described in the UFSAR. Electrical system load changes due to this modification do not affect any of the analyzed accidents or transients described in the UFSAR.

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Safety Evaluation No: 94 0073, Rev 1 UFSAR Revision No. 7 i

l g Reference Document: EDP-26726 Section(s) 3.5,10.1,10.2, and 10.4 l g ECR-26726-1 1 Table (s) 10.1-1 Figure Change X Yes No l

E B Title of Change: Replacement of 7th and 8th Stage Blades and Diaphragms with Pressure Plates in Low Pressure Turbines 1,2 and 3 l

SUMMARY

I This modification removed the 7th and 8th stage rotating and stationary blading at each end of each of the three low pressure main turbine cylinders and installed pressure plates in place of lI the 7th and 8th stage stationary blading. The pressure plates are thick steel plates with a large number of small holes designed to ensure that the steam pressure drops between stages l

g are maintained at levels similar to the original design, thus, allowing safe and reliable operation

,g i during cycle-5. This modification was implemented as a result of root cause investigation and inspection of the turbine generator failure event on December 25,1993.

This modification has resulted in changes to various parameters which impact the plant heat balance and electrical generation capability. The UFSAR figures have been revised to i

incorporate these changes. However, the addition of pressure plates has not affected critical parameters (feedwater temperature, turbine bypass flow, moisture separator reheater steam flow, and steam dome pressure) which are inputs to plant safety analysis. The pressure plate design affects the extraction steam flow and drain flow to feedwater heater #1. The flows are decreased and thus are well within the design capabilities of the system. Since the generator output is decreased by approximately 200 MW, this will impact the performance of condenser in increasing back pressure by about 0.2" Hg. The turbine trip and alarm setpoints will remain the same as before. Exhaust steam to the condenser is now at a higher velocity and enthalpy.

An evaluation was performed to assess the adequacy of the current condenser tube anti-vibration staking at the higher steam flow conditions. The results of this assessment indicated no adverse effects. Operation with pressure plates increases the heat load on the cooling towers and circulating water system by about 200 MW. This additional load causes the average outlet temperatures of the circulating water system and cooling tower to be higher.

I There are no safety-related impacts to such a change. This modification did not impact the function and/or setpoints for LP exhaust spray cooling system.

Safety-related aspects of the operation with pressure plates were evaluated against postulated failures within the power conversion system as described in UFSAR 10.1. Given below are the conclusions of this evaluation.

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SAFETY EVALUATIONS ENGINEERING DESIGN PACVAGES PAGE 90 Safety Evaluation 94-0073, Rev 1 (Continued)

• Radiological consequences of the postulated breaks in the feedwater system that allow the discharge of contaminated water into the turbine building are not affected because:

( l ) feedwater tempeiature and flow has not changed, and ( ii ) radioactive coolant activity

[ remained unchanged because the pump forward drain portion of the feedwater remained unchanged.

( a Radiological consequences of postulated failure of the steam jet air ejector line resulting in discharge of the activity directly into the turbine building is not affected since there is no change in the reactor power and steam flow to the turbine.

• Reanalysis of the turbine missiles with new configuration concluded that all missile structures and barriers are structurally adequate.

• Introduction of contamination into the reactor vessel via the condensate /feedwater system is not affected since condensate is treated in the condensate polisher /demineralizer.

Additionally, the root block material and pressure plate material are comparable to blading and diaphragms which have been removed.

All other safety analyses, including LOCA, containment analyses, reactor overpressure analyses and ATWS were considered and have not been affected by this new turbine configuration.

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SAFETY EVALUATION

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Safety Evaluation No: 94-00'74, Rev 1 UFSAR Revision No. 7 Reference Document: EDP-26660 Section(s) NA

[ Table (s) NA

( Figure Change X Yes No Title of Change: Replacement of Reactor Pressure Vessel (RPV) Bottom Head Drain Line b Thermocouple

SUMMARY

[ This modification replaced the existing RPV bottom head drain line thermocouple G33N042 with two thermocouples of a new design. The old design used one thermocouple with dual elements, and the new design uses two thermocouples, each with a single element. In addition, old mirror insulation was replaced with a longer piece to cover more of the pipe around the thermocouples. The increased weight of the replacement mirror insulation was reconciled with the piping stress calculations and was determined to have negligible effect.

The seismic qualification was also addressed. The new thermocouples use smaller gauge wire and do not require thermowells. The change made to the RPV bottom head drain line thermocouple in this modification does not affect the function of the thermocouple or its l

temperature indication. Two single element thermocouples will provide the same information that was previously provided by one dual element thermocouple. With two separate thermocouples, the design is less susceptible to single failure and loss of indication is less f likely.

L The bottom head thermocouple is classified QA Level 1M and Seismic Category ll/l. The p temperature indication is not safety related. However, because it is used for complying with

' Technical Specifications, it is classified QA Level 1M.

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Safety Evaluation No: 94-0075 UFSAR Revision No. 7

[ Reference Document: EDP-13958 Section(s) 15.7,15.11, A.1.143

[ Table (s) 3.2-1, 3.9 1, 3.9-27

( Figure Change X Yes No Title of Change: Various Modifications in the Off Gas System

SUMMARY

This safety evaluation covers a number of design modifications made in the off-gas system via EDP-13958. The design modifications included: (1) pressure setpoint increase for preheater relief valves (N6200F114A/B) from 195 psig to 210 psig and replacement of existing 3/4" x 1" preheater relief valve _with a larger capacity 1-1/2" x 2" relief valve, (2) installation of orificed bypasses around preheater shell side continuous vents, (3) installation of new screen mesh over the catalyst bed in the recombiner, (4) installation of new relief valve downstream of pressure control station (N11-F400A/B) for steam supply to steam jet air ejector with the relief valve discharge routed to main condenser, and (5) reclassification of the Off Gas System

, design to a non-ASME Code Section til system.

L in addition, sections 15.7.1 and 15.11.1 of the UFSAR have been corrected to reflect that the

' off-gas piping and components were not designed to the uniform building code seismic requirements. For this modification, a gross failum of the off-gas system due to a seismic event more severe than the current design basis ivas analyzed, it is concluded that offsite r dose would not exceed 0.5 rem as specified in Reguictory Guide 1.29. Therefore, the off-gas L system does not require Category I design, and these modifications were performed as non- l Category I modifications. i The physical modification (items 1-4 as stated above) will bring the system to existing code requirements and improve system operational characteristics. The reclassification of system piping, valves and ring water vacuum pumps (item 5 above) does not produce any functional L changes in the system. The reclassification revises the requirements for maintenance and procurement activities. The reclassification of system components is in accordance with currently accepted design requirements for the off gas systems (Regulatory Guide 1.143).

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Safety Evaluation No: 94-0082, Rev 1 UFSAR Revision No. 7 ,,

Reference Document: EDP 26580 Section(s) None Table (s) None Figure Change X Yes No Title of Change: Modification of Water Level Control Loops for Feedwater Heaters 1 and 2

SUMMARY

[ This modification provided enhancements in the water level control loops for the normal and emergency feedwater heater drain lines on the feedwater heaters 1 (N, C, and S) and 2 (N, C, and S). The system enhancements include adding a new independent controller for control of

[ heater emergency drain valve, upgrading the existing controller for normal level control valve, installing a new 3/8" tubing air signal line, and upgrading solenoid valves in the system. This modification also removed feedwater heaters 1 and 2 rack mounted level indicators which are I not used. Removal of these indicators from the associated instrument racks will provide additional rack space for mounting new emergency drain line level controllers.

E Separate controllers for the normal and emergency level control valves eliminate a series operation of these valves. The new scheme also provides finer control and smoother valve positioning. The stability and reliability of heater drain system is also improved by separating the operation of normal and emergency level control loops.

This modification, which is associated with a nonsafety-related system (located in the turbine r building) does not directly or indirectly affect any safety-related or important to safety function L of the plant.

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Safety Evaluation No: 94-0083 UFSAR Revision No. 7 l

Reference Document: EDP-26749 8.3.2 Section(s)

Table (s) NA l

L Figure Change X Yes No Title of Change: Design Modifications for implementation of Hydrogen Water Chemistry

[ (HWC) Systems

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SUMMARY

Hydrogen Water Chemistry (HWC) Systems (Hydrogen injection System and Oxygen injection l System) are planned for implementation during Cycle 5. This modification installed signals from the feedwater flow rate, reactor protection system trip, offgas system trip, and heater feed pumps " running" interlock and brought them to a convenient point for tie into the new HWC l equipment during Cycle 5 operation without requiring a plant shutdown or system outage. In addition, UFSAR sections 8.3.2.1.3.2 and 8.3.2.1.3.3 have been revised to show that 24/48 yde systems A and B also fumish power to other miscellaneous instrument loops. These additional loads have been analyzed and found to have no appreciable impact on the 24/48 vde battery system operation.

The feedwater flow rate signal is taken by addition of a circuit similar to STARTREC input, with an isolator to provide 4-20 made signal. Failure of feedwater flow accident has been evaluated

] in the UFSAR, and this modification is within the envelope of the analysis. The signals for the RPS trip and offgas system trip have been taken from the spare relay contacts indicating the required status. This modification will have no impact on the operation of RPS or offgas system because existence of contact to coil separation. Heater feed pump " running" signal has been taken from the auxiliary contact of the switchgear breaker feeding each heater feed

- pump. ,

r All of the above signals, with the exception of heater feed pump running signal, are wired to relay room panel H11P878 from their respective system panel in the relay room.

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Safety Evaluation No: 94-0084 UFSAR Revision No. ~7 Reference Document: EDP-26790 Section(s) NA Table (s) NA Figure Change X Yes No Title of Change: Addition of a New Valve in Torus Water Management System

SUMMARY

This modification added a 3" nominal pipe size (NPS) branch connection to 6" NPS line M-I 4091 with new isolation valve G5100F084 and a blank-flanged end. Line M-4091 is the Torus Water Management System (TWMS) supply from TWMS pumps to the Condensate Demineralizer System (injecting downstream of the condenser pumps and upstream of the I polishing demineralizer beds). This connection is made to allow installation and removal of a temporary modification (to clean suppression pool water) when needed. Line M-4091 is designated Quality Group D and is built to the ANSI B31.1 Code. The piping is nonseismic and nonsafety-related, and the new connection is located in the turbine building basement pipe tunnel. Design parameters are 350 psig at 195'F with normal parameters of 210 psig at 95'F, The new pipe tap is not required to perform any safety-related function. It has no effect on the function or operation of any other component or system. Flooding associated with postulated failure of the new pipe tap is bounded by postulated header piping failure or larger turbine building pipe breaks. The materials and components used in this modification are of the same quality standards as the original system. The pipe taps and header piping are designed to keep pipe stresses within ANSI B31.1 code limits.

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Safety Evaluation No: 94-0085, Rev 1 UFSAR Revision No. 7 l

I Reference Document: EDP-26664 Section(s) NA Table (s) NA Figure Change X Yes No Title of Change: Adding Pipe Connections in Off Gas, Condensate, and Feedwater Systems

SUMMARY

This modification added piping connections to the off gas, condensate, and feedwater systems.

These piping connections will be used during cycle #5 when hydrogen water chemistry and zinc injection systems are installed without a need of shutting cbwn the reactor. The piping connechons added include a manual valve for isolation and a check valve to prcyont back flow. The end of the I piping was brought out to a low radiation area and was capped. The piping connecbons added include:

• 3/4" connections to inject hydrogen into the suction side of each of the three heater feed pumps (N2003C012, N2003C013 and N2003C014).

• A 2" connection to the 18" off gas manifold to inject oxygen into the off gas system.

• A 1/4" tubing connection into hydrogen monitor tubing for the oxygen monitor.

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• 2" connections to the reactor feedwater recirculation lines for the supply and retum piping  ;

for the zine addition skid to be located in the southeast comer of the first floor of the TB 1 steam tunnel.

• A 1/4" tubing connection on the sample drains collector tank (P3301A001) retum to the condensate system to inject oxygen into the condensate system for corrosion control.

1 The new piping and valves associated with this modification were designed and constructed in I accordance with ANS1 B31.1.0 (1967) Power Piping Code, Group "D" and Quality Assurance Level Non-Q. The piping and the tubing added by this modification is supported in accordance with design specification. The addition of these taps does not make a seismic event (rupture of the i

feedwater or off gas systems) more likely; nor will the taps change the radioisotopic mixture i

, contained in the system. The taps do not affect the performance of any system required to operate j following a transient or an accident. This modification has no impact on Offsite Dose Calculation l Manual and Process Control Program. l I l

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Safety Evaluation No: 94-0087 UFSAR Revision No. 7 Reference Document: EDP-26828 Section(s) None Table (s) 9.3-2 I Figure Change X Yes No Title of Change: Installation of a New Sample Point in the Reactor Water Cleanup System

SUMMARY

This modification installed 's new sample point downstream of valve G3352F119 (second valve outside primary containment) to monitor the effectiveness of Hydrogen Water Chemistry (HWC) System, which will be implemented during Cycle 5 of operation. The sample is routed into RWCU heat exchanger room to a roughing cooler which is cooled by RBCCW. The I sample is then routed to the Reactor Building Sample Station (P33P405A) to a spare sample connection. The addition of this new sample point does not change the function of the RWCU and RBCCW systems.

This modification does not change any chemistry parameter that is measured for maintaining reactor coolant chemistry. The foilure of this new sample point is bounded by analysis covering an instrument line break outside the primary containment (UFSAR 15.6.2).

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Safety Evaluation No: 94-0089 UFSAR Revision No. 7

[' Reference Document: TSR-26917 2.2 Section(s)

[ Table (s) NA

( Figure Change X Yes No Title of Change: Installation of Chemical Storage Tank and Associated Piping and Valves at

( Circulating Water Pump House

SUMMARY

This modification installed one (1) 8200 gallon (nominal) fiberglass reinforced plastic (FRP) multipurpose chemical storage tank and 22' x 22' x 3.5' pad providing 150% retention capacity

( should a rupture of the tank occur. A 2" schedule 80 PVC pipe is routed from the drain connection into the CW Pump House for connection to a pump skid provided and controlled by Chemistry. Two (2) isolation valves are provided on the suction line, one inside the pump

" house and one within the diked area. A tank fill line with an isolation valve and 'camlok' type fitting are also provided. Due to the excessive cost associated with the use of bromo-chloro-dimethylhydantoin (BCDMH) to treat the Circulating Water (CW) System for micro-biological

' fouling of the Main Condenser, the plant instituted daily treatments using sodium hypochlorite (NaOCl) after RFO3. The delivery system consisted of a gravity feed upstream of the CW pumps utilizing either a semi-bulk container or a transport tanker supplied by a local

' hypochlorite vendor as the storage vessel. This treatment method was successful, and the plant decided to install a permanent storage tank that can be used to store hypochlorite, or other chemicals which may be used in the future to treat the CW System or the Main r Condenser. The hypothetical rupture of a circulating water expansion joint at the Main L Condenser, and subsequent flooding of the Turbine Building (UFSAR 10.4.5.3)is unaffected because NaOCl has no deleterious effects on the expansion joint material. Therefore, use of r NaOCl will not increase the probability of an expansion joint failure resulting from a pressure L surge in the CW System. Habitability of the Main Control Room is unaffected by this modification and the use of NaOCl since the control Center HVAC chlorine detectors remain in r operation and are unaffected by this modification. In addition, hypochlorite is not as volatile as L chlorine gas, and generates fumes similar to those resulting from the solution of BCDMH that would be created in the brominators. Should a tank failure occur, the contents would be c retained within the 150% capacity dike surrounding the tank. No gas cloud would be created.

[ Therefore, this change will not increase the probability or consequences of an accident previously evaluated in the UFSAR.

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Safety Evaluation No: 94-0092, Rev 1 UFSAR Revision No. 7 j 1

Reference Document: EDP-26868 Section(s) 9A.4.2 l l

Table (s) None Figure Change l X l Yes l lNo Title of Change: Relocation and Replacement of Fire Dampers in the Control Center Heating, Ventilating, and Air Conditioning System

SUMMARY

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This modification installed a new fire damper (T4100F903) in the common discharge ductwork of the CCHVAC retum air fans and removed the fire dampers (T4100F111 and F112) which had been previously installed improperty. UFSAR section 9A.4.2.15.1 has been revised to describe operator ,

actions that are required to open fire damper T4100F903 in the event of the damper closing due to I fire in the ventilation equipment room. In addition, two fire dampers (T4100F109 and F110) on the suction side of retum air fans were removed and replaced with two new fire dampers that will be

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l located in the plane of the new fire wall.

The new fire damper configurations in the CCHVAC system do not impact the ability of the system l to provide an acceptable environment in the main control room for the control room operators, l equipment and controls as described in the UFSAR. The system parameters, such as air flow rates and duct pressures, are not impacted by this modification. The CCHVAC sy0em will operate as designed and described in the UFSAR. The new fire damper configurations are in compliance with ,

l Fermi 2's Fire Protection Analysis described in the UFSAR (Appendix A). These new damper configurations are installed to meet the requirements of a UL listed 3-hr rated fire barrier. In the l

event of a fire in the plant, the operators will be able to achieve safe shutdown of the reactor. For a fire in the main control room, the fire dampers will keep the fire from spreading to the ventilation equipment room and the plant can be safely shut down from the dedicated shutdown panel. Also, this modification does not result in an increased challenge to any safety system assumed to ,

function in Fermi 2's accident analysis such that any safety system's performance is degraded  ;

belowits design basis.

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{J Safety Evaluation No: 94 0093 UFSAR Revision No. 7

[: Reference Document: EDP-27003 Section(s) ,None Table (s) s.3-2

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s.3-s

[ Figure Change Yes X No Title of Change: . Replacement of Motor on Valve E4150F002, HPCI Inboard Steam Isolatum

[ Valve

SUMMARY

This modification replaced the existing motor (60 ft-Ib, 7.8 hp, 7.3 kW) on the HPCI inboard steam isolation valve (E4150F002) with a similar motor (80 ft-lb,10.3 hp, 9.8kW). The replacement was made to assure that enough torque will be produced under degraded voltage and accident temperature conditions. The larger size motor was evaluated for maximum allowable thrust and valve function. This modification resulted in revising UFSAR Table 8.3-2,

[ " Emergency Diesel Generator System Divisional Connected Loads" and Table 8.3 6,

" Emergency Diesel Generator System: Loss of Offsite Power and Loss of Coolant Accident at Zero to Ten Minutes." The replacement of the motor resulted in an increase of 2.3 kW in electncal loading of EDG-12. It has been determined that total load on each EDG for all

{ conditions is within the short-time rating of diesel generator in accordance with paragraph C.2 l

of Regulatory Guide 1.9 Revision 2 and item 3.7.2 of IEEE Standard 387-1977.

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Safety Evaluation No: 94 0094 UFSAR Revision No. 7 i.

Reference Document: TSR-26607 Section(s) _NA a

[ Table (s) jj h Figure Change X Yes No 7

Title of Change: Removal of Level Switches From the Condensate Phase Separators

SUMMARY

This. modification removed level switches G11N09A-D from the condensate phase separators in the radweste system. The backlit indicators for " low" and "high" sludge level for both phase separators (A and B), located on Radwaste Control Room Panel G11-P001-3 were also removed and replaced by bisnk plates. Level elements G11N179A and B, located in phase 4 separators A and B are abandoned in place. This modification has no impact on the liquid level loops used to decant the supernatant from the phase separators. As a result of the December 25,1993, Turbine Generator failure and subsequent flooding of the Radwaste Building Basement, Condensate Phase Separator A and B sludge level indicatmg switches G11N090A-D were damaged. Repair efforts were attempted, but were unsuccessful due to unavailability of parts as the vendor is no longer in business. The design basis for these devices is to provide the Radweste Control Room operator a qualitative indication of the amount of' resin / sludge contained within the Condensate Phase Separators. During preoperatumal testing of the system, the need to accurately track the quantity of resin / sludge contained within the phase separators was identified. An administrative system was put into place to track backwash volumes from the various filter domineralizers in the plant and initiate j '

processing of the phase separators at a predetermined solids volume. Therefore, the slud'as level indicators became unnecessary components but were not deleted from the design.

Elimination of the level switches and abandonment of the level elements in place has no effect on operation of the nonsafety-related condensate phase separators. There is no safety-related equipment assumed to function in any of the accident analyses in the UFSAR supported by operation of, or interlocked with these devices. This modification has no effect on the Liquid Radwaste System; therefore, the limits imposed by the Offsite Dose Calculation l Manual (ODCM) 3.11.1.1 and 3.11.1.2 and the associated margin of safety remain unaltered.

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I Safety Evaluation No: 94 0096 UFSAR Revision No. 7 Reference Document: EDP-27002 Section(s) NA Table (s) 8.3-2 8.3-6 Figure Change Yes X No Title of Change: Replacement of Existing Motor with Large Size Motor for MOV B3105F031A

SUMMARY

I This modification replaced the existing motor (60 ft-lb, 7.8 hp, 7.3kW) for reactor recirculation system motor operated valve, B3105F031 A with a larger size motor (80 ft-lb,10.3 hp, 9.6 kW).

The replacement was made to assure that enough torque will be produced under degraded l voltage and accident temperature conditions. This modification resulted in revising UFSAR Table 8.3-2, " Emergency Diesel Generator System Divisional Connected Loads" and Table 8.3-6, " Emergency Diesel Generator System: Loss of Offsite Power and Loss of Coolant l Accident at Zero to Ten Minutes."

The replacement of the motor results in an increase of 2.3 kW in electrical loading of EDG-12 and 14. It has been determined that total load on each EDG for all conditions is within the short time rating of the diesel generator in accordance with paragraph C.2 of Regulatory Guide 1.9 Revision 2 and item 3.7.2 of IEEE Standard 387-19T/.

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Safety Evaluation No: 94-0101 UFSAR Revision No. 7 Reference Document: EDP-27064 Section(s) NA Table (s) 8.3-6 I

Figure Change Yes X No Title of Change: Replacement of Existing Motorwith Large Size Motor for MOV P4400F601B

SUMMARY

This medification replaced the existing motor (5 ft-Ib, 0.33 hp, 0.5 kW) for emergency equipment cooling water system motor operated valve P4400F601B with a larger size motor I (10 (t-lb,0.7 hp,1.45 kW). The replacement was made to assure that enough torque will be produced under degraded voltage and accident temperature conditions. This modification resulted in revising UFSAR Table 8.3-6, " Emergency Diesel Generator System: Loss of Offsite I Power and Loss of Coolant Accident at Zero to Ten Minutes."

The replacement of the motor results in an increase of 1.0 kW (actual increase of 0.95 kW) in l electrical loading-of EDG 14. It has been determined that total load on EDG 14 for all conditions is within the short time rating of the diesel generator in accordance with paragraph C.2 of Regulatory Guide 1.9 Revision 2 and item 3.7.2 of IEEE Standard 387-1977.

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f-Safety Evaluation No: 944102 UFSAR Revision No. 7 Reference Docurnent: EDP 27067 Section(s) NA _

[ Table (s) 8.3-2 8.34 Figure Change Yes X No Title of Change: Replacement of Motor for Valve (T4803F601) in the Nitrogen inerting System

SUMMARY

} This modification replaced the existing motor (40 ft-Ib, 5.3 hp, 4.8kW) for the nitrogen inerting

) system valve T4803F601 with a new motor (40 ft-lb, 5.2 hp, 5.1kW). This change in motor introduced additional load on EDG-11. UFSAR Table 8.3-2, " Emergency Diesel Generator System Divisional Connected Loads" and Table 8.3-6, " Emergency Diesel Generator System -

Loss of Off-Site Power and Loss of Coolant Accident at Zero to Ten Minutes" have been revised to incorporate the change.

The new motor was evaluated for feeder size adequacy, starter size, fuse size, maximum allowable thrust, stroke time impact, additional weight impact on piping and the system, and the MOV functxm The new motorwas found acceptable as a result of this evaluation.

As a result of this modification, the load of 2756kW was calculated for EDG-11 with the loss of off-site power and loss of coolant accident at 0-10 minutes when all EDGs are available. This new calculated load is within the short time rating of the EDG in accordance with Paragraph C.2 of Regulatory Guide 1.9, Revision 2 and item 3.7.2 of IEEE Standard 387-1977.

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h Safety Evaluation No: 94-0104 UFSAR Revision No. 7 b Reference Document: TSR-27016 NA Section(s)

EDP-27029 Table (s) NA

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h Figure Change X Yes No Title of Change: Installation of New Pressure Transmitters on Feedwater Heaters #1 (N, C, &

( S) and #2 (N, C, and S)

SUMMARY

This modification installed new transmitters to measure Feedwater Heaters No.1N,' 1C,1S, 2N,2C, and 2S shell pressure. _ These transmitters are referenced to absolute zero pressure,

( utilize capillary tubing with remote pressure seals, are highly accurate and of the latest Rosemount design. The capillary, remote seal feature allows an installation that ensures that the wet leg will remain filled (sealed capillary). The calibrated range of 0 to 15 psia of the new transmitter provides information that is more accurate and easier to use than the existing 30" HgVac to +30 psig calibrated range. New cables were installed to connect the new transmitters to local racks H21-P405, H21-P545, and H21 P546, where the old transmitters N22-N482A, B, C and N22-N500A, B, C with associated pressure indicators N22-R401A, B, C and N22-R403A, B, C were located. The old, nonfuncboning instrumentations were removed.

Now, local, loop-powered digital indicators scaled for 0 to 15 psia were installed on these local r

racks. The existing cables from these local racks to Relay Room panel H11-P872, and from L this panel to the computer room analog 1/O panel C91 P618 for the process computer interface were used. New 24 vde loop power supplies were installed in H11-P872 since the old 52.5 vdc r power supply voltage rating is too high for the new transmitters. The old process computer 1.6 L ohm signal dropping resistors were replaced with 4 ohm resistors to maintain the same my signal input range to the process computer. The process computer points (sequence numbers p A126 through A131) were rescaled to agree with the new transmitter calibrated range L corresponding to O to 15 psia.

p Replacement of the inoperative feedwater heaters 1 and 2 pressure instruments is required for L performance evaluation of the main turbine. The new transmitters utilize the same source tap and root valves as the transmitters and pressure indicators being removed. The connechon between the capillary diaphragm seal and root valve is designed and constructed to the same

(. standards as the connections to the removed instruments. The transmitter output is not used for control, manual or automatic. There is no increase in the probability or the consequences of an accident previously evaluated in the UFSAR.

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Safety Evaluation No: 94-0105 UFSAR Revision No. 7 Reference Document: EDP-27125 Section(s) NA Table (s) 6.2-2

[ Figure Change X Yes No r Title of Change: Elimination of Pressure Locking in Core Spray Division il Inboard Isolation L Valve (E2150F005B)

SUMMARY

To eliminate the susceptibility of pressure locking in valve E2150F005B, this modification added an insulated equivalizing bypass line from the E2150F0058 valve to the local leak rate

[ ' test (LLRT) line on the downstream side to equivalize the pressure in the body cavity of the valve with the piping to the reactor vessel. The existing leakoff connection was also plugged under this modification. The bypass line consists of small bore pipe and elbows, which are insulated.

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Stresses in the existing LLRT piping and the new pressure relief bypass line were evaluated E for the applicable static and dynamic loading conditions to ASME Section ill NC requirements.

The impact to Class 13/4" Sockolet and 12" header piping due to the modification was also evaluated and found acceptable. The materials used in this modification and the system piping that it is interfacing with are QA-1, Group A and are safety related.

This modification does not affect the operation of the E2150F005B valve or of the core spray system. The valve will now be unidirectional in its ability to seal against leakage, but will seal against leakage from RPV in accordance with its isolation function. This valve is local leak rate

] tested, which will ensure all leakage requirements are met.

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Safety Evaluation No: 94-0109 UFSAR Revision No. 7 Reference Document: EDP-26837 Section(s) 5.2,10.4 Table (s) NA Figure Change X Yes No Title of Change: Implementation of Zinc Injection System

SUMMARY

This modification installed skid mounted equipment in the TB-1 steam tunnel that will be used for the injection of depleted zine oxide into the feedwater. The skid includes a zine oxide

,I dissolution column, strainer, filter, flow straightener, flow control valves, other associated piping and valves, and instrumentation. The zine injection system will be tested and placed in operation during Cycle #5. Zine has been shown to reduce radiation fields coming from various primary coolant pipes by occupying the deposition sites that cobalt-60 would occupy.

The cobalt-60 will then be removed by the reactor water cleanup filter /demineralizer instead of deposited on piping surfaces. The zinc injection system is nonsafety-related, QA level Non-Q, seismic category None. The new system is not required to mitigate consequences of an accident or for safe shutdown of the reactor. The system is a passive system with no active components.

The impact of zine addition in the feedwater has been evaluated for nuclear fuel, primary coolant loop materials, feedwater and reactor water chemistry, radwaste accident (UFSAR I 15.7.3), feedwater system operation and postulated double ended break of 2-inch line in the zine addition system. The results of these evaluations indicated either an insignificant impact or no adverse consequances from those already analyzed in the UFSAR.

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l Safety Evaluation No: 94-0111 UFSAR Revision No. 7 Reference Document: EDP-27122 Section(s) 5.5.6 Table (s) NA

.I Figure Change Yes R No Title of Change: Change of Gear Set for RCIC Injection Valve E5150F013

SUMMARY

1 This modification changed the gear set of the motorized actuator for E5150F013 to increase the f I mechanical ratio and thrust output capacity. This minimizes the additional torque required from the motorin order to achieve the revised actuator torque output. The increased output gear ratio of the actuator ensures that the motor can perform its design function during a degraded voltage

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j condition. The consequences of increasing the mechanical ratio is to reduce the stem stroke speed I and increase the valve cycle time. The additional leakage through this line due to 9.4 seconds additional stroke time is insignificant compared to the time required for the control room operators to assess plant conditions, determine that isolation of the RCIC injection line is required, and actuate I the control switch for closure of the valve. In most analyses, operator action is not credited for at least 600 seconds; thus, the change in stroke time represents lees than two percent of the total time the valve would be full open with a postulated need to be isolated. This represents a negligible I change in mass or energy releases which are conservatively calculated in the analyses. Therefore, the increased valve closure stroke time is an insignificant change compared to the current analyzed i

plant configuration and represents no increased potential for, or adverse consequence of, any l event.

The gear set change does not affect any of the piping break evaluations made for the RCIC system. The gear set is a part of the Limitorque standard product line of interchangeable parts for the type SMB size 0 actuator installed on E5150F013, and is bounded by the actuator seismic and environmental qualification reports. The increased torque and thrust capacity that are a result of the gear set change are within the design capacities of the actuator and valve, and do not affect the parameters or the current piping seismic analysis. The extended stroke time of the E5150F013 inject;on isolation valve does not inhibit the RCIC pump from accomplishing the licensed required I makeup injection within 50 seconds. The additional energy draw on the associated de power supply battery is insignificant. The gear set change also does not affect the leakage performance of E5150F013 as required by Technical Specifications Limiting Conditions of Operation 3/4.6.1.1 I and 3/4.4.3.2. This is contmiled by the torque switch adjustment, and the associated MOV testing performed under the GL89-10 response program for MOVs, the plant's 10CFR50 Appendix J Primary Reactor Containment Leakage Testing Pmgram, or the ASME Section XI inservice testing I for pumps and valves program. The setting of the torque switch is not altered by a requirement of this modification.

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Safety Evaluation No: 94-0112 UFSAR Revision No. 7 Reference Document: EDP-27136 Section(s) NA Table (s) 8.3 6 I

Figure Change Yes X No I

Title of Change: Replacement of the Motor for Valve T4803F602 in the Nitrogen inerting System

SUMMARY

This modification replaced the existing motor (40 ft-lb, 5.3 hp, 4.8 kW) for nitrogen inerting system I valve T4803F602 with a similar, but not like-for-like, new motor (40 ft-Ib, 5.2 hp, 5.1 kW). This change in the motor introduced additional load on EDG-11. UFSAR Table 8.3-6, " Emergency Diesel Generator System - Loss of Off Site Power and Loss of Coolant Accident at Zero to Ten I Minutes", has been revised to incorporate this change. The new motor was evaluated for feeder size adequacy, starter size, fuse size, maximum allowable thrust, stroke time impact, additional weight impact on the piping and the system, and the valve function. The new motor was found I acceptable as a result of this evaluation.

With this modification, the new calculated load for EDG-11 is found to be within the short time rating I of EDG in accordance with Paragraph C.2 of Regulatory Guide 1.9, Revision 2 and item 3.7.2 of IEEE Standard 387-1977. The larger size motor ensures the reliable valve and system funcbon, and does not reduce the margin of safety as defined in the basis of the Technical Specifications, I UFSAR, or SER.

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Safety Evaluation No: 94-0113 UFSAR Revision No. 7 Reference Document: EDP-27161 Section(s) NA Table (s) NA Figure Change X Yes No Title of Change: Relocation of Filter Demineralizer influent Strainer in RWCU (G33) System

SUMMARY

Tais modification relocated filter demineralizer influent strainer G3305D005 from influent line to the filter demineralizer bypass line upstream of RWCU bypass valve G3352F044. The RWCU system normally operates with the filter demineralizer bypass valve closed. Thus, in its new location, the strainer will be utilized during normal operation. This will reduce the strainer high differential I pressure problems, which used to exist with the strainer in the influent line. In its new location, the strainer will reduce maintenance frequency and thereby reduce radiation exposure. Moreover, the strainer basket can be cleaned by flushing / draining during a scheduled plant shutdown, which will also result in reduction in radiation exposure. The designed intent for this strainer is not affected by I this modification as it will still meet the intent of preventing back flow of the resins into the reactor pressure vessel.

This modification has no impact on the UFSAR accident analysis since the operation of the RWCU system is not changed by relocating the strainer. The new location of the strainer has been reviewed and a new support added to meet Seismic M requirements. The relocation of the strainer I has no impact on the RWCU System line break analysis. The RWCU System is not required for safe shutdown of the plant and, in general, is a non-Q system outside the boundary for primary containment. The strainer relocation does not impact the primary containment function or degrade reactor water chemistry limits described in the Technical Specifications, UFSAR, or SER.

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Safety Evaluation No: 94-0116 UFSAR Revision No. 7 Reference Document: EDP-27173 Section(s) NA Table (s) NA I Figure Change [XJ Yes l lNo Title of Change: Replacement of RPV Head Vent Manual Bypasc Valves

SUMMARY

[ This modification installed replacement manual valves for the two RPV head vent manual bypass valves (B2100F001 and B2100F002). These valves were found leaking steam during Cycle 4 operation and had eroded beyond repair. The functions of the valves are to provide a

[ path for venting the gas / vapor bubble from the top of the vessel during a refueling outage flood up prior to reactor head removal and for the post refueling vessel hydrostatic leakage surveillance. The RPV head vent bypass valves B2100F001 and B2100F002 were replaced I with 2" stainless steel manual globe stop valves V8-3781 and V8-3782. The replacement i

' valves are 1500# ANSI Rated, ASME lil Class I valves. The stainless steel valves selected were used because they were the only ones available on short notice as spare valves of

{ suitable configuration and with Code documentation for this application. The modification also installed a pipe tap for hydrostatic testing of the modification. A welded pipe cap was installed at the completion of hydrostatic testing of this modification.

I L The RPV head vent valves are normally closed and are locally operated when the reactor is shutdown. The operational use and design functions of these valves are not being changed r by this modification. These valves only interface with the RPV pressure boundary and do not L support, or rely upon, other safety-related systems. The replacement valves are the same size, configuration, manufacturer, and weight of the original valves. Both valves are 1500#

r ANSI rated, ASME lil Class i valves with a similar type of seating arrangement. The only new y potential failure mechanism is the combination of materials with different expansion coefficients. This has been accounted for in the stress report which satisfies the ASME code r requirements. Any postulated line rupture is bounded by UFSAR sec+ ion 15.6.5 (LOCA inside e containment).

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l Safety Evaluation No: 94-0117 UFSAR Revision No. 7 Reference Document: TSR-26816 Section(s) 12.1 Table (s) NA Figure Change R Yes X No 1

E Title of Change: Replacement of Concrete Block with Steel Rolling Door in the Turbine Building

SUMMARY

This modification permanently replaced an 8" hollow / concrete block with a steel rolling door on the north wall on the first floor of the turbine building during the RFO4 outage. The concrete block was temporarily replaced with a steel rolling door for ease of movement of equipment and material at the north end of the turbine building (Temporary Modification 94-0012 and Safety Evaluation 94-0030-Rev 1). The access opening in the north wall is 10' wide x 12' high. The 18" I thick removable concrete shield planks placed in front of the door opening is 11'-8" high. This leaves a slot of 4" high x 10' wide,11'-8" above the first floor which does not have 18" of concrete shielding. The rolling steel door is designed for a wind load of 20 pounds per square I foot (88 mph wind speed). The turbine building does not house any safety-related equipment.

The replacement of 8" hollow concrete block with a rolling steel door does not adversely affect any of the missile barriers designed to protect safety-related equipment.

Replacement of the 8" hollow concrete block with a rolling steel door provides convenient access to the north end of the turbine building during major outages. During operation conditions, the 18" thick concrete plank will still provide adequate shielding to plant personnel in the northwest area of the turbine building from skyshine under post LOCA conditions. Plant Engineering had eariier assessed the dose rates in this area of the plant with the shield planks entirely removed i during a post LOCA condition and determined that access to the PASS system is not adversely l affected. Use of the stairway in the northwest comer of the turbine building may be limited l during post LOCA conditions if the shield planks have been removed. The shield planks are installed for ALARA reasons during operating modes 1,2, and 3. The shield planks serve no purpose during operating modes 4 and 5. The permanent removal of one shield plank 11'-8" above the floor and the replacement of the 8" hollow block with a steel rolling door does not increase the probability or the consoquences of an accident to equipment important to safety previously evaluated in the UFSAR.

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Safety Evaluation No: 94-0119 UFSAR Revision No. 7 Reference Document: EDP-26942 Section(s) NA Table (s) NA I Figure Change X Yes No Title of Change: Replacement of Degraded Flow indicating Switch in Core Spray System (E21)

SUMMARY

I This modification replaces the degraded flow indicating switch, E21N0068 with a Barton 581A-2 flow switch. The old Core Spry System (CSS) *B" flow indicating switch E21N006B was an ITT Barton model 289A mounted on H21P019 and qualified for forty (40) years. However, it required replacement because it could not be calibrated. The old ITT Barton indicating differential pressure I switch needed upgrading to NUREG 0588 Category I per the EQ Upgrade Manual. Regulatory Guide 1.89 Section C.6.E allowed the use of a Category 11 item until a qualife' d upgrade could be obtained and an outage of sufficient duration was available for replacement. The replacement I switch for E21N0068 is an ITT Barton model 581A-2 flow switch qualified to NUREG 0588 Category I requirements and, therefore, meets the upgrade requirement of the EQ Upgrade Manual. The function of the CSS *B" switch, E21N006B, is to provide contacts to open the Core Spray Minimum Flow Valve for core spray flow 5775 gpm and close the valve for a flow >775 gpm.

The new switch has the sama wiring configuration and the same failure modes. It has the same flow setpoint as the old switch. It does not have a local flow indication like the old switch, but that indication is not required for system function. There is no increase in tN probability or consequences of an accident previously analyzed since the new switch is essentially an upgraded version of the old switch There is no reduction in the margin of safety provided by the switch as defined in the bases for any Technical Specification, UFSAR, or SER.

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PROCEDURES, TESTS, AND EXPERIMENTS I

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SUMMARY

I Safety Evaluation No: 90-0107 . UFSAR Revision No. 7

.I DER 89-1125 7.5 Refstence Document: Section(s)

Table (s) NA Figure Change Yes X No Title of Change: Clarification of Role of Dedicated Shutdown Panel for Remote Shutdown with Coincidental Loss of Offsite Power

SUMMARY

This safety evaluation provides UFSAR clarification of the role of the Dedicated Shutdown Panel for I shutdown outside the control room with tne coincidental loss of offsite power by refening to the Dedicated Shutdown Panelinstead of the Remote Shutdown Panelin UFSAR Section 7.5.1.5.3.b.

UFSAR Sechon 7.5.1.5.3.b previously stated that the Remote Shutdown Panel would be used in 3 the event that a remote shutdown was required and offsite power was lost. However, tnree process 1 displays or loops - RPV pressure, drywell pressure, and torus temperature - are ultimately supplied by the offsite grid via SS64 from the 120KV Fermi 1 mat. The 120 VAC power for these displays or I loops is not restored by the diesel generator load sequencer upon recovery from a loss of offsite power. The Dedicated Shutdown Panel is the appropriate panel to provide for remote shutdown coincident with a loss of offsite power. The Dedicated Shutdown Panel and associated procedures have as their design basis, the capability of starting CTG 11 and restoring balance of plant power.

'I The Dedicated Shutdown Panel procedures minus the fire scenarios are equivalent to the Remote Shutdown Panel procedures in achieving shutdown.

This change does not make any physical changes to the plant. The change does not degrade safety systems or affect any postulated accident scenario.

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SUMMARY

I Safety Evaluatix No: 92-0035, Rev 1 UFSAR Revision No. NA Reference D.;cument: DCR 92-1071 Section(s) NA Table (s) NA 1

I Figure Change Yes X No Title of Change: Using the Fire Protection System as Temporary Cooling for the Circulating Water (CW) Pumps

SUMMARY

This evaluation justifies revising plant procedure 23.101, " Circulating Water System" to use the fire protection (FP) system as temporary cooling for the CW pump and motor bearings when the general service water (GSW) system is not available. This is accomplished by running fire hoses from the number 4 fire hydrant to the cooling water supply lines for each CW pump. Thic will allow l l

the CW pumps to continue operation and avoid a plant shutdown. Temporary cooling is required when the GSW system is chemically shock treated to kill any living organisms and zebra mussels in the GSW piping. This avoids the possibility of plugging the CW pump cooling lines due to the migration of freshly killed organisms, mussels, and shells.

This evaluation is contingent on the following administrative controls:

1. The hoses are inspected siong the entire length for leaks once per operating shift.

I 2. The area between the CW pumphouse and the number 4 fire hydrant is closed to vehicular traffic to prevent rupturing a hose by a vehicle driving over it.

l The FP systern is periodically treated to kill new shellfish larvae while the temporary I 3.

connection is installed.

4. The FP system is flushed and shock treated before or after the temporary connection is I removed.

This change enhances plant reliability by providing an allemate method of cooling the CW pumps.

I The temporary connection between the FP system and the CW system does not prevent these systems from performing their design functions. These systems are not safety related and their loss does not adversely affect the plant's ability to achieve and maintain safe shutdown conditions.

The level of fire protection provided for systems addressed in the Technical Specifications is not altered by this temporary connection.

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1 SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 3 SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 92 0080 Rev 1 UFSAR Revision No. 7 Reference Document: NPP-RC3 05 2.4,12.1

(. Section(s)

Table (s) NA r

L r Figure Change Yes X No Title of Change: Storage of Contaminated Equipment in Building No. 42

SUMMARY

This evaluation justifies storing radioactive contaminated equipment in plant yard building No. 42.

This building is located on the north side of the Radwaste Building and across the access roadway.

The stored items include electrical and mechanical tools, extension cords, wooden uibbirig from the turbine deck, and scaffolding. NPP-RC3-05, " Control and Storage of Radioactive Matenal in Building 42", implements the appropnate administrative controls for this storage. The maximum

} activity for the stored materials is 2 Curies. To minimize fire hazards, the combustible matenal is

) limited to that which is incidental to the equipment being stored (i.e., lubncating or hydraulic fluids contained in the equipment). This matenal is stored in strong, tight Wpgas which are suitable for i shipment. The lids are securely fastened with closing mechanisms. This assure the containment of radioactive material under normal storage conditions and provides additional fire pduidkii.

Radiological cMuis will be implemented through existing procedures to maintain radiation exposure ALARA. Using Building No. 42 as a storage area for contaminated equipment removes the demand for storage space in the OSSF. Using Building No. 42 as a contaminated equipment storage area also improves plant housekeeping and provides a safe working environment by reducog the number of storage areas located throughout the plant. It also reduces the amount of contaminated trash generated when useful material is discarded for lack of suitable storage space The use of Building No. 42 as a contaminated equipment storage area required a revision to the UFSAR. UFSAR Subsection 12.1.2.2.6 has been revised to state that plant areas and yard buildings may have radiation levels greater than 0.5 mr/hr providing suitable radiological controls f are in place UFSAR Subsection 2.4.13.3 has also been revised to address the safe storage of contaminated materials in Building No. 42.

The storage of radioactive materials in Building No. 42 does not interfere with any equipment important to safety because no such equipment is located in or near Building No. 42. The UFSAR accident scenarios involving the release of radioactive materials from the reactor and turbine buildings are not impacted by the storage of radioactive material in a separate building.

The offsite doses would still be within ODCM limits. The proposed storage configuration is I enveloped by the current UFSAR missile evaluation because each package is within the class of l potential missiles evaluated that could be generated by a tomado and the Category I reactor / auxiliary buildings are designed for tomado missiles. A fire hazard evaluation indicates that a fire in Building No. 42 will not affect control center habitability nor impact the safe shutdown capability of the plant. The potential liquid release resulting from the probable maxrnum meteorological event site flood is bounded by the 10 Curie source term limit in Technical Specification 3.11.1.4. A seismic event evaluation indicates thatif the packages are tipped over

SAFETY EVALUATIONS I PROCEDURES, TESTS, AND EXPERIMENTS PAGE 4 Safety Evaluation 92-0080, Rev 1 (Continued) during a seismic event the package dosura mechanisms limit the opening and spilling of the contents. The maximum organ dose, assaming a 100% material release, would be bounded by the I dose limits set forth in the Fermi 2 Safety. Analysis Report NUREG-0798.

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Safety Evaluation No: 92 0094 UFSAR Revision No. NA

( Reference Document: RAP EM104 Section(s) ' NA Table (s) NA Figure Change Yes X No Title of Change: Radological Emergency Response Plan (RERP) Exercee Cycle Change

SUMMARY

This evaluation justfies changng the RERP exerose c)* length from five years to six years.

Prevously, RAP-EM1-04 required that all major elements of the RERP plan be demonstrated dunng evaluated exercises in a five-year cycle. However, the Federal Emergency Management Agency (FEMA) requires the State of Michigan and Monroe and Wayne Counbes to demonstrale the mapor elements of their emergency plans every six years. Many aspects of the onsite and offsee pians are common and depend on the exercise scenario to demonstrate if the onsee and offste exerase cycles were allowed to conbnue on different cycle lengths, it would become increasingly difficult to ensure that al elements of both plans could be demonstrated by a common exerase scenano. Changing the onsste cycle to six years ensures that the major elements of the onsde and offste emergency plans can be demonstrMed by a common exercise scenano.

This change contnues to ensure that the RERP organization and au plan elements are tested penodically dunng evaluated exercises. It does not decrease the effechveness or readness of the Fermi 2 emergency response civ r.L.asi. This change does not impact operatng or response procedures nor does it impact plant design, configuraton, or funcbon Changing the exerase cyde from five to six years does not involve any parameter assoaaled with the bases of the UFSAR, Techrwcal Specificabons, or the Fermi 2 Safety Evaluabon Report NUREG 0798. Although NUREG 0654, "Critona for Preparaton and Evaluabon of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", Seebon N requires that the maior elements of the emergency plan be demonstrated within a five-year penod, Regulatory Guide 1.101,

" Emergency Plannng and Preparedness for Nuclear Power Reactors", Revison 3 allows licensees to develop methods and solutons different from those set out in NUREG 0654. Detroit Edson has not formally committed to NUREG 0654 for its emergency plan.

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PROCEDURES, TESTS, AND EXPERIMENTS PAGE 6 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 92-0095 UFSAR Revision No. 7 Reference Document: LCR 92-151-UFS Section(s) 3.7;3.10;8.3 Table (s) NA l

Figure Change Yes X No l

Title of Change: UFSAR Cable Tray Loading Description Revision

SUMMARY

This evaluation justifies revising the UFSAR description for cable tray loading to clarify the I acceptability of cable tray loading with respect to the evaluated plant cable tray load database. The cable trays were originally designed for 40 psf loading with the exception of the relay room area cable trays that were designed for 50 psf loading. After the initial design, the cable tray design l loads were adjusted to reflect the actual loads. The actual loads include the as-built cable, conduit, fire wrap, and airdrop loads. The cable trays and their supports were requalified to address the actual loads, Design specifications require cable trays to be designed for a uniformly distributed I load of 50 psf, a concentrated load of 250 lbs acting concurrently anywhere on the span, a maximum span of 8 feet, and a safety factor of two. A review of the cable trays whose total dead weight load exceeded 50 psf indicates that the excess loading is due to air drop loads. Air drop i

loads result from the weight of vertical cables between the tray and panel located below. The

" calculated weight is added to the tray supports and the tray supports are qualified based on this final load. The impact of air drop loads on cable tray qualification has been evaluated and the

" airdrop loads are acceptable with a safety margin greater than two.

The cable trays and supports are seismically qualified for the actual loaG,rg. In cases where the actual loading exceeds the original design loads, the actual loads have been evaluated through approved design calculations and still meet the safety factor described in the UFSAR. The cable trays and their supports will function as designed during normal and accident conditions and will not J adversely affect other plant equipment.

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i SAFETY EVALUATIONS l PROCEDURES, TESTS, AND EXPERIMENTS l PAGE 7 l SAFETY EVALUATION

SUMMARY

?

Safety Evaluation No: 93-0002 UFSAR Revision No. 7 Reference Document: LCR 93004 UFS Section(s) 7.8 i

f Table (s) NA  !

Figure Change Yes X No j Title of Change: Change in Available Techncal Support Center (TSC) HVAC Cooing Capacty .

SUMMARY

Dunng the evaluaton of NRC Informabon Notice 92-32, " Problems identified with Emergency ,

Ventilaton Systems for Near-Site (within 10 Miles) Emergency Operatons Facilities and Technical  !

Support Centers" it was determined that the required cooling capacdy for the TSC is greater than the available cooling capacity. The actual cooling capacity is 15 tons as opposed to the 20 ton  !

W1y stated in the UFSAR. A Review of the Radiologeal Emergency Response Plan (RERP),

RERP implementing procedures, and the NRC incdent response manual also indicated that the number of TSC personnel could be as many as 38 people as opposed to the 25 TSC personnel stated in the UFSAR. As a result of this evaluation, the following UFSAR changes were made- ,

1. UFSAR Subsecton 7.8.2.4 - The number of personnel accomriiodAd by the TSC has been .

deleted and replaced with a reference to the RERP plan.

i 2. UFSAR Subsecton 7.8.2.5 - The actual available cooling capacdy has been changed from 20

]

tons to 15 tons.  ;

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Removal of the TSC staffing figure is acceptable because the RERP plan and its implementng >

procedures specify TSC staffing requirements in detail. This change does not affect the number of personnel actually assigned or reportog to the TSC and there is no impact on Technical l SpecifeGors.

The change in cooling capacity results in a change in TSC maximum temperature from 75'F to ,

j 85'F drybulb temperature. The elevated temperature may impact certain personnel over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> )

shift. However, administrative actions are in place to ensure that adequate personnel are available.

Procedures EP 301-01 and EP 530 provde the necessary protective acbons for onste personnel L Additonally, all equipment required for continued TSC operation can withstand the maximum

! temperature increase. Sece the TSC will continue to cany out its design funcbon through the use of admmistrative controls, there is no reducbon in the margin of safety as desenbod in the UFSAR, Fermi 2 Safety Evaluation Report NUREG 0798, or the Technical Specirsuons.

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[ PROCEDURES, TESTS, AND EXPERIMENTS PAGE 8 b

SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93 0006, Rev 2 UFSAR Revision No. NA Reference Document: SOE 93-001 NA

{ Section(s)

Table (s) NA

[

Figure Change Yes X No

[-

Title of Change: Feedwater Flow Calibraton Test Evaluabon

SUMMARY

{-

This evaluabon justifies the feedwater flow calibraton test method and tracer chemical used and r evaluates the potential testing and corrective achon problems that may occur. The tests conset of L injectng a lithium twtrate tracer solution upstream of the feedwater flow elements; monitoring main steam line pressure, feedflow venturi differential pressure, and downstream tracer concentrabon, r and collechng process computer balance of plant data. The potential correchve actons include L onhne feedflow recahbraton. This test ubhzes independent feedwater flow measurements to assess feedwater flow loop accuracy and flow element performance, it allows the test data to be evaluated and any conective achons to be formulated it also provides baseline informaton on flow element t

differential pressure, main steam pr**'e+m, and other selected balance of plant parameters.

The proposed test has been analyzed with respect to line breaks. The test equipment is properly

[ rated for the test conditions and any test equipment failure would result in feedwater leakage that is bounded by the UFSAR feedwater line break analysis. The tracer chemmal lithium is not specifically addressed in the UFSAR but the lithium concentratens and the controls apphed to the

[

injechon method prevent feedwater chemistry limits from beng exceeded The effect of the tracer on reactuty is negligible. The effects of lithium achng as an fuel cladding zircaloy oxidizing agent were evaluated and the lithium concentrabons used in the tracer do not challenge the chemstr/

limits in EPRI NP-7077, "PWR Primary Water Chemistry Guidelines". The tracer does not impact H other emergency safety features equipment because it has no effect on control systems and is not injected in sufficient quanbbes to be measured at other plant sampling ponts such as the offgas r system. The design basis funcbon of the reactor water cleanup system, condensate polishng L demnerakzer system, feedflow instruments, the process computer, the rod worth mirumizer, rearculsbon pump limiter #1, and the feedwater control funcbons associated with this equipment r are not affected. The low lithium concentraton in the steam will not affect the main steam line radiaton rnonitors.

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SUMMARY

[ Safety Evaluation No: 93 0008 UFSAR Revision No. NA Reference Document: FlP RC102 Section(s). NA

{

Table (s) NA Figure Change Yes X No-Title of Change: Contamnated Storage in Warehouse B

SUMMARY

This evaluaton justdies storing 137 contaminated control rod drive mecharusms and 11 contamnated safety relief valves in Warehouse B which is located outside the protected area. This equpment was received from the Shoreham Nuclear Plant. Normally, contaminated equipment would be stored in the on-site storage facility (OSSF). Hcmever, this space in needed for the storage of radmactive waste awaiting shipment. Warehouse B is QA level B storage area whidi provides a fire detechon and suppression system; an environment where the temperature is mentained within moderate limits; and protechon from wind and precipitation. Estabbshing Warehouse B as an addibonal radioactive matenal storage area will assure the availability of this equpment to support plant operaticiis while saving radioactive waste storage space.

'f There is no is. pact on safety related systems, structures, or components because the above equement is stored outside the protected area. Storing the equipment in Warehouse B will not affect the radwaste building accident analyzed in the UFSAR. Analyses of the hypothebcal worst case liquid and airborne radioactivity releases ind'cates that the releases will not exceed 10 CFR 20 l Appendix B limits.

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[ Safety Evaluation No: 93 0009 UFSAR Revision No. 7 Reference Document: LCR 93 012 UFS Section(s) 3.9 Table (s) NA Figure Change Yes X No b Title of Change: The Use of Valve Design Seisme Acceleration Values in the Seisme Design Calculabons

SUMMARY

This evaluabon justifies using the valve design seismic acceleration values instead of the 5.0 g honzontal and 3.0 g verbcal accelerabon values stated in the UFSAR. Seismic Category I power actuated valves have been seismically qualified through a combinsbon of tests and analysis The actuator is quahfied by test and the valve pressure boundary and extended structure supporting the actuator are qualified by piping analysis. The valve design specmc.M initially specified that the valve seismic analysis be performed using the 5.0 g horizontal and 3.0 g vertical simultaneously apphed equivalent static accelerabons without any piping analysis input up front. As part of the motor operated valve (MOV) program, valve testing provides a direct indicabon of valve thrust ,

loads in many cases, the measured or predcted thrust load significantly exceeds the thrust value

l that was used in the original seismic qualificabon calculations. The higher thrust loads have been taken into account in subsequent supplemental seismic design calculabons made by the valve.

manufacturer. Whenever necessary, the valve seismic accelerabon values were reduced below the UFSAR values to iuuirir,cdate the higher thrust loads. However, the new design accelerabon p values are greater than or equal to piping stress analysis values and are, therefore, still acceptable Refirwng the valve structural analysis through the use of actual predicted acceleraben'and thrust values has no adverse affect on the valve's ability to perform its design funcbon Using the test thrust values in the valve seismic design calculabons reduces the probability of valve malfuncbon There is no reduction in the margin of safety because the change in valve seismic calculabon methodology does not affect the ASME Code stress limits specified in UFSAR Subsechon 3.9.

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PAGE 11 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93 0011 UFSAR Revision No. NA Reference Document: 35.000.233 Section(s) NA

[

Table (s) NA c Figure Change Yes X No L

Title of Change: Use of Freeze Seal to Block Pressurized Steel Pipes r

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SUMMARY

c This evalumbon justifies the use of freeze seal for temporarily blocking certain pipelines for purposes l of testing or maintenance. The plant maintenance procedure (35.000.233), "Installabon of Freeze Plugs", has been revised to include criteria as to when a 10 CFR 50.59 unique safety evaluabon is necessary.

Piping systems that interface with primary and secondary containment are not penrutted to have freeze seals when containment is operable. Installation and removal of freeze seals, when used, causes no permanent physical piping material property damage such as strength, toughness, or micro structure Further, the loss of the freeze plug is bounded by existing moderate or high energy pipe break analyses. Since the effects of flooding are bounded by existing analyses in the design calculabons, freeze seals not requiring a unique safety evaluation per entena given in the plant maintenance procedure, do not increase the probability or consequences of an accdont previously evaluated in the UFSAR.

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PAGE 12 l SAFETY EVALUATION

SUMMARY

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Safety Evaluation No: 93 0014 UFSAR Revision No. 7 {

-l Reference Document: LCR 93 063-UFS Section(s) 9.5 -

Table (s) NA i Figure Change Yes X No Title of Change: Hi-Com (Pubhc Address) System and Emergency Alarm System UFSAR ,

Desenpbon Changes  !

SUMMARY

i This evaluabon justifies revising the Hi-Com and emergency alarm systems as follows-l.

1. UFSAR Subsechon 9.5.2.2.2 has been revised to state that the Hi-Com system layout i pemwts commurwcabon between the main control room and buildings and areas of the i plant. It previously stated that the layout pemuts communicabon between the main control -

room and al buildings and areas of the plant. This subseebon was also revised to state that speaal arrangements for evacuabon notificabon have been provided for high noise areas .

where emergency evacuabon nvuishi is not broadcasted. I

2. UFSAR Subsecnon 9.5.2.2.4 has been revised to state that the emergency alarm system l is designmf to broadcast distinct signals to the plant. It previously stated that dishnet signals l were broade::,st throughout the plant.

These changes were made to ensure that the UFSAR reflects the actual plant configurabon for the ,

above systems and to define the acbon required to ensure that adequate emergency notificabon will be provxted to su personnel worlang within the protected area during an emergency. This revision is also the final acbon remaining for NRC InspecGon Open item (341/92007-01 (DRSS)).

This item addresses an inspector concem that a lack of communicabon in the non-intemJptable air supply control air compressor room is inconsistent with the system desenpbon in UFSAR' Subsechon 9.5.2.2.4.

The Hi-Com and emergency alarm systems do not directly affect the operation of.any other plant }

system This change did not involve any field modificabons. The existing acodont analyses are not impacted by a malfuncbon or failure of the Hi-Com or emergency alarm systems and no new acadent scenanos are introduced. The above UFSAR changes do not change or affect the refueling p;4Tvnii communicabon system required by refueling Technical SpecirwGens.

l SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 13 SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 93-0017 UFSAR Revision No. NA Reference Document: PCN-RO-F02 Section(s) NA l

Table (s) NA l

Figure Change Yes X No Title of Change: Reverse Osmosis (RO) Liquid Radwaste Process Evaluation L

SUMMARY

This safety evaluation justifies the operation of a pilot scale Pacific Nuclear RO unit to evaluate the

[ compatibility and the effectiveness of the RO process on Fermi 2 liquid radwaste streams. The RO unit processes 500 gallon waste batches from the chemical waste tank (CWT) und floor drain y collector tank (FDCT). It also processes general service water (GSW) collected from the L radiological control area (RCA) and decontamination activity waste water. The RO test unit consists of a feedwater transfer pump,20 micron rated feedwater cartridge filter, RO permeator booster pump, RO unit, process instrumentation, process valves, and skid-mounted control panel. Power is j supplied by a 480/240 VAC maintenance distribution cabinet and 120 VAC normal plant receptacle

^

circuits. The RO skid is located in the northwest comer of the radwaste building basement. Waste from the CWT and the FDCT is transferred to a portable batch tank in accordance with plant E procedures. The GSW water and decontamination waste water are transferred to the batch tank from storage containers using a suitable transfer pump. No chemical pre-treatment of the waste is required. The RO unit processes the waste at a nominal flowrate of 6 to 10 gpm. The process

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effluent or permeate is discharged to the floor drain collection system for normal processing via a

" floor drain and sump. The waste stream or retentate is directed back to the batch tank until the concentration is approximately 20% by weight. The fluid is then passed through an activated

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carbon bed to remove suspended solids and organic material and retumed to the batch tank. The

, batch tank is refilled as required for further processing. Process data and samples of the permeate and retentate are taken during the test runs and analyzed. The RO unit is not connected to the.

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liquid radwaste (LRWS) system during the test period.

The RO unit is not located within the vicinity of and does not interface with equipment important to safety. Administrative programs which control the sampling and disposition of waste water

- processed in the LRWS are not altered by the RO unit operation. There are no significant ALARA concems associated with the RO unit because of the low process stream activity and projected

- dose rates. The source term for the radwaste system accident evaluated in the UFSAR has not

> changed because no new activity is introduced. Any failure and resultant activity release is bounded by the existing accident scenario.

The operation of the RO unit has no effect on the Offsite Dose Calculation Manual (ODCM) or the Process Control Program (PCP). )

SAFETY EVALUATIONS

3 PROCEDURES, TESTS, AND EXPERIMENTS 3 PAGE 14 .

l SAFETY EVALUATION

SUMMARY

l Safety Evaluation No: 93-0024 UFSAR Revision No. 7 Reference Document: LCR 93-087-UFS Section(s) 9A.4 Table (s) NA Figure Change Yes X No Title of Change: Removal of Sign and Rope Barrier Requirements for Area Between Yard Manholes 16946 and 16947

SUMMARY

This evaluation justifies removing the UFSAR Subsection 9A.4.7.7 requirement for having the area I between yard manholes 16946 and 16947 roped off and posted to prevent combustible materials storage in close proximity to the manholes. This administrative control, in addition to the design features described below, was instituted to preclude the possibility of a single exposure fire affecting both safe shutdown divisions as required by 10 CFR 50, Appendix R. Yard manhole lI 16946 provides access to a cable duct run housing division I safe shutdown cables and yard manhole 16947 provides access to a cable run housing division ll safe shutdown cables. Both 3 manhole openings are covered by a tight fitting iron cover fitted with a cast iron ring. The manhole j covers are located 28 feet apart and a 1" to 2" layer of soil and gravel covers both manhole structures and openings. The administrative control is not required because:

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1. The soil and gravel will act as an insulator for the portion of the postulated fire's heat radiated to the ground, and

2. If a liquid fire is postulated, the soil and gravel will act as a flame arrester and lack of oxygen in the soil will not support the combustion process below grade. The tight fitting manhole cover will allow only a small amount of non-buming liquid to pass and pool on the floor of the manhole structure.

Eliminating the requirement for signs and rope barricades between these manholes eliminates i unnecessary inspection and maintenance of the signs and rope baniers.

This change does not affect any plant system, equipment, procedure, process, or program. The only acadent associated with the removal of this administrative control is a fire and the i I

consequences of a fire are discussed in UFSAR Subsection 9A.4. This change does not alter the level of fire protection provided for systems addressed in the Femii 2 Technical Specircations. The probability of a fire damaging the safe shutdown cables within these manholes is not increased despite the increased potential for storing combustible material in their vicinity because the soil and gravel cover above the tight fitting manhole covers is sufficient to prevent an exposure fire from propagating into the manholes.

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SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 15 SAFETY EVALUATION

SUMMARY

f Safety Evaluation No: 93-0030 UFSAR Revision No. NA

! Reference Document: COLR 4 REV1 Section(s) NA Table (s) NA l

Figure Change Yes X No Title of Change: Operating Limit Minimum Critical Power Ratio (OLMCPR) Mid-Cyde Exposure Change

SUMMARY

[ This evaluation justified changing the OLMCPR mid-cyde point exposure from 7000 MWD /ST to 6500 MWD /ST. Data accumulated early in the cyde 4 indicated that the achievable energy would

- be significantly less than originally expected. The core operating strategy was revised to extend the full power life of the core. This new strategy resulted in an exposure distribution outside of the

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bounds assumed in the original COLR 4 reload analysis and, therefore, required a revised reload analysis. The OLMCPR mid-cyde point exposure was changed because the revised analysis, COLR 4 Revision 1, assumed that the mid-cycle exposus would occur 22 full power days eadier than assumed in the original analysis.

This change did not require equipment modifications. The OLMCPR values did not change as a result of the revised operating strategy. There was no change to the methods of implementing the revised strategy or to plant operations as a result of the OLMCPR mid<yde point exposure

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revision. This revision applied more conservative limits earlier in the cyde and the safety limit minimum critical power ratio was unchanged from ine previous cyde.

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Safety Evaluation No: 93 0034 UFSAR Revision No. NA l Reference Document: SOE 93-02 Section(s) NA Table (s) NA l

Figure Change Yes X No Title of Change: Reactor Water Cleanup (RWCU) System Flow Test

SUMMARY

This evaluation justifies performing a RWCU system flow increase evolution to determine setpoint I

changes for the implementation of Engineering Design Package EDP 7291 (RWCU system flow increase). The flow rate will be increased by adjusting the flow rate on RWCU flow controllers FCP K174A and B. The RWCU recirculation pump flow will be increased from 133,000 lbs/hr (354 gpm)

" to 148,200 lbs/hr (395 gpm). The RWCU flow will be controlled as measured at the RWCU filter demineralizers which results in a total demineralizer flow increase from 270 gpm to 296 gpm. The flow rate increase will be in 5 gpm increments until the above specified flow rate is achieved.

i Reactor water temperature and pressure do not change due to this evolution. The RWCU system design pressure and normal operating temperature are not affected. The RWCU system flow rate increase will increase the N16 contribution to the RWCU component dose rates. However,

' calculations indicate that this should not result in significant or adverse increases in radiation levels in accessible areas. The radiological consequences of a RWCU system line break are enveloped F

by the main steam line break outside containment analysis in UFSAR Subsection 15.6.4. The increase in break flow rate will result in a temperature increase in the torus room and the reactor building. However, the torus mom temperature is bounded by a long term temperature profile which

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is limited by the equipment heat load rather than pipe break blow down. The reactor building

" temperature increase is insignificant for equipment qualification because calculations indicate that area temperatures are much higher for DBA-LOCA or other postulated high energy line breaks.

The torus room and steam tunnel equipment environmental qualificationc are bounded by the main

- steam line or feedwater line break analyses. The performance of this evolution does not affect the set point of any RWCU isolation actuation instrumentation and the increased flow will not affect the reactorisolation response time.

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SUMMARY

E 93 0039, Rev 1 7 Safety Evaluation No: UFSAR Revision No.

{ Reference Document: DC 5536 VOLi Section(s) 6.4; 7.1: 7.3; 9.4; 9.5 Table (s) 9.4-2, 9.5-2

. Figure Change Yes X No Title of Change: Relay Roorn, Cable Spreading Room, and Reactor Protechon System (RPS)

Motor-Generator (MG) Set Rooms Heatup Analysis

SUMMARY

Deviabon event report DER 92-0273 documents the investgation of the discovery of a rmqualified relay in the Cardox system for the standby gas treatment system (SGTS) which could prevent the SGTS system from performing its intended safety funcbon The DER evaluation determined that a single failure of a relay in the Halon system could cause a loss of cooling to the relay room, the cable spreading room, and RPS MG set rooms. DC-5536 Volume i provides a timenemperature profile for each room and its results indicate that sufficient time is available for operator acbon to open the smoke /Halon dampers and restore cooling to the relay room and the cable spreading

[ room. The RPS MG set rooms analysis was performed due to the potential for a plant scram However, this is a previously known phenomenon and it is not discussed further A single failure of

" the Halon system could cause the loss of ventilabon to the computer room but there is no adverse affect because the computer room has its own air conditioning system.

' During normal plant opersbon, high relay room temperature whii be annunciated on the process computer. The increased temperature in the cable spreading room will not exceed the maximum ambient temperature of the cables and the increased temperature will be sensed by personnel E entenng the room All necessary acbons to restore cooling to either the relay room or the cable spreading room are contained in plant procedures. Following an accident, smoke /Halm damper ,

possbon may not be available. However, an increase in control center temperature will be sensed  ;

, by control room personnel. Manual actions to restore cooling can be performed in the rentrol  !

I center. Manual acbons to switch control center heating ventilating and air condR;0r#,g (CC'iVAC) l system dmssons are described in UFSAR Subseebon 9.4.1. The loss of ventilabon will not prevent j the CCHVAC system from shifting to the recirculation or chlorine rnodes. Control room pressure will L not be lost because the makeup fan is not affected The duct pressures developed due to the closure of the smoke /Halon dampers are bounded by engineering calculations and duct bypass r leakage is not affected. Smoke /Halon damper closure does not change commitments to 10CFR50 q L Appendix A General Design Criterion 19, Regulatory Guide 1.52, or Regulatory Guide 1.95. 1 r

SAFETY EVALUATIONS

{ PROCEDURES, TESTS, AND EXPERIMENTS PAGE 18 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93 0055 UFSAR Revision No. 7 1.7,12.3

[ Reference Document: FMD RC1 LCR-93-158-UFS Section(s)

Table (s) NA 7

Figure Change Yes X No Title of Change: Revision to FMD RC1 (and UFSAR) to implement 10CFR20 Requirements A-

SUMMARY

This evaluation justifies revision to FMD RC1, " Radiological Protection" (and UFSAR) to implement the new requirements of 10 CFR 20. The pnmary changes are the use of annual rather than quarterly dose lirnits, the use of limits based on the " total effective dose equivalent" rather than separate external and intemal dose limits, explicit regulatory requirements for fetal pihi, explicit regulatory requiroments for maintaining radiation dose "As Low As Reasor. ably Achevable", 1 and revised terminology and administrative requirements to implement the 8tbsve. In Appendix B of 10 CFR 20, the term '1Vlaximum Permissible Conchiuai;cd (MPC) has been superseded by the term, "Denved Air Concentration" (DAC), with some of the values changing to reflect new

[ methodologes for the determination of intemal dose. The DAC and MPC are equivalent terms.

The changes to this FMD assure the effective implementation of the new 10 CFR 20 requirements.

[ They are mostly editorial and administrative in nature, and provide an equivalent degree of radation protecton There are no changes to structures, systems, or components as desenbod in the F UFSAR, there are no new hazards not described in the UFSAR, and there are no hardware,  ;

L funcbonal, or operational changes to any plant system as described in the UFSAR. l L

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SUMMARY

Safety Evaluation No: 93& T7 UFSAR Revision No. NA Reference Document: RERP Plan Section(s) NA Table (s) NA I Figure Change Yes X No Title of Change: Miscellaneous Changes to RERP Plan

SUMMARY

I This evaluation justifies changes made to the RERP Plan. The changes included the following:

I 1. Site Area Emergency and General Emergency classifications based on total effective dose equivalent (TEDE) and adult thyroid exposure. i

2. CHRRM indication of severe core damage based on new UFSAR source term data resulting from power uprate. )

3. Changed the indication for the steam line break outside primary containment without isolation condition to reflect Technical Specification allowable values vice setpoints. j

4. Value for radioiodine concentration indicating a 5 rem /hr thyroid dose rate is based on derived air concentration (DAC) for nonstochastic affects (2,000 DAC-hrs = 50 rem CDE thyrnid).

5. Dose assessment program described as providing output in units of TEDE and adult thyroid exposure.

l 6. Table J-1 based on guidance from EPA 400.

7. " Emergency Director is responsible for authorizing emergency personnel to exceed 10CFR 20 dose limits." Requirement for volunteers now only applies to extreme exposure in accordance with EPA 400.

8. Table K-1 based upon guidance from EPA.

The majority of the changes in the RERP plan were necessary to comply with 10CFR 50.47 and 10 CFR 50 Appendix E requirements and to implement EPA PAG manual, EPA 400 which

, superseded EPA 520. The changes in the RERP plan do not affect the configuration of any systems, structures, or components (SSCs), nor does it affect the operation of any SSC.

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 930079, Rev 1 UFSAR Revision No. 'NA' Reference Document: LER 93 015 Section(s) NA Table (s) NA ,

i Figure Change l l Yes lXlNo l Title of Change: Reactor Reorculation System Valve Confguration Change Dunng Shuldown

SUMMARY

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This evaluation justifies closing reactor recirculation system pump "B" suction valve  !

(B3105F0238) when the pump discharge valve (B3105F0318) did not close fully on a manual signal from the control room when lining up Division 2 of the Shutdown Cooling System Closing the B3105F023B valve is required to declare Division 2 of RHR system operable to perform LPCI and shutdown cooling functions as required under Technical Specification. The l LPCI design basis assume the plant to be at full operating power when a design basis loss of coolant accident occurs. The energy of the plant in cold shutdown is much lower than at full power, and it requires only a single division of RHR for shutdown cooling. With the B3105F0238 valve closed, RHR Division 2 can perform its LPCI function and core spray Divisions 1 and 2 can immediately inject water into the vessel since it is fully depressurized.

Closing B3105F023B valve does not in any way affect the probability of accidents postulated for Fermi 2 design (e.g., loss of coolant) or otherwise represent an accident precursor with the plant in cold shutdown condition. Closure of both the suction and discterge valves around pump *B" does not represent a threat to pump casing integrity since the seal purge was isolated during this configuration and because the seal staging leakoff will continue to flow under pressure for thermal relief, if required. Positioning B3105F023B velve closed does not affect the functioning of any other safety-related systems or instruments required in Operating Condition 4 or 5.

I SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS I PAGE 21 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0001 Rev 2 UFSAR Revision No. NA Reference Document: Work Request Section(s) NA

g 000Z934311

.5 Table (s) NA Figure Change Yes X No i

Title of Change: Radiological Aspects of Using Temporary Water Processing Vessels inside a i Diked Area Outside the CST /CRT Contaminated Area l

SUMMARY

This safety evaluation justified the processing of the contents of the CST or CRT by temporary processing equipment located adjacent to the CST /CRT diked area. The turbine generator failure event on December 25,1993, resulted in contamination of the condensate storage tank (CST), the

.I condenser hotwell, and in flooding of the turbine and radwaste building basements.

I The total activity contained in the CST (i.e., existing contents increased fivefold) was only 1.37 curies which is less than 10 curies allowed by Technical Specification 3.11.1.4. Offsite radiological accident analyses were performed to examine the offsite radiological consequences of an accidental release of the liquid in the processing vessels or transfer lines. The accident sequence, I very similar to that of the UFSAR section 15.7.3, involves spillage of the CST or CRT liquid and penetration into the ground, transit through the liquid aquifer into Lake Erie and thence to the Monroe Water Intake. Resultant calculated isotopic concentrations at Monroe Water inlet were I compared to the NRC's 10CFR 20 MPC values and summed over all isotopes (IC/MPC).

Separate calculations were performed, first using the most recent measured radionuclide analysis l

(in the CST) and other with radionuclide concentration 5 times higher than those measured for I added flexibility. In both calculations, the value of IC/MPC was found below the NRC criterion.

An accident analysis was run to examine the offsite airbome consequences should an airbome release occur. The doses from airbome inhalation at the site boundary were found to be insignificant when compared with 10 CFR 100 limits. l I

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PROCEDURES, TESTS, AND EXPERIMENTS I PAGE 22 SAFETY EVALUATION

SUMMARY

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Safety Evaluation No: 94-0002, Rev 3 UFSAR Revision No. NA Reference Document: Procedures Section(s) NA I NPF 65.000.607 NPP 23.107.02 and Temporary Table (s) NA l

I Modification 94-003 Figure Change Yes V No Title of Change: Temporary Waste Water Storage Locations

SUMMARY

This evaluation justified temporary waste water storages at various locations following the turbine generator failure on December 25,1993, when approximately 1 million gallons of water, including I turt>ine oil, were discharged into the turt>ine and radwaste buildings. The water and oil completely disabled the turbine building floor and equipment drain systems, along with both the solid and liquid radwaste systems. This event necessitated utilization of temporary waste water storage locations.

I Waste water was temporally stored in the following locations prior to processing: main condenser hotwell, up to seven (7) 21,000 gallons capacity fractionating (frac) tanks and oily waste processing i equipment in the southwest comer of first floor of turbine building, and turbine building basement.

The projected source term in the hotwell, turbine building basement and fractionating storage locations was calculated to be less than 7.0 curies which is less than the source term assumed in the UFSAR 15.7.3 for a Radwaste Building tank rupture release.

Offsite radiological accident analyses were performed to examine the offsite radiological consequences of an accidental release of the liquid contained in the hotwell, turbine building I

j basement, frac tanks, oily waste processing equipment, and transfer lines. The accident sequence, very similar to that of the UFSAR Section 15.7.3, postulated seismically induced cracks in the turbine building basement, rupture of storage locations, and spillage of the entire liquid volume was analyzed. Resultant calculated isotopic concentrations at Monroe Water intake were compared to NRC's 10 CFR 20 MPC values and summed over all isotopes (IC/MPC). Two calculations were performed first using the January 12,1994 radionuclide analysis in the radwaste basement water and second with radionuclide concentrations 10 timer higher than those measured on 1/12/94. In both calculations, the value of IC/MPC was found below the NRC criterion.

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{ PROCEDURES, TESTS, AND EXPERIMENTS PAGE 23 b SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 944003 UFSAR Revision No. NA Reference Document: LCR 93040-ODM Section(s) NA

{

Table (s) NA Figure Change Yes V No Title of Change: Use of Condensate Storage Tank (CST) for Liquid Waste Discharges

(

SUMMARY

This evaluation justified the use of condensate storage tank (CST) as an interim liquid waste

( storage and processing tank and discharge to the environment via circulating water decant line during plant shutdown conditions following the December 25,1993 turbine generator failure event.

Revisions to ODCM and related implementation procedures were made in keeping with Technical

[ Specification 6.14.2. These changes ensured that all requirements of Technical Spedfauon 6.8.5e addressing radioactive effluents were fully complied with this usage of CST.

ODCM changes did not reduce the level of radioactive effluent control required by 10 CFR 20.106

( or 10 CFR 20.1302,40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50. Further ODCM changes did not adversely impact the accuracy or reliability of effluent, dose or setpoint calculations.

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' PAGE 24 SAFETY EVALUATION

SUMMARY

E Safety Evaluation No: 944005 Rev 2 UFSAR Revision No. NA Reference Document: DM.OP-053 48309 Section(s) NA

{

Tatile(s) NA Figure Change Yes X No Title of Change: Installabon and Operabon of Temporary Water Pivcess;is Equpment for Condensate Storage and Condensate Retum Tanks

SUMMARY

[ This evaluabon justified installabon and operabon of a vendor supplied liquid processng equipment iviivwirs the December 25,1993 turt>ine generator failure event which precluded the use of the condensate or radwaste system to clean up the Condensate Storage Tank (CST) or Condensate

[ Retum Tank (CRT). The vendor supplied liquid processing equipment constructed to ASME Sechon Vill standard was ubhzed in a diked area west of the CST /CRT.

[ The CST provdes a source of water for HPCI, RCIC, SBFW, Core Spray, and CRD system. The suchon for core spray was transferred to suppression pool during the temporary use of vendor supphed equipment. The other users are not needed during plant shutdown condibon in mode 4 or

5. The CRD system was supplied water from either CST or CRT, wh'chever tank was not being processed.

A separate lined diked area was built to contain inadvertent process spillage or total contents of the equipment and hoses in case of vessel or line rupture Storm drains in the veinsty of this diked area were temporarily sealed to further mibgate radioactive material release to the envronment.

f Accdontal roiease of the contents of this equipment to the environment was also analyzed it was

' determned that release from this process would not violate any of Fermi 2 accident release limits.

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SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 25 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0009, Rev 1 UFSAR Revision No. NA Reference Document: NPP 65.000.607 Section(s) NA Table (s) NA  !

I Figure Change Yes X No Title of Change: Processing of Skimmed Oil in On-Site Storage Facility Using Marine Pollution Control Equipment

SUMMARY

This evaluation covered oil skimming / transfer from potentially contaminated Radwaste/ turbine i building locations to the On-site Storage Facility (OSSF) after the December 25,1993 turbine  !

generator failure event. The oil was processed in the OSSF by marine pollution control oil processing equipment for storage / disposal. This temporary condition required a reevaluation of the I OSSF building's capability to contain a significant source term and prevent adverse radiological consequences should an accident occur while the OSSF building was used as a storage location for potentially radioactive oil.

The gross activity level in the oil was estimated to be less than 4.0 curies which is less than 10 curies allowed by Technical Specification 3.11.1.4 for outside temporary storage tank activity. An isotopic analysis of the unlikely event of liquid transport path to the environment with resultant I release to Lake Erie and transport to potable water intake was performed. EC/MPC for potable water intake was calculated to be less than the 10 CFR 20 limit.

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PROCEDURES, TESTS, AND EXPERIMENTS l

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SUMMARY

Safety Evaluation No: 94-0017 UFSAR Revision No. NA se Reference Document: NA Section(s) NA Table (s) NA Figure Change l l Yes W No Title of Change: Justification of Plant Operation in Modes 4 or 5 with Recirculation System

Valves in Off-Normal Status

SUMMARY

This evaluation justifies plant operation in Modes 4 or 5 by closing valve B3105F023A when I valve B3105F031A is open or its status cannot be ascertained. The safety requirement for closing the recire discharge valve F031A arises from the postulated situation in which the discharge valve fails to close during a recirculation suction line break LOCA, LPCI water could i

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, be lost flowing through the discharge valve and recirculation pump into the downcomer region and then into the broken recirculation loop.

The limiting accident for Fermi 2 is recire line break and is not dependent on whether valve I B3105F031A is closed or B3105F023A is closed. The closure of valve B3105F023A under cold shutdown conditions has the same desired effect as that achieved by closing valve B3105F031A. The repositioning of valve B3105F023A does not affect any other safety-related I system or function. The valve does not have any control or power system interfaces with any other safety related systems. The A loop recirculation pump is protected from any potential I

damage by isolation of the CRD seal purge. The closure of valve B3105F023A in modes 4 and 5 is only required if valve B3105F031A cannot be closed or its status is not known. By closing valve B3105F023A, the shut down cooling and LPCI feature of RHR can be utilized, if necessary, and the plant can be maintained in accordance with the Technical Specification requirements. The closure of valve B3105F023A is required to maintain the reactor coolant pressure boundary integrity, and valve B3105F023A is equivalent to B3105F031A in this attribute.

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I SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS I PAGE 27 SAFETY EVALUATION

SUMMARY

Safety valuation No: 940018, Rev 1 UFSAR Revision No. NA Reference Document: SOP 23.718.01 Section(s) NA Table (s) NA Figure Change U Yes l X l No Title of Change: Attemative Processing of the Waste Surge Tank (WST) Contents

SUMMARY

This evaluation allowed for processing of the WST to the Condensate Retum Tank (CRT) using temporary processing equipment in order to provide a surge volume for incoming liquids I to the Liquid Radwaste System (LRWS). Treatment of the WST was accomplished by the use of temporary equipment consisting of hoses, fittings, valves, pumps and treatment vessels containing activated carbon for filtration and Total Organic Carbon (TOC) removal and/or mixed bed ion exchange bead resin. A suction was taken from the WST drain connection I (G1100F113) and retumed to the CRT via valve G1100F2003 installed under EDP 26303. A temporary pump provided the motive force to treat the contents of the WST and recycle it to

B the CRT. Two temporary treatment vessels containing activated carbon for filtration and TOC removai and/or mixed bed ion exenange dead resin were provided to treat tne water. Each iE was rated at 65 gpm,150 psig and 150'F. The filtration and ion exchange media used was capable of producing an effluent water quality which was acceptable for recycle to the CRT.

The temporary processing equipment was contained in the Radwaste Building and utilized the permanent plant recycle piping routed to the CRT. Any potential release of radioactivity to the ground water aquifer and ultimately to Lake Erie was bounded by the analysis. The liquid processed was compatible with the materials of construction employed in the temporary modification. Process temperatures and pressures were compatible with the pressure retaining components used in the attemate process path. The contents of the WST were retumed to the CRT, and Offsite Dose Calculation Manual (ODCM) Controls which pertain to offsite dose as a result of controlled discharges to the environment did not apply.

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SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 28 SAFETY EVALUATION

SUMMARY

I Safety Evaluation No: 94-0019 UFSAR Revision No. NA Reference Document: SOP 23.701.10 Section(s) NA Table (s) NA Figure Change Yes X No Title of Change: AttemaSve Method to Decant the Condensate Phase Separators (CPSs)

SUMMARY

This evaluation allowed for decanting of the CPSs to the Waste Clarifier Tank (WClarTK) using i I temporary processing equipment in order to provide a backwash volume for the Fuel Pool Cooling and Cleanup (FPCCU) filter demineralizers, Liquid Radwaste System (LRWS) precoat filters, or Condensate Polishing Demineralizers (CPDs). A temporary pump provided the motive force to process the supematant from CPS A or B to the WClarTk. This submersible pump was  !

suspended from the manway of the phase separator such that its suction is at or slightly above the normal decant level (5.0 ft of tank level). This minimized carryover of resin lines similar to the permanent plant piping design.

The temporary processing equipment is contained in the Radwaste building basement and utilizes the permanent plant piping supplying the WClarTk. There is no safety-related equipment located in these areas. Any failure of the temporary processing equipment and subsequent uncontrolled release was bounded by the analysis. Any potential release of radioactivity to the ground water l aquifer and ultimately Lake Erie was also bounded by the analysis. The liquid processed was I compatible with the materials of construction employed in the temporary modification. Process temperatures and pressures were compatible with the pressure retaining components used in the attemate process path. Since the contents of the CPSs are processed to the WClarTk, Offsite Dose Calculation Manual (ODCM) Controls were not applicable.

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i SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 29 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94 0037 UFSAR Revision No. NA Rearence Document: Procedure 23.205 Section(s) NA Table (s) NA Figure Change l l Yes l X l No Title of Change: Leak Test for Pressure isolation Valves E1100F050A&B

SUMMARY

This evaluation justifies the method used for leak testing of valves E1100F050A&B. As a result of Tech Spec Amendment No. 98, E1100F050A&B valves are not to be considered containment isolation valves and, therefore, do not require an Appendix J Type C test.

However, to protect the low pressure part of the RHR piping, these valves are still considered as pressure isolation valves (PlV), and therefore, require a PlV leak test with water as the test medium at 1045 psig. Prior to Tech Spec Amendment No. 98, the maximum allowable leakage was 1 gpm. Amendment 98 increases the PlV allowable leakage to 10 gpm.

I This test was performed during Operational Condition 5 (refueling) with vessel refueled.

During this test, the fuel movement was administrative!y prohibited, and irradiated fuel was

' adequately covered with water. The steps of closing required valves E1150F048 and F003 during this test assured that even an inadvertent start of RHR pump would not result in a water hammer. During this test, the piping, valves, and pump were within their design bases r conditions.

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SUMMARY

[ Safety Evaluation No: 94 0048 UFSAR Revision No. NA Reference Document: Work Requests Section(s) NA

{- C101930223 000Z942342 Table (s) NA

[ 000Z942351 000Z942352 Figure Change Yes X No Title of Change: Use of Freeze Seal

SUMMARY

This evaluation Jushfies the use of freeze seals on the CRD insert / withdraw (1/W) lines to isolate the

[ RPV dunng maintenance work. Three vent valves were cut out and replaced and the bonnets were removed on twenty inlet and outlet HCU isolation valves for wedge inspecbon per SIL-419. The freeze seals were located in the overhead above the HCUs on the first floor reactor building. This

[ maintenance work was done during the fourth refueling outage with the core fully off loaded The CRD pumps were isolated from the freeze seals and the freeze seals were subjected to an elevation head of approximately 90 feet (40 psig).

With the core fully off loaded and the CRD pumps isolated from the freeze seals, the 1/W lines are no longer high energy lines. No more than two freezes were done at one time. Loss of both freeze plugs is bounded by existing moderate energy or high energy line break analyses since the effects

{- of spraying and flooding are bounded by the analyses. There was a long length of pipe between the RPV and freeze locabon so break flow rate would be low (less than 40 gpm for the insert line or

[ 25 gpm for the withdrawline).

The piping of the freeze seal loceuon tends to exhibit less resistance to fracture initiation and l propageuon (due to the low temperature). Procedure 35.000.233 provides for pre-freeze NDE and L does not allow freezing over piping system discontinuities to reduce the chance of an unacceptable ,

I defect in the freeze region. Precautions against impact loading are also included. These measures p siy sificriitly reduce the possibility of bnttle fracture. Procedure 35.000.233 also included provisions j L intended to prevent loss of the freeze plug including temperature monitoring, cuchv;;ing heat epp;k,suon, adequate nitrogen supply and venfication that the freeze plug has formed in addition, 4 7

the freeze seal was conbnuously monitored and a conungency plan was prepared for isolabon in L the event of plug loss as part of procedure 35.000.233. Installation and removal of the freeze seal causes no permanent physical piping material property damage such as strength, toughness or micro structure. Therefore, freeze seals on the CRD 1/W lines per procedure 35.000.233 did not

[ increase the probability or consequences of an accident previously evaluated in the UFSAR.

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SAFETY EVALUATIONS I PROCEDURES, TESTS, AND EXPERIMENTS PAGE 31 SAFETY EVALUATION

SUMMARY

g Safety Evaluation No: 94-0076 UFSAR Revision No. NA Reference Document: Work Request Section(s) NA l I 000Z945687 Table (s) NA l

Figure Change Yes X No Title of Change: Installation of Freeze Seal on Instrument Une

SUMMARY

This evaluation justified installation of a freeze seal on the instrument line from reactor water level tap B21 LOO 8 to isolate the RPV and allow maintenance on excess flow check valve B21F508. The I freeze seal was located between drywell penetration X-54B and the excess flow check valve in the fuel pool cooling heat exchanger room on the 3rd floor of the reactor building. This work was done during the fourth refueling outage with the core fully off loaded. The instrument line is 1" schedule 160 stainless steel pipe. The freeze seal was subjected to an elevation head of approximately 40 I feet.

With the core off loaded, the level instrumentation line is no longer a high energy line. Loss of the I freeze plug is bounded by existing moderate energy or high energy pipe break analyses since the effects of spraying and flooding are bounded by the analyses in the engineering design l

calculations. The freeze seal installation procedure also includes provisions intended to prevent loss of the freeze plug including temperature monitoring, controlling heat application, adequate l nitrogen supply and verification that the freeze seal was continuously monitored and a contingency plan was prepared for isolation in the event of plug loss.

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SUMMARY

g l

Safety Evaluation No: 94-0079, Rev 1 UFSAR Revision No. 7 Reference Document: NPP-RC3-06 Section(s) 2.4 j

, Table (s) NA Figure Change l l Yes [ X_l No Title of Change: New Procedure for Control and Storage of Main Turbine Diaphragms

SUMMARY

I This evaluation justifies a new procedure, NPP-RC3-06, " Control and Storage of Radioactive Material in Turbine Diaphragm Enclosures." Following the December 25, 1993, turbine i

I generator failure event and root cause investigation, the 7th and 8th stage diaphragms wers removed from the three low pressure cylinders of the main turbine. Pending final disposition of these slightly contaminated diaphragms, it was decided to store these diaphragms, tools, and )

I equipmot in special enclosures on a cement pad located north of the radwaste building. The new ,.:rocedure implements the appropriate administrative controls for this storage. The procedure assumes a maximum total activity of 2 curies. Any removable contamination on 1

2 these diaphragms, tools, and equipment will be limited to 500 dpm/100 cm beta-gamma and I 2 20 dp J100 cm alpha, sgd6cd in 49CFR 173.443.

which is well within the limits for transportation in the public domain, as The proposed storage of radioactive material in these enclosures will not increase the probability of any of the accidents analyzed in the UFSAR. Although many of the UFSAR accident scenarios result in the off-site release of radioactive materials, those scenarios involving the Reactor ana Turbine Buildings will not be impacted by the storage of contaminated diaphragms, tools and equipment in a separate building. If it is assumed that the same seismic event simultaneously ruptures all of the radwaste tanks, as assumed in l UFSAR, Section 15.7.3 and also releases all of the radioactive material in these enclosures as an airbome source, this will still not increase the consequence of the UFSAR analysis, since the airbome and liquid scenarios impact different pathways over different periods of time.

$ Even considering the virtually impossible scenarios that both accidents impact the same

'+ person or location, the offsite doses would still be within NRC and/or ODCM limits.

4 I in addition, the diaphragms, tools and equipment are located in these enclosures, which are located on the northeast side of the Category l Reactor / Auxiliary Buildings. The UFSAR, Sections 3.3.2.3.7, "Tomado Failure of Nonseismic Structures", and 3.5 " Missile Protection".

I evaluate the site conditions, the missile types considered in design, and the structural strength of the Category I structures. Each enclosure is included in the type of potential missiles that could be generated by a tomado. The Category I buildings are designed for the impacts of I tomado missi!es. A spectrum of missiles was selected, approved by the NRC, and used as a design basis for these buildings (UFSAR Sections 3.3,3.5, and 3.8). The Category I buildings are designed to perform under harsh tomado and missile loads. The design loads already I

1 SAFETY EVALUATIONS

{ PROCEDURES, TESTS, AND EXPERIMENTS PAGE 33 Safety Evaluation 94-0079, Rev 1 (Continued) considered envelop the impact from potential missiles that may be generated from these enclosures or the tools and equipment. The considerable weight of the diaphragms, and the fact that they are flat, makes it very unlikely that they could be picked up by a tomado.

Also, the fire hazard from storing these diaphragms, tools and equipment in these enclosures was evaluated. The enclosures are located greater than 250 feet from the Auxiliary and Reactor Buildings and even further from other safety-related structures. A fra in these

[ enclosures will not affect Control Center habitability, since the Control Center would be isolated the same as if a chlorine release had occurred (UFSAR, Section 9.4.1.3). Hence a fire in these enclosures will not impact the safe shutdown capability of the plant.

The potential release of the stored radioactive contents during the probable maximum meteorological event (PMME) site flood was also examined. The PMME flood elevation is 586.9 feet (UFSAR, Section 2.4). This is above the cement pad elevation, but in any event, the resulting liquid release, if any, would be bounded by the 10 Ci source term limit of Technical Specification 3.11.1.4.

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SUMMARY

E Safety Evaluation No: 94-0000, Rev 1 UFSAR Revision No. 7 Reference Document: COLR 5 Section(s) 3.9,5.2,6.3,15.0,15.1, 15.2,15.3,15.4,15.6,15.7 Table (s)_ 15.0 01,15.0 02,15.0 03 Figure Change X Yes No Title of Change: Fuel Cycle 5 Core Operatng Limits Report (COLR)

SUMMARY

This evalumbon justmes changes to the COLR for fuel cyde 5. The operatng limits reported in the COLR have been rowsed to reflect the charactenstics of the new core. Specificaly, the mnmum critical power ratio (MCPR) for each bundle type has been updated based on noensing analysis.

Mammum planary knear heat generabon rate (MAPLHGR) limits have been added, and limits for bundles no longer in f5e core have been deleted. A linear heat generabon ratio (LHGR) limit for

[ fresh fuelhas been modod j Dunng the fourth refusing outage, one hundred aghtyeight (188) GE8 bundles, along with forty (40) GE11 and nnety-two (92) GE6 bundles, were replaced with two hundred twenty-eight (228) fresh GE11 fuel bundles and nnety-two (92) reinserted GE6 fuel bundles.

The rod block morator setpoints reman unchanged for cycle 5.

The cycle 5 UFSAR revision was made to update cycle specific operating limit mamum cribcal power ratio (OLMCPR) in Chapter 15. Also, other minor changes have been made in the UFSAR to prowde clanficebon and correct grammabcal errors

SAFETY EVALUATIONS

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SUMMARY

( Safety Evaluation No: 944090 UFSAR Revision No. NA

{ Refemnoe Document: EOP Plan Section 1 Section(s) NA Table (s) NA Figure Change Yes X No.

TIlle of Change: Revision to Plant Specific Technical Guidelines

SUMMARY

This evalumbon jushfied a revision to the Plant Specific Technical Gudelines (PSTG) to incorporate r

i BWR Owners Group Emergency Procedures Committee (EPC) approved items. PSTGs are a porton of Fermi's NRC requred Emergency Procedure Generaton Package. The revisions made indude the feili,v/cs:

L 1 PSTG step PC/H-1.2, second bullet should read: "If drywell oxygen concentrabon is not below 5%, initiate and maximize the drywell air purge flow or nitrogen purge flow." As implemented in

[ the EOP flowcharts, this change provdes another opton for purging the pnmary containment when hydrogen levels are below 1%. This sechon of the emergency procedures only allows vent and purge when within the radoactive release rates for normal day to day opershon.

2. PSTG steps TW/L-5.1, 5.2, 5.3, ovemdes (OR) before RC/L-2, or before conhngency 1 (Allemate Level Control) or before ceduiviicy 4, RPV flooding, or before conhngency 5, Levei/ Power Control, and or before conhngency 6, Primary Containment Flooding The revised  ;

L PSTG provides directons for these steps and ovenides, to terminate injechon into the pnmary H containment from sources extemal to the primary containment. As implemented in the EOP flowcharts, this change provdes better guidance to the operators without changing the intent of  ;

the EPG step. l The actons directed by this PSTG change will help to ensure containment integnty is maintamed and thus would not affect the radiological consequence of accderits evaluated in the UFSAR.

These changes to the PSTG do not involve changes to the facility nor changes in normal operahng procedures Because the EOPs have been based on the industry accepted standard (i.e., rev 4, EPGs) and that these changes have been reviewed and approved by the EPC, it 's concluded that the use of the EOPs will not increase the probability or consequences of an accident previously evaluated in the UFSAR. In additon, the use of the revised steps will minimize the consequences L of acodents which go beyond the design bases of the UFSAR.

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SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 36 SAFETY EVALUATION

SUMMARY

f h Safety Evaluation No: 94-0110 UFSAR Revision No. NA f Reference Document: OM-070-WS Section(s) NA OMM7-WS Table (s) NA Figure Change Yes X No Title of Change: Processing of Rainwaterin CST /CRT Diked Area

(

SUMMARY

This evaluation justified temporary installation of Vectra Technologies, Inc. equipment in the CST /CRT diked area to process rainwater with trace levels of radioactive material. The turbine

} generator failure event on December 25,1993, resulted in low levels of contamination on spots of

) the liner in the CST /CRT diked area. This resulted in approximately 20,000 gallons of rainwater which required to be processed to either remove the licensed radioactive material for release of water or to make the water suitable for use in the plant. Vectra Technologies, Inc. equipment used a reverse osmosis and evaporator system for processing the rainwater.

}

The temporary processing of liquid in CST /CRT diked area did not increase the consequences or

} the probability of causing any of the accidents analyzed in the UFSAR. The processing of water did not interfere, either directly or indirectly with any equipment important to safety. The temporary operation did not reduce any explicit or implied margin of safety as described in the UFSAR or SER.

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l SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 37 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94 0115 UFSAR Revision No. NA Reference Document: Work Request Section(s) NA 000Z948171 Table (s) NA l

Figure Change l l Yes l X l No I

Title of Change: Installation of Freeze Seal on RPV Vent Line l

SUMMARY

This evaluation justified the installation of a freeze seal to isolate a section of the RPV vent line for hydrostatic testing. The freeze seal was located inside the drywell on the 2' NPS riser between the valve nest and the bulk head penetration. The work was done during RFO4 with the reactor in cold shutdown and depressurized. The affected piping is classified as Quality Group A, QA Level 1 and Seismic Category 1.

L With the reactor in cold shutdown and depressurized, a loss of head vent piping integrity will I not cause RPV drain down. Loss of the freeze plug or piping integrity will not cause significant blowdown or flooding due to the short section of piping that is pressurized. Dynamic effects are bounded by existing high energy pipe break analysis. The freeze seal installation procedure also includes provisions intended to prevent loss of the freeze plug including temperature monitoring, controlling heat application, adequate nitrogen supply and verification that the freeze seal was continuously monitored.

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SAFETY EVALUATIONS PROCEDURES, TESTS, AND EXPERIMENTS PAGE 38 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94 0120 UFSAR Revision No. NA Reference Document: 90.TRM.3.8.4.3-1 Section(s) NA Table (s) NA Figure Change Yes X No Title of Change: Correct Typographical Error in the Technical Requirements Manual (TRM)

SUMMARY

This evaluation corrects a typographical error in TRM procedure 90.TRM.3.8.4.3-1. E11-F026 should read E11-F026B. The "B" suffix was not included in the original issuance of Table 3.8.4.3-1 in the Technical SpecIaGons. This table was later transferred to the Technical Requirements Manual with prior approval of the NRC.

The correchon of a typo vi.phical error for E11-F026B is strictly administrative and does not affect plant opersbon The change does not reduce the margin of safety as defined in the basis of any Technical Specificabon, UFSAR, or SER section.

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FERMI 2 b

SAFETY EVALUATION

SUMMARY

REPORT TEMPORARY MODIFICATIONS

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SAFETY EVALUATIONS TEMPORARYMODIFICATIONS PAGE 1 6

SAFETY EVALUATION

SUMMARY

r Safety Evaluation No: 930027, Rev 1 UFSAR Revision No. NA

f. Reference Document: Temporary Section(s) NA Modification 93 0003 Table (s) NA L

p Figure Change Yes X No~

Title of Change: Temporary Installation of High Pressure Coolant injechon (HPCI) Pump Discharge Pressure Transmitter L

SUMMARY

This evaluabon jushfies the installabon of a temporary HPCI pump discharge pressure transmitter to I collect data for the evaluabon of HPCI piping support forces The transmitter and connochng tutung l were installed for the HPCI pump and valve test, 24.202.01, performed in May 1993. The transmitter was connected to the HPCI pump discharge vent line located in the first floor reactor buildng steam tunnel labynnth Test data was transmitted to a cart-mounted analyzer by wnng i

routed through the steam turenel water-tight door. The water-tight door was opened only for penods of 10-20 mnutes for data collechon The transmitter measured the discharge pipng pressure prior to HPCI pump startup, during pump startup, and during the pump trip at the end of the test. Both

, the transmitter and the connectng tubing were rated above the 1330 psig HPCI discharge piping design rabng.

This equipment was used for monitonng purposes only. The vent piping was inspected for leaks prior to placmg the modificabon into service. The containment inboard isolabon valve E4150F006 was closed and doenergized during the test to isolate the pressure transmitter from the reactor vessel This rendered the HPCI system inoperable However, injechon capability could be restored by isolahng the transmitter, reenergizing E4150F006, and opening it. The maximum leakage from the 3/4" vent line was estimated to be 350 gpm. This leakrate is bounded by the foodwater line break analysis in UFSAR Subsechon 15.6.6. The floor drains are capable of draining this amount of water and the leak could be eliminated by stopping the HPCI pump and isolatng the vent line.

The transmitter introduced an addebonal potential missile into the steam tunnel However, this potenbal missile would be constraned by the labynnth walls and would not affect any other equipment required for safe shutdown of the plant. The time penods when the steam tunnel door was open for data collecton are bounded by the eight hour open steam tunnel water-tight door analysis evaluated in Fermi 2 Safety Evaluabon 93 0006. The test cart was attended by test personnel dunng the test. While unattended, it was secured to a sturdy plant structure in such a manner that it could not become a missile. Therefore, no adjacent equipment required for safe l shutdown or to mibgate the consequences of an acodent were affected

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{- TEMPORARY MODIFICATIONS PAGE 2 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 934033 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modl5 cation 934004 Table (s) NA L

r Figure Change Yes @ No Title of Change: Standby Liquid Control (SLC) Pump B Discharge Relief Valve Removal for

/ Mantenance t

SUMMARY

This evaluabon justdies removng SLC pump B discharge relief valve C4100F0298 for maintenance while mantaning SLC pump A operable. The SLC pump B suction and discharge isolabon valves are closed; the relief valve is removed; and blank flanges are installed in the relief valve inlet and outlet pipng. Removal of the relief valve renders SLC pump B inoperable and places the unit in the

} Techrwcal Spec 4c.t;en 3.1.5 seven-day schon statement. The outlet flange restores the common J SLC pump suchon piping integrity and the inlet flange provides a cleanliness boundary.

l The blank flanges meet the original design requirements such that relief valve repairs can take

) place withm the 7-day schon statement penod of Techrwcal Speedicabon 3.1.5. The installabon of the outlet piping blank flange and closmg the SLC pump B suction and discharge valves restores and mantans SLC pump A operability throughout the schon statement penod. There is no increase in the likelihood of a failure of the redundant pump and no new acodent scenanos are created l

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[ TEMPORARYMODIFICATIONS  ;

PAGE 3 0- SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93 0069 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA ,

Modification l l

93 0009 Table (s) NA l

Figure Change Yes X No i l

Title of Change: Temporary Use of Generator to Run CTG-11 Unit 4  ;

SUMMARY

This evaluaton justified the use of a temporary 1.75 MW diesel generator to provxie necessary power for running CTG 11-4 when construebon activity was in progress on CTG 11 Unit 1. In the event constructen actmty on CTG-11 Unit 1 is not completed within seven (7) days, Technical Specificaton 3.7.11 allows to designate an altamative source of power to altamative shutdown bus.

[ The temporary standby diesel generator was provxied with over 600 gallons of fuel sufficient to run the standby generator at full load for more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> The 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> are enough to support CTG-11 Unit 4 operabon and actmbes ===~1=ted with Appendix R dedcated shutdown system or station blackout. A telephone system is provided at the control room, dedicated shutdown panel, and CTG-11 contml staten to enable Opersbons to communcate to ensure timely operabon of the temporary standby generator and CTG-11 Unit 4 and to energize the plant buses within the l

requred time to meet Appendix R and stabon blackout requirements. Addibonally, an operator was statened at the peaker unit dunng the temporary modification period to ensure prompt response as needed Plant Operabons procedures assooated with dedcated shutdown and staten blackout were revised to accommodate the change in configurabon when a standby generator was utilized and CTG-11 Unit 4 was used to power the system instead of CTG-11 Unit 1.

Design and opersbonal aspects of the temporary modificaten were evaluated for impact on the L exisbng Appendix R and stabon blackout requirements and were found to be within acceptable limits. Utilizabon of CTG-11 Unit 4 to support the dedcated shutdown system for Appendix R and r stabon blackout requirements is asbmated to delay startup and loading of the altamate power L source by less than 10 minutes. The total time required to start up and load CTG-11 Unit 4 is less than 27 minutes which is within 45 minutes before reactor core is uncovered as desenbod in c UFSAR 7.5.2.5.4. The additonal ten (10) minutes added to previously determined 38 minutes to L perform the station blackout procedure is within one (1) hour stipulated in UFSAR A.1.55 r

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SAFETY EVALUATIONS

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(. SAFETY EVALUATION

SUMMARY

[' Safety Evaluation No: 93 0078 UFSAR Revision No. NA Reference Document: Temporary NA

[ Modification Section(s) 93 0011 Table (s) NA

> Figure Change Yes l X l No I.

Title of Change: Temporarily Blocking Selected Turbine Building Heat and Smoke Vents

SUMMARY

I This evaluation justifies temporarily blocking closed eight (8) of twenty-six (26) turbine building ,

( heat and smoke vents following the December 25,1993 turbine failure event. The turbine '

building roof is provided with 13 roof vent openings, each with a double vent that opens by spring action. The vents open automatically by a fusible link that melts at a specified temperature or by pressure. Eight of the twenty-six roof vent doors opened from fusible link

( molting on December 25,1993. The temporary modification allowed blocking of the openings using 2 x 4 boards and wire until the fusible links were replaced. The b!ocking of the roof vents was essential to eliminate excessive intrusion of cold air into the turbine building which

( could have a detrimental effect on plant equipment.

The temporary modification did not increase the probability or the consequences of an accident previously evaluated in the UFSAR because the remaining eighteen roof vent doors were fully operable / functional. Additionally, since the plant was in shutdown condition with the turbine generator deenergized, a second fire directly below the secured roof vents was not I

postulated to occur in the short time interval until the secured roof vents were restored to their original design configuration. Even if a second fire were to occur, the remaining eighteen roof vent doors would function as designed and release the resultant smoke and heat as required.

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SUMMARY

' Safety Evaluation No: 93 0080 UFSAR Revision No. NA Reference Document: Temporary NA

( Modification Section(s) 93 0012 Table (s) NA L

r Figure Change Yes. X No Title of Change: Altemate Path of Clean Water into the Reactor Vessei - -

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SUMMARY

This evaluation justifies the installation of a temporary modification which was used following

! the December 25,1993, turt>ine event to supply an attemate path of clean domineralized water -

to the reactor pressure vessel via the control rod drives. Following the December 25,1993 event, normal water supply from condensate domineralizer or condensate storage tank was not feasible. A temporary hose connection was installed between P11F030A and C11F006 valves. Clean water from condensate retum tank using a condensate jockey pump or hatwell supply pump was supplied to reactor vessel using CRD cooling or charging lines (charging lines available until scram is reset). The purpose of the attemate path was to be able to supply  ;

a clean volume of water to the CRD mechanisms and also improve reactor water chemistry. l To ensure that potential back leakage from the vessel through CRD cooling and chTging headers is minimized, a check valve was installed at valve C11F006. The temporary hose I used was rated for a minimum of 125 psig and CRD pumps were isolated and tagged out ,

during the period of modification The temporary modification did not increase the probability or the consequence of an accident i prevk>usly evaluated in the UFSAR. The discharge pressure of the condensate jockey pump 1 (100 psig) is much less than the normal operating discharge pressure from the drive water l pumps (12001500 psig). If a break were to occur in the connecting hose, the effect would be minimal since,1) a check valve installed at valve C11F006,2) a manual isolation of the system by closing either or both P11F030A and C11F006 valves, and 3) the types of breaks postulated in UFSAR 3.6 indicated that a leak from a ruptured hose would be within the bounds of flood analysis for the reactor building.

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[ TEMPORARY MODIFICATIONS PAGE 6

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93 0082, Rev 2 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modirmation 93-0013 Table (s) NA _

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( Figure Change []Yes l X l No Title of Change: Temporary Installation of Reactor Water Letdown Path

SUMMARY

This evaluation justifies the installation of a temporary letdown path from the RWCU system to

) a demineralizer skid for cleaning and demineralization and retuming water to the condensate J retum tank via HPCI test retum line. This temporary modification was used with the plant shut down and depressurized during the forced outage following the turbine generator event on December 25,1993. The temporary modification was made to a nonsafety-related portion of the RWCU system.

No potential for fuel damage existed as a result of the conditions and controls on the temporary flow path since fuel cooling / coverage was maintained. Furthermore, the amount of time available for operator response (7 to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />) was long enough to minimize adverse consequences of any leak. The potential radioactivity release was bounded by steam line break release quantities as described in the UFSAR 15.6.4. Leaks in the temporary equipment in the turbine building were bounded by the flooding analysis for the offsite dose ccnsequences.

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SUMMARY

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Safety Evaluation No: 93 0084 UFSAR Revision No. ,NA Reference Document: Temporary Section(s) NA i Modification l I 93 0014 Table (s) NA I Figure Change l l Yes l X l No i

i Title of Change: Temporary Plugging of Selected Floor and Equipment Drains in the Turbine Building

SUMMARY

This evaluation justifies a temporary installation of plugs in selected floor and equipment drains located in the turbine building. The drains plugged included:

Floor Drains D028-55 through 57, D028-17,18,33 through 35, and 68 through 71 I Equipment Drains D027-10 through 12 i

I As a result of turbine generator failure event on December 25,1993, the radwaste building basement was flooded. The purpose of temporary plugging of selected drains was to isolate the turbine building floor and equipment drains which flow to sumps located in the radwaste 4 building basement in order to facilitate cleanup and recovery of both buildings and the l equipment contained therein. Transfer of water from the turbine building to the liquid radwaste system via the turbine building sumps was not affected by the temporary modification.

Installation of this temporary modification did not increase the probability or consequences of an accident previously evaluated in the UFSAR. Installation of drain plugs had no effect on I the ability of the drain systems to prevent flooding of the reactor building under the probable maximum meteorological event flood scenario described in UFSAR 3.4.4.4. Additionally, the modification did not impact the ability of the plant to cope with a rupture of the circulating water I system as described in UFSAR 10.4.5.3. The drains plugged were not required to support operation of any equipment important to safety. Since the plugs were only installed in Modes 4 and 5, the margin of safety for the standby feedwater pumps (required in Modes 1,2, and 3) remained unchanged.

SAFETY EVALUATIONS TEMPORARY MODIFICATIONS PAGE 8 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0085, Rev 3 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modification 93-0015 Table (s) NA Figure Change Yes X No I Title of Change: Installation of Temporary Attemaia to the Reactor Water Cleanup (RWCU)

System

SUMMARY

This evaluation justified the installation of a temporary modification to the RWCU system to allow processing of reactor waterwith an ion exchange / filtration system supplied by Chem-Nuclear. This

" temporary modification was made following the December 25,1993, turbine generator failure event when radwaste systems became inoperable due to flooding in the radwaste basement. The l lg temporary Chem-Nuclear equipment located on the refueling floor in the reactor building also I i3 provided for the transfer of spent resins directly to a dewatering unit that allowed the spent resins to be stored in the Onsite Storage Facility (OSSF). This temporary modification was used only during l cold shutdown or refueling operation conditions and with reactor pressure vessel at atmospheric I pressure. The dual train skid and demin equipment provided by Chem-Nuclear was designed and constructed under an NRC approved quality assurance program. The unit's piping and valves are l

of stainless steel and are designed and constructed to ANSI B31.1 standards. The demineralizer i pressure vessels are designed and constructed to ASME Section Vill standards. High integrity containers (HICs) used for resin storage are designed to meet the requirements of 10CFR 61, and the cask used for moving the HICs is designed to the criteria of 10CFR71. Additionally, all piping and equipment installed in this temporary modification were supported or secured to meet Seismic II/I criteria. While the RWCU system was operated with this temporary modification installed, the primary containment isolation valves and isolation actuation instruments for RPV water level 2, system differential flow high and demineralizer inlet temperature high were maintained operable.

The changes made to the RWCU system via this temporary modification affected the nonsafety-related portion of the systems, and it did not impact the operation of other safety-related systems.

The consequences of an accident involving a breach in the RWCU system are actually reduced I with the reactor in shutdown condition. Any radiological release and offsite radiological consequences from a break, spill or rupture in the temporary system is enveloped by the analysis in UFSAR 15.6.2, " Instrument Une Break." A break in the line or container carrying the resins did not I present a significant airbome problem since the radionuclides remain in the solution or with the resins. The combination of physical and administrative controls implemented for use of an uncertified cask for onsite movement of the spent resin from the RWCU temporary modification assured a level of safety equal to or better than the use of a certified cask on a public highway.

SAFETY EVALUATIONS

[ TEMPORARY MODIFICATIONS PAGE 9

( SAFETY EVALUATION

SUMMARY

b Safety Evaluation No: 940004 Rev 1 UFSAR Revision No. NA Reference Document: Temporary NA

(' Modifications Section(s) 94001 and Table (s) NA 94-007 Figure Change Yes X No Title of Change: Discharge of Contents in the Condensate Storage Tank Using a Temporanly Modified System

SUMMARY

The evaluation justified using a temporarily modified system for liquid radioactive discharge following the December 25,1993, turbine generator failure when the normal discharge path was not

( available due to radwaste basement flooding.

All discharges into temporarily modified system occurred in accordance with approved plant

[ procedures, the Offsite Dose Calculation Manual (which was revised prior to the discharge) and Fermi 2 National Pollutant Discharge Elimination System permit.

' Offsite dose to the public as a result of any failure (of temporary system) and su%=nt uncontrolled release was bounded by the accident analysis contained in UFSAR 15.7.3,

" Hypothetical Liquid and Solid Radwaste System Accident Analysis."

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Dunng the discharge of the CST contents, the volume of water in the CST was reduced below the level required for operation of Standby Feedwater, Core Spray, HPCl/RCIC systems. This was

[

acceptable as HPCI, RCIC and SBFW are not needed in plant shutdown modes 4 and 5. The core spray system was aligned to the suppression chamber in accordance with Technical Specirse. i 3/4.5.2,"ECCS Shutdown",

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SUMMARY

Safety Evaluation No: 94-0006 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modification

.I 94 002 Table (s) NA I Figure Change Yes X No Title of Change: Temporary Installation to Provide Additional Path for Supplying Clean Water From Torus to Reactor Vessel or From Torus to Condensate Retum Tank (CRT)

SUMMARY

I This evaluation justified a temporary installation to provide additional path for supplying water to reactor vessel or CRT. This temporary modification (i) allowed a higher input of clean water into the reactor pressure vessel to help in reestablishing proper reactor water chemistry in conjuncbon I with other temporary filtering and letdown systems and (ii) provided a method of retuming good torus water to the CRT so that higher than normal torus level can be controlled. This temporary modification was installed after the turbine generator failure event on December 25,1993.

l Installation of this temporary modification and its possible failure was analyzed for reactor pressure vessel draining, torus draining, radionuclide release, and impact on ECCS/ESF equipment due to flooding. It was concluded that reactor vessel water level will not be reduced below Tech Spec 2.1.4, Reactor Water Level Safety Limit. The imposed administrative controls on torus draining provided adequate assurance of continuity of function of the torus per Tech Spec 3.5.3. The analysis of possible radiological release quantities were found to be within UFSAR limits of I Radwaste Building release consequences and Tech Spec curie limits for outside tank storage with leakage collection in a location that will not result in a release in a seismic event. Impact to equipment important to safety was bounded by flooding analysis in UFSAR 3.6.

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SAFETY EVALUATION

SUMMARY

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Safety Evaluation No: 94-0008 UFSAR Revision No. NA l Reference Document: Temporary Section(s) NA I Modification 94-0003 Table (s) NA Figure Change Yes X No Title of Change: Installation of Floor Drain Plugs and Dikes in the Turbine Building

SUMMARY

u This evaluation justified the temporary installation of plugs in the floor drains D013-19,21, and 22 and two track drains located between the railroad tracks in the turbine building rail car bay (grid

- location J/K-1 and J/K-3/4). In addition, this modification installed 5 foot dikes at the turbine building southwest rollup door, southwest personnel access door, south access door to turbine lube oil tank

] room, and personnel access door to the Office Service Building. This temporary modification was used during water recovery plan after the December 25,1993 turbine generator failure event. The floor drain plugs and dikes were installed to ensure that any liquids released in the event of a

] fractionating (frac) tank (s) and/or oily waste processing system (both located in the turbine building first floor) failure are contained within the turbine building. In the unlikely event of failure of frac tank (s) and/or oil waste processing system, the discharges will be routed via floor drains and stairwells to the basement of the turbine building instead of release to the area outside the turbine building.

This temporary modification was utilized in the plant in mode 4 or 5. UFSAR Section 3.4.4.4

] "Intemal Flood Protection" and Section 10.4.5 3, " Safety Evaluation" (for hypothetical rupture of circulating water system piping) analyses were not invalidated due to existing condition of the plant. ]

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SUMMARY

Safety Evaluation No: 94-0012, Rev 2 UFSAR Revision No. NA Reference Document: Temporary Section(s) _NA Modification 94 0005 Table (s) NA Figure Change Yes X No l

Title of Change: Provide Temporary Flow Path to Support Cleanup of Control Rod Hydraulics, Control Rod Drives (CRD), Condensate Retum Tank (CRT), Torus Water, and

{ Main Condenser

SUMMARY

L This evaluation justified temporary installation of attemate flow paths to support the cleanup of the control rod hydraulics, control rod drives, condensate retum tank, torus water, and water resulting from floodup (static hydro) of the main condenser. Flow paths provided included: (1) Condensate Storage and Transfer System (P11-00) to CRD pump suction. Letdown from Reactor Water Cleanup (RWCU) for reactor vessel level control through temporary demineralizers to the Condensate Retum Tank / Condensate Storage Tank (CRT/ CST), (2) P11-00 system to CRT/ CST via temporary demineralizers, (3) Torus water to temporary demineralizers and retum to torus, main condenser or CRT via torus water management system, and (4) Water generated from floodup of the main condenser retumed to torus, CRT, or recirculated back to the main condenser. In support of this temporary modification, EDP-26281 and EDP-26790 installed 3-inch taps on torus water management system (TWMS) retum line, condensate storage supply to radwaste, TWMS supply to l condensate polishing demineralizer and reactor water cleanup letdown line. The temporary

' modification was utilized during operating condition 4 and 5 only with reactor shut down and depressurized following the turbine generator failure event on December 25,1993.

' The temporary modification diverted flow from a nonsafety-related portion of the RWCU system and other nonsafety-related systems. No potential for fuel damage existed as a result cf the r condition and administrative controls established on the flow paths created by this modification L since the fuel cooling and coverage (with water) was maintained. Leaks from the equipment in the turt>ine building were bounded by the flooding analysis for the building presented in the UFSAR r sechon 10.4.5.3. Also, Technical Specifications were maintained as they related to the fuel integrity.

An analysis of isotopic releases using the methodology applied to Radwaste Building tank rupture radioactive release showed the release limits were met.

SAFETY EVALUATIONS TEMPORARY MODIFICATIONS PAGE 13 SAFETY EVALUATION

SUMMARY

[ Safety Evaluation No: SE 944016, Rev 3 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modifications 94 4 009,94-0010 Table (s) NA Figure Change Yes [ X_j No Title of Change: Temporary Installation of Liquid Radioactive Waste Treatment Equipment

SUMMARY

This evaluation allowed installation of a temporary liquid radioactive waste treatment equipment to allow processing and disposition of sub-standard quality water encountered during plant restart that is undesirable to process through the permanent Liquid Radwaste System (LRWS). This arrangement of pumps, hose, fittings, and treatment vessels will treat water from other LRWS tanks forwarded to Waste Sample Tanks A, B, or C in order to recycle or discharge to the environment via the permanent plant liquid radwaste blowdown line. All discharges were in accordance with the Offsite Dose Calculation Manual (ODCM) and the Fermi 2 National Pollutant Discharge Elimination System (NPDES) Permit.

Offsite dose to the public as a result of any failure of the temporary processing equipment and subsequent uncontrolled release was bounded by the accident analysis. The entire processing system and discharge flow path was contained in the Radwaste Building and utilizes the permanent plant discharge piping routed to the Circulating Water Pump House.

Any potential release of radioactivity to the ground water aquifer and ultimately to Lake Erie was bounded by the analysis. The liquid processed was compatible with the materials of construction employed in the te.mporary modification. Process temperatures and pressures were compatible with the pressure retaining components used in the attemate process path.

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SUMMARY

Safety Evaluation No: 94-0027 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modification 94 0002 Table (s) NA Figure Change Yes X No Title of Change: Additional Path for Clean Water to the CRT/ CST or the Reactor Vessel from the Torus

SUMMARY

This evaluabon justified a temporary installation of an additional path for clean water into the reactor vessel from the torus or to the CRT/ CST from the torus. This water path allowed a higher input of clean water into the RPV to help in reestablishing proper reactor water chemistry in conjunction with other temporary filtering systems and letdown systems and provided a method of retuming good torus water to the CRT/ CST so that higher than normal torus level can be controlled.

The temporary modificabon was reviewed for potential leakage along its path and found not to impact safe shutdown equipment or equipment important to safety required for maintaining desired plant cordtions. Sufficent administrative controls existed under this temporary modificabon to preclude loss of torus funcbon as required by the Technica! Speafications. The administrabve L

controls and the extremely long, available operator response time (two days) prior to loss of required torus level prevents the torus draining. The temporary modification does not reduce

[ reactor water level to less than the required values for Reactor Water Level Safety Limit. The imposed administrative controls on torus draining provided adequate assurance of continuity of function of the torus. The analysis of possible radiological release quantities were found to be l

within the limits of Radwaste Building release consequences and Tech Spec curie limits for outside h tank storage with leakage collection in a locabon that will not result in a release in a seismic event.

Impact to equipment important to safety is bounded by the Flooding Analysis.

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SAFETY EVALUATIONS TEMPORARY MODIFICATIONS PAGE 15 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0030, Rev 1 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modification 940012 Table (s) NA Figure Change R Yes X No Title of Change: Insta!!adon of a Rolling Steel Door in the Turbine Building

SUMMARY

This evaluation justified temporary installation of a chain operated rolling steel door in the north wall of the turbine building for the 10' x 12' equipment opening that is normally blocked closed with 8" concrete blocks and 18" concrete shield planks. This provided a construction access for movement of equipment and material at the north end of the Turbine Building needed for retubing the north end of the Main Condenser.

The 18" thick concrete shielding and 8" concrete blockout does not protect any equipment important to safety, it is used to shield plant personnel, specifically operators ingresc/ egress to the main control room via the stairway on the northwest comer of the Turbine Building for skyshine into the Turbine Building under post LOCA conditions. Removal of the shield planks and concrete blockout has no effect on dose rates in the area of the door. The rolling steel door was designed to withstand a wind load of 20 pounds per square foot (88 mph wind loading). The door had a bottom bar weather-strip. This provided adequate seal to protect I against foul weather flooding. This temporary modification did not affect the probable maximum meteorological event (PMME) for site flood analysis.

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SAFETY EVALUATIONS TEMPORARY MODIFICATIONS PAGE 16 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 940077, Rev 1 UFSAR Revision No. NA Reference Document: Temporary Section(s) NA Modif' cation 94 0016 Table (s) NA Figure Change l l Yes lX lNo Title of Change: Temporary Installation of Test Instruments for Turbine Testing

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SUMMARY

This evaluation justified the installation of strain gauges, accelerometers, and pressure transducers on the turbine system for the purpose of obtaining physical data during the power ascension program following the replacement of HP control valve intemais. In the case of the pressure transmitters, a connection to the turbine flow path was made as contrasted with the accelerometers and strain gauges which were mounted extemal to the process pressure retaining envelope. Pressure integrity of the temporary installation was assured by the following actions:

The pressure transduceritself was proof tested to approximately three times the maximum exposed system pressure.

The fittings used to install the transducers relied on metal-to-metal seal and not straight threads with gasket or"O" rings.

Each transducer was installed with complete mechanical integrity. The very small piezoelectric pressure transducers were supported by the pipe fittings alone. The larger strain gauge pressure transducers were connected to the process with short lengths of tubing and were supported independent of the process connection.  ;

l Each temporary attachment to the pressure boundary was pressure tested as an assembly prior to installation and were checked for leaks during the initial plant pressurization.

This temporary modification did not modify or change the operational characteristics of the

( turbine or associated systems. Careful evaluation of the catastrophic failure of any of the Temp Mod transducers resulted in the conclusion that these failures would not create the possibility of an equipment malfunction of a different type than any previously evaluated in the UFSAR. The worst case failure would result in a half trip of the NS logic for the low steam line pressure isolation. This isolation is designed as an anticipatory trip which mitigates the loss of reactor coolant inventory (minimizes reactor depressurization rate) as a consequence of the turbine and bypass valves failing open. The use of anticipatory signals as scram orisolation system inputs are part of a very conservative defense in depth approach but are not required to mitigate the consequences of analyzed transients or accidents in the UFSAR.

SAFETY EVALUATIONS TEMPORARYMODIFICATIONS PAGE 17 SAFETY EVALUATION

SUMMARY

[ Safety Evaluation No: 94 0108 UFSAR Revision No. NA Reference Document: Temporary NA

{ Modification Section(s) 94 0014 Table (s) NA Figure Change Yes X No Title of Change: Temporary Installabon of Plugs in Floor and Equipment Drains in the Turtune Building

SUMMARY

This evalumbon justified temporary installaton of plugs in certain floor and equipment drans in the turtune building basement. This temporary modificaton was necessitated after the December 25, 1993, turtune generator falure event and flooding of the radwaste building basement. The purpose of this modificaton was to isolate those turtune building basement drains that flow duectly by gravity to sumps in the radwaste building This greatly reduced the potenbal for a loss of the radwaste hquid processing capability due to a significant water spillage in the turbine building This modificaton did not impact the turbine building basement drains that flow to the turtune building sumps which then transfer water to the liquid radwaste system by sump pumps. This modificaton allows plugging of the drains under all plant operating modes. l The Standby Feedwater pumps are located in the turbine building basement and would be lost if significant flooding of the basement were to occur. The Standby Feedwater System is not safety related and is not requred to support the safe shutdown of the reactor except for its use in the  !

aitemate shutdown system The altemate shutdown system is used to respond to a fire in the control center complex and selected auxiliary building fire zones. Per UFSAR secton 9.A.5.a.4, a i fire need not be considered concurrent with other plant ::ccioents or severe natural phenomena Therefore, a turtune building flooding accdont need not be considered coincident with a fire requinng the use of the Standby Feedwater System. Flood scenarios desenbed in the UFSAR that result in flooding of the turbine building are described in UFSAR sectons 10.4.5.3 and 3.4.4.4.

b Secton 10.4.5.3 desenbes a hypotthi rupture of a circulating water expansion joint that floods the turtune building basement up to a grade level. Seebon 3.4.4.4 describes a site flood resulbng c from the Probable Maximum Meteorologeal Event (PMME) that would result in flooding of the L basement and 1st floor of the turbine and radwaste buildings and evaluates the potenbal impact on safety systems in the reactor building due to backflow flooding through the drains system. For both of these events, safe shutdown of the plant is assured These events can be considered as

( bounding for flooding in the turbine building. Plugging of the selected floor and equipment drains in the turtune building does not impact these events or change their conclusions as described in the UFSAR.

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REPORT UFSAR i

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l SAFETY EVALUATIONS UFSAR I PAGE1 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0031 UFSAR Revision No. NA I

Reference Document: UFStR Section(s) NA Table (s) NA Figure Change Yes X No Title of Change: Superintendent - Maintenance Qualifications

SUMMARY

$ This evaluation justified the appointment of an individual to the position of Superintendent -

Maintenance who does not meet all of the qualifications for this position as described in 7 UFSAR Chapter 13. This individual meets all the requirements except "an understanding of electrical and pressure vessel and piping codes" as recommended in ANSI 18.1-1971, Section 4.2.3. The Superintendent - Maintenance has over 24 years industrial and power plant experience. This includes nuclear power plant experience. Any lack of understanding of electrical and pressure vessel and piping codes by the Superintendent - Maintenance is adequately compensated for by the expertise and working knowledge of the individuals that J report to him. These individuals include the General Supervisor - Mechanical who has a L working knowledge the pressure vessel and piping codes; the Supervisor- Electrical who has a working knowledge of the electrical codes and standards; and the Maintenance Welding

- Engineer who has expertise and knowledge of the pressure vessel piping codes and has successfully completed various formal training courses pertaining to his job functions.

Additional training will be provided to the Superintendent - Maintenance in electrical and J pressure vessel and piping codes and engineering and technical resources will be made available to strengthen his capabilities in these areas.

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The subject individuars lack of understanding of the electrical and pressure and piping codes does not directly impact plant operation or equipment maintenance because his subordinates 1 possess the required expertise. The ability to consult with these subordinates also ensures I that the performance of the Superintendent - Maintenance as a Onsite Review Organization

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(OSRO) member is not adversely affected.

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I SAFETY EVALUATIONS UFSAR l

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1 l SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0057 UFSAR Revision No. 7 l

Reference Document: LCR-93-151-UFS Section(s) NA Table (s) 9.4-6 I Figure Change l l Yes l X l No l

Title of Change: Correction of Drywell Cooling Coil Capacity 1

SUMMARY

f This evaluation justified correcting drywell coolirig coil capacity in UFSAR Table 9.4-6. The j heat removal capacity required by design specification is 324,000 Btu /hr and this is the value !

used in the applicable design calculations. UFSAR Table 9.4-6 is revised to change 328,000 Btu /hr to 324,000 Btu /hr. This is a documentation change only with no impact to the plant as-built condition.

iI The drywell cooler heat removal capacity is not a parameter used in any accident analyses,

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l l and the drywell cooling system is not required for safe shutdown of the plant. Technical I

l Specification 3.6.1.7 requires that during power operation, the average drywell ambient l temperature not exceed 145'F in order to ensure that this drywell peak ambient temperature j does not exceed its design limit during LOCA conditions. The drywell cooling system l l

adequately meets this requirement.

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i l SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0062 UFSAR Revision No. 7 1

Reference Document: LCR-93-163-UFS Section(s) 3.5, 9.2 Table (s) NA

.I Figure Change Yes X No Title of Change: RHR Cooling Tower Tomado Missile Protection

SUMMARY

This evaluation resolved an open design basis item which identified a lack of a formal design calculation to substantiate the frequency of damaging all four fans of the RHR cooling towers at 5 X 10"' per year, as stated in the UFSAR 9.2.5.2.2. Based upon a systematic review of relevant documents, it has been determined that NRC acceptance of the tomado missile

] protective features of the Fermi 2 RHR service water cooling tower and associated equipment is not the 5 x 10"' per year frequency of damaging all four cooling tower fans presented in the UFSAR. The real basis for acceptance of the design is Edison's ability to achieve safe shutdown, assuming that fans may be damaged and that damaged fans can be replaced prior to their need. The safe shutdown scenario considers the damage of one tower in conjunction with a single failure, which is much more likely than the damage of all four fans by tomado missiles. The probability that both cooling tower divisions can be rendered out of service by

] tomado generated missiles entering the fan discharge stack is calculated in the UFSAR A sequence of events is presented in which tomado missiles damage one cooling tower fan in each division. The probability of this event is calculated as 5.2 x 10"" per year. Since this probability is not an exact probabilistic calculation result, it is more property characterized as

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between 10* and 104 ' per year.

The changes to the UFSAR delete selected descriptive information (i.e., 5 x 10"' valve) that is not related to the basis for acceptance of Fermi 2 plant design and provide minor clarifications.

-. These changes, therefore, do not increase the probability or the consequences of an accident

] previously evaluated in the UFSAR. The NRC's frequency acceptance limit for rendering both divisions of cooling tower fans inoperable is 10 per year. This limit, in itself, contains margin to the NRC's perceived failure limit (margin of safety). Changing the calculated frequency

] documented in the UFSAR from "5 x 10"' per year" (which is not an exact probabilistic calculation result) to a frequency *between 10* and 10"' per year" (which more property characterizes the result) does not affect this latent margin already set forth by the NRC.

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[ UFSAR PAGE 4

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SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0068 UFSAR Revision No. 7 Reference Document: LCR 93-169 Section(s) 13.1 NA

( Table (s)

( Figure Change W Yes R No Title of Change: Radiation Protection Functions and Responsibility Changes

SUMMARY

( This evaluation provided the option for Superintendent - Radiation Protection (RP) to assume the functions and responsibility of the Radiation Protection Manager (RPM). Training and qualifications requirements and reporting levels for both positions, however, remain unchanged. The changes to the UFSAR allow management the flexibility to combine the

( Superintendent - RP and the RPM positions. This organizational flexibility allows better utilization of radiation protection resources and yet maintains an equivalent degree of radiation protection for plant personnel, members of the public, and the environment.

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The changes in the RP organization do not impact the operation of any plant system or the response to any emergency condition.

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{ SAFETY EVALUATIONS UFSAR PAGE 5 b

SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 93-0074 UFSAR Revision No. 7 Reference Document: LCR 93-176- UFS Section(s) 13.4

[ Table (s) NA

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Figure Change Yes X No Title of Change: Safety Engineering R esponsibilities Change

SUMMARY

( This evaluation justified the revision to delete the paragraph in UFSAR sechon 13.4.3.4 requiring Safety Engineering to review all plant LERs and other intemal operating experience reports. The deletion includes the sentence discussing that Safety Engineering may initiate further action even if

( OSRO has completed their review. The subject paragraph in the UFSAR has been deleted because the action is redundant to the provision of UFSAR 17.2, Quality Assurance Program.

UFSAR 17.2.16 describes the corrective action program including independent review by Safety

[ Engineering and Nuclear QA. UFSAR 17.2.1.5 also describes that review of correchve action documents is performed by Nuclear Quality Assurance and Safety Engineenng within their assigned area of responsibility.

k Elimination of the review by Safety Engineering of the resolution of intemal operating experiena information will not increase the probability or consequences of an accident previously evaluated in E the UFSAR because the provisions for evaluations of the experience, review of the conclusions and

' implementation of the changes by the responsible Nuclear Generation group remain unchanged.

The independent review of selected plant activities by ISEG including maintenance, inodificetsoris, E operational problems and analysis as discussed in UFSAR 13 4.3.3 remains unchanged.

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Safety Evaluation No: 94-0024 UFSAR Revision No. 7 Reference Document: UFSAR Section(s) 7.7,15.1 Table (s) NA Figure Change Yes X No Title of Change: Update to UFSAR Sections 7.7 and 15.1 ,

SUMMARY

This evaluation Justifies updating of UFSAR Sections 7.7.1.3.3.6, " Turbine-Driven Feedwater Pump Control" and 15.1.2.3.2, " Input Parameters and initial Conditions." The information

I provided in these sections was out of date. PDC 13112, Rev A, was implemented during RFO3 to change the controlled speed range of each feedwater pump so that a maximum flow demand signal to both pumps would be limited to 117% rated feedwater flow.

PDC-13112, Revision A, changed the upper limit of the automatic feedwater flow control range j from 115% to 117% rated flow. The increase in the maximum feedpump flow capacity to 3 117% rated is bounded by the analysis of feedwater flow controller failure - open trensient event which assumes 130% maximum flow demand. The new higher feed $ vater pumps speed limit setting (5200 RPM) results in a higher feedwater pumps flow capacity. TI'is enhances plant capability to avoid a low reactor water level scram following a single feedwater pump trip and recirculation pumps runback. The increase of the feedwater flow demand shinal to 117%

rated, does not change the operation of the feedwater control system, and has r.o additional impact on plant operation. The increased feedwater pumps speed limit (5200 RPM) allows the operation of the plant with an adequate margin from the maximum feedwater pumps flow capacity analyzed for the transient ioad capacity, and the high water level trip setpoint is not changed.

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SAFETY EVALUATIONS UFSAR PAGE 7 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0029, Rev 1 UFSAR Revisic n No. 7 l Reference Document: LCR 94-051-UFS Section(s) 6.2 94-052-ISI Table (s) NA l

'Agure Change R Yes T No

'E Title of Change: Deletion of Bypass Leakage Requirements from the UFSAR

SUMMARY

This evaluation justifies the deletion of bypass leakage testing requirements for the HPCI and I RCIC test retum valves E4150F008, E4150F011, and E5150F022 as specified in UFSAR section 6.2.1.2.2.3. E4150F008 had been designated as bypass leakage boundary valve for primary containment penetration X 9A. E4150F011 had been designated as bypass leakage boundary valve for primary containment penetration X-9A and X-9B, and E5150F022 for I penetration X-98. Designating primary containment penetrations X-9A and X-98, along with their associated containment isolation valves for the bypass leakage program, is more conservative than using the existing bypass path and its accompanying valves. This is consistent with NUREG 0798 and thus, it is acceptable to use the primary containment isolation boundary valve for the bypass leakage program in lieu of the existing bypass leakage l valves (E4150F008, E4150F011, and E5150F022).

Primary containment isolation valves for penetrations X-9A and X-9B are designed to ASME 111 class requirements (Class I for inboard and outboard containment isolation valves) and meet the requirements for divisional power and diverse isolation signals. These valves are leak tested in accordance with 10CFR 50, Appendix J and meet the design and testing requirements of the existing bypass leakage valves (E4150F008, E4150F011 and E5150F022). Use of the primary containment isolation boundary valve for the bypass leakage program in lieu of the existing bypass leakage valves for the HPCI and RCIC test retum lines is bounded by existing analysis. This change to the bypass leakage program does not degrade the performance of any safety system and does not impact any radiation barrier.

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SAFETY EVALUATIONS UFSAR PAGE 8 SAFETY EVALUATION

SUMMARY

S fety Evaluation No: 94-0052 UFSAR Revision No. 7 i

Reference Document: LCR 94-090-UFS Section(s) 5.5 Table (n) NA Figure Change Yes X No Title of Change: Revision to UFSAR 5.5.7.4

SUMMARY

This evaluation justifies the revision to UFSAR 5.5.7.4, which describes the safety evaluation for the RHR system. The write-up: " Administrative controls would be used to allow energizing and opening valve F608 only when valve F009 could not be opened. Valve F608 would be opened by operator action at the motor control center." has been revised to: " Administrative I controls would be used to allow energizing and opening valve F608 only when valve F009 could not be opened. These administrative controls require dual operation of a keylock switch and a push button switch (in control room) to open the valve. An auditory and visual feedback I is provided by a control room alarm following the key lock switch operation. This is to prevent any inadvertent valve opening." Further, a statement is added in section 5.5.7.4 that during normal plant operation, the power is removed from valve E1150F008 to prevent any spurious action causing any intersystem LOCA.

The above change in UFSAR does not modify any physical hardware design. It merely clarifies the administrative steps necessary to open RHR shutdown cooling suction valve E1150F608 and the additional conservatism provided by removing the power fuses for valve E1150F008. The revised UFSAR write-up describes administrative measures taken to protect the low pressure RHR pipe and a release of any radioactivity to the secondary containment. It I does not reduce the margin of safety as described in the UFSAR or any Technical Specifications.

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I SAFETY EVALUATIONS UFSAR PAGE 9 1

SAFETY EVALUATION

SUMMARY

l Safety Evaluation No: 94-0070 UFSAR Revision No. 7 Reference Document: LCR 94-146-UFS Section(s) 9.3 I Table (s) NA Figure Change l l Yes fXl No Title of Change: Changes to UFSAR Section 9.3.3.2

SUMMARY

I This evaluation justified changes to UFSAR section 9.3.3.2 pertaining to the operation of emergency sump pumps which pump fluids from the oil / water sumps to the liquid waste holding pond. The changes made include the following:

1. UFSAR revision 6 indicates that the sumps are " rapidly

• emptied. An internal QA audit I indicated that " rapidly" refers to the automatic initiation of the pumps and since the pumps are now administratively controlled, the term " rapidly" does not apply. This UFSAR change deletes the term " rapidly."

2. UFSAR revision 6 indicates that the pumps operate "in case of fire." As this does not apply due to the unsusceptibility of the oil that these sumps see to fire, this UFSAR change deletes this statement.

3. UFSAR revision 6 also indicates that the pumps are controlled by level switches, but this

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would only take place after the administrative controls are removed.

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The emergency sump pumps rather than being activated by level switches are now p administratively controlled. Each of these emergency pumps has been valved off, locked, and l

electrically isolated to prevent their use due to the possibility of an uncontrolled release outside of the RRA.

A careful evaluation has determined that there is no danger to other equipment from fire damage associated with not pumping the sumps to the holding pond rapidly, but placing the emergency pumps under administrative control vs. automatic control could result in flooding the sumps. If the sumps on the first floor should flood, they would eventually fill the areas in which they are located. Any excess fluid would overflow to the turbine building first floor where L the floor drains would take it to the sumps in sump alley. It was also determined that the amount of fluid going to sump alley would be small. Even if the amount were extensive or any e

amounts going directly to sump alley were extensive, all safety-related equipment or equipment needed for safe shutdown in the turbine building is located on the second floor and above; therefore, no safety-related equipment would be placed in danger. An extensive review of turbine building flooding impact on safety-related equipment was done and no impact was found.

I SAFETY EVALUATIONS UFSAR I PAGE 10 l SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0072 UFSAR Revision No. 7 Reference Document: NA Section(s) 11.2 Table (s) NA Figure Change Yes X No Title of Change: Addition of a Portable Ultraviolet (UV) Radiation Total Organic Carbon I (TOC) Reduction Tool

SUMMARY

I This safety evaluation justifies the addition of a portable UV radiation TOC reduction system to reduce the organics into constituents ion exchange can remove. The UV portable unit will be installed in radwaste and/or OSSF buildings and operated in the liquid radwaste system. The j

i I organic destruction method will be used, as deemed necessary, to reduce organics from the liquid radwaste process waste streams. This will permit Fermi 2 to more effectively treat organically contaminated water while reducing solid radwaste.

I The portable UV unit is designed to a pressure rating of 150 pounds, which is also the design pressure rating of radwaste systems. Therefore, a breach of UV unit is bounded by existing accident scenarios described in UFSAR section 15.7.3. UV treatment is a passive treatment.

The portable UV treatment unit does not adversely impact applicable sections of the Technical Specification (TS 3/4.11.1.4) and Offsite Dose Calculation Manual (ODCM 3/4.3.7.11, 3/4.11.1.1,3/4.11.1.2 and 3/4.11.1.3).

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SAFETY EVALUATIONS UFSAR PAGE 11 SAFETY EVALUATION

SUMMARY

l Safety Evaluation No: 94-0078 UFSAR Revision No. 7 Reference Document: LCR 94-156-UFS Section(s) 4.4 Table (s) NA

[ Figure Change U Yes lXlNo Title of Change: Adding New References in UFSAR 4.4

SUMMARY

This evaluation justifies adding two new references (35 and 36) in UFSAR 4.4 (Reference pages) and parenthetically in UFSAR 4.4.4.6.3. The references added are:

35: NRC Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal Hydraulic Instabilities in Boiling Water Reactors", July 1994 36: BWROG Letter 94078, "BWR Owners Group Guidelines for Stability Interim Corrective Action", June 1994 l

Implementation of guidance in References 35 and 36 is designed to upgrade Femil 2 instability prevention and response procedures and programs to reflect more recent industry experience.

Implementation of the new guidance will reduce the likelihood of an instability occurring, and L will require more conservative actions in the unlikely event an instability does occur. The ,

changes may result in an increased probability of an unnecessary reactor scram due to the l r more proactive guidance on preemptive scrams for events known to cause Region A entries l and emphasis on early instability detection and mitigation. While an unnecessary reactor l scram is not desirable, it is not classified as an accident, and the benefits of this proactive J

Reactivity Management philosophy outweigh this potential risk. The proposed changes will I improve, rather than reduce, the margin of safety as defined in the bases for the associated Technical Specification, UFSAR, or SER sections.

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SAFETY EVALUATION

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Safety Evaluation No: 94-0086 UFSAR Revision No. 7

[ Reference Document: LCR 94-013-UFS Section(s) NA 5.2-1, 5.5-1 h Table (s)

Figure Change Yes l X l No Title of Change: Revision to the Design Pressure and Temperature for Reactor Recirculation

( System (RRS)

SUMMARY

This evaluation justifies revision to the UFSAR to more clearly describe the design pressure ano temperature values of the reactor recirculation system components. A rev'en of design 4 documentation conducted during preparation of the RRS Design Basis Document indicated  !

that the RRS pumps (B31010001A, B} discharge valves (83105F031A, B) and system l discharge piping were not all purchased to the same design conditions. Design conditions for ]

the discharge valves exceed those for the discharge piping and thus are acceptable. An  !

evaluation of the pump design conditions was performed in accodance with ASME Code equations for determining minimum wall tiickness. This evaluation concluded that the pump design pressure and temperature condNons envelop the system discharge piping design requirements and thus are acceptable. The UFSAR change provides individual entries in Tables 5.2-1 and 5.5-1 for the RRS pumps, discharge valves, and discharge piping so that the i

differing design conditions are clearly presented. A note has also been added to each table to  ;

indicate that the RRS pump design conditions envelop those for the system discharge piping. I r The UFSAR revisions are documentation-only changes to ensure consistency with RRS design L documentation The physical configuration, function, and operation of the RRS and its components are not affected. The UFSAR changes have no impact on Technical Spec (cabon r 2.1.3 and 3/4.4.

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SAFETY EVALUATIONS .

UFSAR PAGE 13

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SAFETY EVALUATION

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Safety Evaluation No: 94-0095 UFSAR Revision No. 7 Reference Document: LCR 94-181-UFS Section(s) NA

( Table (s) 7.3-12(3.3.2-3)

( Figure Change l l Yes lXlNo Title of Change: Revision to UFSAR Table 7.3-12 (3.3.2-3)

SUMMARY

l This evaluation justifies revision to UFSAR Table 7.3-12 (3.3.2-3) and Technical Requirements Manual (TRM) Table 3.3.2-3 to incorporate the response time limit of 5 0.5 seconds for the main steam line flow-high, that initiates the MSIV closure as a result of the main steam line break. This change does not impact plant operation, procedures, tests, or other functions.

The response time testing for the affected instrument channels are performed in accordance with surveillance procedures 44.020.043,44.020.044,44.020.045 and 44.020.046.

The addition of the MSIV actuation time for the Main Steam Line Flow- High instrumentation of 5 0.5 seconds to UFSAR Table 7.3-12 (3.3.2-3) and TRM Table 3.3.2-3 is the bounding assumption included in the UFSAR, Section 15.6.4, Steam Line Break Outside Containment.

i The MSIV actuation time of 5 0.5 seconds is already included in UFSAR Table 15.6.4-1. This change does not modify the physical configuration, function, test, or operation of the plant.

The performance of response time testing is in agreement with UFSAR analysis.

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SAFETY EVALUATIONS I UFSAR PAGE 14 SAFETY EVALUATION

SUMMARY

Safety Evaluation No: 94-0097 UFSAR Revision No. 7 Reference Document: LCR 94-194-UFS Section(s) 5.5, 7.3 j Table (s) NA

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SUMMARY

[ Previous design modifications (EDPs 5544 and 8684) provided a high drywell pressure

] isolation signal to the RBCCW/EECW supply to the drywellisolation valves. A loss of Division 1 offsite power followed by a postulated single failure that activates the high drywell pressure signal, will result in total loss of cooling to the reactor recirculation pump "A" seal and pump

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motor. This safety evaluation corrects the description of the reactor recirculation system (RRS) pump's response to loss of cooling water (UFSAR 5.5.1.4) and adds systems that isolate on a I high drywell pressure signal (UFSAR 7.3.2.2.7.6). Additional systems isolating on high drywell pressure are reactor recirculation pumps seal purge supply lines and EECW Divisions 1 and 2 drywell cooling supply lines (Note: isolation signal from ECCS logic, not from RPS logic).

L Even if all cooling is lost to the recirculation pump, there is at least 10 minutes available for operator action to prevent any seal failure that could result in loss of reactor coolant. Also, the trip of the RRS pump motor (e.g., as a result of loss of cooling) is not an accident, but a H transient event of moderate frequency. As stated in the UFSAR, the RRS pump is not considered essential for the safe shutdown of the plant. The small amount of coolant loss due to gross seal failure can be easily made up by the several independent makeup sources, such u as CRD, SBFW or RCIC.

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Safety Evaluation No: 94-0118 UFSAR Revision No. 7 Reference Document: LCR 94-233-UFS Section(s) 7.5 Table (s) 7.5-4 i Figure Change l l Yes X No -

Title of Change: Revision to UFSAR Section 7.5.2.5

SUMMARY

This evaluation justifies the following changes to the UFSAR section 7.5.2.5.

Table 7.5-4 is deleted since it is not necessary for the understanding of the description in UFSAR 7.5.2.5.2. In place of the table, a description has been added (in UFSAR 7.5.2.5.2) listing the type of equipment being operated during the use of the dedicated I shutdown procedure. This change makes this section consistent with the procedural outline given in UFSAR 7.5.2.5.3.

Revise Section 7.5.2.5.3 to replace step (2) with a step to manually operate certain valves locally. These manual valves are part of plant procedure 20.000.18 steps that have been walked down and verified to be within the criteria given in UFSAR Figure 7.5-10, Maximum Tiime Available for Operator Action.

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• Revise Section 7.5.2.5.4 to indicate that UFSAR Figure 7.5-10 provides time goals for the three phases of the shutdown procedure.

• Revise Section 7.5.2.5.4 to change the time from 17 to 27 minutes that the SBFW will -

be injecting water under the worst case assumptions. An assumed 10-minute restart r time is added to the worst case scenario, based on a normal starting sequence time of L about 6 minutes. The increased time to establish SBFW flow to the vessel from 17 minutes to 27 minutes still leaves a significant amount of water covering the core when the flow is re-established. As additional verification, the procedure 20.000.18 was walked through by Operations. The 13 minute coast down time was assumed and an additional 10 minutes restart time was considered for the time to restart SBFW flow.

This verification included an additional item to walk to and close P5000F154, air supply

[ valve to E41F011. The procedure was accomplished within the guidelines of UFSAR Figure 7.5-10.

Revise Section 7.5.2.5.4 to indicate that the spurious diversion path through E41F011

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is closed by isolating the pneumatic supply valve, and this operation is accomplished during the startup time of CTG Unit 1.

SAFETY EVALUATIONS

(. UFSAR PAGE 16

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Safety Evaluation 94-0118 (Continued)

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• Revise Section 7.5.2.5.4 to indicate that Figure 7.5-10 provides the acceptable criteria for achieving restoration of primary containment cooling and torus cooling and that the procedure can be completed within the time criteria established by this figure.

The above changes to the UFSAR do not change the possibility that a fire occurs or the ability to shutdown the reactor using the dedicated shutdown system. In addition, the changes made L do not impact nor reduce the margin as provided in Technical Specification Bases 3/4.7.11, '

Appendix R, Attemative Shutdown Auxiliary Systems. The criteria that the water injection (to r RPV) capability be restored before the water level reaches the top of active fuel is not impacted by the above changes to the UFSAR.

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SAFETY EVALUATIONS

{ UFSAR PAGE 17

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SAFETY EVALUATION

SUMMARY

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Safety Evaluation No: 95-0012 UFSAR Revision No. 7

[ Reference Documenh LCR 95-035-UFS Section(s) 9.2,11.4 A.1.70

[ Table (s) 1.6-2 Figure Change l X l Yes No

[l Title of Change: Removal of UFSAR Figures

SUMMARY

This evaluation justifies removal of UFSAR Figures 9.2-1 (Sheets 1 and 2) and Figure 9.2-12 (Sheets 1 and 2) from the UFSAR. These are P&lDs for the GSW and TBCCW systems.

Figures 9.2-1 (sheets 1 and 2) and 9.2-12 (sheets 1 and 2) were removed from the UFSAR and the reference to these figures within the UFSAR text of the GSW and TBCCW system desenpbons replaced with the Detroit Edison drawing number. Removal of Figures 9.2-1 and 9.2-12 has no design change impact on either system. This change does not seek to change or modify any of the drawings being removed from the UFSAR. Any future design

{ modifications that impact these drawings will be conducted through the normal plant design change process. Modifications to plant systems, components or structures will continue to be implemented in accordance with the requirements of 10CFR50.59, as required.

The figures removed from the UFSAR were used to enhance the GSW and TBCCW system r descriptions in the UFSAR as recommended by Reg. Guide 1.70. The GSW and TBCCW system's safety evaluation is described in the written portion of the UFSAR, Sechon 9.2.1 and 9.2.7. Eliminaten of these UFSAR figures does not change or reduce the capability of either the GSW or TBCCW system as described in the written text of the UFSAR. Therefore, the L consequences of an accident as previously evaluated in the UFSAR are unchanged.

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UFSAR PAGE 18 b

SAFETY EVALUATION

SUMMARY

b Safety Evaluation No: 95-0019 UFSAR Revision No. 7 Reference Document: LCR-95-052-UFS Section(s) 2.2 Table (s) NA

[ Figure Change l l Yes l X l No Title of Change: Incorporate the Frenchtown Township Water Treatment Facility

SUMMARY

This Safety Evaluation addresses recent construction of the Frenchtown Township water treatment facility and its operation. This Safety Evaluation was written to revise the UFSAR regarding nearby industrial, transportation and military facilities to include this facility and address any potential hazards as a result of its operation. The addition of the Frenchtown L

Township water treatment facility is not a change to the Fermi 2 facility as the water treatment plant is not located on the Fermi 2 site.

I L The operation of the Frenchtown Township water treatment plant and its use of chemicals and other potential hazardous materials was evaluated. The substances used at the water treatment plant are not a hazard to the Fermi site. The chlorine release accident analysis was not impacted by this facility or its use of chemicals. The facility was determined not to possess the potential to cause missiles that could impact the plant. The facility was also determined not to have the potential to create an aircraft hazard.

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SAFETY EVALUATIONS UFSAR PAGE 19 The following Technical Specification Amendments were incorporated into Revision 7 of the UFSAR. The NRC safety evaluation (which is based on the Detroit Edison evaluation supporting the change) that accompanies each amendment provides the basis and justification for the UFSAR revision.

l T. S. Amendment Description UFSAR Section/ Table l 67 Elimination of rod sequence control 7.7 I

l system. Other UFSAR changes were made in a previous UFSAR submittal.

87 Changes to SRV pilot pressure Figure 5.1-03, Sheet 3 setpoints due to power uprate. Figure 7.3-12, Sheet 3 95 Removal of certain 18-month 9.2 surveillances during shutdown

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98 Removal of containment isolation 6.2 I valves classification for valves Tsb!9 6.2-2 E1100F050A and B Table 6.2-15 l

99 and 100 Relocation of Response Time 7.2 Tables 3.3.1-2, 3.3.2-3, and 3.3.3-3 7.3 from Technical Specifications to the Table 7.2-4 UFSAR (Amendment 100) and Table 7.3-11 elimination of response time testing Table 7.3-12 for the isolation actuation system corresponding to the diesel generator start and sequencing of loads (Amendment 99)

- END OF SAFETY EVALUATION

SUMMARY

REPORT -

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a c DETROIT EDISON --FERMI'2 AITTOMATED RECORD MANAGEMENT .

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