ML20217J198

Summary Rept of Changes,Tests & Experiments Completed, for Period 950101-970930

ML20217J198
Person / Time
Site:	Quad Cities
Issue date:	09/30/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20217J193	List: ML20217J198 SVP-98-113, Forwards Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period 950101-970930
References
NUDOCS 9804060224
Download: ML20217J198 (156)

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SUMMARY

REPORT OF CHANGES, TESTS AND EXPERIMENTS COMPLETED ATTACHMENT A SVP-98-113 l

f 9004060224 980331 PDR ADOCK 05000254 R PDR L

SAFETY EVALUATION REPORT INDEX M041/241-7 E04-2-93-192 E04-1-95-050 DCP 9600185 M04-1-85-032A,B E04-2-93-207 E04-1-95-059 DCP 9600189 M041(2)-85-032B E04-2-93-245 E04 1-95-060A DCP 9600205 ,

M04-245 51 E04 2 93-253 E04 1-95-060B DCP 9600206 '

M04-1/2-86-3 E04-2-93-306 E04-1-9540C DCP 9600208 M04-1(2)-87 002 E04-2-93-307 E04-1-95-060D DCP 9600209 M041(2)47-003B E04-0-94-117 E04-195 060E DCP %00224 M04-1(2)-87 59 E04 0-94-126 E04 1-95-060F DCP 9600228 A,B,C,D E04-0-94-192 E04-1-95-060 DCP 9600250 J M04-148 080 E04-0-94 211 A,B,C,D,E,F DCP 9600251 M04-1(2)-88-101C E04-0-94-225 E04-1-95-062 l

DCP 9600282 i M04-248-103D E04-1 94-002 E04-1-95-064 1 M04 0-89-072 E04-1-94-070 A,B,C,D,E,F DCR 4 94159 M04-1(2)-89-074 E04-1-94-ll3 E04-1-95 066 DCR 4-95-041 J M04 2-91-013A E04-1-94-Il7 E04 1-95-067 M04-1-9l-021 A E041-94138 E04-2 95-003 SSCR 9600334 M04-2-92 028A E04-1-94-171 E04-2-95-015 M04-2-93 007B E04-1-94-186 E04-2-95-016 SE-91270 I M04-1-94 007B E04-1-94-193 E04-2-95-020 SE-91408 M04-2-94-007A E04-1-94-195 E04 2 95-023 SE-91-538 M04-1(2)-95-002 E04 1-94-207 E04-2-95-024 M04-0-95-004 E04-1-94-217 E04-2-95-025 SE-95-013 E04-1-94 236 E04 2-95-028 SE-95-043 PO4-1(2)-91 ll4 E04-1-94-237 E04-2-95-037 SE-95455 l P04-1(2)-91-143 E04-1-94-244 E04-2-95-043 SE-95-056 , l E041-94-245 E04-2-95-044 SE-95 057 E04-193-069 E04-1-94 246 E04-2-95-046 SE-95 058 E04-1-93 084 E04-194-247 E04-2-95-048 SE-95-060 E04-193-113 E04 2 94-002 E04-2-95-050 SE-95 062 E041-93131 E04 2-94-007 E04-2 95-051 SE-95-063 E04-1-93 145 E04-2-94-071 E04 2-95-055 SE-95 064 E04-1-93175 E04-2-94-118 E04-0-96-040 SE-95 065 E04-193-183 E04-2-94 128 E04-0-96-003 SE 95 066 E04-1-93185 E04-2-94-130 E04-1-96-004 SE-95 067 E04 1-93-192 E04-2-94-131 E041-96-007&008 SE-95 070 E04-1-93-212 E04-2-94-132 E04-1 96-Oll SE-95 071 E04-1-93 238 E04 2-94-133 E04-1-96 023 A SE 95-072 E04-1 93-244 E04-2-94-134 E04-1-96 023B SE-95 073 E04-1-93-253 E04-2-94-135 E04-1-96 023C SE-95 074 E04-1-93 306 E04-2 94-138 E04-1-96-023G SE-95-075 E01-2-93-ll3 E04-2-94 171 E04-1-96-024 SE-95-077

( E04-2-93-131 E04-2-94 209 E04-1-96-025 SE-95-078 l E04 2-93-145 E04-2-94-220 E04-1-96-031 SE-95 084 E04-2-93-174 E04 2-94-226 E04-1-96-039 SE-96-001 E04-2-93-175 E04-0-95 006 E04196-042 SE-96-002 E04-2 93-184 E041-95-036A E04-1-96-043 SE-96-003 E04-2-93-187 E04-1 95-042 E04 2 96-025 SE-96-004 1

ENODOCS'SEINDX. DOC

p i

SE-96-008 SE-96 098 SE-%181 SE-97 083 SE-96 010 SE-5 100 SE-%-182 SE 97-084 SE-96412 SE-96-102 SE 96-185 SE-97-088 SE-96413 SE-%103 SE-%186 SE-97 090 SE-96417 SE-% 104 SE-%189 SE-97-092 i SE-96 019 SE-%-105 SE-%-190 SE-97 093 L

SE-% 020 SE-96-106 SE-%-192 SE-97-094 SE-%422 SE-% 108 SE- % 193 SE-97 095

- SE-96423 SE.%-109 SE-%-l% SE-97 / M L - SE-96424 SE- E ll! SE-% 197 SE-97 097 SE-96425 SE-%-112 SE-97-002 SE-97-100 SE-96427 SE %-113 SE-97 003 SE-97-101 SE-96 029 SE-%-116 SE-97405 SE-97-102

. SE-96430 SE-96-118 SE-97407 SE-97-121 SE-96 031 SE 96-121, Rev 1 SE-97 008 SE-97-122 i

SE-96432 SE-5 122 SE-97 009 SE-97-123 SE-96-033 SE-%123 SE-97 Oll SE-97-125 SE-96435 SE- % 124 SE-97 013 SE-97-126 SE-96-036 SE-%125 SE-97 014 SE-97-127 SE-96 037 SE-96-129 SE-97 015 SE-97129 SE-96 040 SE %-133 SE-97 017 SE-97-130 I SE 96441 SE %-136 SE-97418 SE-97-131 L SE-96442 SE-% 137 SE-97-019 SE-97-133 i- SE-96-046 SE-%138 ' SE-97 020 SE-97-140 SE-96-050 SE-%139 SE-97 021

!. SE-96454 SE-96-141 SE-97426 SE-96455 SE-%142 SE-97 030

' SE-96-063 SE-%143 SE-97-033

)

{

SE-96-064 SE-%145 SE-97 034 SE-96 070 SE-%-147 SE-97 035 l . SE-96473 SE-%-151 SE-97-036 l SE-%074 SE-5152 SE-97 037 SE-96 075 SE-%-153 SE-97 038 '

SE-96 076 SE-96-154 SE-97 041 SE-96 078 SE-%-155 SE-97 043 i SE-96 082 SE-%-156 SE-97447 SE 96083 SE % 157 SE-97-049 i SE-96 085 SE-%158 SE-97-052 SE-% 086 SE-96-163 SE-97-054 l SE-% 088 SE-96-164 SE-97 056 l SE-96-089 SE-E165 SE-97-064 SE-96 090 SE-%166 SE-97 065 l

! SE-96 091 SE- % 167 SE-97 066 SE-96 093 ' SE-%-168, Rev i SE-97-070 SE-96 094 - SE 96-170 SE-97 073 l SE-96 095 SE- % 172 SE-97 075 i SE-96 096 SE-%-174 SE-97 076 l

- SE-96 097 SE-96-175 SE-97-080 I I

2-ENODOCSWEINDX. DOC i-1-

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Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1/2-81-7

Title:

N 2Supply to Heating Boilers I

Description:

This change installed N2 supply lines to A and B heating boilers for dry lay-up during summer months when the heating boilers are not in use.

1 Result: Unreviewed safety question does not exist. This change was added to an existing N line. There 2  !

are four manual isolation valves and a pressure regulating valve in the existing system line from the bulk j

storage tank to the modification tie-in. There is sufficient isolation between the bulk storage tank and a i

postulated leak in the added lines. The drywell-to-suppression chamber differential pressure is not affected. I This addition will be utilized intermittently and at such small flows that there will be no significant affect on the amount of N2used so Technical Specification 02 concentration will not bejeopardized.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1-85-032A,B

Title:

Drywell/ Torus Differential Compressor IB Control Schematic i

Description:

This change was done to correct deficiencies in the drywell/ torus differential compressor 1B control schematic and associated IwJ panel wiring. The breaker feeding this equipment and the fuse rating on the control transformer were oversized and did not provide adequate protection.

Result: Unreviewed safety question does not exist. The pumpback air compressors maintain drywell pressure greater than torus pressure. This work corrects deficiencies in the existing back-up compressor control and power circuitry. Correcting these deficiencies enhances system reliability which does not have any bearing on the probability of an accident. Enhancing system reliability does not have any effect on -

radiological consequences of an accident or malfunction. However, increasing system reliability assures that the minimum ~ drywell/ torus differential pressure is maintained by providing a back-up compressor.

Downsizing the feed breaker and control transformer fuse provides a more appropriate means of circuit protection. The failure modes remain unchanged. Enhancing the pumpback air capability does not increase the probability of a malfunction of equipment important to safety.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification l Evaluation Reference Number: M04-1(2)-85-032B l

Title:

Feed Breaker for Drywell/ Torus Differential Compressor IB (2B)

Description:

This change downsized the feed breaker for the drywell/ torus differential compressor IB (2B) from 150 A to 100A. Also, a control transformer fuse was downsized. This was done to correct circuit design protection deficiencies. Also, circuit installation deficiencies from the original Station Modification were corrected by complete rework of the control wiring. This provides an operable backup compressor for the pumpback air system.

Attachment A, SVP-98-ll3, Page 1 of 153 i

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Result: Unreviewed safety question does not exist. The changes are to the backup compressor which has  !

never been functional. Correcting design deficiencies will allow the backup system to be operational. This increases the reliability of the pumpback air system. In the event that the primary compressor becomes inoperable, the backup compressor could then be used. Downsizing the feed breaker and control transformer fuse provides a more appropriate means of circuit protection. The failure modes remain unchanged. Enhancing the pumpback air capability does not increase the probability of a malfunction of equipment important to safety.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification j i

Evaluation Reference Number:

M04-2-85-51

Title:

Halon Fire Protection System

Description:

This change revised the original scope of the modification which was to install a Halon Fire Protection System in the Quality Control X-Ray Film Lab. The scope was changed considerably. The Halon System installation has been completed but because the suppression does not meet NFPA code requirements it must be removed. No attempt has ever been made to seal air leaks in the X-Ray Film Lab thus, the Halon, once introduced, cannot be maintained at a 6% concentration for 10 minutes as required by NFPA code. This ECN was generated to demolish portions of the system that no longer were required. The Halon cylinder was removed; its solenoid, associated piping, conduit, supports, and TRI devices were removed. The detection system remains intact including the exhaust fan trip interlocked to the two area smoke detectors. The manual pull station was re-wired for local and remote alarm activation only.

Result: Unreviewed safety question does not exist. There are no new failure modes introduced. The failure modes of the new equipment are identical to that of other XL3 detection equipment already in use.

This new system enhances the plants ability to quickly detect a fire in the film lab and employs Quad Cities defense-in-depth concept in order to ensure that safe shutdown capacity is not impaired by a fire. The new system actuates local alarm bells and remotely alarms in the control room. All interfaces with other systems, structures, and components are unchanged. The probability of an accident is not affected by the addition of the new system.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1/2-86-3

Title:

Security CCTV System

Description:

Four fixed closed circuit television (CCTV) cameras were added to four camera towers along the security fence on which ran and tilt (P&T) cameras already existed. The new fixed cameras allow l the P&T cameras to cover areas along the security fence which have received inadequate coverage in the past. Other changes made include the exchanging of a fixed camera location for a P&T location and the relocation of a fixed camera. Accompanying cables and monitors were added as required.

1 Result: Unreviewed safety question does not exist. This change affects the security of the plant; not the operation of systems related to safety within the plant. No new failure modes have been created. The I

Attachment A SVP-98-113, Page 2 of 153 j

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I security of the plant is upgraded by the additional CCTV's without reducing the margin of safety in the existing security system. I Safety Evaluation Number: NA Type of Safety Evaluation: Modi 0 cation l Evaluation Reference Number: M04-1(2)-87-002

Title:

RilR SW Pump

Description:

This change modified the RilRSW pump internals dampening the vibration amplitudes occurring at vane-pass frequency by angling the volute's inlet edges (cut-water). This will decrease the dynamic forces created by the interaction between the impeller vane pressure wake and the volutes. Thh will improve pump seal life.

Result: Unreviewed safety question does not exist. RIIR service water How is not affected. The probability of an accident is not changed. RilR service water pump provides cooling water to the RiiR lleat exchanger. Modifying the pump internal casing improves performance of the pump and increases reliability of the pump and its components. Consequences of an accident are decreased. The probability of an equipment malfunction decreases.

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Safety Evaluation Number: NA Type of Safety Evaluation: Modification l

Evaluation Reference Number: M04 1(2)-87-003B

Title:

Reactor Building Sample Panels

Description:

This modification replaced the existing sample panels in the Reactor Building for both units with new sample panels which have improved operational features and a modular design.

Result: Unreviewed safety question does not exist. Probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased because this modi 0 cation consists ofinstalling non-safety-related, seismic and non-seismic mounted equipment. This modification does not affect any design basis accident or single failure event scenarios as previously analyzed. No new failure modes are created. Seismic mounting of the panels will assure adjacent safety-related equipment is protected from damage during a seismic event. The margin of safety is not reduced since all conditions applicable to this modification have been previously addressed in the basis for the existing equipment to be replaced.

I Safety Evaluation Number: NA Type of Safety Evaluation: Modification UFSAR Revision l

Evaluation Reference Number: M04-1(2)-87-059A,B,C,D; UFSAR No. 94-22,94-26

Title:

Reactor Vessel Water Level Instrumentation (RVWLIS) l l Attachment A, SVP-98-113, Page 3 of 153 i

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Description:

Changes implemented by these partial modifications included replacement of the existing condensing chamber with a new condensing chamber. The Yarway column was replaced by a new condensate reservoir. The variable reference leg was rerouted inside the drywell. The cold reference leg piping from the new condensing chamber and condensate reservoir was relocated to the reactor building holding the vertical drop in the drywell to a maximum of two feet. The new reference leg piping was routed outside the drywell and tied back into the existing instrument piping upstream of the instrument racks. The existing containment penetrations, X-108 and X-109, were used for the rerouted cold reference legs. The l purpose of this modification is to minimize the effect of reference column water boil-offin a post-LOCA or l high drywell temperature condition on the "A" loop reactor vessel water level instrumentation system. A l

UFSAR change was done to reflect the addition of the new penetrations and a description of the reactor 1 i

vessel water level instrumentation system. This system is used to determine the reactor water levels deGned in Technical Specifications, must meet the operability requirements set forth in Technical Specifications, and must be capable ofinitiating the protective functions delineated in Technical Specifications.

! Result: None cf the safety functions, setpoints, or actuation functions delineated in the Technical Specifications for this system are changed or modified as a result of the implementation of this modification. Plant operation is not changed as a result of this modification. The function of the reactor vessel water level instrumentation system remains unchanged. This system will be more accurate during certain postulated Design Basis Accidents. All components in this modification are passive. There are no new failure modes. Therefore, there is no basis for a postulated increase in the probability of an accident.

The consequences of this accident is dictated by pipe size. This parameter remains unchanged by these partial modifications. No other RVWLIS failure mechanism could increase the consequence of the small line break accident analysis. Therefore, all postulated failures remain bounded by the existing accident

' analysis. Function of original design is unchanged. The modification will enhance system accuracy by limiting the amount of vertical reference leg piping in the drywell and thereby reducing the potential for reference leg boil-off.

Safety Evaluation Number: 90-582 Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1-88-080

Title:

Dryer / Separator Pit Seal Gate l

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Description:

This change installed the seal gate between the existing reactor cavity bulkhead and dryer separator pit. The new seal gate is equipped with dual inflatable seals and will be used during refueling outages.

Result: Unreviewed safety question does not exist. This change has no effect on equipment important to l safety. The dryer separator pit is non-safety-related and scismic properties of the pit are not compromised.

This change improves the integrity of the dryer separator pit and actually reduces the possibility for accidents due to water damage and leakage. The equipment affected by this modification is not significant to any margin of safety as defined in the Technical Specifications.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1(2)-88-101C Attachment A, SVP-98-113, Page 4 of 153

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Title:

" Signa' Resistors", Thermocouple l

Description:

This change established new computer inputs from existing instrumentation loops, using

\- resistors and established new computer inputs from existing unused thermocouples.

Result:

Unreviewed safety question does not exist. The addition of new input signal points using resistors inserted into non-safety-related circuits does not affect any safety-related circuit. The use of existing, non-l used thermocouples for new computer inputs does not affect any safety-related circuits. Therefore, these l additions have not increased the probability of an accident or malfunction of equipment imponant to safety.

All new cables were installed in seismically mounted conduits to fully contain circuit failures and to mitigate failure effects. Balance of plant failures do not create new accidents or malfunction situations.

Margin of safety remains unchanged. {

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Safety Evaluation Number: NA Type of Safety Evaluation: Modification  :

Evaluation Reference Number: M04-2-88-103D

Title:

RHR and HPCI Supports '

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Description:

RHR large bore piping was evaluated for a 340 degree F shutdown cooling mode j

temperature. This modification will bring the piping system within code design margins. This change j included revising, adding, or removing seventy-seven pipe supports, two wall penetrations and performing various reinforcing welds. l Result: Unreviewed safety question does not exist. This modification does not affect any equipment j failures. This modification alters piping and pipe supports only. Revising piping and supports will not i

affect plant operation when RHR and HPCI functions as intended. These changes have no effect on RHR and HPCI operation or safety function. This modification is required to qualify the piping and supports for higher temperature and hydrodynamic load conditions. It decreases the possibility of failure of these lines j and decreases the probability of failure of RHR. No new failure modes are introduced. The design reduces .

l the possibility of a failure of piping in the RHR & HPCI lines.

l Safety Evaluation Number: NA Type of Safety Evaluation: Modification i

Evaluation Reference Number: M04-0-89-072

Title:

Fire Protection I

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Description:

This change involved tapping into the existing fire protection line near the northwest corner of Warehouse No. land installing a 6-inch diameter branch connection at this point and a position indicating isolation valve. The purpose is to provide an isolation point for the fire protection branch line which serves Warehouse No. 2 and Warehouse No. 3.

Result: Unreviewed safety question does not exist. This modification does not affect any bounding conditions in the accident analysis. No new accidents are introduced. No new systems are being added or modified so the occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated is not increased. The intended function of the modified system is not Attachment A, SVP-98 113, Page 5 of 153

changed and the modified portion is designed to the same or better standards as the existing system.

Therefore, the margin of safety is not reduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1(2)-89-074

Title:

Roll-O-Matic Filters

Description:

This change replaced roll-o-matic filters in various locations in the reactor and turbine buildings for both units with photocells and lights to prevent using up filter rolls before the filter media is depleted. The original timing controls advanced the filter media before filter media became dirty. The auto sensing control senses the amount of dirt buildup on the filter. When the filter reaches a preset value, the photo sensor does not receive light and the advancing motor is energized and the filter rotated. An annunciator light is activated in the control room when the filter runs out or when the photocell light bulb burns out.

Result: Unreviewed safety question does not exist. This system performs a non-safety-related function and is not addressed in the UFSAR. These filters do not impact accident probabilities.

Safety Evaluation Number: NA Type of Safety Evaluation: Moditication UFSAR Revision Evaluation Reference Number: M04-2-91-13A; UFSAR No. 94-6 and 94-17

Title:

liigh Pressure Coolant Injection Sparger l

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Description:==

This safety evaluation is for the incomplete installation of M04-2-91-013 (only). The safety evaluation for the entire modification is unchanged. The portions of the modification that were not installed are: 1) two support upgrades to larger snubWs on the turbine exhaust line and 2) the sparger was not installed inside the torus. A new Engineering Change Notice (ECN) issued has added the following to the modification design: 1) since the current turbine exhaust pipe inside the torus is not removed dering Q2R11, the internal vacuum breaker line and valves are removed and capped; 2) an interference between the  !

X-220 reinforcement and the external vacuum breaker line is corrected and a support found on the IIPCI '

turbine exhaust that is not part of the piping analysis is removed. The reason for this modification is to provide for stable condensation in the HPCI turbine exhaust line. The UFSAR was updated to reflect the new equipment and changes to Group 4 isolation logic. This modification was not an unreviewed safety question (with no Significant llazards Identified) and NRC approval of the licensing amendment was required (and received per Amendment No.130) prior to considering the modified systems to be operable, j when required to be operable by Technical Specifications. The changes requiring NRC review were: 1)the l new containment penetration,2) the isolation valves and their performance parameters,3) the new Group 4 l logic for the vacuum breaker isolation, and 4) the trip setpoint for IIPCI Low Reactor Pressure.

Result: The new vacuum breakers can fail open causing torus pressurization when liPCI is run. Since the l

existing vacuum breakers can cause the same failure, this is not a new failure mode. By upgrading the

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Group 4 isolation logic to incorporate current design criteria, the number of potential failure modes of the  !

l Attachment A, SVP 98-ll3, Page 6 of 153

l containment isolation system has been reduced. The design incorporates functional, electrical, and physical separation principles ofIEEE-384, wherevn rassible, and NRC requirements for containment design.

Safety Evaluation Number: NA Type of Safety Evaluction: Modification Evaluation Reference Number: M04-1-91-021 A

Title:

Replacement of Filter / Dryer Skid For instrument Air Compressor Desca!ption: This partial modification replaced the filter / dryer skid for instrument air compressor with a

larger capacity filter / dryer skid. A coalescing filter was instr.lled on the service air tie-in to the new l

filter / dryer system to remove oil from the service airjust prior to entering the instrument air filter / dryer l skid. These changes increase the air capacity of the instrument air system.

I Result: Unreviewed safety question does not exist. This modification causes no change to the normal operation of the instrument air system. It adds needed air capacity to the instrument air system, thus l

reducing the plant's reliance on service air. The filter / dryer skid will be functionally identical to the existing equipment. The probability ofinstrument air failure is not increased by the changes made since the i

' capacity of the skid is being in:reased. The consequences of a malfunction of equipment important to safety is not increased due to the capacity being increased. These changes do not adversely impact the instrument air system / service air system or functions of any system so as to create the possibility of an l accident or malfunctior: of e. type different from those already evaluated.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification l

l Evaluation Reference Number: M04-2-92-028A i

Title:

Nitrogen Inerting System

Description:

This change is a partial change to the above listed modification but has a separate safety evaluation associated with it. The work invoived includes adding two capped off piping taps to two existing l _ Nitrogen inerting lines, one above the torus and the other in the Reactor Building at Ground Floor elevation.

l Two new pipe spool pieces were routed through existing secondary containment / fire barrier penetration no.

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13. This pipe penetration provides a barrier between the U2 Reactor Building and the 1/2 Diesel Generator Room. The two new pipe spool pieces were capped offon each side of the penetration. This change l ensures that there are redundant nitrogen flow paths to the primary containment and torus free air volume, j The redundant lines ensure that nitrogen can be supplied to the containment during post LOCA accident i- conditions to control combustible gas buildup.

Result: Unreviewed safety question does not exist. The pipe taps and spool pieces are non-functional and

!' have no affect on the existing Nitrogen inerting system or secondary containment penetration. No new failure modes are created. The taps and spool pieces are designed and installed in accordance with all original design requirements to prevent a failure of existing nitrogen supply piping and components. The probability of an accident is not increased. The taps and spool p!eces do not interact with any other instrument lines. Operation or failure of these components therefore will not cause a failure ofinstrument lines that could lead to an accident.

Attachment A, SVP-98-i l3. Page 7 of 153 j

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Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-2 93 007B

Title:

Temperature Indication for RVLIS

Description:

His partial mod added temperature indication for the RVLIS instrument rack !stions.

Current indication used by emergency operating procedures comes from Standby Gas Treatment System (SBGTS). This partial adds a temperature channel to the Reactor Building Area Temperature (Leak Detection) Monitoring Panel for the QGA emergency operating procedures. The current Panalarm Annunciator Panel is replaced with a Yokogawa chart recorder and auxiliary alarm unit. This temperature indication will be used by the operator to implement QGAs for determining when the RVLIS instrumentation may be inoperable due to reference leg flashing. Reference leg flashing could cause the static head in the reference leg to change due ta water being boiled off following an accident.

Result: Unreviewed safety question de not exist. The 902-21 panel indication of RVLIS area temperatures will provide an improved area temperature indication for use with the emergency operating procedures. The new chart recorder and auxiliary alarm unit will be much easier to maintain because this model is a standard recorder type used in several applications in the Main Control Room. Spare parts are readily available. The reliability of the Reactor Building Area Temperature Monitoring system is being substantially improved. There are no adverse system interactions. The new therr ,occuple cnd conduit are seismically mounted. The chart recorder and alarm unit are seismically mounted. The changes have no impact on any equipment required to mitigate the consequences of an instrument line break outside of containment.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-1-94-007B

Title:

Core Shroud Head & Separator Assembly

Description:

Two of the four shroud head / separator asse by support legs which protrude twelve inches )

below the assembly were located such that they would intenere with the Core Shroud Repair being installed under M04-1-94-007A, To preclude this potential interference, this partial mod trimmed away the two Shroud Head and Separator assembly support legs at azimuths 103 and 283. This partial mod also provided for permanent installation of two new 3" diameter support legs at azimuths 95 degrees and at 275 degrees replacing the modified / trimmed legs. The new support legs were welded to the Separator bottom flange.  ;

To facilitate work activities, the Shroud Head was placed on four temporary aluminum stanchion supports j in the Dryer Storage Pit. The temporary stanchion supports beared directly on the Dryer Storage Pit support i ring and completely supported the Shroud Head during work activities associated with this partial mod.

(

Result: Unreviewed safety question does not exist. There is no increase in the probability of an accidem.

j The Shroud Head & Separator assembly will not rely on the modified support legs to secure the shroal head l to the core shroud nor to direct the steam-water mixture through the Separators. This assembly is l mechanically secured to the core shroud by hold down bolts around its entire circumference. Removal of  !

the existing support legs and installation of two new permanent replacement legs has no effect on nearby i safety-related equipment. Consequences of malfunctions of safey-related equipment are not affected.

1 At:achraent A, SVP-98-113, Page 8 of 153 O

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l Standpipes and steam separators are welded to the shroud head which is secured to the core shroud by hold down bolts around its entire circumference. The structural adequacy of the permanent replacement legs has been evaluated in accordance with ASME, Section 111 requirements. This demonstrates that the new leg have adequate capacity to support the separator durir.g refueling operations. Temporary aluminum stanchion supports were utilized in supporting the Separator during associated work activities. Thus, there are no adverse impacts to systems or functions so as to create the possibility of an accident or malfunction different than described in the UFSAR.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification Evaluation Reference Number: M04-2-94-007A, Revision 1 added to Addendum 1

Title:

Core Shroud Head & Separator Assembly

Description:

This partial modification trimmed away the lower portion of two lifting lugs which connect to the Shro% 4ead & Separator assemb.y. A portion of one of two attachment welds was reduced in size.

The trimming also resulted in complete removal of the lower weld between the lifting rod and the lug.

Additional welding was performed to ensure that each modified lifting lug assembly has an equivalent load carrying capacity to the existing configuration design bases. The lugs are utilized as support points when the Separator is moved into the equipment pool for storage. Temporary supports were used to support the weight of the assembly. During installation of the Shroud Head & Separator assembly an interference was discovered at the 103 and 283 azimuths. The interference was from the Shroud repair hardware installed by this modification. Two of the four Shroud Head & Separator assembly lifting lugs which protrude twelve inches below the assembly are resting on top of the shroud repair hardware.

Result: Unreviewed safety question does not exist. There is no increase in the probability of an accident.

The Shroud Head & Separator assembly will not rely on the modified lifting pins to close off the core outlet and force the ceam-water mixture through the separators. This assembly is held in place by bolts which extend from the flange to above the separators. Removal of a portion of the lifting lug eliminates interference of the !ug with the core shroud repair. Consequences of malfunctions of safety-related equipment are not affected. Standpipes and steam separators are welded to the shroud head forming an assembly which rests on .he core shroud. This assembly is hdd in rhce by bolts which extend from the flange to above the separator. The evaluation demonstates that the modified lifting lugs have adequate capacity to support the separator during lifting and moving operations. Havever, results showed that additional welding was required to ensure the modined assembly has an equivalent load carrying capacity to the existing configuration design bases. The lifting rod bracket attachment welds at these two locations have been reevaluated and are below ASME Section 111 limits. Temporary aluminum block stands were used to provide stable support of the Separator and also to protect the pool liner during the current refueling period. Thus, there are no adverse impacts to systems or functions so as to create the possibility of an accident or malfunction different than described in the UFSAR.

Safety Evaluation Number: NA Type of Safety Evaluation: Modification l Evaluation Reference Number: M04-1(2)-95-002 l

Title:

TIP Isolation Ball Valves Attachment A, SVl'-98-113, Page 9 of 153 l

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j

Description:

This modification incorporates a relay, pushbutton, and indicating light Sto both unit's Tif logic. A contact off the new relay opens when a Group 2 isolation signal is present. Upon rewiting of the isolation signal this contact remains open, preventing TIP valves from re-opening The valves may be l

opened after the new Group 2 TIP Isolation Reset pushbutton is depressed. Modi 6ed system requires at l least two operator actions before any of the subject TIP valves can be re-opened. The new indicating light l provides positive indication that the TIP isolation circuitry has been reset. These new components were  ;

installed in the 901(2)-13 panel in the Main Control Room. This modification was issued in response to a i

! degraded condition regarding Primary Containment isolation System (PCIS). Presently, if the TIP valves I are receiving an open signal, and a Group 2 isolation signal is received, the valves will cloue but may re-open following the general isolation signal reset. I Result: Unreviewed safety question does not exist Normal operation of the TIP system is not affected by i

this mod. A failure of the new relay contact in the open position would initiate an isolation of the TIP system, closing the valves to a safe position. This failure is analogous to loss of 120 VAC control power. A l failure of the contact in the closed position would allow the valves to open after a general isolation reset. l l This failure would not prevent an isolation, but would act to return the system to the current configuration.

The conclusion is no new failure mode is introduced in either case. The failure ofconcern for the l

pushbutton is for it to fail in the closed position. This would not prevent an isolation, but would act to i

return the system to the current configuration. Failure of the pushbutton in the open position simply would l i not allow a TIP reset after isolation. Conclusion is no new failure modes are introduced. This change maintains the TIP isolation valves closed following a Group 2 reset, therefore is an improvement to the j PCIS system, and consequences of an accident are decreased. l i

1 Safety Evaluation Number: NA Type of Safety Evaluation: Modification l

Evaluation Reference Number: M04-0-95-004 I

Title:

Gatehouse Security Card Readers and Hand Readers

Description:

This change involved the addition of hand profile (hand readers) and associated equipment l to the present security card reader system. Six reader units were installed in the gatehouse on the station  ;

entry card readers. These hand readers work in conjunction with the presently installed card readers to allow access to protected areas. A hand reader was also installed in the badge fabrication office to allow for the addition of hand profiles to the system as personnel are added at later times.

Result: Unreviewed safety question does not exist. Failure of the reader system will not impact plant operations, shutdowns, or create any new unit failure modes. The gatehouse security equipment is not required for safe shutdown or accident mitigation. Therefore, there is no increased risk in the probability of an accident due to this modification. Consequences of a malfunction of equipment important to safety does not exist.

Safety Evaluation Number: NA Type of Safety Evaluation: Minor Design Change Evaluation Reference Number: P04-1(2)-91-114

Title:

HPCI Motor Gear Unit Switch Attachment A, SVP-98-il3, Page 10 of 153 L- _-

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==

Description:==

This change replaced the HPCI Motor Gear Unit (MGU) Switch with a break-before-make switch, This will provide additional contacts for isolation of the MGU signal converter during manual operation of the MGU. The relay which is currently used to isolate the signal converter was eliminated.

The new switch removes the possible cross feeding of both automatic and monyl control of the HPCI MGU.

Result: Unreviewed safety question does not exist. The probability of a malfenction of equipment important to safety is decreased by this modification. The consequences of a failure of the llPCI MGU will not be affected since the operation of the MGU is not being changed. The installation of this new switch al.eviates a malfunction of cross connecting the manual and automatic feeds for the HPCI MGU controller.

No other failure modes are applicable.

Safety Evaluation Number: NA Type of Safety Evaluation: Minor Design Change Evaluation Reference Number: P04-1(2)-91-143

Title:

HVAC Ducts

==

Description:==

This change involved adding supports to the Reactor Building Supply ductwork to bring it within UFSAR allowable stresses.

Result: Unreviewed safety question does not exist. These supports do not affect the normal operation of I the Reactor Building Ventilation System. Therefore, this change does not create the possibility of an accident. This change was done because this area of the reactor building ventilation is considered to be a  ;

secondary containment boundary and must be seismically qualified. This change will bring the supports up l

to seismic qualifications. This change does not increase equipment failures, but ensures the ductwork

{

remains intact following a seismic event. j l

Safety Evaluation Number: NA- Type of Safety Evaluation: Exempt Change l

Evaluation Reference Number: E04-1-93-069

Title:

Limitorque Motor Gearing and Spring Pack for Reactor Recirculation System (RRS) Pump  :

Discharge Valves

==

Description:==

This change provides for the replacement of the Limitorque motor gearing and spring pack for the Reactor Recirculation System Pump Discharge Valves. The overall gear ratio (OAR) will change from 61.5 to 88.56 in response to a GL 89-10 design review which identified unacceptably low design margins. This will increase the valve stroke time from 29 seconds to approximately 42.6 seconds. The #14 AWG outboard cable for the 5B valve was replaced with #2 AWG cable.

Result: Unreviewed safety question does not exist. The increased valve stroke time is still within the maximum allowed stroke time of 45 seconds used in the LOCA analysis. The change will improve the reliability and motor capacity of the Limitorque actuator during postulated design basis accidents or transients. The change will not affect previously evaluated failures.

Attachment A, SVP-98-113, Page i1 of 153

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-084

Title:

High Pressure Coolant injection (HPCI) Room, Fire Suppression Piping

Description:

This change levels up HPCI Gre suppression piping, adds additional pipe supports to hold the piping level and to upgrade piping to meet NFPA 13 span criteria. The HPCI Gre suppression piping currently rests directly on the HPCI Gland Seal Leakoff (GSLO) Line. This change relocates the GSLO line to provide adequate clearance of suppression piping.

Result: Unreviewed safety question does not exist. The changes provide clearances to reduce the system interaction betwem '.ae HPCI subsystem and the Gre suppression system for the HPCI room and upgrades the piping to meet NFPA 13 span criteria. The change has no effect on the operation of any system. The new piping reroute does not significantly affect piping head losses, How characteristics, controls, access to equipment, or fire suppression nozzle spray patterns. The changes are all inside the HPCI room and do not affect any room penetrations, seals, or Secondary Containment. There are no new hazards created by the l installation of new and modiGed supports.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-113

Title:

Isolation Valves to Allow Isolation of EHC Fluid

Description:

This change involved installation of two manual isolation valves, a new local pressure indicator, and relocated four existing instrument connections upstream of the new isolation valves. This will allow isolation of the EHC fluid to the Main Stop Valves, Control Valves, Combined Intermediate Valves and front standard while EHC fluid to the Main Turbine By-Pass Control Valves (BPVs) remain unisolated. This will allow the BPVs to remain in operation while work is being performed on other EHC components. The relocation of the instrument connections will allow for continuous monitoring of the oil header pressure and proper operation of the two positive displacement pumps when the system is isolated.

The local pressure indicator is to have local verincation that the isolation valves have isolated the system prior to performing work.

Result: Unreviewed safety question does not exist. This change does not change the operating parameters of the EHC system. During normal system operation the isolation valves will be in the locked open position. Therefore, system function will remain the same and plant operation is unaffected. This change actually enhances the system, by allowing system isolation during normal operation. It does not increase the likelihood of equipment failures. This change may eliminate a potential source of failure for the EHC system by allowing the system to be worked on during normal system operation. The change enhances system performance which will result in added assurance that the affected equipment will function as designed. Therefore, this installation doa not increas- the probability of a malfunction of equipment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-131 Attachment A, SVP-98-113, Page 12 of 153

I

Title:

High Pressure Pump Seal Cooling Lines Rerouted i

Description:

This change involved rerouting the high pressure pump seal cooling lines such that the first stage booster pump will be the source of cooling water. A balancing line was added between the first stage casing and the inboard stufYing box to minimize the differential pressure between the inboard and outboard seals. Vent valves were installed on the first stage casing and each of the mechanical seal cooling lines.

The original arrangement produced high flow rates through the seals causing premature seal failure due to the seal flush recirculation header previously supplying the seal cooling water.

I Result: Unreviewed safety quotic does not exist. Flow from the RHRSW pump mechanical seals will be reduced to prevent premature seal failure. Operating modes are not changed. The function of the pumps is the same. By rerouting the seal cooling lines, seal life will be extended. There are no new types of failure modes caused by this design. The function of the RHRSW system is not affected. The probability of an accident is not increased. The RHRSW pump will be more reliable and the probability of equipment malfunction will decrease. Seal failure will be less likely to occur. The consequences of seal failum has not increased.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-145

Title:

Post-Accident Sample System

Description:

This change involved replacing the existing Whitey VREL valves with new Sentry VREL valves in the Post-Accident Sample System (PASS). The reason for the change is because the existing  ;

valves cannot provide adequate flow control to achieve the desired flow rate at the Post-Accident Sample '

System.

Result: Unreviewed safety question does not exist. This replacement will not change the operating parameters of the PASS or the normal operation of the plant. The new valves will operate in the same manner as the old valves. These valves are operated manually and do not rely on any type ofoperator that could possibly fail. Upon the extremely remote possibility that these valves do fail, they could be replaced very quickly and easily due to their location and compression-type connections. This replacement is placing the system back to its original design with respect to the Sentry type valve. The replacement of the valves does not increase the probability of a failure to the PASS. The valves do not increase the consequences of a malfunction of safety-related equipment. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-175 l

Title:

RHR Pump Discharge Pressure Switches

Description:

This change involved replacing the eight RHR Pump Discharge Pressure Switches. These switches provide an interlock into the ADS system. The existing Static-O-Ring pressure switches are l

l Attachment A, SVP-98-ll3, Page 13 of 153 i

i L

subject to excessive setpoint drift, and were replaced with new SOR pressure switches which maintain acceptable setpoint drift values.

Result: Unreviewed safety question does not exist. The function and operation of the new switches are the same as the existing switches. Plant operation is unaffected. No new failure modes are created. The new pressrre switches maintain an acceptable margin of setpoint drift as evidenced by engineering calculations.

u -hange does not change the assumptions associated with any accidents already evaluated. Enhancing system performance by providing a switch with an acceptable setpoint drift margin does not increase the 3

conscquences of an accident and actually decreases the probability of a malfunction of equipment important I to safety from occurring.

I Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-183

Title:

Reactor Recirculation Motor-Generator Set Megawatt Transducers

Description:

This change replaces the 1 A and 1B reactor recirculation motor-generator set megawatt transducers with a functionally equivalent transducer. The only major difference is that the new transducer does not have an internal power supply and, hence, requires an external 120VAC power source.

1 Result: Unreviewed safety question does not exist. The new transducers have been procured as a direct i replacement, functional equivalent, of the original components. Tb e reliability of the power level indication will be improved. The failure mode of the new transducers is the same as the old ones. The 120VAC power source is non-safety-related and adequately sized. These circuits do not interface with or affect safety-related circuitry. There is no change in plant operation.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Refereace Number: E04-1-93-185

Title:

Thermal Detectors for Unit 1 Main Transformer

Description:

This change replaces the existing thermal detectors for the Unit 1 Main Transformer with a i linear-type detection cable (Protectowire). The old detectors were a frequent source of 125 VDC grounds.

Result: Unreviewed safety question does not exist. The new detectors are less likely to become grounded and, hence, will improve system reliability and reduce the possibility ofinadvertent actuations. No new failure modes are introduced. The probability of a malfunction of fire detection equipment is reduced. The new system is operated in the same manner as the old and utilizes existing deluge piping, supports, and spray nozzles. All interfaces with existing structures, systems, and components are unchanged.

1 Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change i

Evaluation Reference Number: E04-1-93-192 Attachment A, SVP-98-ll3, Page 14 of 153

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Title:

Turbine Bypass Control Valves ,

' l i

Description:

' ~ This change removed the Fluid Jet Supply tubing from all the Turbine Bypass Control  !

. Valves (BPV) and installed a manifold assembly between each BPV and its associated servo valve. This - I change was installed to comply with the recommendations made in General Electric Technical Information i

Letter 841-3A. These recommendations were made due to a high failure rate of the Fluid Jet Supply tubmg. .;

. Result: Unreviewed safety question does not exist. This change removed the source of failure for Turbine '

- BPVs. This provided added assurance that the affected valves will function as designed. Consequences of '

an accident are not increased. The function and operation of these valves did not change. Probability of an accident is not increased. This change enhances system performance. Removing a potential failure source l

'actually' decreases the potential for failure of equipment.  !

1 l

a Safety Evaluation Number: NA ' Type of Safety Evaluation: Exempt Change Evaluation Reference Namiber: E04-1-93-212

Title:

Diesel Generator (DG) Cooling Water Flow Measurement Instrumentation

Description:

_ h This change replaces local flow instrumentation (annubar Flow Element FE l-3941-28 and  ;

Flow Indicator FI l-3941-28) which measures DG cooling water flow to the Emergency Core Cooling i

System (ECCS) room coolers, with higher range models. This allows readings to be taken under higher  ;

flow conditions. Also, a three valve manifold was added to provide a cross-connect between the high and -j low pressure side of the flow indicator to equalize DP across the cell when the flow indicator is removed from service, returned to service, and during back flushmg. i

' Result: ' Unreviewed safety question does not exist. Increasing the range will provide better monitoring of cooling water flow to the ECCS room coolers during testing. The function of the DG cooling water system 1 and interacting equipment will not change. De replacements are seismically qualified to maintain the - 'i

. integrity of the pressure boundary. The replacement will mduce the probability of equipment failures. .

Failure modes for the additional three valve manifold have been evaluated and accepted. The manifold increases reliability due to its protective capabilities during back flushing. Overall, this change will l improve the reliability of the instrumentation to operate as originally intended. The instrumentation is not j used for any safety-related applications. Original design basis requirements have been met or exceeded. '

I Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change

- Evaluation Reference Number: E04-1-93 238 - '

Title:

RHR, CS, A HPCI Indicating Lights l1

Description:

This change installed resistors in series with the indicating lights for RHR, CS, and HPCI l located on the 901-3 and 2212-4 panels. The resistors were plate mounted and attached to the existing L

j; . .  : unistrut in the referenced panels. The new resistors are 580 ohm 25 watt wire wour.d and are manufactured by Ohmite. The wiring for each affected indicating light was reconfigured to conn:ct the resistor in series -

with the indicating light on the hot side of the light. Removal of burnt out lamps has caused fault currents  !

resulting in blown fuses or damaged equipment thus rendering the subject control circuit inoperable. The -i 6, e

'E h . Attachment A, SVP-98-113, Page 15 of 153  ;

)

addition of the resistors in series with the indicating lamps prevents short circuits in the event ofindicating lamp or socket failure.

Result: Unreviewed safety question does not exist. No plant operations are affected by this change.

Addition of resistors and replacement ofindicating lamps has no impact on system interactions. The addition of resistors in series with the indicating lights decreases the probability of a malfunction of equipment. This change increases reliability of the RHR, Core Spray, and IIPCI systems without functionally changing the systems. The new resistors have no impact on system operation.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-244

Title:

Off-Cas 11old-Up Pipe Flow Elements

==

Description:==

This change replaced the existing Off-Gas hold-up pipe flow elements with new, more accurate flow elements. The existing elements do not work and are obsolete.

Result: Unreviewed safety question does not exist. The new system performs the same function as the system being replaced. The linearization circuitry for the new element is more accurate than the existing, out-dated circuit. The new elements improve system reliability and enhance measurement accuracy. The equipment is non-safety-related, classified as non-seismic and, there is no safety-related equipment near by that could be affected by failure of the elements in a seismic event. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-253

Title:

Electro-Hydraulic Control System

==

Description:==

This change involved installing a new pressure balanced Pall duplex filter assembly on the 1 A and 1B Electro-11ydraulic Control (EllC) high pressure piping system and adds a new commuter valve to each skid. The installation of the new filter assembly allows on-line filter element replacements.

Result: Unreviewed safety question does not exist. This change does not change operating parameters of the EHC system. This new filter enhances the system allowing filter replacements to be done while EllC system is in service. During normal system operation, the EHC system will not operate any differently than before. Therefore, system function will remain the same and plant cperation is unaffected. This change does not increase the likelihood of equipment failures and does not increase the probability of a malfunction of equipment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-93-306 Attachment A, SVP-98-113 Page 16 of 153

Title:

Replacement of RHRSW Hx Flow Control Valves

Description:

This change replaced the RHRSW Hx Flow Control Valves with an Anchor / Darling globe '

valve with anti-cavitation trim. The existing 15ft-lb motors, gear ratio, and thermal overload heater were replaced with a 25 ft-lb motor. The stroke length of the valve changed from 3 inches to 5.5 inches and the stroke time increased from approximately 40 seconds to approximately 87 seconds. The downstream 1

orifices were removed. Valve closure changed from limit close to torque close. The existing valves had a '

history of component problems. Replacement of the valves was determined to be more cost-effective than i

retrofitting the existing valve after refurbishment and repair.

i Result:

Unreviewed safety question does not exist. This change does not adversely affect the operation at any plant system or component. The designed function and operation of the valve does not change. It improves the valves' reliability to open and close. The throttling capability of the valve is not altered by the anti-cavitation trim nor is flow capacity reduced. The new trim eliminates cavitation induced erosion damage and reduces RHRSW piping vibration, thus reducing equipment failures. The increased stroke time improves the margin in the thrust windows for the valves. No new failure modes are created.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change i Evaluation Reference N=.ber: E04-2-93-113

Title:

Isolation Valves to Allow Isolation of EHC Fluid i

Description:

This change involved installation of two manual isolation valves, a new local pressure indicator, and relocated four existing instrument connections upstream of the new isolation valves. This will allow isolation of the EHC fluid to the Main Stop Valves, Control Valves, Combined Intermediate Valves and front standard while EHC fluid to the Main Turbine By-Pass Control Valves (BPVs) remain unisolated. This will allow the BPVs to remain in operation while work is being performed on other EHC components. The relocation of the instrument connections will allow for continuous monitoring of the oil header pressure and proper operation of the two positive displacement pumps when the system is isolated.

The local pressure indicator is to have local verification that the isolation valves have isolated the system prior to performing work.

Result: Unreviewed safety question does not exist. This change does not change the operating parameters of the EHC system. During normal system operation the isolation valves will be in the locked open position. Therefore, system function will remain the same and plant operation is unaffected. This change actually enhances the system by allowing system isolation during normal operation. It does not increase the likelihood of equipment failures. This change may eliminate a potential source of failure for the EHC system by allowing the system to be worked on during normal system operation. The change enhances system performance which will result in added assurance that the affected equipment will function as designed. Therefore, this installation does not increase the probability of a malfunction of equipment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-131

Title:

High Pressure Pump Seal Cooling Lines Rercuted l Attachment A, SVP-98-113, Page 17 of 153 l

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Description:

This change involved rerouting the high pressure pump seal cooling lines such that the first stage booster pump will be the source of cooling water. A balancing line was added between the first stage casing and the inboard stuffing box to minimize the differential pressure between the inboard and outboard seals. Vent valves were installed on the first stage casing and each of the mechanical seal cooling lines.

The original arrangement produced high flow rates thrvugh the seals causing premature seal failure due to i the seal flush recirculation header previously supplying the seal cooling water.

Result: Unreviewed safety question does not exist. Flow to the RHRSW pump mechanical seals will be reduced to prevent premature seal failure. Operating modes are not changed. The function of the pumps is ,

the same. By rerouting the seal cooling lines, seal life will be extended. There are no new failure modes l

caused by this design. The function of the RHRSW system is not affected. The probability of an accident is '

not increased. The RHRSW pump wi!! be more reliable and the probability of equipment malfunction will decrease. Seal failure will be less likely to occur. He consequences of seal failure has not increased. l' Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-145

Title:

Post-Accident Sample System

Description:

This change involved replacing the existing Whitey VREL valves with new Sentry VREL valves in the Post-Accident Sample System (PASS). The reason for the change is because the existing valves cannot provide adequate flow control to achieve the desired flow rate at the Post-Accident Sample System.

Result: Unreviewed safety question does not exist. This replacement will not change the operating parameters of the PASS or the normal operation of the plant. The new valves will operate in the same

manner as the old valves. These valves are operated manually and do not rely on any type of operator that could possibly fail. Upon the extremely remote possibility that these valves do fail, they could be replaced very quickly and easily due to their location and compression type connections. This replacement is placing the system back to its original design with respect to the Sentry type valve. The replacement of the valves does not increase the probability of a failure to the PASS. The valves do not increase the consequences of a malfunction of safety-related equipment. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-174

Title:

Drywell Pressure Switch Replacements i

Description:

This change replaced sixteen static-o-ring drywell pressure switches which were subject to I

excessive setpoint drift. The replacement switches demonstrate an acceptable margin of setpoint drift per the setpoint error analysis provided in NDIT No. EIC-94-013-2 prepared by Comed Nuclear Engineering Department. These switches have also demonstrated the ability to maintain acceptable setpoint values in their Unit 1 application.

Attachrnent A SVP-98-113, Page 18 of153

1 l

l Result: Unreviewed safety question does not exist. The function and operation of the switches remains the I same, liowever, the ability to maintain acceptable setpoint drift values increases. The replacement I switches will not affect any interactions with other structures, systems, or components. No ne v failure modes are created nor do the switches affect any equipment failures. A setpoint error analysis has been performed. The quality and reliability of the switches has been enhanced. There is a decrease in the probability of a malfunction of equipment important to safety by providing acceptable setpoint drift of the switches.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-175 I l

l

Title:

RHR Pump Discharge Pressure Switches l

Description:

This change involved replacing the eight RHR Pump Discharge Pressure Switches because of excessive drift. These switches provide an interlock into the ADS system. The existing Static-O-Ring j pressure switches are subject to excessive setpoint drift, and were replaced with new SOR pressure switches which maintain acceptable setpoint drift values. In addition, junction boxes were added to accommodate installation of terminal blocks.

i Result: Unreviewed safety question does not exist. The function and operation of the new switch are the '

same as the existing switches. Plant operation is unaffected. No new failure modes are created. The new pressure switches maintain an acceptable margin of setpoint drift as evidenced by engineering calculations.

This change does not change the assumptions associated with any accidents already evaluated. Enhancing system performance by providing a switch with an acceptable setpoint drift margin does not increase the consequences of an accident and actually decreases the probability of a malfunction of equipment important to safety.

l l

l Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-184

Title:

RHRSW Sump Pump Discharge Check Valves I

Description:

This change involved replacing the RilRSW sump pump discharge check valves located in the RHR Service Water vault sump pump discharge headers. 'Ile replacement Mission DUO-CHECK check valves have two hinged plates which provide a less torturous flow path through the valve. This design will minimize collection of debris in the seat area of the valve and thereby reduce the likelihood of leakage past the valve seat. This modification also installed stainless steel wire mesh cloth around the outside of each sump pump intake strainer which will prevent pieces of wood and other debris from being pumped into the piping and getting trapped in valves and other components. The existing valves have exhibited significant valve corrosion.

Result: Unreviewed safety question does not exist. The failure modes of the new valves are the same as for the existing valves. Therefore, consequences of an equipment malfunction are the same. Reliability of l the check valves will be increased since it will be less likely that debris will collect and allow leakage past l the valve seats. With the valve hinge pin positioned in the vertical direction, the resistance of the new Attachment A, SVP-98-113, Page 19 of 153

valves is less than the existing valves. Process parameters are not being altered and operating modes are not being changed. Derefore, there are no new accidents or failure modes created by this design.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-187

Title:

Reserve Auxiliary Transformer nermal Detectors

Description:

This change replaced the existing thermal detectors for the Unit 2 Reserve Auxiliary Transformer with a linear-type detection cable. The existing detectors have proven to be a frequent source of 125 VDC grounds. The new detectors improve system reliability and reduce the possibility of inadvertent actuations.

Result: Unreviewed safety question does not exist. The new system performs the same function as the system being replaced. No new failure modes are introduced. All interfaces with other systems, structures, and components are unchanged. Failure modes of new equipment are identical to ecyipment being replaced. The new detectors are less likely to become grounded thus improving the reliability of the system.

The probability of a malfunction of fire detection equipment is actually reduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-192

Title:

Turbine Bypass Control Valves

Description:

This change removed the Fluid Jet Supply tubing from all the Turbine Bypass Control Valves (BPV) and installed a manifold assembly between each BPV and its associated servo valve. This change was installed to comply with the recommendations made in General Electric Technical Information Letter 841-3A. Dese recommendations were made due to a high failure rate of the Fluid Jet Supply tubing.

Result: Unreviewed safety question does not exist. This change removed the source of failure for Turbine BPVs. This provides added assuraN that the affected valves will function as designe' : ' sequences of an accident are not increased. The function and operation of these valves did not ch ..e. Probability of an i accident is not increased. This change enhances system performance. Removing a potential failure source actually decreases the potential for failure of equipment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-207

Title:

Lateral Supports for RiiRSW Pump Crossover Pipes

Description:

This change installs a lateral support at the mid-span of each RiiRSW pump's crossover pipe to reduce pump vibration during operation, it also adds spring type supports to each suction and discharge pipe to funher reduce the pump nozzle loads and provide a mechanism to allow minor Attachment A, SVP-98-113, Page 20 of 153 l

t l

l

adjustments in the piping for pump alignment during maintenance activities. It also deletes "up-lift" supports associated with the cross over lines which were determined to be not required.

Result: Unreviewed safety question does not exist. Reconfiguration of the supports increases the reliability of the RHRSW pump and piping system by reducing pump vibrations and nozzle loads. No new failure modes are created. All pipe supports are seismically designed. Process parameters have not been altered.

Primary and secondary containment is not affected. No penetrations, seals, or doors are modified.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-245

Title:

New 480 VAC Panel

Description:

This change replaced the existing panel with a 10 circuit distribution panel and installed six  ;

60 amp welding receptacles powered from this new panel. Five receptacles were mounted on the outside of l

the turbine shield wall west of the new circuit panel. A sixth was installed on the inside of the turbine shield j

wall. This new configuration provides a safer and more efficient means for providing power on the turbine deck.

1 Result: Unreviewed safety question does not exist. This change does not electrically interact with plant I equipment. Calculations were performed to determine that structural loads are acceptable. Therefore, this change does not affect equipment failures. The addition of a circuit panel and welding receptacles increase equipment reliability over the existing configuration. Therefore, failure mode of the 480 VAC panel is lessened in severity. The probability of a deaign basis accident is unchanged. The change does not affect equipment important to safety, therefore consequences of an accident are not changed.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2 93-253

Title:

Electro-Hydraulic Control System

==

Description:==

This change involved installing a new pressure balanced Pall duplex filter assembly on the 2A and 2B Electro-Hydraulic Control (EHC) high pressure piping system and adds a new commuter valve to each skid. He installation of the new filter assembly allows on-line filter element replacements.

Result: Unreviewed safety question does not exist. This change does not change operating parameters of the EHC system. This new filter enhances the system by allowing filter replacements to be done while EHC system is in service. During normal system operation, the EHC system will not operate any differently than before. Therefore, system function will remain the same and plant operation is unaffected. This change does not increase the likelihood of equipment failures and does not increase the probability of a malfunction ofequipment.

Sxfety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Attachment A, SVP-98-ll3, Page 21 of 153

Evaluation Reference Number: E04-2-93-306  !

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Title:

Replacement of RilRSW IIx Flow Control Valves

Description:

This change replaced the RHRSW Hx Flow Control Valves with an Anchor / Darling globe valve with anti-cavitation trim. The existing 15ft.lb motors, gear ratio, and thermal overload heater were replaced with a 25 ft-lb motor. The stroke length of the valve changed from 3 inches to 5.5 inches and the stroke time increased from approximately 40 seconds to approximately 87 seconds. The downstream orifices were removed. Valve closure changed from limit close to torque close. The existing valves had a history of component problems. Replacement of the valves was determined to be more cost-effective than retrofitting the existing valve after refurbishment and repair.

Result: Unreviewed safety question does not exist. This change does not adversely affect the operation of any plant system or component. The designed function and operation of the valve does not change. It improves the valves' reliability to open and close. The throttling capability of the valve is not altered by the anti-cavitation trim nor is flow capacity reduced. The new trim eliminates cavitation induced erosion damage and reduces RHRSW piping vibration, thus reducing equipment failures. The increased stroke time improves the margin in the thrust windows for the valves. No new failure modes are created.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-93-307

Title:

Inboard Suppression Pool Coolingrfest Valves

Description:

This change upgrades or replaces components of the Inboard Suppression Pool Cooling / Test Valves (M02-1001-36A) to reduce cavitation induced valve and piping vibration and improve overall operation and capabilities of the valves.

Result: Unreviewed safety question does not exist. This change will not create new failure modes. This change will enhance the effectiveness of the valves by reducing the potential for vibration-related component failures. The new components are seismically and structurally qualified and actually increase l the seismic and structural capabilities of the valves. The additional valve weight has been analyzed and l accepted. An evaluation of this change concluded that it does not cause Emergency Core Cooling System I (ECCS) pump suction plugging concerns. The valve is not being modified in function or application.

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! Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-94-117

Title:

Diesel Generator Cooling Water Pump Discharge Header Pressure Sensing Line Dercription: This change replaced the 3/4" angle globe valve with a new 3/4" globe valve and associated instrument tubing, the instrument isolation valve, a new 3/8" instrument calibration connection tee, and a new snubber fitting for pressure indicator. The valve is located in the pressure indicator's sensing line, which comes directly off the Diesel Generator Cooling Water Pump discharge line. The valve was frozen open. This is a normally open valve. The problem arose when maintenance performed a calibration on the Attachment A, SVP-98-il3, Page 22 of 153

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l pressure indicator. The root valve, along with the existing instrument valve, is used to provide isolation of i the sensing line during calibration.

]

i Result: Unreviewed safety question does not exist. This change does not modify the operation of the safety-related diesel generator cooling water pump (DGCWP) or the function of monitoring DGCWP discharge pressure. The only safety-related function the new components are required to perform is maintaining pressure boundary. The new sensing line and associated components are designed to maintain the pressure boundary during all operating modes. This change does not introduce any new failure modes.

The entire sensir line has been seismically qualified in order to maintain structural and pressure integrity during a seismic event. The system will function the same as previously. Consequences of a malfunction of equipment important to safety is unchanged.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change l Evaluation Reference Number: E04-0-94-126 '

Title:

Intermediate Radwaste Storage Facility (IRSF) Crane

Description:

This change installs a new keylock switch to provide for locking out the crane's rotating feature in order to prevent misorientation ofloads which could lead to grapple deformation and load drop.

Result: Unreviewed safety question does not exist. The new keyed switch reduces the probale ity of a container drop accident due to misorientation. The crane's rotating feature is not required for moving radwaste cylinders. Crane operation is otherwise unaffected. No new failure modes are introduced. There is no change in amount of radioactivity that can be stored in a container. There is no affect on the potential off-site dose consequences resulting from a container drop accident .

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Refern.ce Number: E04-0-94-192

Title:

Plant Radio System Anten ta

Description:

This change will add one additional transmitting antennae to enhance the coverage of the plant radio system and add 8 new non-transmitting antenna to propagate radic signals into selected areas of the plant that cannot be covered by any transmitting antenna.

Result: Unreviewed safety question does not exist. Installation of the new antennas will enhance radio coverage for the plant, eliminating existing " dead spots". A Motorola survey has shown that no equipment in the vicinity of the new atennas will be affected by RFI or EMI. The new antennas do not interact with any safety related equipment or safety barriers.

Safety Evaluation Number: NA Type of Safety Evak aa: Exempt Change Evaluntion Reference Number: E04-0-94-211 l Attachment A, SVP-98-113, Page 23 of 153 l

Title:

Replace Purgemaster Meter in the Radioactive Waste System

Description:

This change replaced the existing purgemaster meters on the floor drain filter, spare floor drain filter and waste collector filter. The existing throttle valves internal to the purgemaster meters on the floor drain filter would not fully close.

Result: Unreviewed safety question does not exist. This change of purgemaster meters will not functionally change the way the radioactive waste system operates. The function of the replacement meter is identical to the existing meter. No system interactions or operating modes will be affected. No new failure modes are created. The system and its components are non-safety-related and non-seismic. The probability of an equipment malfunction important to safety will not increase.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-94-225

Title:

Hazard and Alarm Relays for Emergency Diesel Generators

Description:

This change replaced the Hazard and Alarm relays for all three Emergency Diesel Generators (EDG) with seismically qualified relays. This significantly reduces the system's susceptibility to seismically induced contact " chatter" and eliminates operator actions to restore the EDG HVAC following a seismic event. A PIF investigation had determined that several relays located in the CO2 Fire Protection System could disable the HVAC for all three EDGs during and after a seismic event. This problem had been identified at Cooper Nuclear Station.

Result: Unreviewed safety question does not exist. This change does not affect plant operations. The replacement relays are identical in function to the original relays. However, the new relays are also seismically qualified so that they will function during and after a seismic event. No new failure modes are introduced. The failure mode of the new relays is the same as fbr the original relays. The new relays increase the reliability of the C02 suppression system for the EDGs by reducing the susceptibility to unexpected operation due to chatter. This makes the EDG HVAC systems more reliable because they are less likely to be tripped by spurious operation of the CO2 system. Therefore, the consequences of the accident are reduced due to increased reliability of the EDGs.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-002

Title:

Reactor Building Equipment Drain Tank System

Description:

This change installed clean out ports in various locetions of the Reactor Building Equipment Drain Tank system, the Reactor Building Floor Drain system, the Clean Up Phase Separator Tank Drain line and a floor drain line running from the second floor of the Reactor Building. This change allows for hydrolasing. These systems have developed high radiation hot spots in high traffic areas of the Reactor Building. The hydrolasing will reduce the high radiation hot spot areas.

Attachment A, SVP-98-113, Page 24 of 153

Result: Unreviewed safety question does not exist. The addition of the clean out ports does not have any operating affect on any of the drain systems. This change does not alter how the system will function during operation. The drain line systems do not have any interfaces with any systems required for safe shutdown of the plant. This includes any safe shutdown equipment or components and any containment systems. The drain systems do not have any direct operating impact on any safety-related systems. The installation of the clean out ports meets all design criteria of the referenced systems. Therefore, any changes to these drain systems do not increase the probability of an accident.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-070

Title:

Residual lleat Removal (RHR) Pump Minimum Flow Valves

Description:

This change replaces the 27.99 Overall Ratio (OAR) Limitorque Motor Gearing on the RHR Pump Minimum Flow valves (M01-1001-18A(B)) with a 47.85 OAR to increase the motor gearing capacity in the opening and closing directions which will result in a larger thrust window for the valves.

Result: Unreviewed safety question does not exist. This change will improve the reliability of the valves to open and close. The resultant longer stroke time (additional 12 seconds) is acceptable because there are no Tech Spec or UFSAR requirements for maximum stroke time and because the change will not reduce the capability of RHR to perform its design function. No new failure modes are created. Pressure boundary integrity is maintained. The affects of changes in weight and center of gravity on the pipe stress analysis maintain the system within code allowables.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-113

Title:

MSIV Guide Liners

==

Description:==

This change provided for corrective actions for both the Main Steam Isolation Valve (MSIV) guide liners and the MSIV actuator springs on the inboard and outboard isolation valves. This change provided for an alternate method to retain and position the lower guide liner. This redesign involved new, modified upper ar d lower guide liners which will be mechanically interlocked, utilizing locating pins to prevent rotation, and a belleville spring between the seat ring bore and the lower liner to prevent vertical movement of the lower liner during plant operation. The upper liner was supplied with an oversized O.D.

for field fitting. The lower liner was supplied with an oversized O.D. to allow for field fitting which may require reboring the valve body to obtain the required concentricity. The lower liners will have beveled lead-ins on all four ports, eliminating the need for specific orientation after the initial installation. The lead-in bevels reduce flow turbulence through the valves.

Result: Unreviewed safety question does not exist. Installation of the newly designed MSIV guide liners and actuator springs will not affect plant operation when the valve functions as designed. This design provides improvement to the valves' operation in that it prevents the lower liner from rotating which caused panial flow pon blockage and the new actuator springs are designed so that they will not experience excessive stress due to compression. These improvements allow the valves to function as designed and all Attachment A, SVP-98-113, Page 25 of 153

system interactions remain as they are presently. Consequences of an accident will not increase. The probability of a malfunction ofequipment important to safety will not increase since the change provides MSIVs that will function as designed with no change in system interactions.

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Safety Evaluation Number: NA Type of Safety Evaluatie=: Exempt Change

.l Evaluation Reference Number: E04-1-94-117

Title:

Diesel Generator Cooling Water Pump Discharge Header Pressure Sensing Line l

Description:

This change replaced the 3/4" angle globe valve with a new 3/4" globe valve and associated instrument tubing, the instrument isolation valve, a new 3/8" instrument calibration connection tee, and a new snubber fitting for pressure indicator. The valve is located in the oressure indicator's sensing line, which comes directly off the Diesel Generator Cooling Water Pump discharge line. The valve was frozen open, which from a system operation standpoint is acceptable, since this is a normally open valve. The problem arose when maintenance performed a calibration on the pressure indicator. The root valve along with the existing instrument valve, is used to provide isolation of the sensing line during calibration.

Result: Unreviewed safety question does not exist. This change does not modify the operation of the safety-related diesel generator cooling water pump (DGCWP) or the function of monitoring DGCWP discharge pressure. The only safety-related function the new components are required to perform is maintaining pressure boundary. The new sensing line and associated components are designed to maintain the pressure boundary during all operating modes. This change does not introduce any new falhwe modes.

The entire sensing line has been seismically qualified in order to maintain structural and pressure integrity during a seismic event. The system will function the same as previously. Consequences of a malfunction of equipment important to safety are unchanged.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-138

Title:

Drilling Hole in HPCI Injection Valve Disc to Reduce the Potential for Bonnet Over-Pressurization

Description:

This change drilled a hole in the feedwater side of the valve disc for the HPCI Injection Valve. Drilling a hole in the feedwater side of the disc permits the pressure in the bonnet to equalize with the piping system, thereby reducing differential pressure and eliminating pressure locking susceptibility.

Pressure locking can occur in flexible-wedge disc valves when fluid becomes pressurized within the valve bonnet and the actuator is not capable of overcoming the additional thrust requirements resulting from the j differential pressure created across both valve discs by the pressurized fluid in the valve bonnet.

Result: Unreviewed safety question does not exist. As discussed in SOER 84-7 and Comed Study Report, the drilled hole will improve the reliability of the valve by preventing any potential trapped liquid binding.  ;

- This will improve the reliability of the valve to perform its function to open for emergency core cooling and i to isolate the HPCI system from the feedwater system. This change does not alter the function of the valve.

It is entirely internal to the valve and does not reduce the valve's capability to stroke or maintain the pressure boundary. The valve disc is not part of the system pressure boundary and the ability to isolate the  !

Attachment A, SVP-98-113, Page 26 of 153  !

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reactor coolant has not been reduced. No new interfaces are created betwecn the valve and other systems.

The change does not create any new failure modes or system interfaces for the valve.

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t Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-171 1

Title:

RCIC Injection Valve

Description:

This change drilled a hole in the feedwater side of the valve disc for the RCIC injection valve. This modification is deemed a high priority design change based on the valve function, the change in ambient temperature near the valve following an accident, Generic Letter 89-10 requirements for motor-l operated valve testing, and its Risk Achievement Worth. A study evaluated various ways to prevent i

pressure lock binding of the valve. Trapped liquid binding of a gate valve can occur when a large differential pressure is created between fluid in the valve bonnet and fluid in the process pipe. This binding can cause the vdve to fail to open due to the differential pressure applying a force to hold the valve disc in the closed position. This pressure can be created when fluid trapped inside the bonnet is heated due to  ;

increasing ambient temperature. Drilling a hole in the valve disc is the recommended solution for this '

problem. This will equalize pressure between the valve bonnet and process pipe on the feedwater system side of the gate valve.

Result: 'Unreviewed safety question does not exist. As discussed in SOER 84-07 and the Study Report, the drilled hole will improve the valve reliability by preventing any potential trapped liquid binding. This will improve the reliability of the valve to perform its function to open for emergency makeup water flow in the event there is a loss of normal feedwater. The required isolation function is for flow from feedwater (and reactor) to RCIC. In this flow direction, the valve will be leaktight. Drilling the hole in the disc does not create any new failure modes of the valve. It does change the valve from a bi-directional valve to a uni-directional valve. Reverse flow, from RCIC to feedwater can still occur. When RCIC is running and the injection valve is closed, there will be a small leakage through the valve. This flow rate is limited by the size of the hole and the differential pressure across the valve. There are no known post-accident scenarios where this small ieakage could pose a safety hazard. These effects are not considered significant and are more than offset in terms ofimpact on safety by the improvement in valve reliability. This change will make the RCIC injection valve less vulnerable to failure due to pressure locking. The changes are entirely internal to the valve so there are no new interfaces created between the valve and any other systems.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-186

Title:

Reactor Recirculation Pumps

Description:

This change installed a vibration monitoring cabinet on elevation 595' of the reactor l

building and provided 120V AC power from panel RLC-8 to the vibration monitoring cabinet. An accelerometer and proximity probes were also installed on the reactor recirculating pumps and motors. The change installed panels in the drywell to mount the probe transmitters, and routed cables from the transmitter panels through drywell penetration X-1028 to the monitoring equipment in the vibration monitoring cabinet. This equipm.ent will record the change in vibration over time which will indicate l Attachment A, SVP-98-113, Page 27 of 153 t

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condition of each motor, pump, and shaft. Based on this data, the equipment may be replaced or repaired prior to a failure occurring.

Result: Unreviewed safety question does not exist. This change will not atTect any existing failure modes nor will it produce any new failure modes of the reactor recirculation pump, shaft, or motor. This system is strictly a monitoring system.' It does not interface with the control scheme of the existing equipment nor does it affect any existing plant safety equipment. Therefore, there is no affect on the consequences of an accident.

Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change Evaluation Reference Number: E04-1-94-193

Title:

Main Condenser Anti-Vibration Stakes

Description:

This change installed approximately 10,200 Cradle-Lock anti-vibration tube stakes in the Unit One Main Condenser. This eliminates vibration-induced tube cracks. Tube vibrations occur during periods when circulating water temperatures are low. This results in lower backpressure and a higher steam specific volume causing an increase in steam velocity. Increase in steam velocity results in vibration of the condenser tubes.-

Result: Unreviewed safety question does not exist. Plant operation is unaffected. The Main Condenser will perform the same design function as before. The stakes do not alter the design function of the Main Condenser. All interfaces with other systems, structures, and components are unchanged. No new failure modes are introduced. The anti-vibration stakes are installed between the condenser tube rows making the tubes more rigid, thus, preventing vibration induced tube cracks. There is no increased probability for any accident to occur. Probability of a malfunction of equipment important to safety is not affected. The Main Condenser is not required fo'r safe shutdown of the plant or to perform any reactor safety function.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation . Reference Number: E04-1-94-195

Title:

1B2 Low Pressure (LP) Feedwater Heater

Description:

This change repaired the 182 LP Feedwater Heater Shell. The shell was repaired by welding a structural plate to the outside of the heater shell. A stainless steel liner conforming to the shape of the shell was placed between the original shell and new plate. The addition of the new plate is designed to meet or exceed original design requirements of the feedwater heater and is a permanent repair. Repairs

' were necessary due to wall thinning caused by flow-assisted corrosion. The thinning was discovered during ultrasonic inspection.

Result: Unreviewed safety question does not exist. There are no new failure modes created. The new repair involved welding a structurally qualified steel plate to the outside of the shell that enables the Feedwater heater to perform its design function for the remainder of plant life. Installation of this change restores the heater shell to original design margins and meets or exceeds the requirements of the original design and code of construction. No active components are added to the Feedwater system that would Attachment A, SVP-98-113, Page 28 of 153

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increase the consequences of a malfunction of equipment important to safety. The repair does not change or  !

alter the function of the Feedwater heaters or the Feedwater system.

I Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change 1

Evaluation Reference Number: E04-1-94-207

Title:

Emergency Diesel Generator Neutral Grounding Transformer

Description:

This change will replace the existing oil filled Emergency Diesel Generator Neutral Grounding Transformer with a General Electric (GE) model 9T28YS601 dry type transformer as part of the PCB reduction program.

Result: Unreviewed safety question does not exist. GE has recommended this transformer as an acceptable replacement for the existing transformer. The replacement transformer meets the seismic and environmental requirements and has the same electrical characteristics as the existing transformer. No new failure modes are created. There will be no affect on the operation of the diesel generator.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-217

Title:

Tool Cage and Storage Cabinets in the South End of the Unit 1 Turbine Building

Description:

This change installs a tool cage and storage cabinets in the south end of the Unit 1 Turbine Building.

Result: Unreviewed safety question does not exist. This change will meet an identified need for a centralized tool issue and storage facility inside the Radiological Protected Area (RPA). There is no safety-related equipment in the nearby area. This change will not affect existing non-safety-related equipment in the area. This change does not modify plant systems or equipment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-236

Title:

New Power Feed to New Maintenance Tool Crib

Description:

This change installed a new power feed to the new maintenance tool crib. Power was supplied from MCC 16-3-1. A 208 VAC 50A main feed breaker was installed in compartment B3. This feeds new regular lighting cabinet in the too! crib area. It provides 120 VAC power to lighting, outlets and l other miscellaneous equipment including weld rod ovens. The new tool crib area does not have adequate lighting and power available.

Result: Unreviewed safety question does not exist. No equipment failures are affected by this change. The equipment is riew and not replacing any existing equipment. The feeder breaker will protect the MCC from Attachment A, SVP-98-113, Page 29 of 153 I

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an overload /short circuit condition. No accident scenarios apply. Power to the tool crib area will not affect any design basis accidents. This power is non-safety-related and will not affect any class 1E circuits. No equipment utilized to mitigate the consequences of a design basis accident will be powered from this area nor will it affect any equipment that is important to safety.

l Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-237

Title:

Containment Cooling (Drywell Spray) Isolation Valves

Description:

This change installs a 1/4" diameter, l'0" long (max), pipe and cap into the bonnet of the Residual Heat Removal (RHR) system Containment Cooling (Drywell Spray) Isola *. ion Valves (1-1001-26A&B) to provide a method for local leak rate testing.

Result: Unreviewed safety question does not exist. By providing a method for local leak rate testing of these valves, this change will help eliminate the need for an integrated leak test on the primary containment.

This change is being installed per applicable codes and specifications and will not affect RHR system function or interactions with other systems. This change does not increase the probability that the valves will fail, one new failure mode was evaluated and accepted. Containment Cooling would function as designed even if the installed pipe were to fail completely. The amount ofcontaminated water that would result from such a failure would be minimal, have no adverse affect on other equipment, and be within the boundary of the secondary containment.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-244, Addendum 1

Title:

Residual Heat Removal Service Water (RHRSW) Flow Reversing Valve

==

Description:==

This change replaces the existing valve (MO l-1001-187B) with a new Anchor Darling Gate Valve. The new valve will require a gearing change in the actuator to maintain a comparable stroke time (the existing valve has a stroke time of 65 seconds, the replacement valve will have a stroke time of 58 seconds). The existing spring hanger associated with this valve is also being modified to ensure that the piping is adequately supported.

Result: Unreviewed safety question does not exist. The new valve is lighter, which will allow analyzed piping stresses to be reduced to within allowable values. This change will improve the valves reliability.

The change will not reduce the capability of the valve to stroke, maintain pressure boundary, or support the operation of the RHR system under accident conditions. This change does not affect the function and operation of the valve, the RHRSW system, or other interactions with RHR or any other system. The failure modes for the new gate valve are the same as for the old gate valve. The new valve has been seismically qualified.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Attachment A, SVP-98-113, Page 30 of 153

Evaluation Reference Number: E04-1-94-245 -

Title:

Residual Heat Removal Service Water (RHRSW) Flow Reversing Valve

Description:

This change replaces the existing valve (MO l-1001-185B) with a new Anchor Darling Gate Valve. The new valve will require a gearing change in the actuator to maintain a comparable stroke time (th; existing valve has a stroke time of 65 seconds, the replacement valve will have a stroke time of 58 seconds). The existing spring hanger associated with this valve is also being modified to ensure that the

. piping is adequately supported.

ResuN Unreviewed safety question does not exist. The new valve is lighter, which will allow analyzed piping stresses to be reduced to within allowable values. This change will improve the valves reliability.

The change will not reduce the capability of the valve to stroke, maintain pressure boundary, or support the operation of the RHR system under accident conditions. This change does not affect the function and operation of the valve, the RHRSW system, or other interactions with RHR cr any other system. The failure modes for the new gate valve are the same as for the old gate valve. The new valve has been seismically qualified.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-246

Title:

Residual Heat Removal Service Water (RHRSW) Flow Reversing Valve

Description:

This change replaces the existing valve (MO l-1001-186B) rMi a new Anchor Darling Gate Valve. The new valve will require a gearing change in the actuator : Tin a comparable stroke time (the existing valve has a stroke time of 65 seconds, the replacement ' N * - have a stroke time of 58 seconds). The existing spring hanger associated with this valve is also l> % r fied to ensure that the pipmg is adequately supported.

Result: Unreviewed safety question does not exist. The new valve is lighter, which will allow analyzed l

piping stresses to be reduced to within allowable values. This change will improve the valves reliability.

The change will not reduce the capability of the valve to stroke, maintain pressure boundary, or support the operation of the RHR system under accident conditions. This change does not affect the function and operation of the valve, the RHRSW system, or other interactions with RHR or any other system. The failure modes for the new gate valve are the same as for the old gate valve. The new valve has been seismically qualified.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-94-247

Title:

Residual Heat Removal (RHR) Suppression Pool Suction Valves

Description:

This change replaces existing Crane 14",150# gate valves (M01-1001-7A,B,C) with Quality Assured Products wedge-type gate valves. This modification will also provide for the replacement Attachment A, SVP-98-113, Page 31 of 153

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of the Overall Gear Ratio (OAR) for M01-1001-7A only (giving it the same OAR as the other two) and the motor for M01-1001-7C only (like-for-like).

{

kesult: Unreviewed safety question does not exist. This change will eliminate valve internal leakage, increase the available thrust window, and improve the operation and reliability of these valves. The i seismic, structural and torque capabilities of the valves are being enhanced. The replacements have been qualified by weak link and seismic reports. No additional failure modes are introduced. RHR system requirements will be maintained.

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Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-94-002

Title:

Reactor Building Equipment Drain Tank System

Description:

This change installed clean out ports in various locations of the Reactor Building Equipment Drain Tank system and the Reactor Building Floor Drain system. This change allows for hydrolasing. These systems have developed high radiation hot spots in high traffic areas of the Reactor )

Building. The hydrolasing will reduce the high radiation hot spot areas.

Result: Unreviewed safety question does not exist. The addition of the clean out ports does not have any operating affect on any of the drain systems. This change does not alter how the system will function during operation. The drain line systems do not have any interfaces with any systems required for safe j shutdown of the plant. This includes any safe shutdown equipment or components and any containment  !

systems. The drain systems do not have any direct operating impact on any safety-related systems. The installation of the clean out ports meets all design criteria of the refenneed systems. Therefore, these changes to these drain systems do not increase the probability of an accident.

l Safety Evaluation Number: NA Type of bafety Evaluation: Exempt Change Evaluation Reference Number: E04-2-94-007

Title:

Instrument Air Supply Line

==

Description:==

This change installs one additional support on an instrument air supply line (2-47138-1/2"-

H) located in the Unit 2 Torus.

Result: Unreviewed safety question does not exist. This change improves the support condition of the instrument air line and increases its reliability. It facilitates the upgrading of the line to safety-related status which is required by reanalysis of the affects of Torus shell hydrodynamic loads and pool swell loads. The function of the line will not change. No additional interactions with other systems or new failure modes will be created.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-94-071 Attachment A, SVP-98-113, Page 32 of 153

Title:

Limitorque Motor Gearing for Residual Heat Removal (RHR) Cross-tie valves

Description:

This change replaces the 72.42 Overall Ratio (OAR) Limitorque Motor Gearing on the .

RHR Cross-tie valves (M02 1001-19A(B)) with a 88.40 OAR to increase the motor gearing capacity in the -

opening and closing directions which will result in a larger thrust window for the valves; Result: Unreviewed safety question does not exist. His change will improve the reliability of the valves to open and close. The change will not reduce the capability of RHR to perform its design function. ' No new failure modes are created. Pressure boundary integrity is maintained. The effects of changes in weight and center of gravity on the pipe stress analysis maintain stresses within code allowables.

Safety Evaluation Number: NA' Type ofSafety Evaluaticn: Exempt Change

. Evaluation Reference Number: . E04-2-94-118

Title:

High Pressure Coolant Injection (IIPCI) Piping Supports

Description:

This change cuts piping associated with the HPCI Pump to determine cold spring (possible source of pump vibration), installs flanges for ease of future maintenance, installs 5 new supports (3 on discharge line and 2 on supply line) to reduce nozzle loads, and demolishes one existing discharge support -

to accommodate the flange.

L

' Result: Unreviewed safety question does not exist. The additional supports and flange make the pumps -

more reliable since they reduce vibration and lower the stresses to which the equipment will 1.m subject. The i function of the HPCI system is not affected. No new failum modes will be introduced. The flange is rated

[

for the same operating characteristics as the discharge piping.

- Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Refe.vace Number: E04-2-94-128

Title:

Residual Heat Removal (RHR) Containment Isolation Valves

Description:

This change replaces the valves' Limitorque operator, yokes, discs, stems, stem nuts and L

power cables. It also installs new thermal overload relays / heaters and changes the circuit breaker settings in the Motor Control Centers (MCCs) to accommoda'e t the larger motors.

Result: Unreviewed safety question does not exist. These changes improve the valve reliability by increasin'g the design margin of the operator using MOV sizing methodologies developed for NRC Generic

' Letter (GL) 98-10. nese changes make the containment irolation valves close faster so any delay in establishing primary containment due to valve stroke time a reduced. Component sizing, seismic  !

qualification, structural loading, and the fire loading chu veristics of the new power cables have been evaluated and accepted.- l

. Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change  !

I l

Attachment A, SVP-98-113, Page 33 of 153 j

Evaluation Reference Number: E04-2.M-130

Title:

Reactor Core Isolation Cooling (RCIC) Inboard Primary Containment Isolation Valve (M02-1301-16)-

Description:

This change upgrades the RCIC Inboard Primary Containment isolation Valve's existing actuator, motor, gearing, spring pack, disc, disc guide, yoke, circuit breaker and thermal overload (TOL) heater.

Result: Unreviewed shfety question does not exist. These changes assure that the valve actuator has sufficient thrust to close against a design basis event in accordance with Generic Letter 89-10. These chages do not alter the function of the valve, but improve its reliability to open and close. No new failure modes aw created. Seismic and structural integrity have been verified. All new parts and components have been evaluf.ed and found suitable for this application.

i

- Safety Evaluation Number: ~.NA Typeof SafetyEvaluation: Exempt Change

.. Evaluation Reference Number: E04-2-94-131

Title:

HPCI Inboard Containment Isolation Valve

Description:

This change replaced the actuator / motor and valve hardware for the HPCI Inboard '

Containment Isolation Valve. These changes were performed to assure that the valve actuator has sufficient thrust to close against a design basis event in accordance with Generic Letter 89-10. The disc and disc -

. guide were replaced to improve valve factors and the structural integrity ofinternal components during steam blowdown conditions. The stiffeners are being added to pipe support to accommodate the increased MOV loadings.

I Result: Unreviewed safety question does not exist. These changes do not functionally change the way HPCI' system operates to respond to an accident. The changes increase the capability of the isolation valve

. to operate during design basis cond_itions. No new failure modes are created. The increased thrust j

generated by the actuator is within the valve's structural limit supported by calculations. The valve's 1 F reliability to operate is improved since this change does not alter the function of the valve. Probability of a malfunction of equipment important to safety does not increase.

i Safety Evaluation Number: -NA Type of Safety Evaluation: _ Exempt Change L

Evaluation Reference Number: E04-2-94132 i

Title:

High Pressure Coolant Injection (HPCI) Inboard Primary Contaimrent Isolation Valve (M02-2301-5)

Description:

This change replaces the HPCI Inboard Primary Containment isolation Valve's existing .

actuator, motor, gearing, spring pack, yoke, circuit breaker, thermal overload (TOL) heater cables, disc and disc guide.

i Attachment A, SVP-98-ll3, Page 34 of 153 1

e

i i

i i

Result: Unreviewed safety question does not exist. These changes assure that the valve actuator has sufficient thrust to close against a design basis event in accordance with Generic Letter 89-10. The new disc and disc guide improve valve factors and the structural integrity ofinternal components during steam i blowdown conditions. These changes do not alter the function of the valve, but improve its reliability to operate. No new failure modes are created. Seismic and structural integrity have been verified. All new parts and components have been evaluated and found suitable for this application.

Safet/ Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-94-133

Title:

Reactor Core Isolation Cooling (RCIC) Outboard Primary Containment Isolation Valve (M02-1301-16)

Description:

This change replaces the RCIC Outboard Primary Containment Isolation Valve's existing, motor, disc, and disc guide.

l Result: Unreviewed safety question'does not exist. These changes assure that the valve actuator has l sufficient thrust to close against a design basis event in accordance with Generic Letter 89-10. The new disc and disc guide improve valve factors and the structural integrity ofinternal components during st'eam blowdown conditions. These changes do not alter the function of the valve, but improve its reliability to operate. No new failure modes are created. Seismic and structural integrity have been verified. All new parts and components have been evaluated and found suitable for this application.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change l

Evaluation Reference Number: E04-2-94-134

Title:

2C3 Low Pressure (LP) Feedwater Heater Shell

Description:

This change repaired the 2C3 LP Feedwater Heater Shell. The shell was repaired by welding structural plates to the outside of the heater shell. Stainless steel liners conforming to the shape of the shell were placed between the original shell and new plates. The addition of the new plates is designed to meet or exceed original design requirements of the feedwater heater and is a permanent repair. Repairs were necessary due to wall thinning caused by flow-assisted corrosion. The thinning was discovered during ultrasonic inspection.

Result: Unreviewed safety question.does not exist. There are no new failure modes created. The new repair involved welding structurally qualified carbon steel plates to the outside of the shell that enables the Feedwater heater to perform its design function for the remainder of plant life. Installation of this change restores the heater shell to original design margins and meets or exceeds the requirements of the original design and code orconstrucHon. No active components are added to the Feedwater system that would ,

increase the consequences of a malfunction of equipment important to safety - The repair does not change or  !

alter the function of the Feedwater heaters or the Feedwater system.

l Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Attachment A, SVP-98 113, Page 35 of 153

_ l Evaluation Reference Number: E04-2-94-135 I

Title:

2B1,2B2 Low Pressure (LP) Feedwater Heater Inlet Nozzles

Description:

This change repaired the 2B1 and 2B2 LP Feedwater Heater East and West Inlet Nozzles,

respectively. The nozzles were repaired by welding stainless steel liud carbon steel sleeves over the existing feedwater heater extraction steam inlet nozzles. Stainless steel liners are highly resistant to .,

erosion / corrosion; The liners halt the effects of erosion / corrosion on the new nozzle sleeves if the original i nozzles erode through. The thinning was discovered during ultrasonic inspection.

~

' Result Unreviewed safety question dor not exist. There are no new failure modes created. The new repair is designed to meet or exceed the original design requirements of the Feedwater heater nozzles and is a permanent repair. This repair does not change plant operation. The repair does not change or alter the function of the Feedwater heaters or the Feedwater system. '

Safety Evaluation Number: NA- Typeof SafetyEvaluation: Exempt Change Evaluation Reference Number: E04-2-94-138 .!

Title:

Drilling Hole in HPCI Injection Valve Disc to Reduce the Potential for Bonnet Over-Pressurization

Description:

i This change drilled a hole in the feedwater side of the valve disc for the HPCI Injection . j Valve. Drilling a hole in the feedwater side of the disc permits the pressum in the bonnet to equalize with the piping system, thereby reducing differential pressure and eliminating pressure locking susceptibility.

Pressure locking can occur in flexible-wedge disc valves when fluid becomes pressurized within the valve

-' bonnet and the actuator is not capable of overcoming the additional thrust requirements resulting from the f differential pmssure created across both valve discs by the pressurized fluid in the valve bonnet.

Result: Unreviewed safety question does not exist; As discussed in SOER 84-7 and Comed Study Report,

~

- the drilled hole will improve the reliability of the valve by preventing any potential trapped liquid binding..

This will improve the reliability of the valve to perform its' function to open for emergency core cooling and to isolate the HPCI system from the feedwater system. His change does not alter the function of the .

valve.' It is entirely internal to the valve and does not reduce the valve's capability to stroke or maintain the

. pressure boundary. The valve disc is not part of the system pressure boundary and the ability to isolate the

)

reactor coolant has not been reduced. No new interfaces are created between the valve and other systems.

l The change does not create any new failure modes or system interfaces for the valve.  !

.i

Safety Evaluation Number: .NA- Type of Safety Evaluation: Exempt Change Evaluation Referesee Numberi E04-2-94-171.

Title:

RCIC Injection Valve ll I

Description:

, This change drilled a hole in the feedwater side of the valvc disc for the RCIC injection 1 valve. This modification is deemed a high priority design change based on the valve function, the change in

ambient temperature near the valve following an accident, Generic Letter 89-10 requirements for motor-Attachment A, SVP-98-113, Page 36 of 153 o

4

I operated valve testing, and its Risk Achievement Woith. A study evaluated various ways to prevent pressu e locking of the valve. Trapped liquid binding of a gate valve can occur when a large differential l~ pressure is created between fluid in the valve bonnet and fluid in the process pipe. This binding can cause

the valve to fail to open due to the differential pressure applying a force to hold the valve disc in the c!osed L position. This pressure can be created when fluid trapped inside the bonnet is heated due to increasing

! ambient temperature.- Drilling a hole in the valve disc is the recommended solution for this problem. This will equalize pressure between the valve bonnet and process pipe on the feedwater system side of the gate valve.

t Result: Unreviewed safety question does not exist.' As discussed in SOER 84-07 and the Study Report,'the drilled hole will improve the valve reliability by preventing any potential trapped liquid binding. This will improve the reliability of the valve in performing its function to open for emergency makeup water flow in f the event there is a loss of normal feedwater. The required isolation function is for flow from feedwater (and reactor) to RCIC. In this flow direction, the valve will be leaktight. Drilling the hole in the disc does not create any new failure modes of the valve. It does change the valve from a bi-directional valve to a uni-directional valve. Reverse flow, from RCIC to feedwater can still occur. When RCIC is running and the injection valve is closed, there will be a small leakage through the valve. This flow rate is limited by the size of the hole and the differential pressure across the valve. There are no known post-accident scenarios where this small leakage could pose a safety hazard. These effects are not considered significant and are I

' more than offset in terms ofimpact on safety by the improvement in valve reliability. This change will

' make the RCIC injection valve less vulnerable to failure due to pressure locking. The changes are entirely internal to the valve so there are no new interfaces created between the valve and any other systems, i

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Refmace Numbert E04-2-94-209 l

Title:

Turbine Control Valve Accumulator Addition j-

Description:

This change installs accumulators at the Fluid Actuator Supply (FAS) connections to the Main Turbine Cor. trol Valves (TCVs) as recommended by General Electric.  ;

L Result: Unreviewed safety question does not exist. This change will reduce pressure pulses in the Electro-Hydraulic Control (EHC) lines. The accumulator / control manifold assembly is passive and does not affect

)

~

the method ofoperation of the TCVs or the EHC system. The pressure dampening of the modified system l

should enhance reliability of EHC components. The availability of the EHC system will be enhanced by reducing the likelihood of fluid leaks. No new failure modes are created. The system will operate exactly ,

as it did before, and there will be no interactions with other equipment important to safety as a result of the  ;

instellation.-

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change

- Evaluation Reference Number: E04-2-94-220

Title:

1500KVA Reactor Building Outage Power Transformer i

i Attachment A, SVP-98-113, Page 37 of 153

Description:

This change removes an existing 480V,1200A, disconnect from the Reactor Building and ,

installs two new 1500KVA 13.8kV:480V dry type transformers. Two disconnects will be provided on the secondary side of the aew transfctmer.

Result: Unreviewed safety question does not exist. This change prevents overloading of the existing supply transformer and 480V feed cable and provides additional capacity for upcoming outage work. The transformer will not electrically affect operation of any plant equipment (per FSAR the 13.8kV system is not used for any essential plant equipment). The transformer and associated coniponents will be seismically mounted and will not mechanically or structurally affect the operation of the plant. Penetrations used to bring the primary feed into the Reactor Building will be sealed so that they function the same after installation.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-94-226

Title:

= Emergency Core Cooling System (ECCS) Torus Suction Valves

Description:

This change replaces six ECCS Torus suction butterfly valves. The seals on the existing valves have deteriorated and cannot provide a safe, positive isolation from the Torus when maintenance is required.

Result: Unreviewed safety question does not exist. The new valves will f.metion the same as the existir g valves. They have the same end dimensions and will be inr* ied in the sanne location. No new syste.r interactions are introduced. The new valves will have safety-related manual operators, whereas the cid valves had non-safety-related manual operators. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-95-006

Title:

' Main Control Room Office Furniture

Description:

This change installs new office furniture in the Main Control Room (consisting of file and

! storage cabinets along the North wall of the Main Control Room back panel area)

Result:- Unreviewed safety question does not exist. Tis change will provide increased storage space in the

~ Main Control Room. It does not interfere with any structu es, systems or components (SSCs) that are important to safety. The possibility of cabinets falling and seismic concerns were taken into account in designing anchors. Floor loading was also evaluated and accepted.

n l- , Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-036A

Title:

Standby Liquid Control (SBLC) Check Valve Attachment A, SVP-98-113, Page $8 of 153 -

l-b t

i-

y. ,

N

Description:

This change replaces a primary containment isolation SBLC check valve (#1-1101-15)in the injection line to the reactor. The existing valve had been experiencing excessive leakage.

~

Result: Unreviewed safety question does not exist. The replacement valve is hydraulically equivalent or

.better than the existing vnive, passing full flow with a lower pressure drop. It is properly sized for this

.-application. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-042 i - . .

(

f

Title:

Lisega Hydraulic Snubbers fr.r Main Stearn Piping in the Unit 1 Drywell  !

Description:

j This change installs new Lisega Hydraulic snubbers on the Unit 1 Main Steam system within the drywell to replace the existing mechanical snubbers.-  !

<~ .

Result: - Unreviewed safety question does not exist. The replacement snubbers will reduce the number of failures that occur. This change will not affect function (t operation of the Main Steam system. There are no interactions with other structures, systems, and compo. ents as a result of this change. Industry .

experience indicates that Lisega hydraulic snubbers are mo'e reliable than any type of mechanical snubber.

The replacements will use existing structural attachments.1 tilure modes are identical to those of the existing snubbers. The likelihood of an accident due to snubb:r failure is reduced. Compliance with Tech  !

Spec requirements for functional testing in conjunction with a revised visual inspection schedule assures ,

continued operability of the new snubbers.

l Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change

' Evaluation Reference Number: E04-1-95-050 .

Title:

l A & IB Off Gas Recombiner Vial Sample Pumps  !

[-

Description:

This change installed vent valves in the suction piping of the l A and IB Off Gas Recombiner Vial Sample Pumps. These pumps do not self-start properly due to vacuum at the pump inlet causing a motor locked-rotor condition. Manual assistance to start the pumps is often required, which is a safety concern. When.the pumps are shut down, a high vacuum condition could exist at the inlet of the  !

pumps. This condition can cause oil to be drawn into the pump's cylinder causing a hydraulic lock which in turn causes difficulty in re-starting the pump. I Result: Unreviewed safety question does not exist. This change does not affect plant operation or the function of the sample pumps. Installing vem valves will improve the starting capability of the pumps and will provide a means to break vacuum at the suction of the pumps prior to shutting off the pumps. They improve the reliability of the pumps. No interfaces with other systems, structures, or components are -

created by the installation of this change. There are no new failure modes generated by this change.

Consequences of a malfunction of equipment important to safety are not increased.

l l

Attachment A, SVP-98-113, Page 39 of 153

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-059

Title:

Service Water Radiation Monitor Drseription: This change installs a flexible hose in the inlet piping to the Unit One Service Water Radiation Monitor (SWRM).

Result: Unreviewed safety question does not exist. This change is being installed to decrease the vibration being felt by the Unit One SWRM which, in turn, should decrease the frequency of spiking and increase the reliability of the system. The possibility that the hose might become loose or detached was evaluated and accepted. The SWRM does not perform any safety function and is not needed for safe shutdown af the plant. This change will not affect the flow rate of the fluid or the design function of the SWRM.

Safety Evaluation humber: NA Type of Safety Evaluation: Exempt Change Evaluation Referesee Number: E04-1-95-060A

Title:

480 VAC MCC 18-1 A Feed Cables

Description:

This Wange upgraded the existing power feed cables from Switchgcar 18 to MCC 18-1 A by replacing the existing 250 MCM cables with new 500 MCM cables. The existing cables were abandoned in place. Routing from the reactor building to turbine building switchgear 18 required a fire barrier and secondary containment breach. This change was issued in response to a poter.tial for exceeding feed cable rated ampacity. Such a condition may shorten the cable life due to insulation degradation.

Result: Unreviewed safety question does not exist. When the upgraded power feed functions as intended, plant operation under normal and accident conditions remains the same. The new feeds allow greater operator flexibility in operating loads without concerns of cable degradation. There are no changed interactions wit other systems, structures, or components. This change simply changes the existing power feed cables with ones of higher ampacity rating. No new failure modes are introduced. The breach of secondary containment and fire barrier were controlled per Station Procedures ard Approved Fire Protection Program. The new power feed eliminates concerns with cable degradation. The new cables function the same as the existing cables. The new cables will not adversely affect terminal voltage, available short circuit current, and will have a long established life.

' Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change

- Evaluation Reference Number: E04-1-95-060B

Title:

L 480 VAC MCC 18-1B Feed Cables

Description:

' This change upgraded the existing power feed cables from Switchgear 18 to MCC 18-1B by replacing the existing 250 MCM cables with new 500 MCM cables. The existing cables were abandoned in place. Routing from the reactor building to turbine building switchgear 18 required a fire barrier and Attachment A, SVP-98-113, Page 40 of 153

1

\

.y secor.dary containment breach. This change was issued in response to a potential for exceeding feed cable j l rated ampacity. Such a condition may shorten the cable life due to insulation degradation.

Result: Unreviewed safety question does not exist. When the upgraded power feed functions as intended, plant operation under normal and accident conditions remains the same. The new feeds allow greater j operator flexibility in operating loads without concerns of cable degradation. There are no changed '

l interactions with other systems,' structures, or components. This change simply changes the existing power

feed cables with ones of higher ampacity rating. No new failure modes are introduced. . The breach of L secondary containment and fire barrier were controlled per Station Procedures and Approved Fire Protection i

Program. The new power feed eliminates concerns with cable degradation. The new cables function the

! same as the existing cables. The new cables will not adversely affect terminal voltage, available short L circuit current, and will have a long established life.

L Safety Evaluation Number: NA Typeof SafetyEvaluation: '

l Exempt Change i

i Evaluation Reference Number: E04-1-95-060C l-

Title:

480 VAC MCC 18-3 Feed Cables l_ '

Description:

This change upgraded the existing power feed cables from Switchgear 18 to MCC 18-3 by.

l- replacing the existing 500 MCM cables with new 750 MCM cables. The existing cables were abandoned in place. Routing from the reactor building to turbine building switchgear 18 required a secondary containment penetration and fire seal. An existing breaker was also replaced. This change was issued in -

I

. response to a potential for exceeding feed cable rated continuous ampacity. Such a condition may shorten the cable life due to insulation degradation.

Result: Unreviewed safety question does not exist. When the upgraded power feed functions as . intended, plant operation under normal and accident conditions will be less restrictive. The new feeds alleviate the need for the present administrative load controls with respect to ampacity, provide future load growth margin and allow the MCC feeder breaker at Switchgear 18 to be set to provide improved cable protection. .,

Replacement of the breaker provides interrupting capability that exceeds available short circuit. There are l no changed interactions with other systems, structures, or components. This change simply changes the existing power feed cables with ones of higher ampacity rating. No new failure modes are introduced. The j breach of secondary containment and fire barrier were controlled per Station Procedures and Approved Fire j Protection Program. The new cables function the same as the existing cables. The new cables will not adversely affect terminal voltage, available short circuit current, and will have a long established life.  !

i

' Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change i-Evaluation Reference Number: E04-1-95-060D l

Title:

480 VAC MCC 19-1 Feed Cables h

Description:

This change upgraded the existing power feed cables from Switchgear 19 to MCC 19-1 by t

replacing the existing 250 MCM cables with new 500 MCM cables. The existing cables were abandoned in i place. Routing from the reactor building to turbine building switchgear 19 required a fire barrier and 4

Attachment A, SVP 98-113, Page 41 of 153

secondary containment breach. This change was issued in response to a potential for exceeding feed cable rated ampacity. Such a condition may shorten the cable life due to insulation degradation.

Result:

Umeviewed safety question does not exist. When the upgraded power feed functions as intended, plant operation under normal and accident conditions remains the same. The new feeds allow greater operator flexibility in operating loads without concerns of cable degradation. There are no changed interactions with other systems, structures, or components. Th's change simply changes the existing p feed cables with ones of higher ampacity rating. No new failure modes are introduced. The breach of secondary containment and fire barrier were controlled per Station Procedures and Approved Fire Protection '

Program. The new power feed eliminates concerns with cable degradation. The new cables function the same as the existing cables. The new cables will not adversely affect terminal voltage, available short circuit current, and will have a long established life.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-060E

Title:

480 VAC MCC 19-2 Feed Cables

Description:

This change upgraded the existing power feed cables from Switchgear 19 to MCC 19-2 by replacing the existing 250 MCM cables with new 500 MCM cables. The existing cables were abandoned in place. Routing from the reactor building to turbine building switchgear 19 required a fire barrier and secondary containment breach. This change was issued in response to a potential for exceeding feed cable l rated ampacity. Such a condition may shorten the cable life due to insulation degradation.

Result: Unreviewed safety question does not exist. When the upgraded power feed functions as intended, plant operation under normal and accident conditions remains the same. The new feeds allow greater operator flexibility in nperating loads without concerns of cable degradation. There are no changed interactions with other systems, structures, or components. This change simply changes the existing power feed cables with ones of higher ampacity rating. No new failure modes are introduced. The breach of secondary containment and fire barrier were controlled per Station Procedures and Approved Fire Protection Program. The new power feed eliminates concerns with cable degradation. The new cables function the same as the existing cables. The new cables will not adversely affect terminal voltage, available short circuit current, and will have a long established life.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-060F

Title:

480 VAC MCC 18-2 Feed Cables

Description:

This change upgraded the existing power feed cables from Switchgear 18 to MCC 18-2 by replacing the existing 250 MCM cables with new 500 MCM cables. The existing cables were abandoned in place. Routing from the reactor building to turbine building switchgear 18 required two new secondary .

containment penetrations and four fire barrier breaches. This change was issued in response to a potential for exceeding feed cable rated ampacity. Such a condition may shorten the cable life due to insulation degradation. i

]

I l Attachment A, SVP-98 il3, Page 42 of 153 I

4 Result:

Unreviewed safety question does not exist. When the upgraded power feed functions as intended, plant operation under normal and accident conditions remains the same. The new feeds allow greater operator flexibility in operating loads withok concerns ofcable degradation. There are no changed interactions with other systems, structures, or cenponents. This change simply changes the existing pow feed cables with ones of higher ampacity rating. No new failure modes are introduced. The breach of secondary containment and fire barrier were controlled per Station Procedures and Approved Fire Protection Program. The new power feed eliminates concerns with cable degradation. The new cables function the same as the existing cables. The new cables will not adversely affect terminal voltage, available short circuit current, and will have a long established life.

Safety Evaluation Number: NA- Typeof SafetyEvaluation: Exempt Change Evaluation Reference Number: E04-1-95-060A,B,C,D,E & F

Title:

480 Volt Switchgear Feed Breakers to MCCs 18-1 A,18-1B,18-2,18-3,19-1,19-2

Description:

The settings for the RMS-9 devices on the main feed breakers for MCCs 18-1 A,18-1B,18-2,18-3,19-1 and 19-2 were changed. The kng time delay values were increased to prevent inadvertent tripping of the breaker during a maximum loading condition. The short time delay settings were also adjusted to compensate for the new long time delay settings. The setting is bas 'd on MCC connected load.

It eliminates administrative load controls.

Result: Unreviewed safety question docs not exist. The change is to the trip setting of the breakers. The-operation of the breaker remains the same. The settings were chosen to provide the maximum cable protection and to prevent a nuisance tripping of the breaker. Failure modes remain the same since the same type of breaker was used. Change of setting does not change the failure mode of the cable or the breaker.

The MCC connected load is less than the cable ampacity thereby preventing a cable failure. Thus, the probability of a malfunction of equipment important to safety has not increased.

Safety Evaluation Number: Nt Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-062 i

Title:

l A and IB RER Pump Suction Relief Valve Discharge Flange

Description:

This change added a bolted flange to the existing relief valve discharge piping to ease future removal of the valves. There is already a flange on the inlet of the relief valve. The 1 A and 1B RHR pump suction relief valves are removed periodically for testing and is very labor intensive. Also, two supports were removed and one support had its U bolt removed.

Result: Unreviewed safety question does not exist. There is no change to the operation of the system due to installation of these flanges. The flanges only facilitate removal of the RHR pump suction relief valves.

The system parameters are not altered. Plant operatim is unaffected. No new failure modes are introduced.

The flanges meet or exceed current design of the relief valve discharge piping. Prior to authorization for operation, the installation was demonstrated to be leak tight per current codes and standards. No accidents

- were identified where the RHR pump suction relief valves are required. They are installed for shutdown Anachment A, SVP-98 113, Page 43 of 153

i cooling mode of operation only. The addition of the flanges will not create any new accident or equipment failures, i

f Safety Evaluation Number: NA Type of Safety Evaluation:

i Exempt Change

- Evaluation Reference Number: E04-1-93-064A through F

Title:

RHR Heat Exchanger Support Steel

Description:

. This change upgraded the Residual Heat Removal (RHR) Corner Room Structural Steel.

This change restored the design basis design margins for the structumi steel. The existing steel was found to meet operability requirements, but did not meet the original design requirements. Structural connection

]

reinforcement and additional bracing to the existing RHR corner room structural steel, and horizontal ..

j supports on the RHR Heat Exchangers were installed. This structural steel provides support for the RHR j Heat Exchangers as well as attachment points for safety-related pipe supports utilized to support RHR and l

RHRSW piping as well as components supporting RHR and RHRSW operation. '

' Result: Unreviewed safety question does not exist. This change will not have an affect on the operation of the RHR or RHRSW systems or equipment; it strengthens the existing structural steel. Strengthening the existing structural steel will reduce the likelihood of a structural failure by ensuring that the design margins i are within the UFSAR allowable limits. This change will not create any new interactions with other i structures, systems or components, nor will any existing interactions be changed. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation:

Exempt Change Evaluation Reference Number: E04-1-95-066 l

Title:

RIIR LPCIInboard Injection Valve i

Description:

This change drilled a hole through the reactor recirculation piping inlet side of the valve flex-wedge disc. Drilling a hole in the high pressure side of the disc permits the pressure in the bonnet to

)

equalize with the piping system, thereby reducing differential pressure and eliminating pressure locking  ;

- susceptibility. Pressure locking can occur in flexible wedge disc valves when fluid becomes pressurized within the valve bonnet and the actuator is not capable of overcoming the additional thrust requirements

. resulting from the differential pressure created across both valve discs by the pressurized fluid in the valve l

bonnet.' '

l.

L Result: Unreviewed safety question does not exist. As discussed in SOER 84-7 and Comed Study Report, the drilled hole will improve the reliability of the valve by preventing any potential pressure locking. This will improve the reliability of the valve to perform its function to open for LPCI injection and close for the required primary containment isolation function. This change does not alter the function of the valve. It is

[ entirely internal to the valve and does not reduce the valve's capability to stroke or maintain the pressure l . boundary. No new interfaces are created between the valve and other systems. The change does not create I- any new failure modes or system interfaces for the valve.

Attachment A, SVP-98-113 Page 44 of 153

s Safety Evaluation Number: 'NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-067

Title:

IIPCI Drain Valve Solenoid Valve

Description:

This change modified the mounting bracket for the HPCI drain valve solenoid valve so it is I

- in an upright and vertical orientation. The existing mounting bracket was removed and a new one installed that will allow proper operation of the solenoid valve during normal and seismic conditions. The instrument

- air lines and conduit were reworked to accommodate the new configuration. The current mounting configuration is not in'accordance with the manufacturer's instructions and may not function properly during or after a seismic event. The solenoid valve is currently mounted in a horizontal direction, however, it needed to be upright and vertical.'

L I

. Result: Unreviewed safety question does not exist. This change has no effect on the operation of the HPCI j valve or the HPCI system. Equipment failures are unaffected. No new failure modes are introduced. The l solenoid valve will be more reliable with this new mounting configuration during and after seismic events.

A calculation was performed to verify that the new bracket adequately supports the solenoid valve under  !

I normal and seismic conditions. The solenoid valve will function better in its new configuration and will not affect the consequences of an accident. By mounting it as it is intended to be, the probability of a l malfunction of equipment important to safety will decrease.

I Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change j Evaluation Reference Number: E04-2-95-003

Title:

ECCS Room Cooler Inlet and Outlet Piping l

Description:

This design installed seven pipe taps that will be used for hydrolazing and cleaning ECCS Room Cooler Inlet and Outlet piping. Each pipe tap included a drain valve and threaded pipe cap. One tap was installed on each line. The ECCS room coolers have experienced reduced flows and high differential  ;

pressures. Current system configuration is comprised of approximately 600 feet of pipe that ranges in )

elevation from 564 feet to 580 feet. Low points within this configuration are not conducive to draining and are, thus, conducive to dirt and silt accumulation. q i

Result: Unreviewed safety question does not exist. Installation of this design will improve system reliability. The new valves will remain in their normally closed position. There will be no impact to plant operating modes or system interactions due to this design. Once hydrolasing and installation activities were completed, the drain valves were closed and a threaded pipe cap was installed. The new pipe taps do not create any failure modes different than the vents and drains that already exist in the system. Stresses are i within allowables. These lines are not a primary system and their failure cannot cause an accident to occur.

There is no increase in the probability of an accident occurring due to this design.

i Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-95-015 Attachment A, SVP-98-113, Page 45 of 153

I 1

Title:

AVCo Solenoid Valve for Target Rock Safety / Relief

Description:

This change removed the Marathon 1500 Series terminal b' lock in the junction box for the  ;

AVCo solenoid valve of the Target Rock safety / relief valve and replaced it with an Environmentally l Qualified splice. The terminal block is currently used to connect the field routed power cable to the  !

solenoid pigtails. The new splices were completed in accordance with existing approved procedures

' intended for the application. The Marathon terminal blocks have been known to exhibit excessive leak current under post Design Bases Event conditions that could affect the performance of the Safety / Relief

! Valve.

- Result: Unreviewed safety question does not exist. The replacement of the terminal block with EQ  :

approved splices will enhance the reliability of the solenoid circuit to ensure the valve will function as

' - intended. Plant operations are not affected. This change significantly reduces the chances ofequipment failures. No new failure modes are introduced. This change does not functionally change the power or '

l control circuit for the SRV, affect how the SRV interacts with other equipment, or a%ct the performance of any other equipment.

Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change Evaluation Reference Number: E04-2-95-016 l i

Title:

Test Logic Circuitry for Turbine Main Test Valves

Description:

This change modified the test logic circuitry for the #2 Main Stop Valve (MSV) to -

eliminate the potential for a reactor scram due to the single failure of the #2 MSV SVOS limit switch. The change eliminates the master / slave relationship between #2 MSV and the remaining #1, #3, and #4 MSV's which has been associated with past turbine trips. This change will improve turbine and reactor reliability by eliminating scrams fmm a single failuir of the limit switch, and the subsequent closing of the other three MSV's which are slaved off the #2 MSV in the test circuitry.

Result: Unreviewed safety question does not exist. This change has no e%ct on operation of the system, or the interactiori with other systems since the modified circuitry eliminates the potential for reactor scrams caused by failure of the #2 MSV SVOS limit switch. No new failure modes are created by this change and only existing non-safety-related relays were re-wired during the installation of this change. Since this change modifies test logic circuitry to eliminate the potential for a turbine trip, the probability of an accident decreases. Operation of the MSV's during an actual Turbine Trip remains unancted. Therefore, consequences of an accident are not increased. Since this change eliminates the possibility of a turbine trip w caused by the master / slave relationship, it will decrease the probability of a malfunction of equipment important to safety. There was no physical work performed on safety-related components. Operation of MSV's during an actual turbine trip remains unaffected. This change increases the reliability and l availability of the turbine.

Safety Evaluation Number: - NA Type of Safety Evaluation:- Exempt Change l Evaluation Reference Number: . E04-2-95-020 - l

! 1

Title:

480V SWGR 25,26,27, MCC 2A-1

! I i

Attachment A, SVP-98-113. Page 46 of 153

i

Description:

- Dataloggers (Load Facts LF-1 units) were installed at Switchgear 25,26,27, and MCC 2A-

1. Load profilers were also installed at various locations on Switchgear 25 breakers. The total assembly was mounted to the lower rear cover of the main feed section of the switchgear and MCC. Installation is non safety-related. He weight added to the bus and mounting hardware was evaluated and found to be acceptable. These load profilers were installed to measure current at the breaker for the individual loads and i

was either tie wrapped to the side of the cubicle or placed on the floor of the cubicle. In either position, it was not able to come in contact with any energized busses. The data was used for two purposes: (1) to perform a calculation to check the setting of the under voltage relay, and (2) to update the Electrical Load Monitoring system with real information. Bus 25 was de-energized to allow safe installation of the profilers. The associated bus was backfed from another bus for the same reason.' Equipment on these l busses is not required during unit shutdown conditions.

u

Result

Unreviewed safety question does not exist. Switchgear 25,26,27, and MCC 2A-1 are non-safety-l  !

related busses. They are not required to be operable per Tech Specs. Normal operation of equipment in and I l powered by these busses was not changed. Since operating parameters of the busses have not changed, there is no affect on any structure, system or component required to be operable by any Tech Spec. No new failure modes are added to the busses as a result of this change. The inherent nature of the clamp on current transformer prevents a fault from being transmitted from a Load Profiler to the bus. These devices will not

[

affect equipment powered from these busses that could cause an accident. A failure of the added equipment will not cause failure of the associated bus. The operation of the bus is not changed. Thus, there is no change to the consequences of a malfunction of equipment important to safety.

1

!- Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-95-023

Title:

Feedwater Heater and Flash Tank Level Indicating Controllers

Description:

. This change replaced 21 existing level indicating controllers, Foxboro Model 43A, with .

new level indicating controllers, Foxboro Model 43 AP. Parts for the obsolete model are no longer available from the original equipment manufacturer.

Result: Unreviewed safety question does not exist. The upgrade of the new model of the level indicating controller will improve :nalntenance through the improved availability of spare parts. The new model has

- some improvements in both accuracy and reliability. A system functional and failure modes review has not identified any reason why the component replacement is not acceptable. The consequences oflevel indicating controller failure has not changed. The function and purpose of the controllers has not been -

changed. Dere are no new system interactions or drain valve failure modes created. This new model has been used in the nuclear industry for approximately 14 years with no significant operational issues identified.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-95-024

Title:

RHR LPCIInboard injection Valve Attachment A, SVP-98-l l3, Page 47 of 153 l

j i l*

Description:

This change was entirely internal to the valve and involved drilling a hole through the reactor recirculation piping inlet (high pressure) side of the valve flex-wedge disc. This permits pressure in the bonnet to equalize with the piping system thereby reducing the differential pressure and eliminating pressure locking susceptibility. His improves the reliability of the RHR LPCI Inboard Injection Valve.

Pressure locking can occur in flexible-wedge valves when fluid becomes pressurized within the valve

~ bonnet and the actuator is not capable of overcoming the additional thrust requirements resulting from th differential pressure created across both valve discs by the pressurized fluid in the valve bonnet. Valve bonnet pressure might be higher than anticipated, causing pressure locking under certain conditions.

Result:

Unreviewed safety question does not exist. As discussed in SOER 84-7 and Comed Study Report, the drilled hole will improve the reliability of the valve by preventing any potential pressure locking. This will improve the reliability of the valve to perform its function to open for LPCI injection and close for the l required primary containment isolation function. This change does not alter the function of the valve. It is .

entirely internal to the valve and does not reduce the valve's capability to stroke or maintain the pressure l boundary. No new interfaces are created between the valve and other systems. The change does not create any new failure modes or system interfaces for the valve.

Safety Evaluation Nunnber: NA' Type of Safety Evaluation: Exempt Change Evaluation Reference Nunnber: E04-2-95-025

Title:

1 Recirc Motor Vibration Switch

Description:

. His change replaced the existing recirc motor vibration switch with a new model switch.

The existing reset pushbutton on the 902-4 panel was removed. The new model switch did not require a reset pushbutton. The new switch required a continuous power source of 120 VAC The existing switch utilized a 125 VDC power source to reset the switch. The new switch is only approved for operation for one cycle. A new permanent vibration monitoring system will need to be installed during the next refuel outage or the switch will need to be replaced with a like-for-like switch. Fuses were added to protect primary ,

containment penetration from a short circuit condition. Due to the setting of the existing switch at the limit i

'ofits range, GE recommended that the existing switch be replaced with a switch more suitable to the

. desired range. The new switch is adjustable from .05 in/sec to .75 in/sec. The existing switch is set to .3 in/sec which is at the lower limit of the switch setting. -ne setpoint of.3 in/sec is in the midrange of the new' switch.

Result: Unreviewed safety question does not exist. The new vibration switch has the same function as the

. original. Further evaluation will be required to use this switch for an installed period of greater than one operating cycle. The function of the switch is to increase the reliability of the recire motors by diagnosing -

o vibration problems as early ::: pWW. *ne new switch is expected to be more reliable than the existing

. switch since the range for the new switch has been matched to the setpoint. The probability of a j malfunction of equipment important to safety is not increased. The vibration switch is measuring excessive

' vibration, which is only an indicator of potential problem. The consequences of a failure of the vibration switch will not be changed.

l=

Safety Evaluation Number: ~ NA Type of Safety Evaluation: Exempt Change  !

Attachment A, SVP-98-113, Page 48 of 153  !

i l

a i

%w .

h: l 2

Evaluation Reference Number: E04-2-95-028 -

Title:

U2 Emergency Diesel Generator (EDG) Fuel Oil Storage Tank Low Level Alarm Switch

Description:

. This change replaced the U2 Emergency Diesel Generator (EDG) Fuel Oil Storage Tank Low Level Alarm Switch with a Static-O-Ring (SOR) switch. This required mounting the new SOR swi ch

- seismically and connecting it to the process tubing and existing terminal block. The original switch was a non-safety-related instrument that utilized a Mercoid level switch to provide alarm indication in the control room'. The existing Mercoid switch caused nuisance alarms in the control room. De diaphragm sensing  ;

element of the new SOR switch is more reliable than the bourdon tube element of the existing switch.

Result: Unreviewed safety question does not exist. The replacement level switch will not cause a change

,' in plant operation or conditions. The failure mode of the new switch is the same as for the existing switch. .

Because the new switch is more reliable than the existing switch, the ability to respond to any postulated '  !

accident will be increased, and the consequences of the accident will be reduced. The new switch does not

' directly interface with safety-related equipment. Its mounting is such that it will not indirectly affect safety- ,

related equipment or components during a seismic event. Therefore, the probability of a malfunction of -l safety-related equipment is unchanged. He switch does not contribute to any postulated event.~ l Safety Evaluation Number: NA . Type of Safety Evaluation: Exempt Change j Evaluation Referesee Number: E04-2-95-037 j

Title:

' Main Steam Line Drain inboard Isolation Valve I

Description:

This change involved replacing the Limitorque motor gearing from a 33.5 to 68.0 overall ]j actuator ratio. Additionally, the inactive mechanical dial position indicator mechanism (MDPI) was removed and the opening on the actuator housing was blanked off. As a result of design margin reviews for l Generic Letter 89-10, it was determined that a gearing change was needed to increase the motor gearing -l capacity of the valve. The MDPI was not electrically connected and removal eliminated a potential actuator ~

. grease leak path.  !

Resmit: . Unreviewed safety question does not exist. This change has no impact on existing failure modes.

The installation of the new gearing will increase the thrust window for the Limitorque actuator.

• Replacement of the gearing will result in a negligible change in both the weight and center of gravity and -

does not require a revision to the piping stress analysis. Therefore, the change will not introduce any new l failure modes or increase the probability of equipment failures. This change enhances the valves' capability l

to isolate on a Group 1 signal thereby, maintaining the release rates below the analyzed limits. This change . j will not alter the function of the valve. An increase in reliability will result from the improved motor  ;

ge'aring capability. -1 i

- Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-95-043 Titlei Feedwater Heater"B" Level Indicating Controller (LIC) High Limit Relay Addition  !

l Attachment A, SVP-98 113, Page 49 of 153

Description:

This change added a high limit relay downstream oflevel indicating controllers and upstream of air operators on emergency drain valves. The pneumatic relays were instal!ed on local racks.

The adjustable high limit relays will provide the capability to prevent the valves from going closed. The capacity of the feedwater heater drains from the "A" drain coolers to condenser are undersized. This results in cycling (open and closed) of the e ~ mergency drain valves. This cycling has resulted in broken bolts on the valve operator. The new high level relays will limit the pneumatic signal from the LICs to allow modulation of the emergency drain valves but prevent them from going closed. This change is an interim solution until a cost-effective solution can be found.

Result: Unreviewed safety question does not exist. The LICs are designed to fail the emergency drain valves open on loss ofinstrument Air. The new relays do not have failure modes that prevent the LICs from

_ performing this function. The addition of the relay is to decrease the probability of breaking ifolts on the -

valve operator. This change is to increase the reliability of the Feedwater Heater drain system. The operation and basic function of the Feedwater Heater Drain system is not changed. This modification was j

not designed to perform any enhancements, other than the replacement ofobsolete equipment. Capacities, response times, and other characteristics are not changed. The function and purpose of the controller has not changed. There are no new system interactions or drain valve failure modes created.

Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change l Evaluation Reference Number: E04-2-95-044

Title:

HPCI Room Cooler Fan Motor 1

Description:

i His change replaced an obsolete G.E. motor that has failed with a Siemens Electric motor.

The motor could not be replaced like for-like.

j Result: Unreviewed safety question does not exist. There is no change in the .HPCI room cooler design function. The new motor is similar in electrical characteristics to the older motor. Therefore, the HPCI -

system will function the same. He new motor meets all the requirements of the original motor and is safety-related and Environmentally Qualified (EQ) which goes beyond the requirements of the original l i

motor. There are no new interactions or failure modes introduced by this change. The HPCI room cooler . i fan motor is important to safety. It must be operable for the HPCI system to be operable. The probability of a malfunction remains the same or decreases with the new motor. Consequences of a malfunction of any

]

equipment important to safety remains unchanged. No new failure modes are introduced.

- Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change Evalestion Reference Number: E04-2-95-046 l

Title:

Electro-Hydraulic Control System

Description:

This change replaced the existing primary and secondary pressure amplifiers and added two i steamline resonance compensator circuit boards to the Electro-Hydraulic Control (EHC) System. The purpose for this was to improve the response of the EHC system to pressure oscillations in the Main Steam Piping. During recent plant startups excessive position changes in the Turbine Bypass Valves were noted, i Attachment A, SVP-98 113, Page 50 of 153 1

j

!l l

1 l

These valves could not adequately control reactor pressure when in the automatic mode above 900 psig.  !

Afler being placed in automatic mode, the valves were opening and closing in a divergent manner and had I to be placed under manual control to maintain reactor pressure. The new circuit boards will help to dampen pressure oscillations by controlling the response of the Turbine Bypass Valves and Turbine Control Valves.

l l

Result: Unreviewed safety question does not exist. As a result of this change, the EHC system will be able l to control reactor pressure more efficiently than in the past by reducing system-induced pressure ,

oscillations. The installation of new circuit boards does not increase the probability or consequerces of an equipment failure. If the new boards were to fail, the consequences would not be any different than if an existing board failed. Operation of the new circuit boards have been reviewed to ensure that the probability of equipment failure has not been increased. The addition of the new boards increase system performance during transients.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change l l Evaluation Reference Number: E04-2-95-048

Title:

Reactor Feed Pump Motor (RFPM) Ventilation Fan Discharge Damper Change l

Description:

This change disconnected the pneumatic operator from dampers 2A-5772-32 and 28-5772-

32. The existing lever arm was modified to be used as a handle for manual operation of the damper. I Manual action will be required to open and close dampers prior to manual starting of a fan. Operating procedures were revised to not allow automatic starting of a fan. This change was issued in response to a i degraded material condition regarding RFPM fan shut-offdampers. This design change is an interim change until a permanent solution can be designed. The damper is currently designed to open when the fan is started and close when the fan is shut down. The fan discharge pressure during startup is causing the damper blades to bend prior to the damper going full open and during fan coastdown the damper goes closed prior to the pressure decrease which is causing the damper to bind and not function properly.

Result: Unreviewed safety question does not exist. The change from automatic operation of the dampers ,

to manual operation does not change the probability of an accident. The RFPM ventilation system is not required to mitigate the consequences of an accident. This change was performed to reduce the recurring problems with these dampers and to lessen the chance of having to shutdown the plant, e.g. decrease the probability of challenging a safety system due to a malfunction of this non-safety system. Failures in this

! non-safety system do not impact any safety system and therefore, do not change the consequences. The function of the RFPM ventilation system has not changed.

l Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-2-95-050

Title:

2A & 28 Off Gas Recombiner Vial Sample Pumps

Description:

This change installed vent valves in the suction piping of the 2A and 2B Off Gas Recombiner Vial Sample Pumps. These pumps do not self-start properly due to vacuum at the pump inlet causing a motor-locked rotor condition. Manual assistance to start the pumps is often required, which is a safety concern. When the pumps are shut down, a high vacuum condition could exist at the inlet of the Attachment A, SVP-98-113, Page 51 of 153

i pumps. This condition can cause oil to be drawn into the pump's cylinder causing a hydraulic lock which in turn causes difTiculty in re-starting the pump.

Result: Unreviewed safety question does not exist. This change does not affect plant operation er the function of the sample pumps. Installing vent valves will improve the starting capability of the pumps and will provide a means to break vacuum at the suction of the pumps prior to shutting off the pumps. They improve the reliability of the pumps and eliminate the need for manual assistance to start the pumps. No interfaces with other systems, structures, or components are created by the installation of this change. There are no new failure modes generated by this change. Consequences of a malfunction of equipment important to safety are not increased.

Sdety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Numbcr: E04-2-95-051

Title:

Service Water Radiation Monitor Descriptiou: This change installed a flexible hose in the inlet piping to the Unit 2 Service Water Radiation Monitor (SWRM). The hose was installed in the 2-17103-1"-0 inlet pipe and attached with hose nipples and pipe clamps. This change was installed to decrease the vibration being felt by the SWRM detector. The decrease in vibration might decrease the frequency of spiking currently being seen.

Result: Unreviewed safety question does not exist. The decrease in vibration from installing the flexible hose should increase the reliability of the system. The probability of an accident will not be increased because the SWRM does not initiate any UFSAR accident or perform any safety function. The failure of the SWRM has no effect on safety-related equipment, therefore the consequences of a malfunction of equipment important to safety is not increased. The change in piping will not impact the flow rate of the fluid and will not have any affect on how the SWRM system functions.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04 2-95-055

Title:

Unit 2 250 Volt Battery Charger

Description:

This change replaced the 250 VDC battery charger #2 feed and load breakers which were undersized for their application. The AC breaker at MCC 29-2 and the DC breaker at MCC 2 were replaced. Analysis indicated that the need for cable replacement was not immediate and could wait until Q2R14. The charger was inoperable at the time of this design change. The charger is rated for 250 amps DC and the DC output current was 275 to 287 amps DC due to the current limit being set for 110 to 115%

of the rated value. The corresponding AC current is 130 amps for 287 amps DC; the size of the AC breaker is 125 amps, if the charger is in current limit for more than 10 minutes, it is likely that one or both of these breakers will trip.

Result: Unreviewed safety question does not exist. The breakers are being replaced to prevent inadvertent tripping of the chargers By increasing the size, the breaker is less likely to trip. The failure modes of the new breakers are the same as the existing breakers. The battery charger is not needed during a LOCA. The Attachment A, SVP-98-113, Page $2 of 153

breakers are properly sized to protect the associated MCCs from a fault at the charger. Thus, the consequences of a malfunction of equipment have not changed.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-96-040

Title:

Reactor Building Siding

Description:

A patch was designed to cover three access holes cut in the inner panels of the Reactor Building siding. The siding has an inner and outer panel with insulation between. The outer panels were loosened due to high winds. The 9-inch square access holes were cut in the inner panels in order to tie off the outer panel to safe support points. This change was issued to provide repair details for the holes. A patch approximately 12 inches by 14 inches was installed over each of the three holes made of 18 gauge galvanized steel and secured with sheet metal screws. The patch and screws were caulked to prevent leakage.

Result:

Unreviewed safety question does not exist. The repair detail is a passive repair designed to restore secondary containment integrity for all modes of operation assured without interaction with other structures, systems, or components. The repair detail does not affect any equipment or equipment failures. A calculation evaluated the structural seismic issues and found the repair to meet applicable requirements.

The probability of an accident is not increased nor are the conseo tences of a malfunction of equipment important to safety.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-96 003, DCP 9600007

Title:

Safe Shutdown Make-up Pump System Flow Control Valve

Description:

This change replaced the Limitorque worm, worm gear, motor pinion, and worm shaft gear for the Safe Shutdown Makeup Pump System (SSMP) Flow Control Valve. The installed worm gear is the hammerblow type which allows worm gear movement without valve movement. This causes slower response times and erratic flow control. The installation of the different worm gear, motor pinion and worm shaft gear allows smoother operation of the salve and more positive flow control.

Result:

Unreviewed safety question does not uxi:t. The design function of the SSMP remains unchanged as a result of the installation of this change. Replacing the gearing in the valve operator will improve the flow control capability of the valve by allowing smoother valve operation. This change has no impact on existing failure modes. The installation of the new gearing in the valve's Limitorque operator will increase the reliability of the valve by allowing smoother valve operation and more positive flow control capabilities.

The change in gearing in the valve operator will not introduce any new failure modes or increase the probability of any equipment failures. The change in the gearing of the operator of the valve will not increase the consequences of any accident.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change I

Attachment A. SVP-98-ll3, Page 53 of 153

Evaluation Reference Number: E04-1 96-004

Title:

RHR LPCIInboard injection Valve

Description:

This change drilled a hole through the reactor recirculation piping inlet side of the valve Hex-wedge disc. Drilling a hole in the high pressure side of the disc permits the pressure in the bonnet to equalize with the piping system, thereby reducing differential pressure and eliminating pressure locking susceptibility. Pressure locking can occur in flexible-wedge disc valves when fluid becomes pressurized within the valve bonnet and the actuator is not capable of overcoming the additional thrust requirements resulting from the differential pressure created across both valve discs by the pressurized fluid in the valve bonnet.

Result: ' 1 reviewed safety question does not exist. As discussed in SOER 84-7 and Comed Study Report, the drilled hole will improve the reliability of the valve by preventing any potential pressure locking. This will improve the reliability of the valve to perform its function to open for LPCI injection and close for the required primary containment isolation function. This change does not alter the function of the valve. It is entirely internal to the valve and does not reduce the valw's capability to stroke or maintain the pressure .

boendary. No new interfaus are created between the valve and other systems. The change does not create j any new failure modes or system interfaces for the valve.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-007 and 008

Title:

Update Reactor Recirculation Pump Discharge Valves Descrip; ion: These changen for the reactor recirculation pump discharge valves eliminate the existing triple packing configuration and replaces it with a single live load configuration via a packing iasert kit which consists of a packing insert, retainer, gland, gland flange, studs, nuts, bolts, lockwashers, and gasket.

The packing was replaced with the station standard ARGO 5-ring live load packing sized for the new packing insert kit. The overall length of the stem was reduced. The yoke was shortened consistent with the reduced packing height, the yoke material was changed, yoke upper flange is thicker, the piping, temperature element, and sight glass for the stem leak-ofiline was eliminated, and bonnet vent line piping, valves, and sight glass were eliminated. Conduits were shortened and for the SA Valve, thejunction box support design / location changed. Various interferences were temporarily removed and reinstalled or eliminated. The original three-stage packing design has exhibited a history ofleakage. The change to single packing configuration was based on various recommendations and results from industry / Comed experience.

Result: Unreviewed safety question does not exist. The design changes to the valve yoke, stem and gland kit do not affect any plant operating function. The stem leak-offline and vent line are both considered non-functional and therefore, elimination of this line will not affect operations but enhance operations by eliminating the need to verify valve position. Pressure locking was determined not to be applicable.

Electrical and steel changes were installed to original design criteria and will not introduce any new type of failure. The valve yoke was upgraded to withstand higher seismic accelerations than required to improve design margins and eliminate flexing of the yoke and cracking of fillet welds. The design change from triple to single packing is to provide an improved design to reduce possibility of packing failure. No new Attachment A, SVP-98-ll3, Page 54 of 153 L

failure modes have been introduced. The new valve yoke is designed to be more rigid and have a greater {

thrust capacity. Therefore, consequences of a malfunction of the RPS pump discharge valve is not increased.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change l

Evaluation Reference Number: E04-1-96-011

Title:

Electro-Hydraulic Control (EHC) System l

Description:

This change replaced the existing primary and secondary pressure amplifiers and added two Steamline Resonance Compensator (SLRC) circuit boards to the Electro-Hydraulic Control System. The  :

purpose of this change was to improve the response of the EHC System to pressure oscillations in the Main  !

Steam Piping. During recent plant startup activities, excessive position changes in the Turbine Bypass Valves were noted. The valves could not adequately control reactor pressure when in the automatic mode I above 900 psig. After being placed in automatic mode, the Bypass Valves were opening and closing in a divergent manner and had to be placed under manual control to maintain reactor pressure. The new circuit boards will help to dampen pressure oscillations by controlling the response of the Turbine Bypass Valves and Turbine Control Valves. They sense changes in Main Steam Line pressure and direct the Bypass Valves and Control Vab>es to respond accordingly.

Result: 'Unreviewed safety question does not exist. The new circuit boards will improve EHC System control of reactor pressure by filteeir pressure pulses which are of a frequency that matches the main steam line resonant frequency thm avoiding oscillations. Thus, the EHC System will perform more etliciently.

The installation of the circuit boards does not increase the probability or consequences of an equipment failure. The operation of the new boards have been reviewed to ensure that the probability of equipment failure has not been increased. Within the complexity of the existing design, the addition of the new boards will increase system performance during transients without significantly decreasing reliability. No other equipment important to safety will be affected by the installation of the new circuit boards. No new failure modes are created.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-023 A

Title:

Reactor Recirculation Pump Suction Piping

Description:

This change performed a weld overlay on a weld located on the 'B' Loop Reactor Recirculation Pump Suction Piping. Ultrasonic examination of this weld identified linear indications at various locations around the weld circumference. As a corrective measure, this weld overlay will ensure that this piping will continue to function as designed.

Result: Unreviewed safety question does not exist. Applying a weld overlay on suction piping of the Reactor Recirculation pump does not affect the operation of the system. The application of the weld overlay on the suction piping weld returns the piping to its design criteria. There are no new interactions with any other system, structure, or component introduced. The weld overlay provides an additional pressure boundary around the c;rcumference of the welded area; therefore, increasing the integrity of the Attachment A, SVP-98-113, Page 55 of 153

Reactor Recirculation suction piping. The addition of the weld overlay will reduce the possibility of a failure of the suction piping in the area of the problem weld. There are no new failure modes introduced.

Probability of a malfunction of equipment important to safety is not increased by the application of this weld overlay.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: . E04-1-96-023B 1

Title:

Reactor Recirculation Pump Suction Piping

Description:

This change performed a weld overlay on a weld located on the 'A' Loop Reactor Recirculation Pump Suction Piping. Ultrasonic examinatica of this weld identified linear indications at various locations around the weld circumference. As a corrective measure, this weld overlay will ensure that this piping will continue to function as designed.

Result: Unreviewed safety question does not exist. Applying a weld overlay on suction piping of the Reactor Recirculation pump does not affect the operation of the system. The application of the weld overlay on the suction piping weld returns the piping to its design criteria. There are no new interactions with any other system, structure, or component introduced. The weld overlay provides an additional pressure boundary around the circumference of the welded area; therefore, increasing the integrity of the Reactor Recirculation suction piping. The addition of the weld overlay will reduce the possibility of a failu.e of the suction piping in the area of the problem weld. There are no new failure modes introduced.

Probability of a malfunction of equipment important to safety is not increased by the application'of this weld overlay.

Safety Evaluation Nurnbe'r: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-023C

Title:

Reactor Recirculation Pump Suction Piping l l

Description:

This change performed a weld overlay on a weld located on the ' A' Loop Reactor  !

Recirculation Pump Suction Piping. Ultrasonic examination of this weld identified linear indications at j various locations around the weld circumference. As a corrective measure, this weld overlay will ensure  :

that this piping will continue to function as designed.

Result: Unreviewed safety question does not exist. Applying a weld overlay on suction piping of the Reactor Recirculation pump does not affect the operation of the system. The application of the weld overlay on the suction piping weld returns the piping to its design criteria. There are no new interactions with any other system, structure, or component introduced. The weld overlay provides an additional pressure boundary around the circumference of the welded area; therefore, increasing the integrity of the Reactor Recirculation suction pipirg. The addition of the weld overlay will reduce the possibility of a l failure of the suction piping in the area of the problem weld. There are no new failure modes introduced.

Probability of a malfunction of equipment important to safety is not increased by the application of this weld overlay.

Attachment A, SVP-98-l l3, Page 56 of 153

)

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-023G

Title:

Reactor Recirculation Pump Suction Piping

Description:

This change performed a weld overlay on a weld located on the 'A' Loop Reactor Recirculation Pump Suction Piping. Ultrasonic examination of this we!d identified linear indications at I various locations around the weld circumference. As a corrective measure, this weld overlay will ensure that this piping will continue to function as designed.

Result: Unreviewed safety question does not exist. Applying a weld overlay on suction piping of the Reactor Recirculation pump does not affect the operation of the system. The application of the weld overlay on the suction piping weld returns the piping to its design criteria. There are no new interactions  ;

with any other system, structure, or component introduced. The weld overlay provides an additional {'

pressure boundary around the circumference of the welded area; therefore, increasing the integrity of the Reactor Recirculation suction piping. The addition of the weld overlay will reduce the possibility of a failure of the suction piping in the area of the problem weld. There are no new failure modes introduced.

)

Probability of a malfunction of equipment important to safety is not increased by the application of this i weld cverlay.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1 96-024 t

Title:

Control Rod Drive (CRD) Flow Control Valve

Description:

This change replaced the existing 1* trim set with a 3/4" trim set for CRD Flow Control ,

Valve. This valve trim set change consisted of replacement of the existing seats (2), ball, cage, and guide .

pin. The control valve had a history of oscillating and cycling open then back closed while in the automatic mode. It was found that the oscillations were caused by internal binding of the valve. Also, the existing valve trim set was found to be incorrectly sized, causing flow instability during low system flow requirements due to a flow capacity greater than two times maximum. The 3/4" trim set will allow the valve to be opened further during both low flow and normal flow conditions.

l Resuit: Unreviewed ,afety question does not exist. The installation of the trim set will help to reduce flow instabilities inside the valve and improve the rangeability of the valve and will greatly reduce the possibility .

ofinternal binding of the valve. The design function of the valve is not changed. There are no new l interactions with any other system, structure, or component created by the installation of this change. The  ;

change will increase the reliability of the valve by reducing the possibility of flow instabilities inside the j valve. No new failure modes are created. The flow control valves do not interact with any equipment 1 required for the safe shutdown of the plant. Therefore, probability of a malfunction of equipment important j te safety is not increased by the installation of this change.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Attachment A, SVP 98-113, Page 57 of 153 l

-1 l

ll Evaluation Reference Number: E04 1-96-025 l

Title:

Reactor Feedwater Pump Discharge Lines i p

Description:

. This change installed a Leading Edge Flow Meter (LEFM) flow measurement system that-.

l provides an independent and more accurate method for measuring feedwater flow. The system uses modern state-of the-art uitrasonic flow measurement technology and provides a unique and diverse measurement technique. It is insensitive to operational effects such as fouling which affect the accuracy of the venturis.

A detailed analysis was performed by an outside agency to verify the accuracy of the system. Overall, the analysis, lab-testing and site-specific testing as well as operating experience at other utilities provides high j confidence in the accuracy of the LEFM feedwater flow measurement system. ' A comprehensive flow

]

testing program was recently performed at QCNPS using this system to es whether venturi fouling or

{

other calibration shifts were present causing reduced electrical output of the unit. This assessment '

confirmed a measurement bias in the venturi flow indication higher than actual for total feedwater flow with '

one line indicating a higher disparity in flow than the other for the duration of the test.' This indicates that QCNPS is operating below its authorized licensed power level.

Result: Unreviewed safety question does not exist. Since the installation of the LEFM flow measurement system is adequately isolated from the Feedwater and electrical distribution systems, there will be no affect to equipment failures. Use of the LEFM to correct feedwater flow measurement used in the reactor thermal I power calculation will not introduce any new or different failure modes to systems or equipment. This change does not add or delete any plant hardware on safety-related structures or components. This activitiy .

' does not' create the possibility of a malfunction of equipment important to safety of a different type thanyan evaluated previously in the UFSAR. The LEFM is a non-intrusive fluid temperature and volumetric flow rate measuring and indicating device. The LEFM transducers are securely mounted on each feedwater flow line using a mounting apparatus that maintains transducer alignment. The determination of core thermal ,

power through use of the LEFM provides greater accuracy and less' uncertainty than the original ventun '

instrumentation system. I o

1 Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96 031, DCP 9500075

Title:

. Standby Liquid Control Pump Discharge check Valve i

Description:

This change replaced the lift-check valve located on the discharge of the IB Standby Liquid :f

- Control Pump with a piston check valve. The discharge check valve was not seating to prevent backflow. l Since replacement' parts do not exist for the existing valve, a new valve was installed. The new valve is  !

hydraulically equivalent to or better than the existing valve passing full flow at 89% of stroke with a lower  !

. pressure drop.  !

- Resulti Unreviewed safety question does not exist. There are no new failure modes associated with the

. new check valve. Failure of the check valve in the open position would still allow two pump flow. Failure .i of the check valve in the closed position would still permit the system to inject sodium pentaborate with one  !

pump. The consequences of a reactor accident are not affected by failure of the new check valve during mitigation of an' ATWS or other event. Pump operability is verified monthly using demineralized water. If wear' problems with the new check valve occur, they would most likely be discovered during the monthly f surveillance as the problem with the existing check valve was discovered.-

I

.' Attachment A, SVP-98-113, Page 58 of 153

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change l Evalcation Reference Number: E04-1-96-039 1

Title:

Reactor Water Cleanup System Outlet Valve (Outboard)

I J

Description:

This change rewired the field wiring to the red light indication for the Reactor Water Cleanup System Outlet Valve. This change was to the field wiring tenninations at terminal block TB9. A. 1 new wire was added frbm TB9 to the red light indication. The existing wire was abandoned in place. The i light indication did not work. )

1 Result: Unreviewed safety question does not exist. The change is only to wiring terminal points for the I indicator light. The operating parameters have not changed. The change does not affect any other i components during normal operation since it is only changing the points where the wire was landed on the I terminal strip. The operation of the light indication and associated control circuit have not changed. The equipment failure modes of the changed circuit have not changed. The end of the abandoned cable was taped to avoid contact with any other circuits. The other end of the connection does not have any lead. The existing lead is soldered at the back of the terminal strip in the panel. The consequences of a failure of this valve to perform its safety function have not changed.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-042

Title:

Emergency Diesel Generator HVAC Vent Fan i

Description:

This change replaced the existing 75 VA Control Power Transformer (CPT) in MCC 19-1 l cubicle G1 with a 200 VA CPT. The CPT provides 120 volt control power for the Unit I feed to the diesel .

HVAC Vent Fan. With the existing CPT, there is insufficient voltage for the relays to pick up to start the motor during a LOCA with offsite power available. The pickup voltage will be increased with the new transformer.

Result: Unreviewed safety question does not exist. The new transformer has a higher rating with a lower impedance and better turns ratio. The result is a higher available voltage for the control relay and the motor contactor relay. The new transformer was evaluated by a calculation. The voltage evaluated is acceptable '

1:

with suffficient margin to pick up the affected relays. The new transformer was seismically mounted and wired using the same wire as the existing transformer. There are no changed interactions with other l structures, systems, or components. There is no adverse affect on plant operations. There are no new l failure modes as a result of this change. A failure of the new transformer is not any more likely than the existing transformer and meets the same design criteria.

Safety Evaluation Number: NA Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-043 l

Attachment A, SVP-98-ll3, Page 59 of 153 I-l

Title:

RHR Inboard Shutoff Valve to Loop Injection

Description:

This change installed two control relays and one terminal block in the back of each motor control center 18/19-5 cubicles B4 and E4. The relays were wired in the control logic for the RHR Inboard Shutoff Valve to Loop Injection Valves in place of the existing 42/MC and 42/MO (close & open starter contactor relays). A contact off of the respective relay will pull in or drop out the existing 42 contactor relay. In the event of an undervoltage condition there is insufficient voltage available at the 42/MO and 42/MC relays to assure that the relays will function properly to energize the RHR Inboard Shutoff Valves in the present configuration. By placing an interposing relay in the circuit and moving the 42 relays to another position in the circuit, both relays will function properly under all plant operating conditions.

Result: Unreviewed safety question does not exist. The new relays require less current to pickup than the existing relay. A calculation was performed to analyze the new minimum voltrage available for the new relays and 42M contactors. There are no new interactions with c?her systems. The new relays do not electrically connect to any other circuits, thus there will be no advem effect on the control circuit operation and no new interaction with other structures, systems, or components. TN failure mode of the new relay is the same as for the existing control circuit. A failure will most likely result m e failure of the valve to open or close. This was the same failure mode of the existing circuit. The change does . ot affect the consequences of an equipment malfunction. Since the coils will have adequate voltage, the valves will perform their intended design function. Thus the probability of malfunction of equipment has decreased.

1 Safety Evaluation Number: NA Typeof SafetyEvaluation: Exempt Change Evaluation Reference Number: E04-2-96-025

Title:

Reactor Feedwater Pump Discharge Lines Pescription: This change installed a' Leading Edge Flow Meter (LEFM) flow measurement system that provides an independent and more accurate method for measuring feedwater flow. The system uses modern state-of-the-art ultrasonic flow measurement technology and provides a unique and diverse measurement technique. It is insensitive to operational effects such as fouling which affect the accuracy of the venturis.

A detailed analysis was performed by an outside agency to verify the accuracy of the system. Overall, the analysis, lab-testing and site-specific testing as well as operating experience at other utilities provides high confidence in the accuracy of the LEFM feedwater flow measurement system. A comprehensive flow testing program was recently performed at QCNPS using this system to assess whether venturi fouling or other calibration shifts were present causing reduced electrical output of the unit. This assessment }

confirmed a measurement bias in the venturi flow indication higher than actual for total feedwater flow with one line indicating a higher disparity in flow than the other for the duration of the test. This indicates that QCNPS is operating below its authorized licensed power level.

l Result: Unreviewed safety question does not exist. Since the installation of the LEFM flow measurement i system is adequately isolated from the Feedwater and electrical distribution systems, there will be no effect to equipment failures. Use of the LEFM to correct feedwater flow measurement used in the reactor thermal i power calculation will not introduce any new or different failure modes to systems or equipment. This change does not add or delete any plant hardware on safety-related structures or components. This activitiy

. does not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the UFSAR. The LEFM is a non-intrusive fluid temperature and volumetric flow rate measuring and indicating device. The LEFM transducers are securely mounted on each feedwater flow l

Attachment A, SVP-98-ll3, Page 60 of 153 l

I

line using a mounting apparatus that maintains transducer alignment. The determination of core thermal l

power thro igh use of the LEFM provides greater accuracy and less uncertainty than the original venturi instrumentation system.

Sakty Evaluation Number: NA Type of Safety Evaluation: Design Change f

Evaluation Reference Number: DCP 9600185

Title:

Emergency Diesel Generator (EDG) Fuel Oil Transfer Pump  !

Description:

This change re;ilaces the existing 150 VA Control Power Transformer (CPT) in Motor i Control Center (MCC) 19-1 cubicle C5 with a 300 VA CPT. The CPT provides 120 volt control power for the Unit I feed to the Diesel Fuel Oil Transfer Pump (1-5203). This change also removes an unused control relay that is wired in parallel with the motor contactor relay.

Result: Unreviewed safety question does not exist. This change increases the pick up voltage to start the motor during a Loss-of-Coolant Accident (LOCA) with offiste power available and operating at minimum acceptable operating voltages. This will assure that the pump performs its design function during this scenario. The existing CPT provided insufficient voltage to start the pump in this scenario. The new transformer will be seismically mounted and use existing wiring. There are no changed interactions with other structures, systems and components and no adverse impact on plant operations. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600189, DCN 001362M

Title:

Pump Discharge Line for the Unit 1 High Pressure Coolant Injec' tion (HPCI) Pump

Description:

This change installs one additional support on the pump discharge piping line for the Unit 1 HPCI pump.

Result: Unreviewed safety question does not exist. The additional pipe support will reduce the nozzle loadings on the HPCI discharge pump which will, in turn, reduce the vibrations and wear in the pump and its components. Therefore, the effect of this change is to increase the reliability of the HPCI Pump and system. The system operates the same t.s before the change. There are no new or changed interactions with other structures, systems, and components. No new failure modes are introduced.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change i

Evaluation Reference Number: DCP 9600205

Title:

Unit 2 Reactor Core Isolation Cooling (RCIC) Motor Operated Valves (MOVs 2-1301-16,17,22, 25,26,48,49,53,60,61, and 62)

Attachment A, SVP-98-113, Page 61 of 153 l

Description:

This change revises the electrical configuration of the control circuit wiring for thhbove RCIC MOVs in response to NRC Information Notice 9218.

Result: Unreviewed safety question does not exist. This change does not affect how the valves function or operate electrically. Permissive devices have been electrically rearranged to ensure that the required limit and torque switches protect the valves from over torque conditions postulated to occur from " hot shorts" in the event of a fire. This change does not introduce new failure modes. This change does not affect the ability of the valves to perform their intended design function. The probability of a malfunction of I equipment is decreased in the event of hot shorts induced by fire, where these valves are required to operate to ensure safe plant shutdown. )

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600206

Title:

Modification on Electrical Configuration of Control Circuit Wiring for RCIC MOVs 1-1301-16, 17,22,25,26,48,49,53,60,61, and 62

Description:

This change does not affect how the valves function or operate electrically. Permissive devices have been electrically rearranged to ensure that the required limit and torque switches protect the valves from over torque conditions postulated to occur from " hot shorts" in the event of a fire, for cases where these valves are required to operate for safe shutdown.

l Result: Unreviewed safety question does not exist. This change decreases the effect of postulated hot  !

shorts on the valve control circuitry but does not functionally change the operation of the valve. This will preclude valve or operator damage from over torque conditions. This change does not introduce new failure modes. This change does not affect the ability of the valves to perform their intended design function. He probability of a malfunction of equipment is decreased in the event of hot shorts induced by fire, where  ;

these valves are required to operate to enwre safe plant shutdown.

l Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600208

Title:

Modification On Electrical Configuration ofControl Circuit Wiring for RHR MOV's 1-1001- l SA&B; 7A, B, C&D; I 6B; 28A&B; 29A&B; 34A&B; 36A&B; and 43 A, B, C&D

(

Description:

In addition to the change in the electrical configurations of the MOVs described above, this change also involves the replacement of control room valve control hand switch 1-1001-5A&B from an open/close contact spring return to auto configuration to a position maintaining open/off/close contact arrangement. Le current control circuit design leaves the valves subject to a control circuit hot short that could bypass 1he limit / torque switch assemblies and damage the valve due to over torque. The new configuration will allow the throttle control of the valve but will eliminate the previous hot short failure mode by maintaining a control power open circuit downstream of the limit / torque switch assemblies. These changes are required to remove a potential mode of failure for the referenced MOV's in the event of an Appendix R fire where control circuit hot shorts are postulated to bypass torque and limit switch permissive devices.

Attachment A, SVP-96-il3, Page 62 of IS3

Result: Unreviewed safety question does not exist.' This change will not adversely affect the RHR system, and will in fact increase system reliability by eliminating a potential mode of failure for the referenced valves. The control characteristics of the SA & 5B valves will remain unchanged (throttle open/close) and the previous automatic closure capability will be climinated due to the hot short potential previously discussed. This change does not introduce any new failure modes. This change does not affect the ability of the valves to perform their intended design function and enhances the capability by making the valves more reliable. Elimination of a potential failure mode (hot shoas) enhances the equipment's reliability to function properly, when required.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP %00209

Title:

Modification On Electrical Configuration of Control Circuit Wiring for RHR MOV's 2-1001-SA&B; 7A, B, C&D; 28A&B; 29A&B; 34A&B; 36A&B; 43A, B, C&D; and 50

Description:

The current control circuit design leaves the valves subject to a control circuit hot short that could bypass the limit / torque switch assemblies and damage the valve due to over torque. These chardes are required to remove a potential mode of failure for the referenced MOV's in the event of an Appendix R fire where control circuit hot shorts are postulated to bypass torque and limit switch permissive devices.

Result: Unreviewed safety question does not exist. This change will not adversely affect the RHR system, and will in fact increase system reliability by eliminating a potential mode of failure for the referenced valves, This change does not affect how the valves function or operate electrically This change does not introduce any new failure modes. This change does not affect the ability of the valves to perform their intended design function and enhances the valves' reliability. Elimination of a potential failure mode (hot shorts) enhances the equipment's reliability to function properly, when required.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change 1

Evaluation Reference Number: DCP 9600224

{

i

Title:

Upgrade To Existing Power Feed Cables From Switchgesr 28 to MCC 28-1 A

Description:

This change is being issued in response to a potential for periodically exceeding the MCC ,

28-1 A feed cable rated continous ampacity during plant operations. Such a condition may shorten the cable l life due to insulation degradation. The fact that the magnitude of the connected MCC load requires that the MCC feed breaker at SWGR 28 be set substantially higher than the rated continuous current of the feed cable dictate that the cable replacement is needed. Routing of the new power cables will utilize the same  ;

secondary containment penetration (modified as necessary) used for the original cables. The loads on the l

existing cables are being administratively controlled until this replacement is complete. 1 1 Result: Unreviewed safety question does not exist. The functional interactions of the new feeder cables feeding MCC 28-1 A with other systems, structures and components remain unchanged. The only change is the ampacity rating of the cables. Installation of the new power feed cables will not introduce any new failure modes because the failure modes of the new cables are identical to the currently installed equipment.

Attachment A, SVP-98-113, Page 63 of 153 l

\

The consequences of an accident will not be increased because eliminating the concerns with cable degradation and increasing the interrupting rating of MCC 28-1 A only increases the reliability of the loads that may be required to decrease the consequences of a LOCA.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600228

Title:

480 VAC MCC 29-2 Feed Cables

==

Description:==

This change upgraded the existing po$ver feed cables from Switchgear 29 to Motor Control Center (MCC) 29-2 by replacing the existmg 250 MCM cables with new 500 MCM cables. The new 1/c

1. 1/0 cable was installed to provide a neutral for emergency lighting cabinets. The existing and new power feed cables are classified as Safe Shutdown (SSD) cables, but the neutral cable is not classified as a SSD cable. The existing power feed cables and neutral were removed and pulled back into the Reactor Building and abandoned in place. The feeder breaker for MCC 29-2 was reset and four of the breakers at this MCC required changes. Two were placed with GE type THED breakers, one was removed and not replaced, and a current limiter was added to the fourth. This change was issued in response to a potential for periodically exceeding the MCC 29-2 feed cable rated continuous ampacity during plant operations. This could shorten the cable life due to insulation degradation.

Result: Unreviewed safety question does not exist. The new cables and breakers installed function the same as the ones being replaced. When the upgraded power feed functions as intended, plant operation under normal and accident conditions will be less restrictive. No new failure modes are introduced.

The new feeds allow greater operator flexibility in operating loads without concerns of cable degradation.

There are no changed interactions with other systems, structures, or components. This change simply changes the existing power feed cables with ones of greater ampacity rating. No new failure modes are  ;

introduced. The breach of secondary containment and fire barrier were controlled per Station Procedurcs and Approved Fire Protection Program. The new power feed eliminates concerns with cable degradation. I The new cables function the same as the existing cables. These changes enhance the ability of the components to perform their designed function. Eliminating concerns of power feed cable degradation and increasing the short circuit rating of MCC 29-2 only increases the reliability of the loads powered by this MCC. l Safety Evaluation Number: NA Type of Safety Evaluation: Design Change l

Evaluation Reference Number: DCP 9600250

Title:

Check Va?ve Into HPCI Keep Fill Line for Unit 1

==

Description:==

A check valve was installed due to concerns regarding the potential formation of vapor pockets in the HPCl discharge lines. Vapor pockets are undesirable because they can lead to water hammer during HPCI startup. In addition, if the Essential Service System (ESS) fill pump used to feed the keep fill system were to fiil, a flow path exists from the HPCI suction piping to the torus through the keep fill system. Installation of the new check valve will prevent backflow through the keep fill system during this lineup.

Attachment A, SVP-98-113, Page 64 of 153

Result: Unreviewed safety question does not exist. There are no new failure modes associated with this installation. The new check valve and associated IST test taps have been analyzed in a design basis '

calculation. The new check valve has been placed on the IST program to ensure valve operability. This change does not change the function of the keep fill system. The installation prevents backflow in an l alternate lineup which does not increase the probability of an accident, and will maintain the HPCI suction and discharge lines filled, thereby preventing water hammer. The installetion of the new check valve does i not increase the probability of a malfunction important to safety.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600251

Title:

Check Valve into HPCI Keep Fill Line for Unit 2 l

Description:

A check valve is being installed due to concerns regarding the potential formation of vapor pockets in the HPCI discharge lines. Vapor pockets are undesirable because they can lead to water hammer during HPCI startup. In addition, if the Essential Service System (ESS) fill pump used to feed the keep fill

{

system were to fail, a flow path exists from the HPCI suction piping to the torus through the keep fill  !

system. Installation of the new check valve will prevent backflow through the keep fill system during this lineup.

Result: Unreviewed safety question does not exist. There are no new failure modes associated with this installation. The new check valve and associated IST test taps have been analyzed in a design basis calculation. The new check valve has been placed on the IST program to ensure valve operability. This change does not change the function of the keep fill system. The installation prevents backflow in an alternate lineup which does not increase the probability of an accident, and will meintain the HPCI suction and discharge lines filled, thereby preventing water hammer. The installation of the new check valve does not increase the probability of a malfunction important to safety.

Safety Evaluation Number: NA Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600282

Title:

Radwaste Mixing Tank Ultrasonic Level Monitor

Description:

This change replaced the existing ultrasonic level monitor and detector in the Radwaste Mixing Tank with new ultrasonic liquid level monitor and detector devices. The existing ultrasonic level monitor does not function and because the model is obsolete an exact like-for-like ultrasonic level monitor cannot be procured. The level detector was also repLeed because the existing one will not communicate I with the new monitor.

Result: Unreviewed safety question does not exist. The new ultrasonic level monitor will detect level in the same manner as the existing one did before it failed. The new monitoring device provides the same outputs to the rest of the system as the existing one did. There are no changes to the overall system created by the installation of the new monitor. No new failure modes are created. The new monitor and detector ratings are the same. A failure of the new monitor and detector is not any more likely than the existing monitor and detector. 'Ihis replacement does not change the probability of a design accident because

, Attachment A, SVP-98-ll3, Page 65 of 153 l

neither the monitor, detector, or radwaste tank that is monitored interfaces with equipment considered in accident scenarios. The new equipment meets the same design criteria.

Safety Evaluation Number: NA Type of Safety Evaluation: Document Change Request Evaluation Reference Number: DCR 4-94-159

Title:

Master Equipment List Updated

Description:

This change updated the Master Equipment List (MEL) by deleting equipment / devices from Failure Mode and Effect Analysis (FMEA) Section of the MEL. During component classification verification program, it was observed that components listed in this section are also listed in the component classification binders. To avoid the duplication of the same data, it was determined that such components should be deleted from FMEA section of the MEL. However, if replaced with a non-like-for-like replacement, an evaluation is required to verify that the basis for exclusion remains applicable. This change also was a result of the verification program.

Result: Unreviewed safety question does not exist. This is a document change only and no changes have been made to the structure, system, or component. He updating of the MEL will provide correct classification information, which will enhance the procurement of correctly classified components. These changes do not create any changed interactions with other systems, structures, or components. This in turn will reduce the potential for component failure. The updating of the MEL will neither affect component failure nor create any new failure mode.

Safety Evaluation Number: NA Type of Safety Evaluation: Drawing Change Request Evaluation Reference Number: DCR 4-95-041

Title:

Various Diagram Changes

Description:

This change updated wiring diagram 4E-1816J to change the color of conductor from i

"ORG" to "WHT" of cable routed between thejunction box and valve bellows pressure switch. Wiring i diagram 4E-1819A was updated to de-terminate "BLU-2" and " RED-2" conductors of cable 17493 from l terminals 53 and 54 respectively in thejunction box and re-terminate at terminals 57 and 58 respectively in l

the samejunction box. Schematic and wiring diagrams were also updated in various outbuildings and plant i

rooms in the HVAC system and a drawing was updated for the 125VDC battery and charger system to ,

change a cable size and identify battery cells. These drawing changes were made to agree with as-built  !

conditions.  !

i Result: Unreviewed safety question does not exist. The information on the drawings revised per this DCR agree with the modified condition of the plant. Plant operations associated with these systems will improve, j as the revised information shown on these drawings will agree with the latest plant condition. No system j interactions are changed by updating the drawings. Showing the latest plant conditions will improve maintenance of the systems, thereby reducing the potential for equipment failure. No changes were made which could degrade the form, function, or reliability of the affected structure, systems, and components. ,

Therefore, these changes do not create any new failure modes. l Attachment A, SVP-98113, Page 66 of 153

i g

l l  %

b .. . .

Safety Evaluation Number: NA Type of Safety Evaluation: Setpoint/ Scaling Change Request Evaluation Reference Number: SSCR 9600334 i l

Title:

2A 125 VDC Battery Charger.

Description:

- The existing AC supply circuit breaker at Motor Control Center (MCC) 28-2 cubicle Cl l was replaced with a larger size circuit breaker. The lowest setting the current limiter for the battery charger

! is able to be set is .117.5% ofits rated output. At this setting nuisance tripping of the AC supply breaker occurs when the charger is in current limit for approximately one hour. Currently, Tech Specs do not have .

l specific requirements for battery charger operation at rated load. Upgraded Tech Specs (TSUP) require the battery charger to be able to operate at rated load for four hours. Thus, in order to allow the battery charger to meet TSUP requirements, the breaker was replaced.

I Result: Unreviewed safety question does not exist. The failure modes of the new circuit breaker are the .

l same as the existing breaker. Since the setting of the current limiter is below the cable ampacity, thermal protection for the cable is assured. For a short circuit in the cable feeding the charger, the new breaker still provides adequate protection for the cable. The new breaker is safety-related, therefore it meets the same

! equipment qualification requirements of the existing breaker. Thus, replacing the breaker will not affect the likelihood of equipment failures or the' severity of an equipment failure. Replacing the breaker will make it less likely to trip, therefore assuring that it will be available during a LOCA to recharge the battery and -

supply 125 Volt DC loads. Thus, the consequences of the accident have not increased. The breaker will

! . meet the safety qualification requirements for installation in the MCC. Thus, it is not any more likely to fail l than the existing breaker and the probability of a malfunction of equipment important to safety does not L increase.

l - Safety Evaluation Number: ' SE-91-270 Type of Safety Evaluation: Minor Design Change Evaluation Reference Number: PO4-2-91-060 l

j' .

Title:

Unit 2 Main Generator l

Description:

Filter circuit was added to Main Generator backup fault detector relay to prevent false trips -

j- of the relay caused by electrical noise.

Result: Unreviewed safety question does not exist. The filter circuit does not affect the relay's intended function.

Safety Evaluation Number: SE-91-408 - Type of Safety Evaluation: Modification

, Evaluation Reference Number: M04-2-88-080

Title:

Installation of Seal Gate Supports in Dryer / Separator Pit i

n Attachment A, SVP-98-l13, Page 67 of 153

Description:

This modification installed supports or alignment brackets for allowing temporary installation of a seal gate between the Dryer Separator storage pool and the reactor refueling cavity. The purpose is to provide a more reliable water seal between the Dryer Separator storage pool and the reactor refueling cavity. This seal is only needed for maintenance in the reactor cavity while the Dryer Separator is stored in the Dryer Separator storage pool.

Result: Unreviewed safety question does not exist. This modification will not impact c: create any new equipment / operation failure modes. The Dryer Separator storage pool is not safety-related. The change will not impact the function of any other structure or equipment. The new seal gate supports are designed to facilitate a more reliable water seal adjacent to an existing bulkhead made up of removable shield blocks between the Dryer Separator storage pool area and the reactor fuel handling area. There is no other equipment in this area which could be affected. The modification will not introduce the potential for any new accidents or failure modes or affect those already evaluated.

Safety Evaluation Number: SE-91-538 Type of Safety Evaluation: Modification Evaluation Reference Number: M04-0-81-007

Title:

N Piping 2 Abandoned in Place

==

Description:==

N 2pi ing P which supplies heating boilers with dry lay-up capabilities is being cut and capped and being replaced by the wet lay-up system.

Result: Unreviewed safety question does not exist. There is sufficient isolation between the areas that were cut and capped and the bulk N2 storage tank.

Safety Evaluation Number: SE-95-013 Type of Safety Evaluation: New Procedure Evaluation Reference Number: QCTS 930-11

Title:

Control Rod Diagnostic Test For Cooling Water Check Valve & Directional Control Valve Leakage

==

Description:==

This was a new procedure to develop diagnostic testing for the cooling water check valve and/or directional control valve on the Control Rod Drive (CRD) Hydraulic Control Units. There was no previously documented methodology for testing the CO cooling water check valve and directional control valve.

Result: Unreviewed safety question does not exist. There is no effect on current equipment failures. The control rod will be at 00. There may be new failure modes by the manual isolation valves sticking closed.

However, the rod will be at 00 and will still perform its safety function of maintaining the reactor in a shutdown condition. There are no accidents affected.

Safety Evaluation Number: SE 95-043 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: 95-020E Attachment A, SVP-98-ll3, Page 68 of 153

Title:

Motor Control Center (MCC) Overload Heater Setpoint Change

Description:

This change replaces existing overload relay heaters in MCC 19-4 Cubicle A-2 and MCC 29-4 Cubicle A-4 with new General Electric Thermal Overload Heaters (TOH).

{

l Result: Unreviewed safety question does not exist. This change responds to an identified problem with overload relay heater sizing. This change only affects the overload relay trip curves. Component sizing was done in accordance with acceptance criteria. This change will not affect normal standby gas operation or  !

operation of the associated MCC. No new failure modes have been introduced.

Safety Evaluation Number: SE-95-055 Type of Safety Evaluation: System Engineering Service Request I Evaluation Reference Number: SESR 4-3098-1

Title:

Repair of Standby Gas Treatment (SBGT) Demister Hold Down Latch j

Description:

This change replaces one of four Demister filter hold-down latches for SBGT train 1/2A with a shop fabricated hold-down latch of similar material and dimensions. A like-for-like replacement part for the broken existing latch is not available.

Result: Unreviewed safety question does not exist. The replacement latch performs the same function as the other three latches. There are no changes being made that will affect the design function of the SBGT system or create interactions with any other structure, system, or component. No new failure modes are created.

Safety Evaluation Number: SE-95-056 Type of Safety Evaluation: Procedure Change Evaluation . Reference Number: QCAP 1500-1 and 1500-13

Title:

Procedure Change for QCAP 1500-1 and 1500-13

Description:

This change relaxes the frequency of testing of certain detection systems from every six months to annually and provides a method and requirements for controlling Appendix R Safe Shutdown Light packs that become inoperable.

Result: Unreviewed safety question does not exist. The reduction in testing will save resources while maintaining compliance with NFPA code and insurance requirements. There will be no adverse effect any portion of the fire protection system. The controls on inoperab' light packs will ensure that appropriate compensatory measures are taken to ensure lighting is availab' , for the safe shutdown of the plant when required for Appendix R Safe Shutdown. The changes will have no effect on the probability of a fire occurring and will not impair the ability to safely shutdown the Unit in the event of a fire. There are no l changes being made to plant systems.

I Safety Evaluation Number: SE-95-057 Type of Safety Evaluation: Temporary Alteration Attachment A, SVP-98-113, Page 69 of 153 I

Evaluation Reference Number: NA l

Title:

Temporary Support for Unit I and Unit 2 Residual Heat Removal (RHR) Corner Room Heat .

Exchangers .

Description:

This change installs temporary supports for Unit I and Unit 2 RHR Corner Room Heat

- Exchangers (four exchangers total) to ensure adequate vertical support for these heat exchanges while an ongoing evaluation to determine actual stress levels is performed.

Result: U' nreviewed safety question does not exist. The additional supports will decrease the probability of

' failure of the RHR heat exchangers during a seismic event. The supports will increase the margin of safety l

for this equipmer,t. This change will not adversely affect the heat exchangers, surrounding equipment, or any plant operations. This change will not affect equipment failure modes. The loading of the supports has been evaluated and accepted.

Safety Evaluation Number: SE-95-058 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: QCIPM 200-4

Title:

. Reset Reactor Recirculation MG Set Scoop Tube Limits To Lower Maximum Core Flow Value

]

Description:

This change reduces the Reactor Recirculation MG Set Scoop Tube demand limiter from a j maximum core flow of 100.8 M#/hr. to a maximum core flow of 99.5 M#/hr. This change is not to enforce I safety analysis but to maintain prudent limit to licensed power level.

Result: Unreviewed safety question does not exist. This change will prevent 102% Core Thermal Power

- (CTP) from being exceeded for 1 pump run-up events from less than or equal to 102% Flow Control Line '

. (FCL). He setpoint is within normal, expected bounds oflimiter settings. The probability of reactor recirculation flow runup transient is unaffected because the change does not affect the reactor recirculation i speed controller circuitry. Tnere is no reduction in the margin of safety because the change is in a more j

conservative direction.

Safety Evaluation Number: SE-95-60 . Type of Safety Evaluation: System Engineering  !

Service Request Evaluation Reference Number: SESR4-3138

Title:

Repair of 2C Residual Heat Removal Service Water (RHRSW) Room Cooler

Description:

' L This change repairs service water leaks in the 2C RHRSW Room Cooler by installing plugs 'l in 4 tubes.

g, Result: Unreviewed safety question does not exist. It was concluded that this change will not affect plant .

operation based on a 1992 study which determined that the RHRSW room coolers had twice the required l capacity for removal of heat generated by pump operation to maintain temperatures below the equipment qualification limit (120*F). The study concluded that the maximum number of tube circuits that could be I l_

, Attachment A, SVP-98-113, Page 70 of 153 y

p

v plugged was 12 of 24. Sufficient margin is maintained so that the RHRSW system will remain available to

- perform its design function. The new pressure boundary will be as strong as before.

l Safety Evaluation Number: SE-95-062 Type of Safety Evaluation: Parts Evaluation Evaluation Reference Number: 95-068-00

Title:

Battery Replacement for Radwaste and Unit 1/2 Service Water Radiation Monitors L

Description:

' This change provides approval, per the station procedure on parts evaluations, to replace the obsolete model battery used in the Radwaste and Service Water Radiation Monitor DAM-4 Panel for batter backup with a new model battery.

L l_ Result: Unreviewed safety question does not exist. The specifications of the replacement battery meet or L exceed those of the original battery. Differences in the mountmg details of the battery in the DAM-4 Panel have been addressed. Seismic mounting of the battery in the DAM-4 Panel is not required and, in a seismic l

cvent, the battery would be confined to within the panel and would not be able to interact with other equipment.

l.

l l Safety Evaluation Number: SE-95-06'l Type of Safety Evaluation: Test Procedure Evaluation Heference Number: ,QCTS 930-13

Title:

Scram Functional Test i

Description:

This change creates a new procedure to' test / actuate the scram valves and solenoids to prove that the control rod drive is not locked and demonstrate the integrity of the Scram Discharge Isolation Valve by passing flow to the Scram Discharge Isolation Volume.-

Result: Unreviewed safety question does not exist. This test will prove that the safety-related equipment of l an individual. control rod drive will properly operate prior to startup. The control rod will remain inserted

. (00 to past full in) during the entire test and, hence, will have already met its safety function. Should the .

l control rod become stuck past full in, the test procedure directs Operators to the appropriate approved  !

l procedure for a stuck rod. The test procedure also provides instructions to prevent reactor water from bemg j

! drained to the Scram D.ischarge Volume if a scram valve becomes stuck open upon reset from the test. The procedure bumps the rods past 00 with the accumulators only partially charged. This applies less force than when the mode switch is moved to SHUTDOWN with the accumulators fully charged. Therefore, this l procedure is bounded by (will result in less wear on the seals) normal movement of the mode switch. There are no new accidents or malfunctions caused by this procedure.

Safety Evaluation Number: SE-95-0M Type of Safety Evaluation: Parts Evaluation i

Evaluation Reference Number: 95-06f-00

Title:

Alternate Replacement Timer Attachm.mt for Radwaste Radiation Monitor Relay R2 l

l Attachment A, SVP-98113, Page 71 of 153 l

l (

1

Description:

This change provides approval, per the station procedure on parts evaluations, to replace the Timer Attachment for Radwaste . Radiation Monitor Relay R2, with a newer model timer attachment.

Result: Unreviewed safety question does not exist. The replacement timer is identical to the original j except that the allowable range for the Time Delay De-energize (TDD) Setpoint changes from 0.3-30 j seconds to 0.6-60 seconds. The TDD Setpoint will remain at 2 seconds. With a tolerance of +1.2/-0 seconds, the relay will still provide adequate control room notification of a monitor problem while prever. ting unnecessary alarms. The size and weight of the replacement are identical to the original. There l are no seismic mounting requirements Safety Evaluation Number: SE-95-065 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: EPN 1/2-3904

Title:

110VAC Submersible Sump Pump

Description:

This change installs a 110VAC submersible sump pump to supply sample water to the j discharge bay composite sampler piping.

1 Result: Unreviewed safety question does not exist. This change compensates for the loss of the existing pump (EPN:l/2-3904) which has stopped working. The submersible pump will provide adequate sample flow to the discharge bay composite sampler using local 110VAC power. It has no controlling or interacting functions with other equipment. It creates no new failure modes, although a short in the pump could trip the circuit for local lighting. The possibility of the pump falling into the diffuser header and rendering the sampling system inoperable was considered. There would be no operational concerns with this pump in the 16' diameter diffuser header.

Safety Evaluation Number: SE-95-066 Type of Safety Evaluation: T:mporary Alteration Evaluation Reference Number: 95-1-25

Title:

Connect Gould Chart Recorder and PC Data Acquisition System to 5 Inputs to Process Computer l

Description:

This change connects a Gould chart recorder and PC data acquisition system to 5 inputs to the process computer.

Result: Unreviewed safety question does not exist. This change facilitates the collection of data to support determination of the cause of spurious Reactor power spiking observed on the Average Power Range Monitors (APRMs). The recorder is designed so that it does not affect the circuit it is monitoring. The circuits being monitored will be isolated by the high impedance of the recorder and signal isolators.

Potential failure modes if the recorder were to introduce interference into the system, were evaluated and accepted. The recorder does not introduce any new failure modes nor affect plant response to a transient or accident.

Safety Evaluation Number: SE-95-067 Type of Safety Evaluation: Procedure Change Attachment A, SVP-98-113, Page 72 of 153

L 1

Evaluation Reference Number: QCTS 0310-03 l

Title:

Unit 2 Emergency Core Cooling System (ECCS)

{

Description:

This change rewrites the procedure to test the auto-initiation of the ECCS and Emergency Diesel Generator (EDG) systems in response to a Loss-of-Coolant Accident (LOCA) with a loss of offsite power (LOOP) to include changes for the Tech Spec Upgrade Program (TSUP) and to include j enhancements identified during the performance of the test.

]

Result: Umeviewed safety question does not exist. The changes allow greater flexibility and add Acceptance Criteria, References, and Steps which will assure compliance with TSUP. The changes do not-affect the actual performance of the test. The scope of this change does not include any modification to the facility or the deletion of any previously performed test. No new accident or malfunction is created. As was the case previously, the test is performed with the reactor shutdown and all possible precautions are in 4 place to assure safe performance of the test. l Safety Evaluation Number: SE-95-070 Type of Safety Evaluation: As-built Drawing Change j I

Evaluation Reference Number: ASBL DCR 4-95-205

Title:

Note for Drawing M-365, Sheet I

Description:

This change adds a note to Drawing M-365, Sheet 1 indicating that only one 5/8" diameter 3 bolt per leg is required to adequately support the Standby Gas Treatment (SBGT) Units. He original i drawing shows one anchor bolt per SBGT leg while the actual installation has a location available for four  !

bolts per leg (2 legs have 3 bolts with the fourth hole empty). 1 Result: Unreviewed safety question does not exist. This change clarifies acceptable SBGT Unit anchorage which both the original design and the current configuration meet. The change is to the drawing only and does not affect plant operation.

Safety Evaluation Number: SE-95-071 Type of Safety Evaluation: Fire Protection Report Updates Evaluation Reference Number: N/A

Title:

Update the Fire Protection ' Reports To incorporate Two Submittals From Sargent and Lundy

Description:

These updates include changes in the combustible loadinE nventoryi and corrections of typographical errors.

Result: Unreviewed safety question does not exist. No systems, structures or components are being changed, and there is no effect on plant operation. The changes in combustible loading do not increase the i fire loading above the allowable limit for any of the fire zones. The safe shutdown analysis is not adversely l- affected. The editorial corrections do not affect plant procedures or operations and do not change the results or conclusions of any analysis.

Attachment A, SVP-98-113, Page 73 of 153

.l L

f- w, Safety Evaluation Number: SE-95-072 Type of Safety Evaluation: Parts Evaluation k

( Evaluation Reference Number: 95-074-00

Title:

Replacement of Reactor Level Signal Transmitters (RX LT-1-646A) '

1

Description:

This change replaces an existing General Electric model 555 transmitter with a Rosemount j model 1152 transmitter. This transmitter provides a Reactor level signal to the feedwater control system, station computer, control room indicator and recorder.

Result: Unreviewed safety question does not exist. . Critical characteristics (such as pressure rating, L  : calibration pressure and current ranges, power supply voltage, etc.) of the Rosemount transmitter are the s same as for the GE transmitter. Accuracy is better for the Rosemount model. The new transmitter is IEEE L qualified for nuclear use while the existing one was not. No new functions or failure modes are created.-

Fit, form, process connections and mounting are very similar and do not create any differences that would affect neighboring equipment or interfaces. The new transmitter weighs less and will be seismically.

mounted. Environmental qualifications will be enhanced.

l Safety Evaluation Number: SE-095-073 Type of Safety Evaluation: Setpoint Change

j. Evaluation Reference Number: 95-050E '

i 1

Title:

- Motor Control Center (MCC) Overload Heater Setpoint Change l .

Description:

~ This change replaces existing overload relay heaters in MCC 18-1 A Cubicle C-with new General Electric Thermal Overload Heaters (10H).'

Result: Unreviewed' safety question does not exist. This change responds to an identified problem with  !

, overload relay heater sizing. This change only affects the overload relay trip curves. Component sizing ~ was

l. ~ done in accordance with acceptance criteria.' This change will not affect normal Emergency Core Cooling l System (EC.CS)' operation or operation of the associated MCC. No new failure modes have been iintroduced.

Safety Evaluation Number: SE-95-074 Type of Safety Evaluation: Temporary Alteration  !

l' t Evaluation Reference Number: WR 95010918302 1

Title:

Alternate replacement for Service Water Radiation Monitor Flow Indicator (FI 1-1741-25)

~

Description:

Replace the existing Kobold model DAA5225 (with rotor) Service Water Radiation Monitor Flow Indicator with a Kobold model DAA6125 (without rotor).

Result: Unreviewed safety question does not exist. This change allows operability to be irstored to the j

- Unit 1 Service Water Radiation Monitor (SWRM). The indicator without rotor will be less susceptible to sample flow blockage caused by debris or crud stuck in the smaller cross-sectional area where the rotor is in place. The function of the flow indicator to provide visual indication of flow is still fulfilled by observation Attachment A, SVP-98-113, Page 74 of 153

)

L

of the flow glass (although visual verification is easier with the rotor than it is without). The control room SWRM low flow alarm will detect inadequate flow independent of this Temporary Alteration.

Safety Evaluation Number: SE-95 075 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: 95-052E (DCR 4-95-17)

Title:

Motor Control Center (MCC) Overload Heater Setpoint Change

Description:

This change replaces existing overload relay heaters in MCC 29-1 Cubicle G1 with new General Electric Thermal Overload Heaters (TOH).

Result: Unreviewed safety question does not exist. This change responds to an identified problem with overload relay heater sizing. This change only affects the overload relay trip curves. Component sizing was done in accordance with acceptance criteria. This change will not affect normal Diesel Generator Exhaust Fan operation or operation of the associated MCC. No new failure modes have been introduced.

Safety Evaluation Number: SE 95 077 Type of Safety Evaluation: GSEP Manual Change Evaluation Reference Number: 95-G4

Title:

Generic GSEP Change

Description:

This change revises references to the Emergency Planning (EP) Director as appropriate to either the Nuclear Regulatory Services Manager or the EP Staff.

Result: Unreviewed safety question does not exist. Station systems, equipment and functions will not be affected by this change which reflects elimination of a corporate EP positioned that 'is mentioned in the UFSAR. The duties of this position will be carried out by the Nuclear Regulatory Services Manager or the .

EP Staff without any change in the actions being performed.

Safety Evaluation Number: SE-95-078 Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-95-052

Title:

Pneumatic Control Link in Reactor Building HVAC

Description:

This change instalis a pneumatic control link between the Unit 1 and 2 Reactor Building HVAC differential air pressure control loops, such that a single unit's differential pressure controller controls both units Reactor Building differential pressure. The opposite unit controller shall remain as an active, in-place, spare.

j Result: Unreviewed safety question does not exist. This change stabilizes the Reactor Building HVAC l differential pressure control by isolating the second differential pressure controller; thereby, eliminating the common refueling floor's redundant control reference. No other systems are affected by this change. No Attachment A SVP-98-113, Page 75 of 153

credit is taken for the function of the HVAC differential air pressure control loops in the UFSAR accident analysis.

Safety Evaluation Number: SE-95-084 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCOP 6900-2S, Rev. O and QCTS 0230-06, Rev.1

Title:

Swap of Altemate and Normal Unit i 125V Batteries and Discharge Testing of the Batteries

Description:

This change creates a new procedure to provide instructions for placing the Alternate 125V battery in service and taking the Normal 12Sv battery out of service and revises an existing procedure to utilize the lineup created to perform discharge testing of the batteries.

Result: Unreviewed safety question does not exist. This swap supports the performance ofdischarge testing and recharge as required by Tech Specs. This change represents an improvement over the method previously used (a Temporary Alteration). All control room instrumentation and alarms will remain operable, there will be no change to schematic or key diagrams, and all operating and emergency procedures can be used without change. Use of the Alternate battery has no effect on system function, plant operation or the possibility of an accident or malfunction. Once the swap is in place, no new failure modes are introduced. Several barriers are in place to prevent the worst case failure which could occur during performance of the swap connecting the Alternate Battery in reverse polarity.

Safety Evaluation Numbert SE-96-001 Type of Safety Evaluation: Parts Evaluation Evaluation Reference Number: ER9502443, NWR #940100816, Parts Eval. #96-001-00

Title:

Replace Terminal Strip

Description:

This change replaces the terminal rtrip at a localjunction box. Junction box and mounting fas'eners were replaced as necessary. An evaluation was performed for both units and all Reactor Water Cleanup area temperature sensors. Terminal strip has developed too much resistance with aging. This prevents good calibration. Original vendor no longer provides spare parts for this equipment. This evaluation is for suitable commercially available replacement parts.

Result: Unreviewed safety question does not exist. The change is intended to be like-for-like replacement and has not degraded the instrument's ability to detect and warn the operator ofincreasing area temperatures. These non-saldy related, non-code, non-Environmentally Qualified, and non-seismic components were originally supplied as commercial items. No new failure modes or system interactions are created. The change does not affect performance, function, accuracy, or reliability of the modified sensor. 1 Terminal strip and junction box are passive in that they are required to perform their function to maintain circuit integrity, but they do not change states or initiate any mitigating response to an accident.

l l

Safety Evaluation Number: SE-96-002 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCTS 920-14, QFP 300-38 j

i I

\

Attachment A, SVP-98-113, Page 76 of 153 L

Title:

Revise Requirement For Full Time Nuclear Engineer Coverage in Control Room During Core Alterations

Description:

Procedures were changed to revise the requirement for " full time" presence of a Nuclear Engineer (NE) in the Control Room during core alterations. The new revision requires the NE to monitor / overview progress of core alterations from various locations, including the control room, refuel floor, or desk / office. The present procedure required the NE in the control room exclusively. The overview function of the NE is best served by allowing observation of each of the parts of the process.

Result: Unreviewed safety question does not exist. The change is administrative, defining the role of the Nuclear Engineer in providing an overview function. There is no change to requirements for moving or handling of nuclear fuel. None of the protected systems or means for mitigation of a dropped fuel bundle or heavy load drop that result in fuel failures are affected.

Safety Evaluation Number: SE-96-003 Type of Safety Evaluation: Work Request Evaluation Reference Number: WR950100362

Title:

GERIS Inspection of the Unit 1 Reactor Pressure Vessel (RPV) Welds

Description:

The subject of this evaluation is a temporary modification to the operation of Unit I wherein GE will be providing an Ultrasonic inspection service associated with the welds in the Unit 1 RPV.

The work was performed during a scheduled outage, following which the apparatus was removed from site.

The mechanical computer controlled manipulator system, coupled with a GE designed Ultrasonic Test (UT) system has the capability to conduct UT inspection of the inner surface of all pressure vessels of the GE BWR fleet above the shroud baffle plate. All of the lifting equipment is designed to meet the requirements of NUREG-0612. In evaluating operation of the GERIS 2000, the following considerations were made within the context of 10 CFR 50.59: 1) Seismic analysis; 2) An assessment to evaluate the consequences of Lost Parts: 3) Radiological Assessment; 4) Fire Protection; and 5) Material compatibility. All work was perfumed with no fuel in the vessel.

Result: Unreviewed safety question does not exist. The GERIS equipment is a passive device intended to l

perform UT examinations of the inside of the RPV shell. As such, the equipment does not have an active interrelationship with the functioning of either the primary system or any other safety system. The sole considerations associated with the GERIS operation deal with the handling of the equipment r.round the open core and the compatibility of the materials of construction and operation with the primary water. Both of these aspects of operation have been evaluated, and no new modes of malfunction have been determined with respect to discussions previously evaluated in the UFSAR. Operations of the GERIS does not create a l new failure mode. Nor does its operation place Quad Cities Unit 1 in a condition of operating outside ofits l design limit.

l Safety Evaluation Number: SE-96-004 Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-0-96-005

Title:

Resupport Water Supply Piping to B CR HVAC Refrigeration Condensing Unit Condenser Section l

l Attachment A. SVP-98-113, Page 77 of 153

l I

l

Description:

This change orients and supports water supply piping to the B LR HVAC refrige ation condensing unit condenser section to allow for proper installation of the front water head. The nmv piping l conforms to the current piping codes for this section and is seismically qualified. Front water head of l condenser section is misaligned by one bolt hole, causing the divider plates between the different passes to

! incorrectly channel flow through the condenser causing excessive erosion of the head and reducing efficiency of the condenser.  ;

~ Result: Unreviewed safety question does not exist. Probability of a malfunction of equipment important to safety is decreased. By reorienting the piping and the front water head for the condenser section, the condenser will become more efficient. This will reduce the loading on the cooling water pumps that supply the condenser section and will also make the compressor more efficient. In addition, the rate of erosion of the condenser and its water heads will be slowed which should extend the life of the component. All piping work is being done to existing codes and is being seismically qualified. Therefore, the probability of the l new installation failing is the same as original piping.

Safety Evaluation Number: SE-96-008 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number:

Title:

Removal of Core Spray Corner Room Equipment Drain Check Valves

Description:

This change removes the core spray corner room equipment drain check valves 1-1399-129 (for Reactor Core Isolation Cooling and 1 A Core Spray) and 1-4899-126 (for 1B Core Spray) at their

respective flanged connections and replace them with a flange fitted with a hose routed to the Reactor l

' Building Floor Drain Sump. The purpose of this temporary alteration is to facilitate draining Core Spray and Reactor Core isolation Cooling (RCIC) systems during refueling outage QlR14. This temp alteration was installed prior to reactor shutdown to facilitate maintenance activities as soon as respective systems were no longer required to be operable.

Result: Unreviewed safety question does not exist. Given a flooding condition in the torus area, for this temporary alteration to cause a failure of the protected equipment, either the closed piping associated with equipment drains in the room must fail or the pipe plug placed in the RCIC pump bedplate drain must fail.

l At the design flooding level in the torus area of 11 feet, the pressure experienced by the piping and the pipe plug will not exceed 5 psig. The design pressure of this class 150, schedule 80 piping is at least 150 psig. A failure of the 1" RCIC pump bedplate drain pipe plug at this pressure is also considered to be unlikely. This is because the plugs are installed as a tight fit and are checked for tightness by pulling on the plug after l installation. Therefore, probability of flooding in corner rooms is not increased by this temporary alteration.

i 3:

Safety Evaluation Number: SE-96-010 Type of Safety Evaluation: On-Site Review Evaluation Reference Number: OSR 96-06

Title:

Fuel Pool Cooling

Description:

His 50.59 evaluates acceptability of performing QCOP 1900-20," Operation of the Fuel Pool Cooling System Using the Reactor Cavity for Suction / Discharge", with reactor cavity to fuel pool gates installed. The operating mode is fuel pool cooling dedicated to the Unit i reactor cavity, which will Attachment A, SVP-98-113, Page 78 of 153

i r not align cooling to the Unit I fuel pool, which has discharged fuel in the pool Fuel pool cooling cleanup filter demins are needed to cleanup the reactor cavity. Pulling the gate,s would spread the cobalt to a much larger water volume, diluting it, and spreading the problem further. Reactor Water Cleanup is out-of-service, and there is no return flowpath for Reactor Water Cleanup.

Result: Unreviewed safety question does not exist. The specified condition involves only discharged fuel in Unit 1 pool. Because Unit 2 fuel pool is connected to Unit 1 pool via the transfer canal being open, the Unit 2 pool cooling system is functioning to cool both pools. The probability of a single unit's fuel pool cooling system tripping is not changed, and neither unit's cooling was credited for the availability of the opposite unit in the SAR analyses. The condition does not add any significant heat load. Therefore, there is 1 no significant change in head loads, or any other operating condition that could increase the probability of a I loss of pool cooling. Unit 2 fuel pool cooling system operation is maintained. No abnormal system

, conditions exist which could increase stresses, operating temperatures, or environmental conditions of any l equipment. Actions in abnormal procedures have progressively strong steps that will ensure that cooling is achieved if temperature of Unit I fuel pool reaches 125 degrees F or Unit 2 fuel pool cooling system trips L and cannot be started.

Safety Evaluation Number: SE-96-012 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number:

Title:

Refuel Bridge Crane i

Description:

Changes the point at which the 100 pound compensation for the Gleason reel tension is applied via the refueling platform programmable logic controller (PLC). The' compensation will now occur

, . at 200 inches below the mast normal up position. This change alleviates spurious " Hoist Loaded" indication I

while not affecting platform operations in the spent fuel pool. There is no change to the PLC logic.

Result: Unreviewed safety question does not exist. Bridge operation and interlock operability are not affected by this change. No new failure modes are created. The PLC logic remains the same. The position .

where the application of the tension compensation is applied is conservative. The change decreases the hoist loaded setpoints (a conservative change) over a wider length of mast travel (including area of travel prior to the change). Change improves the function of the existing hoist-loaded interlocks. The interlocks associated with the hoist-loaded circuitry were in place as required. The application of the compensation at the new position enhances operation by creating independence between the mast transition and tension compensation. The consequences of malfunction of equipment important to safety are not increased.

l l

Safety Evaluation Number: SE-96-013 Type of Safety Evaluation: Work Request l

Evaluation Reference Number: WR #960016565 01

Title:

Diesel Generator Cooling Water Pump Constant Level Oilers Replacement

Description:

This change replaces the Unit 1 Diesel Generator Cooling Water (DGCW) pump constant '

level oilers, Type EB with Type SS (different styles). An inadvertent change in oil level due to human error  !

or vibration would be less likely with the new style of oiler because it is maintained by an internal  ;

mechanism. '

Attachment A, SVP-98-113, Page 79 of 153

Result: Unreviewed safety question does not exist. Both styles of oilers provide the same function for the DGCW pump in that they provide make-up oil to the bearing oil reservoir as needed. Therefore, the operation of the DGCW pump is not affected by replacing the style ofconstant level oiler. The change does not include any change to bearing housing or piping connections and therefore, there is no increase in risk of loss oflubricating oil to the DGCW pump bearings. The consequences of an accident will not increase since the availability of the DGCW pump is not affected by the replacement. The reduction in risk is based on the height adjustment being internal to the oiler and not disturbed during oil addition.

Safety Evaluation Number: SE-96-017 Type of Safety Evaluation: Work Request Evaluation Reference Number: WR #Q21296 and Q21297

Title:

Replace Diaphragms on Scram Solenoid Pilat Valve in the Control Rod Drive System

Description:

This change installs a Buna-N diaphragm and Buna-N style endcap onto a Viton style Scram Solenoid Pilot Valve. The Buna-N diaphragm is on the exhaust post side of the 305-118 valve.

Some of the replacement diaphragms required the needle to be removed from them prior to installation.

This change ensures that scram times remain within Technical Specification limits. The existing diaphragms have developed a sticky phenomena that appears to increase with time and causes a delayed -

start of motion of the control rod during the scram function. Replacing the diaphragm will recover the scram time lost.

~

Result: Unreviewed safety question does not exist. The ability of the control rods to scram will not be affected. The change is functionally the same as the current configuration. The use of Buna N diaphragms in a Viton valve body has been bench-tested and met all of GE/ASCO qualifications for safety-related Scram Solenoid Pilot Valves. The CRD/RPS system failure modes or possible failures are not affected.

This change does not violate Reactor Protection System channel separation or introduce any common mode failures because full functional testing (i.e. scram timing) is specified before reactor operation.

Safety Evaluation Number: SE-96-019 Type of Safety Evaluation: Interim Procedure Change Revision to UFSAR Evaluation Reference Number: IP 96-0046, QCTS 0220-02, Parts Evaluation 96-020-00 UFSAR 97-RS-021

~

Title:

Revision to Discharge Test Current Rate and Update to UFSAR

Description:

This change revises the discharge test current rate from 38 amps to 36 amps. Replacement batteries installed during QlR14 have a 36 amp discharge rate to 1.75 v/ cell over 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The previous battery had a 38 amp rate. The UFSAR is also being updated in Section 8.3.2.3 to change 190-Ah to 170-Ah or greater to reflect the replacement cells installed in the Unit-124/48 volt system. Change to the UFSAR reflects both units. The new battery is identical to the battery previously installed on Unit 1 and currently installed on Unit 2.

Result: Unreviewed safety question does not exist. Operation is not changed. The actual loads require 11 amps to be supplied for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The change in capacity from 38 to 36 amps still leaves a large design Attachment A, SVP-98-113, Page 80 of 153

margin. The batteries are the same in all other respects. There are no new failures. Batteries sit in the same rack, are wired the same, and provide the same loads as before. If a loss of offsite power should occur, these new batteries far exceed the' circuit load requirements and offer the same degree of reliability as the old batteries.

Safety Evaluation Number: SE-96-020 Type of Safety Evaluation: Exempt Plant Change Evaluation Reference Number: E04-1-96-022

Title:

Condenser Vacuum Pressure Switch Replacement

Description:

This change replaces the currently installed Barksdale pressure switches used for detection ofloss of condenser vacuum, with new pressure switches manufactured by Static-O-Ring. The new switches are physically larger than currently installed switches and therefore, were mounted at different locations on the instrument rack. Associsted process tubing and electrical conduit were modified accordingly. The currently installed switches have a history of excessive setpoint drift. This change satisfies the corrective actions associated with LER 2-95-003.

Result: Unreviewed safety question does not exist. This change involves only a replacement of currently installed condenser vacuum pressure switches with new switches by a different manufacturer. The replacement switches are more reliable than the current switches. Functional configuration of the system is unchanged. The new switches improve reliability of the Reactor Protection System which would result in a higher confidence level that the SCRAM will occur as designed to mitigate the loss of Main Condenser Vacuum. Probability of malfuncticin of equipment importent to safety will decrease due to this change. The new pressure switches are less susceptible to setpoint drift.

Safety Evaluation Number: SE-96-022 Type of Safety Evaluation: TemporaryAlteration Evaluation Reference Number: Temp Alt # 96-2-006

Title:

Disable Thermal Overload Alarm on HPCI Auxiliary Oil Pump

Description:

The annunciator for thermal overload for the HPCI Auxiliary Oil Pump (AOP) was disabled to prevent nuisance alarms. The overload was alarming without an overload condition, and the overload relay was alarming without an actual pump overload.

Result: Unreviewed safety question does not exist. This temporary alteration does not affect any of the protective logic installed to automatically trip the AOP to protect the MCC. No new failure modes are introduced. AOP motor current is monitored during surveillance testing. If an overload condition occurs, s the operator is still procedurally directed to trip the AOP. The amber light for the pump on the 902-3 panel would still indicate a potential overload condition. If an overload condition were to occur when the AOP was needed during an accident, the operators would continue to operate the pump per procedure. Operator response to an actual condition that requites HPCI operation are unchanged, thus disabling the alarm will t not create the possibility or malfunction different from those evaluated in the UFSAR. ,

Safety Evaluation Number: SE-96-023 Type of Safety Evaluation: Procedure Change Attachment A, SVP-98-il3, Page 81 of 153

L Evaluation Reference Number: QCTS 360-1

Title:

Unit 1 Feedwater Level Control System

Description:

This procedure revision for Unit 1 feedwater level control system testing is generated based upon the Unit 1 Initial Startup Test Instructions, STI-23. The level setpoint test was performed by moving the level setpoint controller down and up about 6 inches as quickly as possible and the transient recorded until steady-state is achieved. The dynamic adjustment of the system was determined by observation ofits response to level setpoint disturbances at each test point scheduled during 1 and 3 element control mode of .

operations. In addition to the level setpoint changes testing specified in STI-23, other test activities were f.

performed in accordance with this revised startup test procedure. The reasons for the testing were to demonstrate that newly installed feedwater regulator valve hydraulic actuators were capable of providing automatic response to the initiated feedwater system transients, and verify that transient response of the feedwater system in I and 3 element control modes provide fast and smooth reactor water level control at all operating modes.

f Result: Unreviewed safety question does not exist. Performance of this procedure does not impact safe  ;

operation of the plant nor degrade adequacy of system, structure, or component of the feedwater level l control system. This test does not increase possibility of malfunctions that resulted in a maximum or zero l

l. feedwater flow since size of flow demand and level step inputs were limited during the test,1-element and i manual feedwater regulating valve control was available to the NSO at all times throughout the period of performing this procedure, and testing involved in this procedure does not have the potential to increase' the

. probability of feedwater controller failure. The low flow controller is designed to maintain reactor level during low power operation. Feedwater system is not utilized to mitigate consequences of any accident.

Test performed specifically to verify that control components associated with feedwater level control system meet their design requirements.

i Safety Evaluation Number: SE-96-024 Type of Safety Evaluation: Procedure Change j l'

t Evaluation Reference Number: QCAP 1500-1 i.

L

Title:

Changes Made to Fire Watch Requirements

[

Description:

Changes have been made to fire watch requirements when fire barriers are inoperable.  !

l Current fire protection program requires continuous fire watches for inoperable fire barriers regardless of . l' L the provision of operable detection on at least one side of affected fire barrier. Existing procedure and

[-

program requires a continuous fire watch regardless of the detection provided. Fire barriers do not separate ,

safe shutdown fire areas. Purpose is to reduce probability of occurrence of a design oasis fire by limiting )

" fire damage to a small portion of the fire area and allow higher probability to rapidly octect and suppress -I fire prior to becoming a design basis fire. These barriers separate fire hazards from other portions of the l same fire areas. A breach of this barrier alone does not substantially increase probability of rapid fire j spread due to other portions of the defense in depth including rapid suppression and detection which are inherent to these types of hazards. A once per hour fire watch was established to offset this probability.

Result: Unreviewed safety question does not exist. Changes made to procedures are administrative in nature. They describe the type ofcompensatory fire watches required when fire barriers are inoperable.

They do not affect plant operation. It is already assumed that the fire barrier is inoperable and therefore, Attachment A, SVP-98-ll3, Page 82 of 153

I I

these changes do not introduce the potential of an accident or malfunction of a different type already .

evaluated. Changes in Gre watch frequency does not change ability to safely achieve and maintain a safe i shutdown condition.

Safety Evaluation Number: SE-96-025 Type of Safety Evaluation: Drawing Change Request Evaluation Reference Number: DCR 4-96-077 -

Title:

Residual Heat Removal Service Water (RHRSW) Cubicle Cooler Stop Valves

Description:

Drawings M-37 and M-39 were revised to show the normal position of the RHRSW cubicle t cooler stop valves as normally opened valves. The reason for this change was to allow the dp across the room coolers to be normally readable during surveillance testing. The existing RHRSW quarterly procedure QCOS 1000-4 assumes that these valves are open. The QOMs also show these valves are normally open.

1 Result: Unreviewed safety question does not exist. Change does not adversely impact RHRSW system or create the possibility of an accident different than those already evaluated in the UFSAR. Past practice has been to operate with these valves already open. Operation is not affected. Design had already been taking credit for non-safety related tubing to function as pressure boundary during the accident. The gauge is located at the end of 1/2 inch stainless steel tubing that was already within the pressure boundary. Failure of this tubin,g had already been considered.

Safety Evaluation Number: SE-96-027 Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-96-023F (DCP 960067)

Title:

Weld Overlay on Weld 02AS-S12 I

Description:

Ultrasonic examination of weld 02AS-S12 on 'A' Reactor Recirculation Pump suction  ;

piping identined linear indications at various locations around the weld circumference. As a corrective j measure, a weld overlay was applied to the weld to ensure the 'A' Loop of the Reactor Recirculation suction i piping will continue to function as designed.

Result: Unreviewed safety question does not exist. Applying a weld overlay on weld on suction piping of

'A' Reactor Recirculation pump does not affect operation of the system. Application of weld overlay will return piping to its design criteria. No new interactions with any other system, structure, or component are introduced with the installation of this weld overlay. Application of weld overlay provides an additional pressure boundary around the circumference of the welded area therefore, increasing integrity of Reactor Recirculation suction piping. No new failure modes are introduced.

Safety Evaluation Number: SE-96-029 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: Setpoint Change No. 96-0231 & 0241

Title:

RCIC Instrument Setpoints Attachment A, SVP-98-ll3, Page 83 of 153 l

Description:

This change revised the instrument setpoint for PS-001(2)-1360-9A through D from 50 psig to 77 psig to support new Technical Specification Upgrade Program (TSUP) LCO Limit. Per a setpoint uncenainty calculation which was revised to support TSUP, uncertainties associated with these instruments required the process setpoint to be adjusted to a more conservative setting. Per the setpoint calculation, calibrating the instruments'setpoints to a more conservative setting mitigates setpoint drift from creating the possibility of an unsafe condition.

Result: Unreviewed safety question does not exist. Component's function remains the same. Revising the instrument's setpoint does not affect the instrument's function. System function is improved. Moving the setpoint to a more conservative setting means that if the setpoint does not drift during the surveillance interval, the safety actuation shall occur earlier. Should the setpoint drift during the surveillance interval, the total uncertainties quantified by the setpoint calculation are bounded by this time interval, such that the setpoint shall not exceed the Allowable Value (Tech Spec Limit). Therefore, the consequences of an accident decrease.

Safety Evaluation Number: SE-96-030 Type of Safety Evaluation: Shielding Package Evaluation Reference Number: Shielding Package 96-100

Title:

Lead Shielding

==

Description:==

This shielding package requested up to 3000 lbs oflead be hung from the 1 A and IB Residual Heat Removal (RHR) Heat Exchangers. Shielding was hung from the flange of the heat exchangers in a uniformly loaded pattern. This shielding reduces the area dose rates for work associated with RHR corner room steel upgrade modification. The original request indicated the heat exchangers would be out-of-service while the lead was in-place. Station requested to return one loop to service to proceed with loading fuel into the RPV. However, corner room steel mods were still in progress and the lead shielding was required.' This evaluation addresses the operation of the l A and 1 B RHR Heat Exchangers while lead shielding is present.

Result: Unreviewed safety question does not exist. Operation of the RHR system is not affected by this change. The added weight is considered negligible and will not have an adverse efTect on the heat exchangers or their supporting structures. This change will not affect any failure modes, nor will it create any new failure modes. The heat exchangers will still be able to perform their design function for each plant mode. The lead will not alter the operational or functional characteristics of the RHR and RHR Service Water Systems.

Safety Evaluation Number: SE-96-031 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: Setpoint Changes 96-0201 (U1) & 96-0211 (U2)

Title:

Revised Setpoint for Refuel Floor Radiation Monitors

==

Description:==

This evaluation changes high radiation trip setpoint for Refuel Floor Radiation Monitors from 80 mR/hr to 25 mR/hr which is being done to support implementation of the Technical Specification Upgrade Program (TSUP). The new TSUP limit, as well as evisting Tech Spec limit is less than or equal to 100 mR/hr for the trip setting for these monitors. A setpoint enor analysis calculation was performed that Attachment A, SVP-98-ll3, Page 84 of 153

e l

t.

determined that a negative margin of error existed between the Tech Spec limit and the instrument setpoint for the current setpoint of 80 mR/hr. Therefore, based upon error associated with instrument performance during calibration interval and error associated with test / calibration equipment, it was possible for instrument trip to occur non-conservatively with respect to the Tech Spec limit, Per the calculation, the new setpoint of 25 mR/hr has sufficient allowance for errors such that it is ensured that the TSUP trip setting limit will not be exceeded.

Result: Unreviewed safety question does not exist. This change improves the instruments' performance with regard to the Tech Spec trip setting limit. Automatic actuations required, isolation of Reactor Building l & Control Room Ventilation and start of Standby Gas Treatment, are expected to occur as soon/ sooner with the new setpoint of 25 mR/hr as with the current setpoint of 80 mR/hr. Instrument trip and subsequent automatic actuations to mitigate the consequences of an accident will occur as soon or sooner as a result of this change providing better response to mitigation.

l Safety Evaluation Number: . SE-96-032 Type of Safety Evaluation: Procedure Change

Evaluation Reference Number

QCOP 0202-12

Title:

Reactor Recirculation System MG Set Scoop Tube Lock and/or Local Manual Operation l

Description:

This evaluation addresses the practice oflocking the Reactor Recirculation (RR) Pump MG Set Scoop Tube. The constraints implemented in the procedure and the procedure references are being improved to provide added assurance that the necessary protection (RR pump NPSH) are evaluated, and the l

evaluation is being expanded to assure that all of the pertinent SAR issues are addressed. Normally the -

Reactor Recirculation pumps are controlled manually from the control room. This procedure locks up the scoop tube positioner on the fluid coupler in place at the MG Set. The procedure then resets the power source to the scoop tube positioner. This configures the scoop tube positioner to be reset from the control room if manual operation of the pumps is necessary.

Result: Unreviewed safety question does not exist. The manual operation mode of the RR pump / speed flow controls is an analyzed and normal mode of flow changes for needed core flow / pump speed changes.

- The introduction'of the added step of manually resetting the controller (s) to the manual speed adjustment does not change the effectiveness of manual speed changes. Substitution of a manual action for pump l runback / cavitation limitation is not introduced as a new accident or malfunction because the manual action is effectively achie'.ed by procedural requirements. Possible time for manual action is still far shorter than time required to have a detrimental effect on the RR system / core components. ~ All systems are being operated with installed, originally designed equipment.

I Safety Evaluation Number: SE-96-033 Type of Sciety Evaluation: Procedure Change Evaluation Reference Number: QCTS 0310-01

Title:

Testing of Auto-Initiation of Emergency Core Cooling System (ECCS) and Emergency Diesel

' Generator Systems

Description:

Procedure to test auto-initiation of ECCS and Emergency Diesel Generator (EDG) systems l._ in response to a Loss-of-Coolant Accident (LOCA) concurrent with a Loss of Oft' site Power (LOOP) has Attachment A, SVP-98-ll3, Page 85 of 153 l

been rewritten to include changes for the Technical Specification Upgrade Program (TSUP) as well as j enhancements to the procedure identified during performance of the test on Unit 2. This procedure also

{

performs various tests of the EDG required by the station reliability program which include single load

{

rejection, synchronizing test, and unit transfer capability test (1/2 EDG only). Due to testing of 1/2 EDG in '

this procedure, affects on Unit 2 were also evaluated.

Result: Unreviewed safety question does not exist. Adequate steps are included to assure that reactor temperature is monitored appropriately and to provide for natural recirculation. Other changes include installation and removal ofjumpers and blocks to allow or prevent equipment from operating to minimize affect of transient on plant equipment which does not adversely impact equipment. Steps are taken to ensure that 1/2 EDG is available to Unit I and 2 during testing of division 11 equipment for those accidents already analyzed with exception of LOCA on Unit 1. Only one division was tested at a time in order for other division to be available should it be required. No actions taken in procedure would increase probability of an accident.

Safety Evaluation Number: SE-96-035 Type of Safety Evaluation: Procedure Change Evaluatlan Reference Number: QCTS 930-12

Title:

Control Rod Drive Scram Time Testing During Vessel Hydro Testing Descript' ion: This evaluation changes the procedure to perform Control Rod Drive scram time testing during vessel hydro (solid) testing. The change also connects and disconnects the amphenol connectors from the directional control valves so the control rod can be moved. The Code does not allow for air to be in the reactor vessel during the hydro test (to prove scram capability prior to startup).

Result: Unreviewed safety question does not exist. There is no change in the function of the control rod.

Scram operation is not changed. Method for obtaining scram time data is not changed from approved methods in Site procedures. This change is the reactor being solid water. Procedure still meets requirements in Tech Specs. Change does not affect failure modes or equipment failures. Introduction of water from scramming of a control rod increases vessel pressure by approximately 3 psig, therefore probability of an over-pressurization of vessel is not increased. Safety valves are operable during hydro and scram timing.

Safety Evaluation Number: SE-96-036 Type of Safety Evaluation: Ul Cycle 15 Reload NFS-BSSs901 Evaluation Reference Number: OSR 96-11  ;

I

Title:

Unit 1 Cycle 15 Reload

==

Description:==

This evaluation addresses the Unit 1 Cycle 15 reload and associated transient analyses. In addition, this 50.59 evaluates the Unit 2 COLR changes, made as a result of Technical Specification Upgrade Program (TSUP) implementation (which are editorial). The reload fuel for QlC15 was i GE8X8NB-3 type (GE10) and the number of bundles in the reload was 232. The bundles contain 3.32 w/o U235 (144 bundles) and 3.33 w/o U235 (88 bundles). There are no LTA's or special operating provisions.

The QlCIS reload is the second reload ofGE10 for Quad Cities Unit 1.

Attachment A, SVP-98-113, Page 86 of 153

1 Result: Unreviewed safety question does not exist. The GE10 fuel design has been reviewed and (

generically approved by the NRC and is incorporated into GESTAR-II (NEDE-24011-P-A-11). This is the second reload of GE10 fuel for Unit 1 and the enrichments are not substantially different from previous cycles. The GE10 fuel design with interactive channel is very similar to GE9 which has been loaded at Quad Cities Station for Q2Cl1 through Q2Cl3 for Unit 2 and Q1Cl2 through Q1Cl3 for Unit 1. Since there have been no mechanical design changes to the reload bundles there will be no impact to operations, because the new reload bundle design is not significantly different to previous cycles. For the QlC15 core, the bundle to control blade clearances are similar to values seen for C-lattice cores, and hence is acceptable.

This second reload of the GE10 fuel design will not adversely impact operations nor will it have an adverse .

impact on all systems,' structures, or components. The GE10 fuel design with 40 mil offset will not l adversely impact any in-core instrumentation or components. The 40 mil offset effect on in-core j instrumentation was accounted for in the CMC input deck. The GE10 reload design 40 mil offset will not l

adversely impact equipment failures. GE10 fuel has been found not to have any significant dynamic loading l changes on the reactor, internals, or fuel assemblies. None of the UFSAR assumptions have been invalidated with this reload.

Safety Evaluation Number: SE-96-037 Type of Safety Evaluation:  ;

Evaluation Reference Number:

Title:

Q1R14 Reactor Vessel Foreign Material

Description:

Foreign material has inadvertently entered in the reactor vessel or is assumed to have been lwt in the reactor vessel. This also applies to the identified foreign material that was not recoverable. Only items 2 and 3 from a summary list were potentially significant with respect to possible migration pathways or equipment interactions. Item 2 (cotter pin from fuel pool /Rx cavity gate) is not known to be present in the reactor. It was noticed that it was missing, and not found either in the pool or the reactor after extensive searches in the reactor. A lost parts evaluation was performed by Nuclear Fuel Services (NFS).

Result: Unreviewed safety question does not exist. All of the evaluated parts are very low mass and cannot l initiate any equipment / component failures that could breach the primary system boundary. NFS concluded that there were no nuclear safety concerns associated with operation of Unit I with the ider tified parts missing in the primary system. The introduction of foreign material into the vessel or the potential for resulting incremental fuel failures does not affect any of the mitigating safety systems and therefore, will not increase the consequences of an accident.

Safety Evaluation Number: SE-96-040 Type of Safety Evaluation: Problem Identification Evaluation Reference Numbcr: PIF 96-1368

Title:

Local Leak Rate Testing

Description:

This evaluation addresses the operation of the plant in all modes with Reactor Core Isolation Cooling (RCIC) System Steam Supply Valves (1-1301-16 and 1-1301-17) volume with a combined leakage of 10.0 scfh. This volume is a containment leakage path with individual administrative valve leakage limits ofless than 10.0 scfh each. Maintenance on the 1-1301-16 valve identified a degraded Attachment A, SVP-98-ll3, Page 87 of 153

disk sealing * ._ Aplacement parts are not available at this time (3 to 4 month lead time). Overall containment leakage is currently 227 scfh (including the 10.0 scfh for the 1-1301-16 & 17 volume). The Tech Spec limit is 293.75 scfh (0.6 La).

Result: Unreviewed safety question does not exist. In the event of a LOCA, these two valves are required to isolate to provide primary containment integrity. There have been no physical changes made to the volume. Again, this 50.59 was to evaluate the effect of allowing the valves' volume to operate one more refuel cycle (Q1RIS) with the current combined leakage of 10.0 scfh. The maximum path leakage is 10.0 attributed all to either the 16 or 17 valve. In this case the other valve's leakage would be zero and

)

containment integrity is maintained. Minimum path leakage splits leakage evenly between the two valves 1 (5.0 scfh each). Given this, even if one of the valves fails to isolate, (or leak tightness degrades over this operating cycle) the other valve would maintain containment integrity with an acceptable leakage of 5.0 scfh. Permitting the 16 and 17 volume to operate until QlR15 does not increase the possibility of a LOCA l occurring nor increase the probability of a malfunction of equipment important to safety.

Safety Evaluation Number: SE-96-041 Type of Safety Evaluation: Engineering Request Evaluation Reference Number: ER9603058

Title:

Use of Sight Glass Standard & Oiler Standard j

==

Description:==

This evaluation addresses the use of two newly revised standards, Oiler Standard and Sight Glass Standard on various equipment that is equipped or should be equipped with oilers or sight glasses. l Procedures also provide acceptable oil levels and the basis for those levels. The Oiler Standard was created to ensure consistency in installation and level markings. The older standards allowed installation of various styles of oilers, including some models which have exhibited problems (improper level settings) in the past.

The new oiler standard recommends the sole use of the Trico Opto-matic Type LS oiler. The new Sight Glass Standard was condensed into one standard for the same reasons. Also, the method used for marking sight glasses is further defined to include standby level, minimum and maximum operating levels to minimize confusion on whether markings were for operating or standby conditions.

Result: Unreviewed safety question does not exist. Plant operation is not affected. Operation of such equipment is not affected by replacing the style of corstant level oiler. Equipment failures remain unchanged. Use of these standards help to standardize the type of oiler and sight glass markings used. This does not increase consequences of a malfunction of equipment. This increases reliability and minimizes confusion for setting levels, filling reservoirs, ar.d inspecting setpoints. It also reduces time and dose required for sampling. All equipment will continue to perform as designed. Replacement oilers and sight glasses will perform the same as those they replace.

Safety Evaluation Number: SE-96-042 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number:

Title:

Source Range Monitor (SRM) 24 for Unit 2

==

Description:==

This evaluation addresses a temporary alteration which was used to splice a small piece of cable which could be used with the current connector type which was in site stock pending the Attachment A, SVP-98-ll3, Page 88 of 153 I

f l

[ determination and procurement of the proper cable connector. The currently installed RG-108 cable was l

pulled out ofits cable connector. Upon h:pection, it was determined that another type of connector was required to properly terminate the cable. This tempo.m alteration provided for cable 27905 to be transition l

spliced to RG-59 to allow installation of an available connector for termination at Panel 902-36.

l Result: Unreviewed safety question does not exist. There is no change in component function or intent. '

The detector signal is not affected by installing the splice. By correcting the cable-to-connector contact, the reliability is increased. Channel failure would remain in the safe direction. While operating, the splice would pass the detector signal without altering it. Only change in failure modes would be if the splice were to fail, in which case the channel would fail downscale as designed. SRM system has redundancy designed l l into it in case of a failure of one channel. Further, SRMs are not considered as mitigating equipment in any accident or transient. Probability of an accident is not increased.

l l

4 Safety Evaluation Number: SE-96-046 Type of Safety Evaluation: Shielding Package Evaluation Reference Number: Life-of-Plant Shielding Package LOP-046 & LOP-047 l

Title:

Lead Shielding

Description:

This evaluation addresses the life-of-plant shielding package LOP-046 which requested up to 1760 lbs oflead be hung from the supporting steel associated with the 1 A and 1B RHR heat exchangers. .

The lead was hung in a uniform pattern to provide shielding to the lower head of the two heat exchangers. I Shielding package LOP-047 requested up to 2245 lbs oflead be hung from the upper flange of the same two heat exchangers. It was hung from the flange of the heat exchangers in a uniformly loaded pattern. The Radiological Protection Department determined that these heat exchangers are a major contributor to the general radiation dose rates in the particular rooms and concluded that the shielding v as the best solution to  !

minimize exposure to plant personnel involved in work activities around these heat exchangers.

Result: Unreviewed safety question does not exist. Comed calculation determined the revised stress levels in the heat exchangers and supporting steel were still below design allowable values. Structural integrity is i not compromised. The lead is a passive component that will not interact with any other plant system or component. Therefore, plant operations are r.ot affected. Existing failure modes are unaffected and no new failure modes are created by adding lead shielding. The function of the heat exchangers is not changed by the addition oflead blankets.

Safety Evaluation Number: SE-96-050 Type of Safety Evaluation: Unique Test '

Evaluation Reference Number: Unique Test 1-182 '

Title:

Standby Gas Treatment Flow Control Valve

Description:

This Unique Test gives guidance on the method to test the performance of the Standby Gas Treatment Flow Control Valve (SBGT FCV)in the normal control mode and when failed full open. The test also determines the effect of Turbine Building and Reactor Building ventilation on the performance of l

SBGT with the FCV operating normally and failed full open. No test procedure existed.

Attachment A, SVP-98-113, Page 89 of 153

Result: Unreviewed safety question does not exist. This test did not affect any equipment failures. It merely determined the effect of the FCV on SBGT failing open. Since SBGT was not required to be operable, the failing of this equipment does not cause any adverse effects. After the completion of the test, the system was placed back into its original condition. Therefore, no new failure modes were created.

l Safety Evaluation Number: SE-96-054 Type of Safety Evaluation: Temporary Alteration l Evale alon Reference Number: Temp Alt #96-1-119 l  !

Title:

Reactor Water' Clean-up i

)

Description:

This change bypassed low reactor water level isolation for 1-1201-2 valve and also bypassed high temperature at exit of regenerative heat exchanger isolation for the 1-1201-5 and 1-1201-80 valves. This change allowed work to proceed on the I A Reactor Protection System Bus Breakers while still keeping reactor water clean-up system running to provide core cooling.

Result: Unreviewed safety question does not exist. The valves affected by this change are important to safety but not required to be operable with reactor in cold shutdown condition. Probability of malfunction  !

i of equipment important to safety is not increased. This change increases the probability of reactor water ,

I clean-up system performing its core cooling function. Since system is more likely to perform its core  !

cooling function, consequences of a malfunction do not increase. I Safety Evaluation Number: SE-96-055 Type of Safety Evaluation: Unique Test Evaluation Reference Number: Unique Test 1-184

Title:

Standby Gas Treatment (SBGT) Special Flow Control Orifice Test

Description:

This evaluation was for a Unique Test to test the new flow control orifice installed on both trains on SBGT. The orifices were installed under Design Change No. E04-0-96-045. The orifices are sized to limit flow through the trains to less than 4400 cfm if there should be a loss ofinstrument air. This test determined if the SBGT flow control orifice would limit flow to 4000 +-10% cfm. Test was performed when SBGT and Secondary Containment were not required to be operable.

Result: Unreviewed safety question does not exist. Test did not cause any new equipment failures. The SBGT system was returned to normal configuration upon completion of test. Failure of the system does not change as a result of performing this test procedure. During the test, the plant is in a condition that does not require SBGT trains or Secondary Containment. Test does not prevent any plant equipment from performing its design function. SBGT trains cere already inoperable due to maintenance.

Safety Evaluation Number: SE-96-063 Type of Safety Evaluation: Drawing Change Request Evaluation Reference Number: DCR 960177

Title:

Remove Note 3 on HPCI P&lDS (Check Valve Requirement for Keep Fill Line)

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i Attachment A, SVP-98-113, Page 90 of 153

i I

Description:

This change removes Note 3 on P&lDs M-46 Sheet I and M-87 Sheet 1, which reads,

" Valve 1(2)-2301-107 is locked closed. There is no check valve in line 1(2)-2381-3/4"-LX to prevent j backfiow and therefore this line will remain isolated via valve 1(2)-2301-107 until a check valve is in the line." Check valve is not required for proper system operation. Keep fill system with jockey pump is designed to maintain header pressure at 45-85 psi to ensure that system is properly pressurized. This pressure is greater than torus and CCST head, therefore backflow could not occur into keep fill system.

Result: Unreviewed safety question does not exist. Absence ofcheck valve to prevent backflow into ECCS keep fill system will not adversely impact systems or functions as to create possibility of an accident or malfunction of a type different from those evaluated. Keep fill system is designed to be at a higher pressure than 11PCI suction header at all times, thus ensuring that header is always maintained full.

Pressure differential ensures that a check valve is not required when valves are open. Per Tech Specs, in l the unlikely event that keep fill system p essure is less than 40 psig or greater than 90 psig, an alarm will go offin control room, and an orderly shutdown must be initiated if situation is not corrected within twelve hours.

1 Safety Evaluation Number: SE-96-064 Type of Safety Evaluation:

Parts Evaluation Evaluation Reference Number: Parts Evaluation Number 96-040-00 i

Title:

Reactor Building Siding I

Description:

This change reinstalls outer sheet panels on east side of Reactor Building. The existing / original siding is no longer manufactured. The vendor recommended this replacement. This change restores the design function of the reactor building siding. Performance of replacement siding is equivalent or superior to the original. The east side of the Reactor building siding was damaged by a tornado on May 10,1996.

Result: Unreviewed safety question does not exist. Components of replacement siding function the same as the original components. Change does not adversely affect plant operation in any operating mode.

Restores original design function and requirements, and thus, reactor building will continue to perform its secondary containment design function. No new accidents or malfunctions different from those evaluated are introduced by the replacement siding. Ability ofrestored siding to limit release of radioactive materials remains the same.

j Safety Evaluation Number: SE-96-070 Type of Safety Evaluation: Procedure Change

! Evaluation Reference Number: QCTS 920-01 l

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Title:

Change in Shutdown Margin Requirement

Description:

This evaluation changes shutdown margin requirement from 0.25% Ak to 0.38% Ak. In addition, an administrative 0.3% Ak penalty is incorporated to account for increased uncertainty in PANACEA local critical calculations. This change implements the standard SDM LCO.

Result: Unreviewed safety question does not exist. Implementation of the standard shutdown margin LCO and GE-recommended penalty do not affect any plant systems, only the calculation performed in QCTS Attachment A, SVP 98-ll3, Page 91 of 153 l

1 1

T 920-01 is modified. The standard LCO is conservative to the current TS LCO, and the penalty is administrative in nature. Because changes do not affect any plant equipment or systems, no new accidents are created.

l; Safety Evaluation Number: SE-96-073 Type of Safety Evaluation: Exempt Change Evaluation Reference Number: DCP 9600166

Title:

Unit 2 EDG Fuel OilTransfer Pump i .

l

Description:

This change disconnects leads on an auxiliary control relay that is not being used and rcplaces existing 150 VA Control Power Transformer (CPT) with a 300 VA CPT in Motor Control Center (MCC) 29-1 cubicle B2. The CPT provides 120 V control power to the contrW circuit of the Diesel Fuel Oil T ransfer Pump 2-5203. In the event of an undervoltage condition the existing CPT cannot provide sufficient voltage to the motor contactor and parallel control relay to assure that they wil! function properly

[ to energize the 2-5203 pump. By replacing the CPT, the voltage across the starting contactor will provide sutricient current for the contactor to function properly under all plant operating conditions.

Rest.lt: Unreviewed safety question does not exist. No new failure modes created. New transformer has a higher rating than existing transformer. New transformer is scismically mounted in the MCC cubicle. A failure of the new transformer is not any more likely than the existing transformer. New transformer meets

same des

gn criteri: as existing transformer. Removing control relay and changing transformer increase the running voltage, thereby ensuring the transfer pump performs its intended design function resulting in decrease in probability of a malfunction with no change in consequences.

Bsiity Evaluatiin Number: SE-96-074 Type of Safety Evaluation: Exempt Change Evaluation Refes ence Number: E04-2-96-043

Title:

. RHR Inbcard Shutoff Valve to Loop Injection 29A and 29B

Description:

. This change installed two control relays and one terminal block in the back of each Motor  !

Control Center (MCC) 28/29-5 cubicles B4 and E4. Relays were wired in control logic for valves 2-1001- .

29A and 2 1001-29B in place of existing 42/MC and 42/MO (close & open starter contactor relays). A l contact off of the respective relay pulls in or drops out the existing 42 contactor relay. Currently, in the  !

event of an undervoltage condition, there is insufficient voltage available at the 42/MO and 42/MC relays to ,

L assure the relays function properly in the present control circuit configuration. By placing an interposing  !

relay in the circuit and moving the 42 relays to another position in the circuit, both relays will function l

properly.

L 1 Result: Unreviewed safety question does not exist. Failure mode of new relay is same as for existing I

! control circuit. A failure will most likely result in a failure of the valve to open or close which is the same ;

failure for existing circuit, Function of control circuit remains unchanged. New relays are qualified for this j application. By adding interposing relay, the voltage level will be higher which ensures the valve to open  ;

and carry out its design function as intended.  !

Attachment A, SVP-98-113, Page 92 of 153 i

Safety Evaluation Number: SE-96-075 Type of Safety Evaluation: Exempt Change Evaluation Reference Number: E04-1-95-061

Title:

AVCo Solenoid Valve for Target Rock Safety / Relief Valve

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Description:

This change removes the Marathon 1500 Series terminal block in thejunction box next to AVCo solenoid and replaced it within Environmentally Qualified splice. These Marathon terminal blocks

, have been known to exhibit excessive leakage current under post Design Bases Event conditions that could affect performance of the Safety / Relief valve. Replacement of the terminal block with EQ approved splices ,

enhances reliability of the valve.

]

Result: Unreviewed safety question does not exist. Replacement of the terminal block does not affect plant operation. No new failure modes introduced. Splice enhances reliability of the solenoid circuit to ensure the valve will function as intended since terminal blocks have been identified as potential failure points.

This change does not functionally change the power or control circuit for the SRV, affect how the SRV interacts with other equipment, or affect performance of any other equipment. Consequences of a malfunction of equipment important to safety will not increase.

Safety Evaluation Number: SE-96-076 Type of Safety Evaluation: Problem Identification Evaluation Reference Number: PIF #96-2205

Title:

Fire Damper Not Installed

Description:

This evaluation' describes an exception to the 1-hour fire rating described in the Updated Fire Hazards Analysis. Fire barrier Q01-TB-547-WO23-FB is rated as a one-hour fire barrier, A penetration in the HVAC duct goes through the 1-hour fire barrier. There is no fire damper presently installed in this HVAC duct and it is proposed to keep the as-built condition of the HVAC duct without a 1-hour fire damper. Installing a 1-hour fire damper in fire barrier HVAC penetration F-199-13 would not .

provide any additional protection in reducing the spread of fire.

Result: Unreviewed safety question does not exist. There is no affect on equipment required to safely  !

shutdown the plant in the event of a fire. The fire barrier in question is not an Appendix R fire barrier, does not provide separation between fire areas, and is not required for the plant to be able to safely shutdown.

Lack of a fire damper would not increase travel of smoke and products of combustion from a fire on the '

CRD or condensate pump levels due to size of other mechanical openings and penetrations in the area. Not  ;

installing the fire damper is within current code limits. The small HVAC duct penetration in the 1-hour -

rated fire wall will not result in an increase of a malfunction important to safety.

i Safety Evaluation Number: SE-96-078 Type of Safety Evaluation: Exempt Change i

Evaluation Reference Number: E04-2-95-064, Addendum 3 i

Title:

Residual HeatRemoval (RHR) Heat Exchanger Support Steel Attachment A, SVP-98-113, Page 93 of 153 U

Description:

This change reinforces structural steel connections of the existing RHR corner room structural steel and adds horizontal supports to the RHR 11 eat Exchangers. Interferences such as conduit, piping, and/or supports were removed and reinstalled as required to facilitate the structural steel modification. Implementation of this Exempt Change has brought the stress levels of the steel framing within UFSAR allowable stress limits.

Result: Unreviewed safety question does not exist. ChanFe does not have any effect on operation of the RiiR or RHRSW systems; it strengthens existing structural steel to ensure that the design margins are within UFSAR allowable limits. Change does not create any new interactions with other structures, systems, or components, nor changes any existing interactions, and does not alter the operation of any RHR/RHRSW equipment. Likelihood of a failure of RHR/RHRSW equipment supported by the steel framing is reduced.

Safety Evaluation Number: SE-96-082 Type of Safety Evaluation: Design Change  !

Evaluation Reference Number: DCP # 9600220

Title:

Modification to replace Westinghouse FA type breaker with a Westinghouse HMCP type breaker at MCC 2B cubicle P02.

Description:

The existing FA type breaker is obsolete. The breaker is being replaced with a seismic and environmentally qualified breaker. The changed breaker is an equivalent replacement to the existing breaker as evaluated by calculations. The breaker settings were determined using approved acceptance criteria. The operation of the breaker will remain the same.

Result: Unreviewed safety question does not exist. This change does not introduce any new failure modes.

The change of the breaker rating from a 5 amp to a 3 amp breaker was evaluated by calculation and has no effect on equipment failures. The valve does not affect the initiating events for the accident, thus the probability of an accident has not increased. The new breaker is not any more likely to fail than the existing breaker.

Safety Evaluation Number: SE-96-083 Type of Safety Evaluation: Interim Procedure Evaluation Reference Number: Interim Procedure for QOS 6500-01,04;DCPs 9600202,9600203

Title:

Operability Testing of the EDG 1/2 Output Circuit Breakers l

Description:

Procedures were revised via the interim procedure process and used for operability testing of the EDG 1/2 output circuit breaker. These procedures are currently used to test undervoltage relay operation and load shedding upon loss of Buses 13-1 and 23-1 as well as verify that the EDG 1/2 starts and

, operates properly on an undervoltage auto start signal. The interim procedure removed the load shed l verification since the above referenced DCPs only required verification of EDG 1/2 closing times and added I a notification to inform the operations department to secure loads at their discretion. De-energizing bus 13-1 (23-1) and letting EDG 1/2 aute start and load to respective bus is proper method of verifying that interposing relay has not caused EDG 1/2 to exceed its Tech Spec limits.

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Attachment A, SVP-98-ll3, Page 94 of153 l

Result:

Unreviewed safety question does not exist. The changes made did not affect the way the plant behaved under the original procedure. The operations depaitment was noti 6ed that certain loads would be lost and they may secure them as desired. Performance ofinterim procedures do not affect equipment failures nor plant operations. Load shedding verification steps not required for proper plant operation and their removal for purposes of this IP do not introduce any new failure modes.

Safety Evaluation Number: SE-96-085 Type of Safety Evaluation: Drawing Change Request Evaluation Reference Number: DCR 960192 '

Title:

Remove Safety Boundaries for HPCI Keep Fill Line Nos. 1(2)-2381-3/34"-LX

Description:

This evaluation removed safety classification flags on 1(2F2381-3/4"-LX between valves 1(2)-2399-13 and 1(2F2301-107 so that the entire lines are safety-related. The keep fill system with the i

jockey pump is designed to maintain header pressure at 45-85 psig when the HPCI water supply system is lined up to the torus to ensure that the system is properly pressurized to mitigate water hammer loading effects. Calculations have been performed to seismically qualify the piping and supports under all loading conditions, therefore the safety breaks are not required.

Result: Unreviewed safety question does not exist. The piping and supports have been analyzed to ensure the system will function under all loading conditions. HPCI will continue to perform as designed.

Reclassifying the keep fill lines as safety-related will help ensure that the system functions as designed. <

Accident scenarios and their consequences remain unchanged. I l

i Safety Evaluation Number: SE-96-086 Type of Safety Evaluation: Interim Procedure Evaluation Reference Number: IP for QCOP 1000-05 i

Title:

Shutdown Cooling Operation with MCC 28/29-5 De-energized

Description:

This evaluation allowed the issuance of an interim procedure that allowed the operation of shutdown cooling when MCC 28/29-5 is de-energized. With this procedure in place, the shutdown cooling would be available if required. No current procedure exists to give guidance on the method to operate I shutdown cooing when MCC 28/29-5 is de-energized. This change results in valves 1001-28A/B and 1001 29A/B being without power and not going closed when operating in shutdown cooling mode and a group 2 isolation signal is received. These valves will need to be operated locally manually, rather than from the control room. This change only applies when the reactor is cold. This condition is allowed by Tech Specs.

Result: Unreviewed safety question does not exist. This change does not create any new accidents besides those already considered. This change does not increase the susceptibility to flooding and will not result in the l'UR system failing in a different mode. The change to the shutdown cooling evolution will not cause the system to be operated outside of the design or testing limits of the system. LPCI is considered

. inoperable during this evolution but available with the manual actions, which is consistent with Tech Spec requirements during these current conditions which have been considered by the NRC and evaluated as .

acceptable risk when the unit is in cold shutdown.

Attachment A, SVP-98-113, Page 95 of 153

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Safety Evaluation Number: SE-96-088 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: PIF 96-2358

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Title:

Manual Isolation of the Control Rod Drive Pumps during an SBO or LOOP l

Description:

This evaluation added steps to manually isolate the Control Rod Drive Pumps during a Station Black-Out (SBO) or Loss of Off-Site Power (LOOP). NRC Information Notice 90-78 identified previously unidentified radiation release path in the CRD system. Isolation of the CRD pumps is required to ensure operability of the CRD system during Design Basis Accident Conditions with an SBO or LOOP as y discussed in the operability determination for this PIF.

Result: Unreviewed safety question does not exist. This procedure change helps to mitigate the consequences of a Design Basis Accident. The change isolates the CRD system manually from the reactor primary system which prevents the possibility of vessel inventory or radionuclides leaking backwards through the CRD system into the CCSTs. This prevents an increase in dose rates, and therefore, the

~

consequences of an accident are not increased. The safety-related function of the CRD system is unaffected, because the scram capability is contained completely in the hydraulic control unit. The CRD system can also be used per existing operating procedures for alternate injection or normal system operation.

Safety Evaluation Number

SE-96-089 Type of Safety Evaluation:

Design Change Evaluation Reference Number: DCP #9600202 and DCP #9600203

Title:

Mod.fication to add interposing relay and resistor to the closing circuits of the EDG 1/2 output circuit breakers (ACB 152-1321, ACB 152-2329 to buses 13-1 and 23-1).

Description:

A new interposing relay will be energized whenever a close signal is initiated, which in turn will energize the breaker's closing coil. To allow the upstream synchronizing HACR-1V relay to operate properly, a 500 ohm resistor will be placed in parallel with the interposing relay. The high currents produced by the existing closing coils cause significant voltage drops along the control circuits. The new relay, with its high coil resistance, will reduce the voltage drop and allow proper operation'of the closing circuits under plant conditions. The new resistor is required to allow the proper current to flow through the HACR-IV's silicon controlled rectifier.

I Result:' Unreviewed safety question does not exist. This new arrangement will have no effect on the breaker's closing initiation signals, although it will affect the way they are processed (ie interposing relay i actuates closing coil which closes the breaker). The failure modes on the new relay would be the same as the existing one (ie open coil, short to ground, failure to actuate), with the same results. Since the new interposing relay will eliminate a failure mode by allowing proper voltage at the existing closing coil,

)

L equipment failures will be reduced. The EDG 1/2 will function as it did before the modification, therefore i the consequences of an accident will not increase. The probability of a malfunction of equipment irt.portant

' to safety will decrease because the new relay can tolerate a much lower battery voltage than the existing closing coil and still perform its design function.

I s

Safety Evaluation Number: SE-96-090 Type of Safety Evaluation: Temporary Alteration Attachment A, SVP-98113 Page % of 153

Evaluation Reference Number: Temp. Alt. 94-1-154 and 94-2-034

Title:

Increase Dissolved Oxygen Levels in Condensate /Feedwater Systems

Description:

These temp alts provided a means to increase dissolved oxygen levels in the I

condensate /feedwater systems. The oxygen supply line connects to the Turbine Building Sample Panel recovery header, which connects to the Condensate Pump suction header below the hotwell. This change was intended to allow dissolved oxygen levels in the condensate /feedwater system to be maintained in the optimum 20-60 pph range.

Result: Unreviewed safety question does not exist. No new system interactions and no new failure modes are created for the condenser, condensate, or feedwater systems. The change does not increase the potential of a new accident more serious that what has been previously evaluated. Reducing corrosion in the condensate and feedwater piping can only have positive implications for the reg valves. Since the temp alt cannot do anything that would increase the likelihood of a feedwater reg valve spuriously opening, there can be no increase in the probability of a transient. This temp alt adds a new system that interfaces with the Condensate system and the Turbine Building Chemistry Sample System, only. These other systems are not degraded in any way that would prevent their normal function or their ability to function following an accident. These other systems are not relied upon in any accident analysis to mitigate the consequences of any accident or transient.

Safety Evaluation Number: SE-96-091 4 pe of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temp. Alt. 96-2-34

Title:

Supporting of the Jet Impingement Steel Plate

Description:

This temp alt was installed en the HPCIjet impingement restraint to support the impingement steel plate until a permanent design can be installed. This change installed new wall attachment plates to existing bolts and embedment plates. Threaded rods were then installed on the attachment plates and the impingement plate for temporary support.

Result: Unreviewed safety question does not exist. Operation of the HPCI system 12 unaffected by this change. Thejet impingement support ensures that nearby equipment will not be affected during a sudden break in the HPCI steam supply piping. The support is not physically attached to the HPCI piping system.

The originally designed support will function as designed and will not impact analyzed consequences of either a line break inside or outside ofcontainment, or inadvertent injection. It is qualified for all applicable loading conditions. The accident scenario assumptions are thus unchanged and therefore the probability of an accident is not increased.

Safety Evaluation Number: SE-96-093 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temp. Alt. 96-2-35

Title:

Outboard Reactor Head Seal Leakoff Valve Attachment A, SVP-98-113, Page 97 of 153

f l

Description:

This temporary alteration removed power to the solenoid valve associated with the outboard

- reactor head seal leakoff valve by lifting wires at the BB-92 terminal on the 902-4 panel in the control room.

This valve fails closed on loss of power. Reactor vessel seal ring leakage has been detected by the reac seal ring leakage detection system on Unit 2. Following this detection, the control switch for the 2-0220-51 and -52 valves was placed in the " DRAIN" position, the 51 valve is open and 52 valve is closed. GE S L l Number 42 recommends that once leakage through the inner reactor vessel seal ring is verified, further

! operation of the leakage collection system should be avoided to prevent increasing leakage. Unit 2 ran with this configuration for several months. During a Drywell inspection in August of 1996, resdings indicated ,

! that the drain piping indicated leakage past the 2-0220-52 valve. By installing this te;np alt, the 2-0220-51 valve is closed and provides an additional barrier to minimize additional seal ring degradation.

! Result: Unreviewed safety question does not exist. The desired effect of this temp alt is to f - reduce / eliminate leakage into the DWEDS from the head seal detection system. It is possible this leakage l could be redirected past the outboard reactor head seal. This would be considered ur. identified leakage.

L Since requirements for unidentified reactor coolant leakage are much more stringent tisan for identified l leakage paths any changes in leakage path resulting from this temporary alteration are conservative. This temp alt does not introduce the potential for a reactor pressure leakage rate that is higher than the analyzed i

for a DBA LOCA and do not affect any of the systems designed to mitigate the accident therefore consequences of the accident are not increased.

Safety Evaluation Number: SE-96-094 Type of Safety Evaluation: Procedure Changes Evaluation Reference Number: QCOP 6900-24; QCTS 0230-06; QOP 6900-14; QOP 6900-18 l

Title:

Transfer U-2125 VDC Bus, Normal & Alt Battery

Description:

This new procedure, QCOP 6900-24, provided instructions for placing the U-2 Alternate 125v battery in service and taking the U-2 Normal 125v battery out of service. The revision to QCTS 0230-06 utilized the lineup created to perform discharge testing of the batteries. This swap of batteries had previously been done by Temp Mod. The swap is done in order to perform discharge testing and recharge

of the Normal battery as required by Tech Specs. This swap prevents any loss of 125 vde to the two operating units, and the increased risk of having only one 125v battery available. Two batteries are required by Tech Specs while either plant is operating. At the completion of the swap, the Alternate battery is connected to the DC bus by its installed battery leads. To accomplish the swap, temporaryjumper leads are used.

! Result: Unreviewed safety question does not exist. For the swap of the U2125v batteries, the changes do j

not alter system function or the possibility of accident or malfunction. The change does not introduce any ,

i additional failure modes, in that the quality of thejumpers and fuses is consistent with the size and quality j of the cables and fuses already in use. The system is analyzed for loss of 125v dc. Such an event would -

. result in the loss of one of the two 125v DC systems,1/2 scrams on any operating unit, and loss of control '

ofimportant functions, however the Operators have procedures in place for crossconnecting to the operable 125v system, and performing an orderly shutdown.' Changes made by this procedure are designed to supply the same DC loads as the normal 125v battery, including those important to safety. Therefore, consequences of malfunction of equipment important to safety are no greater than using the normal battery

( and distribution path.

I l> -

1 Attachment A, SVP-98-113, Page 98 of 153 i

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I Safety Evaluation Number: SE-96-095 Type of Safety Evaluation: Temporary Alterr. tion Evaluation Reference Number: Temp. Alt. 96-1-133

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Title:

Floor Drain Plug in HPCI Interlock

Description:

Comed Drawing shows that the floor drain in the HPCI interlock is directly connected to the floor drains in the HPCI access tunnel. This temporary alteration installed an expandable plug in the drain to restrict air flow. The floor drain allows an opening in the secondary containment when the Reactor Building (inner) door is opened. The drain plug will restrict air flow to ensure that the required negative pressure is maintained in the Reactor Building. The temp alt will be replaced by a permanent design change that will eliminate the problem.

Result: Unreviewed safety question does not exist. Installation of a plug in the floor drain will ensure that secondary containment is maintained. Therefore, plant equipment will function as designed when secondary containment is maintained. The plug restricts air flow and ensures that the margin of safety is not reduced.

Safety Evaluation Number: SE-96-096 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP # 96000117/ DCN #001378M I

Title:

Modification to the HPCI steam supply pipe whip restraint 9B i

Description:

The proposed change shortens the existing W30x108 columns as shown on drwg. Nos. B- 1 I

1046 and B-1047 to avoid existing contact with the Unit I torus. The sleeve for existing whip restraint 9B l for line # l-2305-10" is currently touching the Unit I torus. The torus requires a minimum clearance of 1 1/8" due to thermal expansion and potential dynamic loading. The restraint needs to b: modified to avoid impacting the torus under all loading conditions.

l l .

l- Result: Unreviewed safety question does not exist. Operation of the HPCI system is unaffected by this l change. The fun 6 tion of the pipe restraint is unchanged. There are no new failure motes associated with shortening the pipe restraint. The restraint will still function as designed and will not negatively impact the

' torus or the HPCI steam supply piping system or supports. The restraint is being shortened to avoid potential contact with the torus, which will lesson the probability of damage to the torus in the unlikely l event of a pipe rupture.

l Safety Evaluation Number: SE-96-097 Type of Safety Evaluation: Work Request Evaluation Reference Number: WR 950084280

Title:

Main Chimney SPING Fail External Alarm '

DescAption: This work request disabled the FAIL EXTERNAL alarm fun.: tion for the Main Chimney

{

Separate Particulate Iodine & Noble Gas (SPING) Monitor. This change bypassed the flow switch contacts  :

by connecting the wire from the SPING Alarm Board directly to GROUNI', thereby disabling the Main l l

Attachment A, SVP-98-113, Page 99 of I 33 L

I Chimney SPING FAIL EXTERNAL, such that no FAIL FXTERNAL alarm lights are received should flow i

drop below the low flow setpoint or exceed the high flow satpoint. By design, the Main Chimney sample

! system internally switches its sample flow to its Victoreen semple pump when the SPING MEDIUM or l

HIGH radiation ranges are entered during an accident situatioe; by design this sample flow via the Victoreen sample pump is much lower than normal sample flow, and a resulting LOW flow condition occurs. This results in a FAIL EXTERNAL and the A-Model reports channel status as " BAD" even though flow through the Victoreen sample pump is proper per design. This change eliminated incorrect reporting of channel status as " BAD" during accident situations when flow has switched to the Victoreen sample l pump.

Result: Unreviewed safety question does not exist. This change spares wires in place or removes them as practical, and is done in such a manner that no interaction between the spared wires and other internal equipment will occur. This change only affects v res for the FAIL EXTERNAL circuitry associated with the flow switch. As the flow switch is being bypassed from the FAIL EXTERNAL circuitry as a result of I

this change, there is no affect on the flow switch. The FAIL light at the Control Terminals is visible only, and the FAIL EXTERNAL function during an accident situation serves only to cause the A-Model to provide false channel status information. The consequences of a failure of the flow switch are therefore unchanged as a result of this vendor-approved change and likewise the consequences of an accident are not increased.

Safety Evaluation Number: SE-96-098 Type of Safety Evaluation: Work Requests Evaluation Reference Number: WR 960056S44,960056846,960056849,960056851

Title:

Drywell Radiation Monitor Test Modes TRIP INHIBIT l

Description:

This change disabled the TRIP INHIBIT function for the TRIP ADJUST and CHECK modes for the Drywell Radiation Monitors by disconnecting a jumper wire that connected the function j switch to the count rate meter board. The Technical Specification Upgrade Program required that a monthly j Channel Functional Test, which includes a verification of required alarm and/or trip functions, be  :

performed. Existing Tech Specs requires only a qualitative instrument check on a monthly basis. The existing jumper provides TRIP INHIBIT such that it does not allow alarm and trip functions to occur when using TRIP ADJUST and CHECK modes of the Drywell Radiation Monitor Function Switch. Because utilization of the TRIP ADJUST and CHECK modes is the most practical method to implement the TSUP requirement so alarm and trip functions can be verified, this change was required.

Result: Unreviewed safety question does not exist. This change has no effect on the OPERATE mode for the monitor, therefore has no effect on the safety function of the monitor. This change enables the trip function for both the TRIP ADJUST & CHECK modes, thereby ensuring proper operation of the monitor's trip functions on a monthly basis. This change eliminates the possible failure of thejumper wire's termination point connections, and thereby eliminates the potential for a resultant unexpected trip while in a test mode. When in test mode TRIP ADJUST & CHECK, this change increases the potential of a spurious l trip by disabling the TRIP INHIBIT. However, spurious trips due to equipment failure during the performance ofinstrument testing are a known risk and testing is conducted accordingly to minimize the consequences of such trips.

l Safety Evaluation Number: SE-96-100 Type of Safety Evaluation: Work Request Attachment A, SVP-98-il3, Page 100 of 153

4 Evaluation Reference Number: WR 960040026-01 I l

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Title:

SRM Shorting Link Removal SCRAM Functional Test l

Description:

This work package tested the non-coincident SCRAM function of the Reactor Protection '

System (RPS) as required by the new Technical Specifications. This was accomplished by removal of a specified shorting link and using the RAMP function of the SRM chassis to create a high-high trip. With the shorting link removed, the SRM trip will cause a 1/2 SCRAM. This is verified for each (4 total) SRM.  !

Also, this package tested IRM SCRAM function by taking the chassis Mode Switch to the "125" position to -

cause a high-high trip. The appropriate channel 1/2 SCRAM was verified. The new Tech Specs required a channel functional test prior to core alterations.

Result: Unreviewed safety question does not exist. Performance of this work package does not create any -

new system interactions either within or external to the neutron monitoring system. The work was performed while the plant was in cold shutdown, which is the condition these scram functions were intended  !

to function. No new failure modes were created. The IRMs and SRMs were operated in accordance with  ;

their design.

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Safety Evaluation Number: SE 96-102 Type of Safety Evaluation: Setpoint Changes Evaluation ' Reference Number: Setpoint Change #9600297, %00298,9600299,9600300,9600301,

% 00302 i

Title:

Replaces Westinghouse FA Type Breakers with Westinghouse HMCP Type Breakers at MCC 2B (Breakers for RCIC Vacuum Pump, RCIC Cond. Pump, MOVs 2-1301-17,22,26,48,49,53,60,61,62)

Description:

This change replaced existing Westinghouse FA type breakers with Westinghouse HMCP type breakers for RCIC motors including MOVs fed from MCC 28. The existing FA breakers are obsolete.

The new breakers are seismic and environmentally qualified breakers. This evaluation is applicable to all ,

above mentioned loads even though all the breakers may not be replaced under the identified setpoint change (s).

Result: Unreviewed safety question does not exist. This c'aange dor.s not functionally change the operation of the equipment or how the equipment operates electrierJiy. The :,nange in the breaker does not affect the ability of the equipment to perform its intended design runctica. The changed breakers are an equivalent replacement to the existing breakers as evaluated Changes in breaker size were also evaluated and found to have no affect. This change does not introduce any new failure modes. The change of the breaker rating has no effect on equipment failures. The operation of the breakers remains the same.

- Safety Evaluation Number: SE-96-103 Type of Safety Evaluation: Setpoint Changes Evaluation Refereace Number: Setpoint Change #9600290,9600291,9600292,9600293,9600294, 9600295,9600296 l:

Titlei Replaces Westinghouse FA Type Breakers with Westinghouse HMCP Type Breakers at MCC 2B i

Attachment A, SVP-98-1I?, Page 101 of 153

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)

l k

(Breakers for HPCI Emerg. Bearin, Oil Pump, HPCI Gland Exhauster, HPCI Turning Gear, HPCI Cond. Pump and MOVs fed from MCCs 2A and 2B) l

Description:

This change replaced existing Westinghouse FA type breakers with Westinghouse HMCP type breakers for HFCI, RHR, and Main Steam system motors including MOVs fed from MCCs 2A &2B.

The existing FA breakers are obsolete. The new breakers are seismic and environmentally qualified j breakers. This evaluation is applicable to all above mentioned loads even though all the breakers may not j be replaced under the identified setpoint change (s).

l Result: Unreviewed safety question does not exist. This change does not functionally change the operation of the equipment or how the equipment operates electrically. The change in the breaker does not affect the ability of the equipment to perform its intended design function. The changed breakers are an equivalent replacement to the existing breakers as evaluated. Changes in breaker size were also evaluated and found to have no affect. This change does not introduce any new failure modes. The change of the breaker rating has no effect on equipment failures. The operation of the breakers remain the same.

Safety Evaluation Number: SE-96-104 Type of Safety Evaluation: Drawing Change Reqcest l Evaluation Reference Number: DCR 960216

Title:

Reactor Building / Secondary Containment

Description:

A DCR was submitted to create a new B-drawing to show alternate siding details.

Additionally, this DCR corrected vendor drawings, which incorrectly showed that group 1 bolts were installed in the parapet channel. This safety evaluation addresses this issue only.

Result: Unreviewed safety question does not exist. Actual plant configuration was field verified and l Calculation QDC-0020-S-0176 proves that these bolts are not necessary. This change to the drawings so that they will accurately reflect plant configuration will not adversely affect plant operation in any operating mode. Equipment failures are not affected. These bolts have never been installed; they are incorrectly shown on the drawings. No new accidents or malfunctions different from those already evaluated are  ;

introduced by correcting these drawings.

Safety Evaluation Number: SE-96-105 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QlP 2000-01,02; QCOS 1600-07; QOA 900-4 A-17, B-17

Title:

Recalibration of Drywell Floor and Equipment Dreins

Description:

Recalibrate the Drywell Floor Drain and Drywell Equipment Drain FT's from 0-388"we input /10-50 ma output to 0-338"we input /10-50 ma output. The existing instrument calibrations are incorrect. Changing the instrument calibration provides correct instrument output, and therefore, more accurate flow totalizer readings.

Result: Unreviewed safety question does not exist. Recalibrating instruments to provide more accurate flow totalizer readings has no effect on system operation nor system interactions during all applicable operating modes. The calibration change makes no physical change in the plant that could impact safety -

Attachment A, SVP-98-113, Page 102 of 153

related equipment. This change merely restores the instrument to its intended accuracy. No new components are added. No new system interactions are created. The change does not affect equipment failures, and no new failure modes result.

Safety Evaluation Number: SE-96-106 Type of Safety Evaluation: Drawir.g Change Request

' Evaluation Reference Number: DCR 960200

Title:

Delete Three Pipe Vibration Restraint Drawings for Ul HPCI and RHRSW

Description:

During pipe vibration restraint walkdowns, it was noted that three suppons were not installed. Since the supports are not part of original construction and nearby supports mitigate pipe vibration, the drawings are being deleted. The support drawings were issued but never installed. The i piping design basis calculations and all piping isometric drawings do not reference the vibration restraints.

Result: Unreviewed safety question does not exist. The piping design basis analyses do not contain the three vibration supports. Deleting the support drawings have no impact on the operation of the HPCI and kHRSW systems since the inpports are not installed or required. There are adjacent supports that will '

mitigate potential pipe vibration, thus the probability of an accident will not increase. The HPCI and RHRSW piping systems meet all Code requirements without the three suppons installed.

Safety Evaluation Number: SE-96-108 Type of Safety Evaluation: Setpoint & Procedure Change Evaluation Reference Number: QCIS 0200-23

Title:

Shutdown Cooling Pressure Permissive

Description:

- The purpose of this evaluation was to suppon the instrument setpoint change and procedure change for the calibration of pressure switch 1(2)-261-23A(B) from 125 psig to 118.5 psig. These pressure switches control the cut-in permissive for Reactor Vessel Pressure - High for RHR Shutdown Cooling Mode Isolation. A calculation was performed to suppon the Technical Specification Upgrade Program (TSUP). .

These switches are being placed in TSUP with a setpoint in Tech Specs ofless than or equal to 135 psig.  !

The calculation error analysis indicates that the switches need to be set at 118.5 psig in order to ensure the l 135 psig Tech Spec is met.

l l Result: Unreviewed safety question does not exist. Changing the setpoint of these instrsments does not l create any new failure modes for the failure of these instruments. Lowering the xctpoint increases the

- conservatism for the protection of the low pressure piping on RHR. This change does not affect the way in

l. which the RHR system can fail.. The consequences of pressure switch failure are the same no matter what l

the setpoint is of the pressure switches. This change does not create any new accidents beside those already considered.

Safety Evaluation Number: SE-96-109 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCTS 0360-01 i

.l '

Attachment A, SVP-98-113, Page 103 of 153 l

Title:

Feedwater Level Control System Startup Test Procedure

Description:

This procedure revision deletes the test steps that evaluate the FRV valve response to small (3 & 5%) and large (7 & 10%) feedwater flow rate changes. These cl.anges in flow rate are difficult to produce without extensive testing equipment and per General Electric are not required to verify the proper operation of the FRV's. The original Unit I starop test instructions did not perform flow rate change

> testing to verify proper operation of the FRV's. This procedure revision for Unit 1 feedwater level control system testing was generated cased upon the Unit 1 Initial Startup Test Instructions, STI-23. The level setpoint test was performed by moving the level setpoint controller down and up about 6 inc .. as quickly.

as possible and the trarisient recorded until steady-state is achieved. The dynamic adjustment i ithe system was determined by observation ofits response to level setpoint disturbances at each test point sch6duled durind 1 and 3 element control mode of operations. In addition to the level setpoint changes testing specified in STI-23, other test activities were performed in accordance with this res ised startup test procedure. The reasons for the testing were to demonstrate that newly installed feedwater regulator valve hydraulic cruators were capable of providing automatic response to the initiated feedwater system transients, and verify that transient response of the feedwater system in 1 and 3 element control modes

._ provide fast and smooth reactor water level control at all operating modes.

Result: Unreviewed safety question does not exist. Performance of this procedure does not impact safe operation of the plant nor degrade adequacy of system, structure, or component of the feedwater level I control system. This test does not increase possibility of malfunctions that resulted in a maximum or zero fHwater flow since size of flow demand and level step inputs were limited during the test,1-element and .

j m nual feedwater regulating valve control was available to the NSO at all times throughout the period of '

performing this procedure, and testing involved in this procedure does not ha've the potential to increase the probability of feedw&r controller failure. The low flow controller is designed to maintain reactor level during low power operation. Feedwater system is not utilized to mitigate consequences of any accident.

Test performed specifically to verify that control components associated with feedwater level control system meet their design requirements Safety Evaluation Number: SE-96-111 Type of Safety Evaluation: Work Requests, Setpoint Changes Evaluation Reference Number: .. WR 960087588-01, %0087571-01

Title:

Drywell Radiation Monitors High Radiation Alarm Setpoint Changes

Description:

The High Radiation Alarm Setpoint (ALERT) and the High Radiation Trip Setpoint for the Drywell Radiation Monitors were changed from 80 R/he to 20 R/hr and from 100R/hr to 30 R/hr, respectively. Per the Technical Specification Upgrade Program (TSUP), the Drywell Radiation High Trip _

' Setpoint shall be less than or equal to 100 R/hr. Previously, this setpoint was not in Tech Specs. The existing High Radiation Trip Setpoint is being lowered te ensure compliance with the TSUP limit throughout a calibration interval, taking into account crort and uncertainties of equipment and test

methodology.

. Result:. Unreviewed safety question does not exist. There are no new failure modes introduced by this change. The new setpoints provide an increase in conservatism to the TSUP limit.' This setpoint change is well within the range of the Drywell Radiation Monitors. This change increases the probability that the H . Attachment A, SVP-98-113, Page 104 of 153

monitor will provide the trip initiation at less than or equal to 100 R/hr as required by TSUP, thereby decreasing the probability that the consequences of an accident increase. This change affects the instrument setpoint only, and therefore has no effect on the consequences of a malfunction of the instrument.

Safety Evaluation Number: SE-96-112 Type of Safety Evaluation: Setpoint Change Evaluation Reference Number: Setpoint Change #95-0321

Title:

APRM Flow Bias Rod Block Setpoint

Description:

This change reduced the APRM Flow Bias Rod Block Setpoint setdown for drift such that the APRM Flow Bias Rod Block is set to the GE recommended nominal value S s (0.58Wd + 50). T purpose of this change is to providejustification for adjusting this setpoint to the GE recommended value.

Presently, the APRM Rod Block Flow Biased alarm initiates at approximately 103% flow control line. This is due to the combined effect of setting the APRM Rod Block Setpoint Setdown down by Instrument Maintenance and the Flow Bias and APRM noise. The setdown produces unnecessary nuisance alarms in the control room which results in increased red motion and fuel duty in order to clear the nuisance alarms.

Result: Unreviewed safety question does not exist. The APRM Rod Block does not provide a safety-related function. The safety-related function is provided by the APRM Reactor Protection Systern trip. Rod blocks generaied by the APRMs are not required for safe operation of the plant. The APRM system will continue to function in the same manner as before. Adjusting the setpoint to the GE recommended value will not adversely impact equipment failures. The change does not result in any new failure modes.

Changing the Rod Block setpoint will have no effect on the SCRAM setpoint. The affect of alarming at a higher flow control line will not adversely increase the consequences or probability of a malfunction of equipment important to safety.

Safety Evaluation Number: SE-96-113 Type of Safety Evaluation: Revision to UFSAR Evaluation Reference Number: UFSAR Tracking No. 96-51

Title:

GE10 Fuel

Description:

Th s evaluation updated the Updated Final Safety Analysis Report (UFSAR) to reflect the use of GE10 (GE8x8NB-3) fuel. The GE10 fuel type was not previously described in the UFSAR. A description of the new fuel type and its associated effects, including the bundle channel design was incorporated into the UFSAR after being reviewed and approved by on-site review.

Result: Unreviewed safety question does not exist. The plant operating conditions / max power levels and flows are not changed with this fuel bundle change, therefore the potential for an equipment malfunction is not changed. The NRC has generically approved GE10 fuel for use as a reload fuel type. The GE10 design is neutronically compatible with the existing fuel types and core components in the QlC14 core. No new l single failures have been created. GE10 fuel has been found not to have any significant dynamic loading changes on the reactor, internals, or fuel assemblies. None of the UFSAR assumptions have been l invalidated with this reload. I Attachment A, SVP-98-113, Page 105 of153 u

l

. Safety Evaluation Number: SE-96-116 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9400076

Title:

Residual Heat Removal Pump Minimum Flow Valves

Description:

This change regeared the existing Limitorque motor niuate.WlO 2-1001-18A(B). These valves were equipped with an Overall Ratio (OAR) of 27.99. Implementation of this change resulted in an OAR of 47.85 for these valves. This change also upgraded spria.g packs to 0101-091 spring packs,if required. This change increases the motor gearing capacity in the opening and closing directions, which will result in an improved thrust window for these valves. The upgraded spring packs provide a grease relieving capability to the spring packs which results in improved performance for the spring packs.

Result: Unreviewed safety question does not exist. This change does not impact existing operating modes or create any new operating modes. Th failure modes for th: regeared motor actuators are the same as those for the existing valve actuators. No new failure modes are introduced and existing failure modes are not impacted. Replacement of the gearing and spring pack results in a negligible change in both the weight and center of gravity and does not require a revision to the existing piping stress analysis. These changes do not add any new interactions between RHR and the equipment initiating anticipated transients. Therefore, the p:obability of an accident is not increased. This change does not alter the function of the valves, but improves the reliability of the valves' opening and closing functions.

Safety Evaluation Number: SE-96-118 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP %00272/DCN 001379S

Title:

High Pressure Coolant injection Steam Supply Pipe Whip Restraint 9B

Description:

This design change shortened the existing W30x108 columns as shown on Station drawings to avoid existing contact with the Unit I torus; The sleeve for the existing pipe whip restraint 9B for line no.1-2305-10"is currently touching the Unit I torus. The torus requires a minimum clearance of 1 1/8" due to thermal expansion and potential dynamic loading. The pipe whip restraint needs to be modified to avoid impacting the torus under all loading conditions.

Result: Unreviewed safety question does not exist. There are no new failure modes associated with shortening the pipe rupture restraint. The restraint will still function as designed and will not negatively impact the torus or the HPCI steam supply piping system or supports. A calculation was performed to qualify the additional steel and welds required to modify the restraint. Since the shortened pipe rupture >

restraint will avoid contact with the torus, the consequences of a malfunction of equipment important to safety will not increase. Shortening the HPCI pipe rupture restraint will not adversely affect the HPCI system since the support is designed to accommodate normal pipe movements. Moving the support away from the torus will allow the torus to function as designed.

! . Safety Evaluation Number: SE-96-121,R1 Type of Safety Evaluation: Design Change I

i Evaluation Reference Number: DCP 9100288 (P04-2-91-114)

I L

Attachment A, SVP-98-113, Page 106 of 153

Title:

U2 HPCI Motor Gear Unit Control Switch -

n

Description:

This change replaced HPCI Motor Gear Unit (MGU)' control sw' itch with a switch equipped with a " break before make" contact configuration. The existing switch is being replaced in order to provide positive isolation of the MGU signal converter during manual operation of the MGU and also eliminate the need for an independent MGU signal converter isolation relay that will be determed and spared. _With existing control switch, a momentary paralleling of control potentials of differing polarities could cause a l~ spurious ground of the 125vde electrical systemL Installation of the control switch with the " break before L make" contact arrangement will eliminate the ground indication and allow operators to assume remote -

l manual control of the HPCI turbine at low speed operation.

Result: Unreviewed safety question does not exist. This change eliminates two potential modes of failure

for the HPCI system. The " break before make" contact configuration of the new switch prevents possible

. electrical shorting of control potentials that could result in malfunction of the MGU and consequently loss

; of HPCI capability. The sparing of a control isolation relay eliminates an unnecessary electrical control

circuit component and therefore, one additional possibility of HPCI malfunction. The probability of malfunction / failure of an important safety system is effectively decreased by implementation of this change.

.]

I-Safety Evaluation Number: SE-96-122 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9300308 (E04-2-93-298)

Title:

' Unit 2 Diesel Generator Pressure Gauges -

Description:

This change added new gauges for Unit 2 Diesel Generator crankcase vacuum, air box, fuel filter inlet, and fuel oil filter outlet. Each gauge was installed on an instrument line with an isolation valve and test tap. The new gauges provide more detailed system monitoring and aid in diagnosing maintenance functional failures. The isolation valve and test tap facilitate the reading, testing, and calibration of the new gauge.

Resulti _Unreviewed safety question does not exist. The indicating function of the new gauges is a n'on- ,

. safety-related function. The failure modes of the new sensing lines are not any diftenmt than the failure  !

modes of the existing in$trument lines. The addition of the pressure indicators do not affect the operation'of

~ _; the diesel. The new instruments and sensing lines are designed to maintain the safety related pressure -

boundary. The purpose of the indicators is to give additional indication for use in trending EDG

- performance. This data will be used for performing preventative maintenance to correct potential problems. j

' Thus, the data provided by these instruments will help make the diesel more reliable and eliminate possible j failures. Thus, the probability of a malfunction of equipment important to safety does not increase.

4 Safety Evaluation Number: SE-96-123 Type of Safety Evaluation: Design Change l

Evaluation Reference Number: DCP 9600248

Title:

Rewire Standby Liquid Control Resistors

Description:

' This change rewires Standby Liquid Control Firing circuit current limit resistors. These resistors were jumpered together with an additionaljumper extending from one resistor to a terminal strip at Attachment A, SVP-98-113, Page 107 of 153

/

the 902-5 panel. These resistors were changed such that each resistar now has its own jumper between it and the terminal strip. The wiring at the meter relay subpanel "ar" located at control room panel 902-5 was field verified and incorporated onto the plant for record drawings. The soldered connections at these resistors were found to be unrW le.

Result: Unreviewed safety question does not exist. This change increases the reliability of the SBLC's firing circuit. By relocating the connection point from the resistors to the terminal strips, the probability of a failure of the connection point becomes less likely due to the stronger connection at the terminal strips.

This change has no effect on the way SBLC functions.

Safety Evaluation Number: SE-96-124 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600009 (E04-2-96-004)

Title:

RHR LPCI Inboard Injection Valve

Description:

This change involved drilling a hole (3/16" diameter) through the Reactor Recirculation piping inlet (high pressure) side of the valve flex-wedge disc. Drilling this hole permits the pressure in the bonnet to equalize with the piping system thereby reducing the differential pressure and eliminating pressure locking susceptibility. This improves the reliability of the RHR LPCI Injection Valve. '

Result: Unreviewed safety question does not exis t. The drilled hole improves the reliability of the valve to perform its function to open for LPCI injection. The function of the valve to close for the required primary containment isolation is not affected by this change. This change improves the RHR system's capability to respond to the long term cooling requirements of the reactor as specified in the UFSAR for anticipated transients or accidents. The change is entirely internal to the valve so there are no new interfaces created between the valve and other systems. The change does not create any new failure modes for the valve.

Drilling a hole in the valve disc does not increase the probability of a loss of pressure boundary integrity.

The valve disc is not part of the system pressure boundary, and the ability to isolate the reactor coolant has not been reduced because the outboard valve disc will remain intact.

Safety Evaluation Number: SE-96-125 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temp. Alt. 96-1-141

Title:

Contaminated Condensate Storage Tank B

Description:

This temporary alteration installed a tygon tube on the level indicator for the B CCST.

Problems were encountered during vent verification for Unit 2 HPCI. The tygon tubing needed to be installed on the level indicator to verify the indicator was working properly and to verify that adequate level was being maintained in the B CCST.

Result: Unreviewed safety question does not exist. Installation of the tygon tube does not effect the design function of the CCST or any of the systems which are fed from it. The tubing is there to provide a verification that the level indicator for the CCST is accurate. The tubing is only valved in when there is someone in attendance. if a failuie of the tube should occur, the valve is closed and the tubing isolated.

f Therefore, no new failure mechanisms are introduced and the probability remains the same. The tubing has l

Attachment A, SVP-98-113, Page 108 of 153

no direct interaction with equipment causing any transients or accidents and therefore cannot affect the probability of the transients or accidents.

Safety Evaluation Number: SE-96-129 Type of Safety Evaluation: Procedure Change

. Evaluation Reference Number: QCOP 1300-01,2300-01,04; QCOS 1300-10,21,2300-09,24 Delete QCOS 1300-09,2300-08

Title:

, Method of Verifying Discharge Header Fill Daily For HPCl/RCIC Systems

Description:

'lhis change provided, as an alternative to the requirement to perform a daily vent of the

' HPCI or RCIC system from the high point vent when the suction for that system is lined up to the Suppression Pool, the allowance to perform a valve lineup verification of the keep-fill feed to HPCI or RCIC. The dose received while performing the verification of discharge fill (venting from the high point vent) is a considerable amount (approximately 40-50 mrem total dose).

Result:

Unreviewed safety question does not exist. The change to the methodology for verifying this fill will continue to ensure that the systems are in the ready condition with a filled discharge header. The accidents and malfunctions in the UFSAR that take credit for HPCI and RCIC assume the system discharge headers are full. Therefore, this change does not change an initial condition or a structure, system, or component used in a UFSAR analysis. No failure frequencies are affected by this change, and no new failures are introduced. Loss of fill due to failures in part of the keep-fill system associated only with HPCI or RCIC would be identified by valve verification surveillance in case of valve mispositions or by normal rounds. Failures in the HPCI or RCIC systems that result in loss of fill would have resulted in loss of fill when systems were lined up to the CCST, also.

. Safety Evaluation Number: SE-96-133 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600380

Title:

"B" Train Control Room HVAC Refrigeration Condensing Unit

Description:

. This evaluation was prepared to add comments that were provided during the review process for SE-96-127. Currently, the crankcase heaters for the "B" Train control room HVAC compressor are fed from MCC 16-3-1 which is non-safety-related. This design change disconnected the heaters from this source and re-fed them from safety-related MCC 18-4. The cables and raceway from this MCC to the compressor's control panel were installed safety-related and will be re-used. However, the crankcase heaters and the raceway that runs between the heaters and the control panel are non-safety-related and will be replaced with safety-related components.

Result: Unreviewed safety question does not exist. With this new configuration, power would be available to the heaters under all plant conditions resulting in a more reliable system. By re-feeding the heaters from a more reliable source, and replacing all non-safety-related components in the heater circuit with safety-

' related ones, no new failure modes are created. The addition of the heaters to MCC 18-4 has been .

evaluated, and it was determined that they are within the capacity of the MCC. If the compressor itself should fail, the heaters would not be required. If MCC 18-4 should fail, the entire "B" train control room Attachment A, SVP-98-113, Page 109 of 153

HVAC would be disabled, and again, the heaters would not be needed. Consequences of a malfunction of equipment important to safety remain the same as a result of this power supply upgrade.

Safety Evaluation Number: SE-96-136 Type of Safety Evaluation: Tech Spec Bases Change Evaluation Reference Number: Tech Spec 3/4.9.A

Title:

Diesel Generator Full Load Reject Test

Description:

This change to the bases for Technical Specification 3/4.9.A clarifies the purpose of the Diesel Generator Full Load Reject Test.

Result: Unreviewed safety question does not exist. This change does not affect plant operation, does not affect equipment failures, and does not introduce any new failure modes. Therefore, this change does not adversely impact systems or functions so as to create the possibility of an accident or malfunction of a type different from those evaluated in the UFSAR. It merely provides a more explicit description for the purpose of a test.

Safety Evaluation Number: SE-96-137 Type of Safety Evaluation: Tech Spec Bases Change Evaluation Reference Number: Tech Spec 3/4.7.D

Title:

Clarifia Meaning of Administrative Control

Description:

This change to the bases for Technical Specification 3/4.7.D clarifies the meaning of the term "adininistrative controls" as used in Note a on page 3/4.7-6. The clarification wording is taken from the improved Technical Specifications, NUREG-1433. Generic Letter 91-08 provided guidance for having a note allowing primary containment isolation valves to be opened under administrative controls. This guidance stipulated that a description of the required administrative controls be included in the Technical Specification bases.

Result: Unreviewed safety question does not exist. This change does not affect plant operation. It provides industry accepted guidance for details of the term " administrative controls" for opening primary -

containment valves. Guidance for the controls required will be available in the Technical Specification bases. This additional guidance does not affect any equipment failnres and will not create any new failure modes.

Safety Evaluation Number: SE-96-138 Type of Safety Evaluation: Design Change Evaluation Referemec Number: DCP 9500044 (E04-2-95-039)

Title:

RHR, CS, & HPCI Indicating Lights

Description:

This change installed resistors in series with the indicating lights for RHR, CS, and HPCI located on the 902-3 and 2212 4 panels. They were plate mounted and attached to existing unistrut in the l referenced panels. The wiring for each affected indiceting light was reconfigured to connect the resistor in Attachment A, SVP-98-113, Page 110 of 153

series with the indicating light on the hot side of the light. Removal of burnt out indicating light bulbs from their sockets have caused fault currents resulting in blown fuses or damaged equipment b rendering the subject control circuit inoperable. The addition of the resistors in series prevent short circuits in the event of indicating light bulb or socket failure.

Result: Unreviewed safety question does not exist. This change does not affect plant operations. The addition of resistors and replacement ofindicating lamps has no impact on system interactions. The addition of resistors in series with the indicating lights decreases the probability of a malfunction of equipment. The resisters ensure that the equipment remains functional in the event ofindicating lamp or indicating light fixture failure.

Safety Evaluation Number: SE-96-139 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600118 j

Title:

Standby Liquid Control (SBLC) Pump Discharge Check Valve

Description:

This change replaced the SBLC pump discharge lift-check valve manufactured by Hancock with a Rockwell-Edwards piston-check valve. The discharge check valve was not seating to prevent backflow. Since replacement parts did not exist for the existing valve, a new valve was installed. i Result: nreviewed safety question does not exist. There are no new failure modes associated with the new valve. Modification of the existing SBLC system by replacing the existing pump discharge lift-check valve with a piston-check valve does not affect the probability of an ATWS occurring. Installation of the new check valve does not affect the reliability of the system and will not be inimical to the health and safety of the public. The new valve docs not create any seismic, loading, EQ, or separation concerns nor does it downgrade performance, affect redundancy, or independence of any systems, structures, or components.

Safety Evaluation Number: SE-96-141 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600229

Title:

MCC 28-3 Feed Cable Replacement

==

Description:==

This change upgraded existing power feed cable from Switchgear 28 to Motor Control Center 28-3 with new cables that have a higher arr.pacity rating and replaced 11 breakers at MCC 28-3. l The existing power feed cables were removed from Switchgear 28 and pulled back into the Reactor Building and abandoned in place. Routing of the six new power feed cables utilized the same secondary containment penetration used for the original power feed cables. The penetration was modified as necessary to accommodate the new cables and rescaled with an approved fire seal to maintain an adequate fire barrier and ventilation seal between the Reactor and Turbine Building zones. This change was done in response to a potential for periodically exceeding the MCC 28-3 feed cable rated continuous ampacity during plant operations.

l Result: Unreviewed safety question does not exist. The functional interactions of the new cables with I other systems, structures, and components remain unchanged. The new cables installed function the same as the originally installed cables. The new cable ampacity rating allows the MCC connected load to be Attachment A SVP-98-ll3, Page 111 of153

{

energized in any combination without experiencing cable degradation. The existing cable ampacity did not allow this operator flexibility, Installation of the new power cables and breakers do not introduce any new failure modes because they are identical to those currently installed. These changes only enhance the a of the components to perform their designed function. Eliminating concerns of power feed cable degradation and increasing short circuit rating only increases reliability of the loads powered by this MCC.

Safety Evaluation Number: SE-96-142 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600227

Title:

MCC 29-1 Feed Cable Replacement

Description:

This change upgraded existing power feed cable frc,in Switchgear 29 to Motor Control Center 29-1 with new cables that have a higher ampacity rating and replaced 3 breakers at MCC 29-1. The existing power feed cables were removed from Switchgear 29 and pulled back into the Reactor Building and abandoned in place. Routing of the six new power feed cables utilized the same secondary containment penetration used for the original power feed cables. The penetration was modified as necessary to accommodate the new cables and resealed with an approved fire seal to maintain an adequate fire barrier and ventilation seal between the Reactor and Turbine Building zones. This change was done in response to a potential for periodically exceeding the MCC 29-1 feed cable rated continuous ampacity during plant operations.

Result: Unreviewed safety question does not exist. The functional interactions of the new cables with other systems, structures, and components remain unchanged. The new cables installed function the same as the originally installed cables. The new cable arr:pacity rating allows the MCC connected load to be energized in any combination without experiencing cable degradation. The existing cable ampacity did not allow this operator flexibility. Installation of the new power cables and breakers do not introduce any new l failure modes because they are identical to those currently installed. These changes only enhance the ability l

of the components to perform their designed function. Eliminating concerns of power feed cable

{

~

degradation and increasing short circuit rating only increases reliability of the loads powered by this MCC. j l

Safety Evaluation Number: SE-96-143 Type of Safety Evaluation: Design Change 1 Evaluation Reference Number: DCP 9600225

Title:

Upgrade of Power Feed Cables for MCC 28-1B

Description:

This change upgraded existing power feed cable from Switchgear 28 to Motor Control Center 28-1B with new cables that have a higher ampacity rating. The existing power feed cables were removed from Switchgear 28 and pulled back into the Reactor Building and abandoned in place. Routing of the six new power feed cables utilized the same secondary containment penetration used for the original l power feed cables. Tue penetration was modified as necessary to accommodate the new cables and resealed i with an approved fire seal to maintain an adequate fire barrier and ventilation seal between the Reactor and Turbine Building zones. This change was done in response to a potential for periodically exceeding the.

MCC 28-1B feed cable rated continuous ampacity during plant operations.

I '

Attachment A, SVP-98 113, Page 112 of 153 I

l

I Result: Unreviewed safety question does not exist. The functional interactions of the new cables with other l systems, structures, and components remain unchanged. The new cables installed function the same as the l originally installed cables. The new cable ampacity rating allows the MCC connected load to be energized in any combination without experiencing cable degradation. The existing cable ampacity did not allow this

)

operator flexibility. Installation of the new power cables do not introduce any new failure modes because ,

they are identical to those currently installed. These changes only enhance the ability of the components to J perform their designed function. Eliminating concerns of power feed cable degradation and increasing short circuit rating only increases reliability of the loads powered by this MCC.

Safety Evaluation Number: SE 96-145 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: DCP 9600406

Title:

Temporary Alteration (TA) on Reactor Feed Pump Motor Windings Damper l

Description:

The ventilation damper 1-5772-24C shaft bushing has broken. A fabricated replacement l will be installed to allow the damper to operate as normal. Installation of this TA will allow continued operation of the RFP ventilation system until such time that the situations can be throughly investigated and ;

proper corrective action can be determined and initiated.

i Result: Unreviewed safety question does not exist. The installation of this TA will not affect any functions  !

of system or plant operations. The installation of this TA will allow for proper function of the damper.

There are no new failure modes presented as a result of this TA. Installation of this TA will allow the damper to operate as designed.

l Safety Evaluation Number: SE-96-147 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temp. Alt. No. 96-1-146 i

Title:

Cable Tunnels in the Aux. Electric Room Scaled

Description:

This change sealed hatchways in the Aux. Electric Room frem the Cable Tunnel below by laying duct tape along the edges. This was done by this temp. alt, until a permanent seal can be designed.

The duct tape will help to reduce the leakage from the Aux. Electric Room when the Control Room HVAC l

'B' Train is in operation. The Temp. Alt. was tested by pressurizing the Emergency Zone to 1/8" water differential pressure.

Result: Unreviewed safety question does not exist. Consequences are not increased because the tape will not prevent the 'B' Train from performing its' design function. The tape helps to reduce air flow through the hatch and allows the 'B' Train HVAC to maintain the required differential pressure in the Emergency zone. Operation of equipment in the Aux. Electric Room is not affected. Hatchways are located at floor level so the tape will not affect operation of nearby equipment. Water drainage from the room is not a concern because no water lines run through the room. No new failure modes are created. No other equipment important to safety is prevented from performing its' design function during an accident.

Safety Evaluation Number: SE-96-151 Type of Safety Evaluation: Engineering Request Attachment A, SVP-98-ll3, Page 113 of 153

3 1

' Evaluation Reference Number: ER9500007

Title:

Main Steam Line Flow Switches i

i

Description:

This evaluation provides technicaljustification for replacing the Barton model 278 {

differential pressure indicating switches, currently installed for Main Steam Line flow switches with Barton '

' model 288A differential pressure indicating switches. The only change is in the switching mechanism. The model 278 is furnished with mercury switches while the model 288A is supplied with snap acting switches.

This evaluation therefore addresses the replacement of the mercury switch with a snap acting switch. The installed model comes equipped with mercury filled switches, which by their nature, are sensitive to vibration. They have been the cause ofmany spurious 1/2 Primary Containment Isolation signals and reactor trips, and this is the reason for this replacement.

Result: Unreviewed safety question does not exist. A calculation has been performed that concludes the replacement model will actuate as designed using the same nominal setpoint as the currently installed model. Therefore, consequences of an accident will not be increased. The new model switches are less i

sensitive to vibration which will improve plant operation by eliminating one potential cause of spurious  ;

reactor scrams. The replacement of switches does not affect any equipment failures. The intended design  !

function remains unchanged. No new failure modes introduced. Consequences of failure of the new switches are the same as those associated with the original switch, therefore, consequences of a malfunction of equipment important to safety will not be increased.

i Safety Evaluation Number: SE-96-152 Type of Safety Evaluation: Adopt Vendor Procedure Evaluation Reference Number: SEP-092-04; SEP-092-02-01

Title:

Perform Neutron Attenuation Testing

Description:

This evaluation addresses adopting vendor procedures to perform neutron attenuation testing in the spent fuel racks. This testing measures the Boron-10 area / density. This is part of work scope to identify any degradation of the spent fuel racks capability to absorb neutrons.

Result: Unreviewed safety question does not exist. Tech Specs require that sealed sources be tested for leakage prior to use and prior to shipment. This vendor procedure requires the source to be leak tested before and after use. Therefore, there is not a chande in the margin of safety. This testing is passive in nature, by measuring the neutron attenuatiori of the fuel racks and will not affect the integrity of spent fuel ,

racks and has no interface with the fuel pool cooling system. It has no impact on any structure, system, or l

component. This testing moves a testing tool into and out of spent fuel rack locations. With the absence of l

fuel manipulation, the probability of a Refueling Accident is unaffected.

Safety Evaluation Number: SE-96-IS3 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCTS 0210-03, Rev. 0

Title:

Station Battery Individual Cell Discharge / Charge J

Attachment A, SVP-98-113, Page 114 of 153 J

l

Description:

_ This new procedure controls the on-line charging and discharging of individual battery cells and details the use of a test device. The charge / discharge unit can be connected to a battery bank without introducing a ground onto the DC system. The procedure is written to provide from 2.33 to 2.5 volts of. '

charging potential, draw the appropriate current for the rated time period, and details the current and time to verify that every station battery will meet its design capacity. This procedure provides the option to charging only single cells, instead of the entire battery and also, to perform discharge tests one cell at a time. Currently, most of the non-safety-related batteries are never tested. This test data is needed to determine replacement schedules.

Result: Unreviewed safety question does not exist. The change has no effect on loads connected to the battery. For the batteries themselves, this change will bring improvements to our ability to predict and prevent cell failures. The single cell charges will reduce the number of total battery equalize charges, thereby increasing both the life of the weak cell, and the life of the strong cells by not burning them out.

This procedure adds no new types of accidents. There is no change in the output terminal voltage. If an accident should occur during the charge, there would be added capacity available from the battery, thereby

- reducing accident consequences.

i Safety Evaluation Number: SE-96-154 Type of Safety Evaluation: Engineering Request Evaluation Reference Number: ER9604988

]

Title:

High Radiation Sampling System Supports I i

Description:

This evaluation adds two new supports on High Radiation Sampling System tubing line in i the 2A RHR corner room in accordance with this Engineering Request. The tubing is poorly supported in its current configuration. This ER requests support details so that the wire can be removed permanently.

Result: Unreviewed safety question does not exist. Plant operation is not affected by the installation of two new supports on the HRSS tubing line. 'Ihe supports are passive and not part of the In-Service Inspection Program. The supports will ensure that the system meets design codes and maintains structural ,

integrity. Equipment failures are unaffected by this change. Loads on the valve (the anchor point for this  !

system) will be reduced and thus will be beneficial for continued operation of the system. The change does  !

not create the possibility of an accident or malfunction of a type different from those evaluated in the 1 UFSAR. j j

i Safety Evaluation Number: SE-96-155 Type of Safety Evaluation: Design Change

]

Evaluation Reference Number: DCP 9600239

Title:

2D1 High Pressure Feedwater Heater Extraction Steam Inlet Nozzle I

Description:

This change repaired the 2D1 HP Feedwater Heater Steam Inlet Nozzle by forming a - i second nozzle around the outside of the first. The repair consisted of welding a stainless steel sheet and a carbon steel shell to the existing nozzle. The new nozzle materials and welds are designed to be pressure boundaries capable of sustaining all original design loads. These repairs were necessary due to wall thinning of the existing nozzle caused by flow accelerated corrosion. The thinning was discovered during ultrasonic inspection.'  ;

Attachment A.SVP-98 ll3,Page 115 of153 .

l 1

I l.

t Result:- Unreviewed safety question does n'ot exist. There are no new failure modes associated with this

. change. The new repair installs a new stainless steel impingement plate and a carbon steel overlay that will enable the nozzle to perform its design function for the remainder ofplant life. Installation of the change restores the Feedwater Heaters' original design margins. The repair conforms to the original code of construction and has been analyzed for increased loading. Thus, there is no increase in the probability of -

malfunction of equipment important to safety. The new repair will not change or alter the operation of the Feedwater heaters or the Feedwater System.

Safety Evaluation Number: SE-96-156 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600242 i

Title:

2B3 Low Pressure Feedwater Heater East Extraction Steam Inlet Nozzle

! j

Description:

This change repaired the 2B3 LP Feedwater Heater East Extraction Steam Inlet Nozzle by forming a second nozzle around the outside of the first. The repair consisted of welding a stainless steel sheet and a carbon steel shell to the existing nozzle. The new nozzle materials and welds are designed to be .

pressure boundaries capable of sustaining all original design loads. These repairs were necessary due to wall thinning of the existing nozzle caused by flow accelerated corrosion. The thinning was discovered l during ultrasonic inspection.

l Result: Unreviewed safety question does not exist. There are no new failure modes associated with this

! i change. The new repair installs a new stainless steel impingement plate and a carbon steel overlay that will '

enable the nozzle to perform its design function for the remainder of plant life. Installation of the change restores the Feedwater Heaters' original design margins. The repair conforms to the original code of construction and has been analyzed for increased loading. Thus, there is no increase in the probability of malfunction of equipment important to safety. The new repair will not change or alter the operation of the j Feedwater heaters or the Feedwater System.' '

l Safety Evaluation Number: SE-96-IS7 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP %00241 'i

Title:

2D1 High Pressure Feedwater Heater Shell 1

~

Description:

This change repaired the 2D1 HP Feedwater Heater Shell. The shell was repaired by welding a structural plate to the outside of the heater shell. A floating stainless steel liner conforming to the shape of the shell was placed between the original shell arid structural plate. The new shell repair materials l

' and welds'are designed to be pressure boundaries capable of sustaining all original design load::. Repairs

were necessary due to wall thinning on the shell side of the 2D1 Feedwater Heater caused by flow

~ accelerated corrosion. The thinning was discovered during ultrasonic inspection.~

Result: Unreviewed safety question does not exist. There are no new failure modes created. The new repair installed a new stainless steel impingement plate and a carbon steel overlay that enables the r

Feedwater heater to perform its design function for the remainder of plant life. Installation of this change restores the heaters original design margms and meets or exceeds the requirements of the original design

[ i l

Attachment A, SVP-98-113, Page 116 of 153 i

l s

and code of construction. No active components are added to the Feedwater system that would increase the consequences of a malfunction of equipment important to safety. The repair does not change or alter the operation of the Feedwater heaters or the Feedwater system.

Safety Evaluation Number: SE-96-158 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600244

Title:

2D2 High Pressure Feedwater Heater Extraction Steam Inlet Nozzle

Description:

This change repaired the 2D2 HP Feedwater Heater Extraction Steam Inlet Nozzle by forming a second nozzle around the outside of the first. The repair consisted of welding a stainless steel sheet and a carbon steel shell to the existing nozzle. The new nozzle materials and welds are designed to be i pressure boundaries capable of sustaining all original design loads. These repairs were necessary due to wall chinning of the existing nozzle caused by flow accelerated corrosion. The thinning was discovered during ultrasonic inspection.

l Result: Unreviewed safety question does not exist. There are no new failure modes associated with this change. The new repair installs a new stainless steel impingement plate and a carbon steel overlay that will enable the nozzle to perform its design function for the remainder of plant life. Installation of the change restores the Feedwater Heaters' original design margins. The repair conforms to the original code of construction and has been analyzed for increased loading. Thus, there is no increase in the probability of malfunction of equipment important to safety. The new repair will not change or alter the operation of the i Feedwater heaters or the Feedwater System.

Safety Evaluation Number: SE-96-163 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600245 l

l

Title:

2B2 Low Pressure Feedwater Heater East Extraction Steam Inlet Nozzle l Deseriptica: This change repaired the 2B2 LP Feedwater Heater East Extraction Steam Inlet Nozzle by l forming a second nozzle around the outside of the first. The repair consisted of welding a stainless steel sheet and a carbon steel shell to the existing nozzle. The new nozzle materials and welds are designed to be pressure boundaries capable of sustaining all original design loads. These repairs were necessary due to wall thinning of the existing nozzle caused by flow accelerated corrosion. The thinning was discovered during ultrasonic inspection.

< Result: Unreviewed safety question does not exist. There are no new failure modes associated with this

[ change. The new repair installs a new stainless steel impingement plate and a carbon steel overlay that will l

enable the nozzle to perform its design function for the remainder of plant life. Installation of the change restores the Feedwater Heaters' original design margins. The repair conforms to the original code of construction and has been analyzed for increased loading. Thus, there is no increase in the probability of malfunction of equipment important to safe'.y. The new repair will not change or alter the operation of the Feedwater heaters or the Feedwater System.

Attachment A,6VP-98-113, Page 117 of 153

i i

Safety Evaluation Number: SE-96-164 Type of Safety Evaluation: Design Change l Evaluation Reference Number: DCP 9600243

Title:

2C2 Low Pressure Feedwater Heater Shell

Description:

This change repaired the 2C2 LP Feedwater lleater Shell. The shell was repaired by welding a structural plate to the outside of the heater shell. A floating stainless steel liner conforming to the shape of the shell was placed between the original shell and structural plate. The new shell repair materials and welds are designed to be pressure boundaries capable of sustaining all original design loads. Repairs were necessary due to wall thinning on the shell side of the 2C2 Feedwater Heater caused by flow accelerated corrosion. The thinning was discovered during ultrasonic inspection.

Result: Unreviewed safety question does not exist. There are no new failure modes created. The new I

repair installed a new stainless steel impingement plate and a carbon steel overlay that enables the Feedwater heater to perform its design function for the remainder of plant life. Installation of this change restores the heaters original design margins and meets or exceeds the requirements of the original design and code of construction. No active components are added to the Feedwater system that would increase the )

consequences of a malfunction of equipment important to safety. The repair does not change or alter the operation of the Feedwater heaters or the Feedwater system.

]

l 4

Safety Evaluation Number: SE-96-165 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600240

Title:

2C3 Low Pressure Feedwater Heater Shell

Description:

This change repaired the 2C3 LP Feedwater Heater Shell. The shell was repaired by welding a structural plate to the outside of the heater shell. A floating stainless steel liner conforming to the shape of the shell was placed between the original shell and structural plate. The new shell repair materials and welds are designed to be pressure boundaries capable of sustaining all original design bads. Repairs were necessary due to wall thinning on the shell side of the 2C3 Feedwater Heater caused by flow accelerated corrosion. The thinning was discovered during ultrasonic inspection.

Result: Unreviewed safety question does not exist. There are no new failure modes created. The new repair installed a new stainless steel impingement plate and a carbon steel overlay that enables the Feedwater heater to perform its design function for the remainder of plant life. Installation of this change restores the heaters original design margins and meets or exceeds the requirements of the original design and code of construction. No active components are added to the Feedwater system that would increase the consequences of a malfunction of equipment important to safety. The repair does not change or alter the operation of the Feedwater heaters or the Feedwater system.

Safety Evaluation Number: SE-96-166 Type of Safety Evaluation: Design Change l

Evaluation Reference Number: DCP 9600429, Temporary Alteration 97-2-28 I

l Attachment A, SVP-98-il3, Page 118 of 153 l

l

Title:

Unit 2 Turbine Building Closed Cooling Water (TBCCW) Temperature Control Valve Bypass Valve

Description:

This temporary alteration isolated the Unit 2 TBCCW temperature control valve bypass valve in order to repair / replace the valve. Service water flow is provided by connecting hoses to the service water d'scharge side of the TBCCW heat exchanger and routing the hoses directly to the Unit 2 and/or Unit 1 king hole. The valve was isolated by closing the TBCCW heatexchanger service water discharge isolation valve.

Result: Unreviewed safety question does not exist.

Safety Evaluation Number: SE-96-167 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temporary Alteration # 96-1-147

Title:

Temporary Power From MCC 42R-2-1 to Portable Welding Machine

Description:

This change provides temporary power from MCC 42R-2-1 to the portable welding machine outside the Reactor Water Cleanup (RWCU) pump room. The power feed is needed for l A RWCU pump repair work, and the reactor building welding receptacles are de-energized pending installation of a permanent design change.

Result: Unreviewed safety question does not exist. This change will not affect plant operation because 42R-2-1 is not used to feed any plant equipment and this feed to the portable welding machine will not be used to feed plant equipment. Therefore, plant operation will not be affected. There are no new failure modes introduced. Combustible loading is affected by the additional cable in the reactor building, however the change is within allowances included in the fire hazards analysis for transient combustibles. Therefore, no accident or anticipated transient described in the SAR is affected.

Safety Evaluation Number: SE-96-168, R1 Type of Safety Evaluation: Design Change l

Evaluation Reference Number: DCP 9600249, DCN 001422E-01 l

Title:

HPCI Banana Jack Adapters

Description:

This change replaced the existing electrical terminal screws for various HPCI system circuits with banana jack adapters in the Main Control Room panel 902-3 and in local terminal box EH at the U2 HPCI turbine. The bananajacks were installed to provide a methodology for the attachment of test or monitoring equipment that will preclude the issuance of a Temporary Alteration each time the equipment is connected to the circuitry. This will also facilitate the trending of system performance parameters by providing access to various electrical signals with the banana jack adapters.

Result: Unreviewed safety question does not exist. Plant or HPCI system operation will not change. The adapters do not change the control or instrumentation wiring scheme for any circuit. Therefore, the HPCI system functional ability to perform the intended design function is unaffected. .All failure modes remain the same as previously analyzed. No new modes of failure are introduced. The adapters perform the same Attachment A, SVP-98-il3, Page 119 of 153

(

1 function as the screw being replaced. All required HPCI system safety responses to accident scenarios remain the same. No probability of equipment malfunction will be increased.

1

\

Safety Evaluation. Number:

SE-96-170 Type of Safety Evaluation: Modification l Evaluation Reference Number: M04-2-85-032B/C, Mechani:al Portion, Addendum #1

{

Title:

Drywellfforus Differential Pressure Compressors

Description:

This evaluation addresses improvements to the installation of the Drywell/forus Difrerential Pressure Compressor identified during the modification testing for the Unit 1 Modification.

While those impavements involved minor reroutes of field installed tubing, all of the improvements were on the 1-8701B tubing. For this Unit 2 mod, the tubing supports are for both the 2-8701 A (primary l

compressor) and 2-8701B (back-up compressor) and some additional tubing is being replaced for the 2-8701A.

l Result: Unreviewed safety question does not exist. The changes improve the reliability of the system. ,

There are no system interfaces with this work so there is no increase in the probability of a malfunction of l any other system. The consequences of a malfunction of eguipment important to safety are actually reduced i by providing a more reliable system. Here are no new functions introduced by this work.

l Safety Evaluation Number: SE-96-172 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600371

Title:

Main Control Room Floor, Personal Computers, Communication l

Description:

This change installed a new raised floor in the Main Control Room and provided new desks I to replace the existing Center Desk, NSO Desks, and Unit Supervisors Desks. The equipment presently mounted on and in the existing desks was relocated to the new desks and under the raised floor. All existing computer equipment and communication equipment functionality remained the same. The new floor, desks, and cabinets are non-safety-related but were seismically mounted. His change also provided for rewiring of computer and communcation equipment at the new location and provided layout and wiring details for the new power receptacles in the main control room. Installation work was completed in three different

phases to minimize the impact on plant operations. This change enhances supervisory overview and j improves operators ability to monitor indications, obtain and use reference materials.

l Result: Unreviewed safety question does not exist. This change has no interactions with any safety-related structure, system, or component. The new raised Main Control Room floor and new desk and cabinets are passive components and have no effect on plant operation during any operating mode. Equipment failures l

~

are not impacted. The intended design function of the Main Control Room remains unchanged and equipment used to achieve the intended design function operates in the same manner as the originally installed equipment. No new failure modes are created. A calculation was performed to ensure that the tie down details would prevem a collapse of the floor or tipping or sliding of desks and cabinets during an SSE event. The relocation of wiring to new equipment will not impact the operation of the equipment nor affect the ability of the operators to control the Units. Consequences of a malfunction of equipment important to Attachment A, SVP-98-ll3, Page 120 of 153 e

e

1 safety are not increased since the new raised flooring, desks and cabinets are designed to function and 1 perform better than the existing furnishings.

Safety Evaluation Number: SE-96-174 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600363, DCN 001425M

Title:

2A Reactor Recirculation Pump

Description:

This evaluation addresses the changes to the 2-0202-A reactor recirculation purap which include removing the furmanite clamp on the pump flange, upgrading the pump cover assembly and rotating l element, removing the pump suction splitters if unacceptable indications are found, and adding pipe flanges to the RBCCW cooling water lines associated with the pump's heat exchanger and unions being replaced by couplings. These new parts have been reengineered through design and material upgrades to reduce the potential for thermal and weld related cracking. Lower carbon content materials are used to reduce the potential for intergranular stress corrosion cracking. Better gasket materials and improved cover fastening will minimize possible future leakage. A rigging plan was developed and an additional safety evaluation written (see SE-97-013) for this part of the DCP.

Result: Unreviewed safety question does not exist. This change does not adversely impact plant operation.

Interactions with other plant components and systems are not changed. The removal of the pump flange furmanite clamp will eliminate a potential future operation concern ifleakage were to develop. The improved design and materials used in the new pump cover and rotating element will increase reliability of the pump by reducing the potential for thermal and weld induced cracking. The inspection and leakoff ports will not have any effect on plant operation and the reph. cement of the flanges with welded caps will reduce the possibility of failure of these components which would result in reactor coolant leakage. The elimination of the unions connecting the heat exchanger cooling lines to the pump will decrease the chance of these connections becoming damaged during seal maintenance. All of these improvements result in a lower potential for malfunction of these components.

Safety Evaluation Number: SE-96-175 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600221,9600261

Title:

Upgrade Reactor Recirculation Pump Suction Valves

Description:

These changes for the reactor recirculation pump suction valves eliminate the existing triple packing configuration and replaces it with a single live load configuration via a packing insert kit which consists of a packing insert, retainer, gland, gland flange, studs, nuts, bolts, lockwashers, and gasket. The packing was replaced with the station standard ARGO 5-ring live load packing sized for the new packing insert kit. The existing stem was replaced with a new identical stem as part of this retrofit. The yoke material was changed, the piping, temperature element, and sight glass for the stem leak-offline was eliminated, and bonnet vent line piping, valves, and sight glass were eliminated. The original three-stage packing design has exhibited a history ofleakage. The change to single packing configuration was based on various recommendations and results from industry / Comed experience.

l Attachment A, SVP-98-113, Page 121 of 153

Result: Unreviewed safety question does not exist. The modified valve has been verified to meet design requirements for structural integrity. These changes increase the reliability of the valve packing and strengthens the valve yoke thereby increasing seismic margin and improving the reliability of the valves to perform their design functions. Operation of these valves in response to plant condities remains within the design basis. The design change from triple to single packing provides an improved desy to reduce the probability of packing failure. This design change will not impact previously evaluated : gipment failures for motor-operated valves. Therefore, no new failure modes have been identified.

Safety Evaluation Number: SE-96-181 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600230

Title:

250V DC Battery Charger #2 Input and Output Cables

==

Description:==

This change was to update the input and output cables at 250 VDC battery charger #2. The existing cables from MCC 29-2 to the battery charger were abandoned in place and replaced by new cables.

This change was in response to a potential for periodically exceeding the 250 VDC battery charger #2 input and output feed cables rated continuous ampacity during plant operations. Such a condition, if allowed to persist, may shonen the cable life due to insulation degradation. A degraded voltage calculation identified that this charger feed cable current was higher than the cable ampacity based on charger loading: A calculation was performed to determine what cable should be used.

Result: Unreviewed safety question does not exist. The new cables installed function the same as the current cables, there fore new no failure modes are introduced. Due to the battery charger input cable upgrade the input vo?tage to battery charger during degraded voltage condition improves. The new cables will be serviceable for the . emainder of the original 40-year design life of the plant. The new power feed 1 eliminates concerns with potential cable degradation due to exceeding rated cable ampacity. Malfunction probability of this equipment should decrease because these changes increase the reliability of the equipment to function as designed.

Safety Evaluation Number: SE-96-182 Type of Safety Evaluation: Parts Evaluation Evaluation Reference Number: Parts Eval. #96-100-00

Title:

C Feedwater Heater Drain Normal Level Control Valve Internals

==

Description:==

This change involved replacing the existing unbalanced trim with balanced trim for the C ,

Feedwater Heater Drain Normal Level Control Valve. Balanced trim will minimize the potential for l actuator failure (broken bolts, yoke or frame) as a result of system pressure transients. The balanced design I provides equalization of pressure across the plug, thereby minimizing the " piston" effect created when pressure spikes below the plug.

Result: Unreviewed safety question does not exist. The valve will not operate differently. The use of  !

balanced trim will reduce system load for this valve, since it is essentially a throttling application. This will improve valve performance and reduce potential operator stress that has lead to bolt fatigue faGures in the i past. No new failure modes are created as a result of this change nor is the potential for increased failures. l The change does not affect the valves ability to control flow from the feedwater heaters. The new trim does .

l l

Attachment A, SVP-98-113, Page 122 of 153 I

1 s

not reduce system stability or flow capacity. As a result, the new trim cannot increase the potential for an accident.

. Safety Evaluation Number: SE-96-185 Type of Safety Evaluation: Work Request 1 Evaluation Reference Number: WR #960076281 '

Title:

Feedwater Level Control

Description:

The work request temporarily installed an electricaljumper to the electrical circuit for the solenoid valve that supplies air to the low flow feedwater regulating Feedwater Regulating Valve (FRV).

The purpose of this work request is to verify that the Feedwater Reg Valve goes to the fully closed position when the solenoid is energized. This test ensures that the feedwater flow into the reactor vessel during the runout-flow-control (ROFC) mode of operation is being controlled by the main feedwater regulating valves, via the runout-flow-control electrical circuit.

Result: Unreviewed safety question does not exist. Performance of this functional test will not impact the safe operation of the plant nor degrade the adequacy of system, structure, or component of the feedwater level control system, since: if performed with the reactor at power and reactor level controlled using the {

main FRV's, the low flow FRV will be isolated and not in-service; if performed with the reactor shutdown, this test will be performed with the low flow FRV isolated with reactor water level controlled by the CRD i and RWCU systems; the electrical circuit that is altered supplies electrical power to the solenoid valve directly and does not alter ROFC logic. Therefore, the installation of this electrical jumper does not introduce the possibility of effecting the control logic of the main FRV's, the CRD, RWCU, or any other plant systems. In addition, no other systems important to safety are affected by the performance of this procedure. No permanent changes are made to the feedwater level control system.

Safety Evaluation Number: SE-96-186 Type of Safety Evaluation: Interim Procedure

! Evaluation Reference Number: IP-96-0195 I

Title:

Control Room Ventilation Boundary Leaks

}

Description:

This interim test procedure temporarily restricted exhaust flow from the Train B Control Room HVAC Equipment Room. This was accomplished by placing a flat plate oflightweight material on .

the return air grille of the Train B HVAC Equipment Room duct. This increased the pressure in the room and allowed for leakage testing. This interim procedure supported identification of Control Room l Emergency Ventilation System (CREVS) boundary leaks. The Train B HVAC Equipment Room does not i currently have sufficient differential pressure to identify leaks.

Result: Unreviewed safety question does not exist. This installation was temporary, and the test could 1 have been terminated as directed by the Unit Supervisor. The installation of the flow restrictor does not l

introduce a new accident or malfunction because this evolution is effectively controlled with the interim '

procedure. The test has no impact on the accident analysis assumptions or initial conditions.

Safety Evaluation Number: SE-96-189 Type of Safety Evaluation: Design Change Attachment A, SVP-98-113, Page 123 of 153

Evaluation Reference Number: DCP 9600105

Title:

Reactor Feed Pump Motor Ventilation Fans & Ductwork

Description:

This change installed a cooling duct from the U2 Reactor Feed Pump Motor ventilation system to the Feedwater Regulating Valve Hydraulic Power Unit (HPU) and the modification of the cooling duct to the Feedwater Regulating Valve Actuators. The HPU is susceptible to high temperature alarms (160 degmen F) due to the lack of air circulation and high ambient temperatures in the vicinity of the equipment.

The installation of this new cooling duct provides adequate cooling air circulation to ensure the hydraulic oil i temperature remains below 145 degrees F, thus eliminating high temperature concerns.

Result: Unreviewed safety question does not exist. This change improves reliability of the Feedwater Regulating Valves and components by enhancing the heat dissipation capability of the components, thus j s

ensuring operating conditions within the vendor temperature recommendations for the Feedwater equipment. This change does not introduce any new failure modes or increase the probability of an equipment or component failure within the Feedwater system.

Safety Evaluation Number: SE-96-190 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9400082 (E04-2-94-113, Addendum 2) I

Title:

Main Steam Isolation Valves

Description:

This addendum revised the scope of the original design change from replacement of guide liners and actuator springs for all eight of the U2 MSIVs te: guide liners and actuator sprinFs were replaced for MSIVs 2-0203-2B and 2-0203-2C during the Q2R13 refueling outage. Actuator springs only will be replaced for MSIVs 2-0203-1B,2-0203-2A, and 2-0203-2D during the upcoming Q2R14 refueling outage.

This change installs new design guide liners for the affected MSIVs which incorporate an alternate method ,

of retaining and positioning the lower guide liner. The new upper and lower guide liners are mechanically I interlocked and a belleville spring between the seat ring bore and the lower liner prevents vertical movement of the lower line during plant operation. This addendum allows for closure of all work performed following installation of the Q2R14 changes. j Result: Unreviewed safety question does not exist. Implementation of this change increases the overall reliability of the MSIVs. A failure associated with the f.ew design guide liners would have the same effect as a failure of the existing liner fillet welds. The liners would not be held in position and may rotate

- resulting in a blockage of the flow port. 'Ihere is a significantly reduced likelihood of any such failure as compared to the existing welded design. No new failure modes are introduced as a result of the replacement of the actuator springs. The new actuator springs will produce at least the same thrust as the existing springs and therefore valve closure times will remain the same. This change has no impact on plant operation. This change does not alter any process parameters for the affected MSIVs. Consequences of an accident are not increased.

I f Safety Evaluation Number: SE-96-192 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 96000267 l-Attachment A, SVP-98-ll3, Page 124 of 153

n

Title:

Unit 2 Emergency Diesel Generator (EDG) >

Deseription: This change removed an unidentified control relay from Unit 2 Emergency Diesel

.' Generitor (EDG) circuitry discovered during a detailed system walkdown. This relay energizes on a start signal and has one normally open contact in the Start Failure (SF) relay circuit. This relay is only found on -

Unit 2. The relay was removed and wiring for the SF relay altered to match that of Unit 1 and 1/2. The

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terminal strip currently utilized for the wiring remained in the panel.

J Resmit: Unreviewed safety question does not exist. This change does not introduce any new failure modes.

All failure modes remain the same as previously analyzed. This change eliminates a potential failure of the SF relay logic. With the existing configuration, if the relay fails to energize, the SF relay would not pick up j on a failure of the diesel to reach 200 rpm in 15 seconds. Removing the relay prevents possible diesel damage during a failure to start. Removing the relay does not create a potential for a diesel failure. Since the removal of the relay and necessary wiring alteration will not affect the operation of the diesel during an accident, the consequences of a malfunction of equipment important to safety does not increase.

Safety Evaluation Number: SE-96-193 Type of Sciety Evaluation: Design Change Evaluation Reference Number: ' DCP 9600020,00021

Title:

' lix Recirculation Pump Discharge Valves l '

Description:

These changes for the reactor recirculation pump discharge valves eliminate the existing

~ triple packing configuration and replaces it with a single live load configuration via a packing insert kit

- which consists of a packing insert, retainer, gland, gland flange, studs, nuts bolts, lockwashers, and gasket.

The packing was replaced with the station standard ARGO 5-ring live load packing sized for the new packing insert kit. The overall length of the stem has been reduced. The yoke material was changed, the

. piping, temperature eternent, and sight glass for the stem leak-offline was eliminated,'and bonnet vent line ,

piping, valves, and sight glass were eliminated. The original three-stage' packing design has exhibited a ' .i history ofleakage. The change to single packing configuration was based on various recommendations and -

results from industry / Comed experience.

Result: Unreviewed safety question does not exist. The modified valve has been verified to meet design l

~ requirements for structural integrity. %ese changes increase the reliability of'he valve packing and i strengthens the valve yoke thereby increasing seismic margin and improving the reliability of the valves to perform their design functions. Operation of these valves in rerponse to plant conditions remains within the ,

design basis. The design change from triple to single packing provides an improved design to reduce the l

probability of packing failure. This design change will not impact previously evaluated equipment failures -!

for motor-operated valves. Therefore, no new failure modes have been identified.  !

-l Safety Evaluation Number: SE-96-196 Type of Safety Evaluation: Design Change  ;

Evaluation Reference Number: DCP 9600046 1

Title:

Replacement of Two Pressure Tap Valves,2-1201-120,2-1201-121 i

Attachment A, SVP-98-113, Page 125 of 153 l

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Description:

The existing defective valves,2-1201-120 & 121, were replaced with Conval globe type valves which are more reliable and maintainable. These valves were replaced due to valve leakage problems. The new Conval valves have an adjustable type packing, which will minimize leakage problems.

They are of the same manufacturer as other valves in the system.

Result: Unreviewed safety question does not exist. The replacement of these valves have no effect on system reliability or plant operations. These valves are intended for use during Local Leak Rate Testing.

The replacement valves are of the same basic type, and their function within the system remains the same as the original design intended. The difference is that the valves have an adjustable packing that allows maintenance activities to be easily performed to reduce the number of future problems. The new valves are at least as reliable as the ones replaced. The basic function of the Reactor Water Cleanup system is not impacted due to this change. The valves and piping are fully qualified and suitable for the application. The l possibility of an accident or a malfunction of a different type is not increased.

Safety Evaluation Number: SE-96-197 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600116

Title:

Extraction Steam Pipeway Electrical Penctration 1

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Description:

This change provided the design documents for the installation of a new electrical penetration in the Unit Two extraction steam pipeway roomi This penetration was used for routing new feed cables between Switchgear 29-2 and MCC 29-2. The scope of this design is to install a new 5" conduit sleeve and seal it with no cables passing through. The remaining cable work was done under a separate

exempt change during Q2R14. The new penetration was scaled to provide a fire barrier. This change was t

issued in response to a potential for periodically exceeding MCC 29-2 feed cable rated ampacity during 7 plant operations. Such a condition, if allowed to persist, may shorten the cable life due to insulation j degradation. J Result: Unreviewed safety question does not exist. This change does not affect equipm:nt failures or introduce any new failure modes. The consequences of a malfunction of the new penetration is the same as existing equivalent penetrations. Consequences of a failure of the floor also remain the same. Penetrations of this natum are common throughout the plant, and provided they are installed to approved design standards, no adverse impact on plant systems is created. The addition of this penetration in no way creates the possibility of an accident or malfunction of a type different than those already evaluated.

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- Safety Evaluation Number: SE-97-002 Type of Safety Evaluation: Engineering Review Evaluation Reference Number: ER9605992 t

Title:

Installation of Supports for U2 HRSS

Description:

Install new supports on High Radiation Sampling System tubing line and a copper air line to valve XSV 2-8941-705 in the 2A Core Spray Room. Tubing is poorly supported in its current configuration, but is not required to meet seismic category 1, l

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. i Attachment A, SVP-98-ll3, Page 126 of 153 1 1

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Result: Unreviewed safety question does not exist. The new supports improve structural integrity and lessen the loads on valve XCV 2-8941-705, the safoty related boundary.

Safety Evaluation Number: SE-97-003 Type of Safety Evaluation:

Evaluation Reference Number: DCP. 960280

Title:

Modincation to replace Square D 0013 series pressure switches with series 9012 pressure switches in the HPCI system.

Description:

The original model of the pressure switches are no longer supplied by the manufacturer.

The following switches will be replacei PS-001-2341-1, PS-001-2341-2, PS-001-2341-3, PS-001-2341-4, PS-001-2341-5, PS-001-2341-6, PS-0J1-2341-7, PS-001-2341-8, PS-001-2341-10, PS-001-2341-11, PS-001-2341-12, PS-001-2341-16, PS-0')l-2341-20. By using these new switches, maintenance of the system will be enhanced through spare parts availability.

Result: Unreviewed safety question does not exist. The new pressure switches are set to the same trip values as the old switches. There are no changes to the way that the MPCI system will function as a consequence of the switch replacements. The replacement switches connect to HPCI turbine oil lines. The switches are procured class IE and are suitable replacements. The change has no impact on the probability of a LOCA or other analyzed accident.

Safety Evaluation Number: SE-97-005 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9400096, DCN 001408E

Title:

Installation of Vibration Monitoring Cabinet on Elevation 595' of Reactor Building i

Description:

Install a vibration monitoring cabinet on elevation 595' of reactor building. Provide 120V -

AC power to the cabinet and install accelerometer and proximity probes on Unit 2 reactor recirculating pumps and motors. Install panels in drywell to mount the probe tran_:mitters, and route cables from transmitter panels through drywell penetration to the monitoring equipment. This equipment is being installed to record the change in vibration over time for Unit 2 reactor recirculation pumps and motors.

Result: Umeviewed safety question does not exist. This system is strictly a monitoring system and does l not affect the safety-related function of any existing equipment. The safety-related function of the l penetration remains as previously analyzed.

l Safety Evaluation Number: SE-97-007 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9500060

Title:

ECCS Emergency Room Coolers Attachment A. SVP-98-113, Page 127 of 153

Description:

This design change installs four venturi type flow indicators on the Service Water return l

piping from the RHR and Core Spray ECCS Emergency Room Coolers. The flow indication is local only and climinates the need for periodic ultrasonic flow testing.

Result: Unreviewed safety question does not exist. The flow indication will not affect performance of the room coolers, it simply provides a more accurate method of trending flow through the coolers.

l Safety Evaluation Number: SE-97-008 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCAP 1500-01

Title:

Administrative Requirements for Fire Protection

Description:

Changes made are to methods used for controlling the fire protection program. Changes include removing the requirement for declaring a firs detection system inoperable if two adjacent smoke

detectors are inoperable, deleting Attachment M, changing surveillance requirements to coincide with E

manufacturer's recommenditions and code requirements and changes to implement corperate guidelines governing Appendix R Em< rgency Lighting. Also changed is compensatory measures required when a fire barrier is declared inoperable.

i Result: Unreviewed safety quection does not exist. Changes being made do not remove any requirements  !

controlling the methods of compliance for the fire protection program and are only to the program itself.

Changes to the procedure will not cause any change in any areas, systems, or programs desiped to reduce the likelihood of a fire occurring and will have no impact on any potential fire initiators. Changes will not affect any equipment in any manner that would cause it to not function as designed during a fire.

Safety Evaluation Number: SE-97-009 Type of Safety Evaluation: Work Request Evaluation Reference Number: NWR 960070156

Title:

Allow Maintenance on Valve 2-2301-6

Description:

The 2-2301-6 valve taken out-of-service in the closed position to allow maintenance to be  ;

performed on it. HPCI pump suction switched to the torus while the 2-2301-6 valve is closed. Test loop on  ;

HPCI not available while torus suction valves are open. QCEMS 250-1 performed.

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Result: Unreviewed safety question does not exist. This change to operation of the 2-2301-6 valve is accounted for in the UFSAR where it states that HPCI Pump suction can be switched to the torus manually by the operator With the 2-2301-6 valve OOS closed, the HPCI system will operate to mitigate the consequences of an accident as assumed by the UFSAR. '

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! Safety Evaluation Number: SE-97-011 Type of Safety Evaluation: Master Equipment List l l

Evaluation Reference Number: DCR 970002 i .

Title:

. CRD Thermal Sleeves Safety Classification -

Attachment A, SVP-98-113, Page 128 of 153 l

Description:

A change has been made to the Master Equipment List (MEL) to reflect the correct safety classification for the thermal sleeves. The MEL incorrectly identifies these CRD thermal sleeves as safety-related components. Per ER9606425, these sleeves are classified as non-safety-related.

' Result

Unreviewed safety question does not exist. No physical changes to the thermal sleeves have taken place. Changing the safety classification does not change the thermal sleeve's function. No new interactions with other systems or the CRD system have been created as a result of this classification.

' Safety Evaluation Nu'mber: SE-97-013 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600363

Title:

Heavy Load Assessment (Rigging Plan)

Description:

This evaluation was for the physical removal and reinstallation of the various pump assembly components required to accomplish the work activity associated with this DCP (rigging of heavy loads).' This DCP removed the furmanite clamp and replaced the cover and internals of the 2A reactor recirculation pump (see SE-96-174). No permanent plant changes occurred. The platform installed inside the drywell was removed, the 2A reactor recirculation system configuration was restored and concrete shield plugs were reinstalled all prior to Unit 2 startup. Also, the existing transition piece in the drywell monorail was reinstalled. NUREG-0612 describes various alternative approaches which provide acceptable measures for the control of heavy loads. The objec:ives of these guidelines are to ensure that either the potential for a load drop is extremely small or por '.ste a load drop. The requirements of this NUREG apply since the weight of these components exceed a heavy load as described in this NUREG. A detailed plan was adhered to for this part of the DCP to which this evaluation applies.

Result: Unreviewed safety question does not exist. A designed based LOCA is not a concern in terms of probability, consequences or malfunctions during this evolution since the vessel will be defueled and the fuel pool gates between the cavity and pool will be closed when the pump motor is moved inside the - .

drywell. A drop of the motor resulting in damage to the recirculation line and draining the vessel is unlikely since all rigging equipment will be designed and/or rated for the motor weight, the drywell monorail system is specifically designed to move the motor, load tests and inspections of the rigging equipment and monorail i performed in accordance with the applicable ANSI standards and Station procedures, and the motor will j only be in the vicinity of the recirculation system for a short time. Also, the 2A reactor recirculation system  !

has been evaluated for a seismic event in its altered configuration. Therefore, during or following a seismic I event, structural integrity of the piping system will be maintained.

Safety Evaluation Number: SE-97-014 Type of Safety Evalu'ation: Procedure Change Evaluation Reference Number: QCTS 0360 01, Rev. 4

Title:

Feedwater Level Control 4

Description:

This procedure revision deletes test sections per PFC-2684 that evaluate Feedwater Regulating Valve (FRV) response to small (3 & 5%) and large (7 and 10%) feedwater flow rate changes.

Per G.E., these changes are not required to verify the proper operation of the FRV's. Also, this revision Attachment A, SVP-98-ll3, Page 129 of 153

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incorporates procedural improvements discovered during Q1R14 startup testing per PFC-2654 and PFC-2678. l Result: Unreviewed safety question does not exist. Level set point change tests were previously specified in the UFSAR as being performed during initial startup test program. As described in UFSAR section 7.7.5.2.3, feedwater level control system is independent of the level scram function. Steady state system performance data are to be collected at an increment of approximately 2% rated power in conjunction with j l

normal plant power ascension. Steam flow and feedwater flow summer gains will be evaluated and adjusted  !

if necessary. All required transient signals are installed and removed in accordance with nuclear work request package.

Safety Evaluation Number: SE-97-015 Type of Safety Evaluation: Nuclear Work Request l Evaluation Reference Number: NWR 960018018 i

Title:

GERIS Inspection of U2 RPV Shell Welds i

Description:

RPV shell welds inspected (circumferential and longitudinal) using GERIS 2000 ID manipulator and UT system to meet the requirements of 10 CFR 50.55a and ASME Boiler and Pressure {

l Vessel Code Section XI,1992 Edition through 1993 Addenda.

i Result: Unreviewed safety question does not exist. The equipment does not have an active interrelationship with the functioning of either the primary system or any other safety system. The principle accident of concern associated with the GERIS equipment is that of the potential for a drop of the equipment over the open core. During Q2R14, the equipment was placed into the vessel after the reactor had been defueled and was remove 3 before refueling commenced; therefore, the concern of dropping the manipulator onto fuel was eliminated. There is nothing associated with this equipment which could create a new or different

" accident".

Safety Evaluation Number: SE-97-017 Type of Safety Evaluation: Engineering Request Evaluation Reference Number: ER9604505

Title:

Packing Configuration Change for Main Steamline Drain Isolation Valves

Description:

Packing configuration of main steam drain isolation valves 1(2)-0220-1 and 1(2)-0220-2 changed to a live load configuration.

Result: Unreviewed safety question does not exist. Valve stroke time was measured after live loading to verify that valves meet Tech Spec and UFSAR requirements for closure time. Local Leak Rate Test was performed to verify valve leakage is acceptable. The change to live loading does not have any adverse effects on the valve function. It is an improvement to minimize packing leakage.

l l Safety Evaluation Number: SE-97-018 Type of Safety Evaluation: Procedure Change l

Evaluation Reference Number: QOS 0020-03 Attachment A, SVP-98-ll3, Page 130 of 153

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Title:

Interim Flood Protection Surveillance J

Description:

Revises surveillance frequency in procedure for Emergency Core Cooling System (ECCS) corner room flood protection from once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for flood protection'and twice per shift for flooding to once per shiR to reduce administrative burden and personnel exposure due to increased frequency testing.

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' Result: Unreviewed safety question does not exist. The change of surveillance frequency does not impact the operation of any equipment. He shiftly verification of plug installation was established to coincide with routine operator surveillance of these plant areas. He reduced frequency of monitoring for corner room flooding will not affect any of the initiators of the accident and has no impact on operation of equipment important to safety.

Safety Evaluation Number: SE-97-019 Type of Safety Evaluation: . Interim Procedure Evaluation Reference Number: 97-0010

Title:

HPCI Subsystem Interlock

Description:

Interim procedure #97-0010 was performed to test the HCPI subsystem interlock which runs the motor speed changer to the high speed stop upon a high drywell pressure initiation signal, nis q interlock had not been tested and is required to be tested per Technical Specification 4.2.B.2. >

Result: Unreviewed safety question does not exist. This procedure makes HPCI subsystem unavailable and then tests a previously untested contact and then restores the system to operable and available status. -)'

- Having HPCI subsystem inoperable and unavailable has already been considered in Tech Specs.' It does not affec't the initiators of this accident. It only tests a function of the HPCI subsystem logic.

Safety Evaluation Number: SE-97-020 Type of Safety Evaluation: Interim Procedures .

- Evaluation Reference Number: 97-001'1,97-0012

Title:

Control Room Train A HVAC Air Distribution Test  !

Description:

These interim test procedures obtained air distribution data with Train A (IP-97-Oll) or  !

Train B (IP-97 012) Control Room HVAC operating in the Isolation and Pressurization Mode and Normal Mode. Quad Cities has committed to pressurizin'g the entire Control Room Emergency Zone to 1/8" W.G.

De data obtained is necessary for engineering to evaluate and establish system balancing criteria. i

- Resmit: Umeviewed safety question does not exist. During performance of this test, the CREVS air balance is affectr.l. Prior to initiation of this test, the applicable LCO was invoked since control room positive pressure may drop below the 1/8" W.G. with respect to adjacent areas. These tests do not affect the manual or automatic isolation of Control Room with respect to a toxic gas accident.

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Safety Evaluation Number: . SE-97-021 Type of Safety Evaluation: Technical Specification

- Basis Change

Attachment A, SVP-98-113, Page 131 of 153 p -

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1 Evaluation Reference Number: Surveillance Requir ment 4.8.D.5.c I

Title:

Control Room HVAC Tech Spec Surveillance i

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Description:

Supports change of Technical Specification (TS) bases to c;arify the TS surveillance requirement for the control room emergency ventilation (CREV) system with respcet to pressurization requirement. TS surveillance 4.8.D.S.c requires, every 18 months, to demonstrate the system is capable of pressurizing the control room to 1/8" water gauge (W.G.) relative to adjacent areas. The TS basis change submitted herein clarifies that the control room is required to be pressurized with respect to potentially contaminated adjacent' areas. A positive pressure does not need to be maintained between the control room

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and an adjacent area that is inside the control room emergency zone, e.g., the cable spreadmg room. i i

Result: Unreviewed safety question does not exist. This change is required w clarify that the control room is required to be positive only to potentially contaminated areas. It does not affect any equipment failures. i It is a clarification which supports the UFSAR requirements for the CR HVAC system. 1 Safety Evaluation Number: SE-97-026 Type of Safety Evaluation: Design Change i

Evaluation Reference Number: DCP 9700047, DCP 9700048

Title:

Clean & Contaminated Condensate Transfer / Reactor Building i

Description:

Removes the 1 %" CCST piping above the refuel floor off ofline nos. l(2) 1920-3" that is used to hose down the dryer / separate pits for Units 1 and 2, remove the supports attached to the Reactor j Building siding, and repair the holes le a by the machine screws. The CCST piping is attached to the  :

Reactor Building siding which is part of secondary containment which inhibits design feature of the blowout I panels. Removal of the supports returns the siding and blowout panels to their original design condition.

Result: Unreviewed safety question does not exist. Piping does not provide a safety function and is only used to wash down the dryer / separator pit walls. 'Ihe machine screws used to attach the four supports to l Reactor Building siding were reinstalled and caulked to meet design basis requirements.

Safety Evaluation Number: SE-97-030 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700062

Title:

Reactor Feed Pump Lube Oil and Seal Cooler Piping  ;

Descripties: A carbon steel spacer with a gasket placed on each side of the spacer was placed between  ;

the mating surfaces of a raised face flange in the return oil piping from the radial bearings to the Lube Oil _.

Reservoir. One face of the flange is welded to hard pipe and the other face of the flange is welded to a braided steel flexible hose. The hose is too short to connect the flange faces welded to the pipe and hose.

The spacer allows flange faces to be bolted together without stretching the braided steel hose (which could cause failure or leakage of the hose over time).

Attachment A, SVP-98-113, Page 132 of 153 l

Result: Unreviewed safety question does not exist. Installation of the spacer will not change the function or .

operation of Reactor Feed Pump Lube Oil and Seal Cooler Piping. Spacer does not change flow requirements of the system since it has the same I.D. as the raised face flange. The installation does not create the possibility of coincident failures of any other equipment important to safety.

Safety Evaluation Number: SE-97-033 Type of Safety Evaluation: Procedure Evaluation Reference Number: QCTP 0950-01

Title:

Reactor Vessel, Nuclear Fuel

Description:

New procedure was created which directs the user on how to perform and obtain approval for a lost parts analysis for a specific lost part. There was no written guidance to do this. Since ATRIUM-9B fuel is being introduced into the reacter, a new generic lost parts analysis has been performed by Siemens that defines maximum changes in Critical Power Ratio (CPR) for both Siemens fuel and GE fuel which corresponds to given percentages of flow area blockage. The procedure directs the preparer of a specific lost parts analysis to use the results of the new generic lost parts analysis in determining safety margins.

Result: Unreviewed safety question does not exist. Procedure directs that if probability of an accident, transient, or malfunction of equipment important to safety could be increased, then a new safety evaluation MUST be performed, in addition to administrative requirements of approving the lost parts analysis. Use of this procedure to calculate the change in CPR as a function of flow blockage does not affect the consequences of equipment malfunctions, because these CPR changes are taken into account in determining the OLMCPR. Procedure requires that potential chemical interactions be evaluated such that there is no impact to any vessel components. This ensures that consequences of any malfunction of any equipment impcrtant to safety will not increase due to chemical degradation of other components.

Safety Evaluation Number: SE-97-034 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temp. Alt. No. 97-1-010 -

Title:

Control Room HVAC 'A' Train, Smoke Purge Damper

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Description:

Removes exhaust air discharge hood (gooseneck) and damper, %-5741-324A, with actuator from the ductwork to allow repairs to improve its' air sealing capability. Attached to the termination of the duct was a blank plate that is the same size as the duct opening including the flange which provided redundant ductwork isolation during repairs. The damper-to-plate swap takes no longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with continuous attendance until the plate is secured. 'A' Train of Control Room HVAC is out-of-service during installation and removal of Temp Alt. 1 Result: Unreviewed safety question does not exist. Installation of a blank plate on discharge ductwork does not increase the probability of a DBA LOCA. No equipment important to safety is adversely affected by this installation. j Safety Evaluation Number: SE-97-035 Type of Safety Evaluation: Temporary Alteration Attachment A, SVP-98-113, Page 133 of 153

i-Evaluation Referwaee Number: Temp. Alt. No. 97-2-014 -

Title:

Equipment Drain Check Valves '

Description:

This change removed core spray corner room equipment drain check valve 2-1399-129 (f RCIC and 2B core spray) at flanged connection and replaced it with a flange fitting with a hose routed to the reactor building floor drain sump. De pump seal leak off valve (2-'1402-39B) was closed and controlled with an out-of-service that states that this valve shall not be opened until this temp alt is removed. Pur was to facilitate draining of core spray and RCIC systems during Q2R14 outage.

Result: Unreviewed safety question does not exist. Operation of plant systems are unaffected by this alt. These valves are no longer relied upon for back flow prevention due to past problems. This temp alt remained in~ place only while the unit was shutdown and RCIC system was not required. ' Ali other lines connecting to this drain piping are closed pipes that will not permit back flow in the unlikely event of a flood in the reactor basement. The 2B Core Spray pump remains operable. Therefore, the safety impact of this temp alt is not significant.

Safety Evaluation Number: SE-97-036 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600210

Title:

Ul/2 Safe Shutdown Make-up Pump Motor Operated Valve

Description:

his change configures the control circuit wiring for SS MOV %-2901-7 to allow limit l

switches to remain functional within the circuit should the control circuit experience a hot short and is '

' required to ensure that torque and limit switch devices protect the MOV from over torque conditions -

postulated to occur from a " hot short"in the event of a fire. Response to NRC Information Notice 92-18.~

{

Result: Unreviewed safety question does not exist. His change decreases the impact of postulated control ,

circuit hot shorts on the MOV but does not change the operation / logic of the valve. There is no change in 1

. system interactions.

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- Safety Evaluation Number: ' SE-97-037 - Type of Safety Evaluation: Design Change

- Evaluation Reference Nuenber: , DCP 9600417 .

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Title:

Ul Safe Shutdown Make-Up Pump Motor Operated Valve

Description:

This change configures the control circuit wiring for SS MOV l-2901-8 to allow limit switches to remain functional within the circuit should the control circuit experience a hot short and is y . required to ensure that torque and limit switch devices protect the MOV from over torque conditions P , postulated te occur from a " hot short" in the event of a fire. Response to NRC Information Notice 92-18.

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Result: Unreviewed safety question does not exist. This change decreases the impact of postulated control circuit hot shorts on the MOV but does not change the operation / logic of the valve. There is no change in system interactions.

.~ j Attachment A, SVP-98-113, Page 134 of 153

Sbfety Evaluation Number: SE-97-038 Type of Safety Evaluation: Design Change

' Evaluation Reference Number: DCP 9400076

Title:

RHR Pump Minimum Flow Valves

Description:

This' change was to regear the existing Limitorque motor actuators MO 2-1001-18A(B).

Currently these valves are equipped with an Overall Ratio of 27.99. Implementation will result in an Overall Ratio of 47.85 for these valves. Increases the motor gearing capacity in the opening and closing -

' directions, which results in an improved thrust window for these valves. Upgraded spring packs provide

grease relieving capability to the spring packs which result in improved performance.

Result: Unreviewed safety question does not exist. Change does not alter the function of the Limitorque actuator, but increases the motor gearing capability of the actuator in the opening and closing directions.

This portion of the RHR system has no interaction with equipment initiating the anticipated transients as i described in the UFSAR.

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n. Q 1 Safety Evaluation Number: SE-97-041 Type of Safety Evaluation: Interim Proc $ dure UFSAR Revision Evaluation Reference Number: ' LP. 97-0028; UFSAR Tracking No. UFSAR-97-R5-038

Title:

Refuel Platform Failure During Fuel Modes

Description:

This change was to implement a procedure to allow an irradiated fuel assembly to be placed I in the fuel prep machine. The refuel platform was moved manually via handwheels versus using the normal I control station. No procedural guidance previously existed for this activity, given that the refueling platform cannot be moved normally. An Updated Final Safety Analysis Report (UFSAR) change has been )

prepared for Section 9.1.4.2.1.3, " Refuel Platform Instrumentation and Control" which states that the refueling platform can also be moved manually by use of handwheels on the bridge.

Result: Unreviewed safety question does not exist. Plant operation was unaffected. The limits of the -

refueling bridge are not changed, only the method of moving the platform is changed. Movement of the refueling platform does not affect the operation of any mitigating equipment.' Specifically, the grapple will not release while fuel is loaded, and the hoist brakes will remain set. The local cell indication is still available while moving the bridge manually. These safety functions prevent the assembly from falling during the performance of the procedure. No new failure modes are introduced by moving the bridge

. manually. The primary method of moving the bridge had failed. However, operation of the grapple is not impacted by moving the refueling platform manually and therefore, the probability of dropping the bundle is  !

not changed. The fuel assembly was not raised above the normal full-up position while performing the procedure. As a result, even while over the core, the assumptions of the accident are not c, hanged. Thus, the consequences of an accident are not increased.

. Safety Evaluation Number: SE-97-043 Type of Safety Evaluation: Procedure Change Attachment A, SVP-98-113, Page 135 of 153 i

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! Evaluation Reference Number: QCOP 1000-36

Title:

RHR Shutdown Cooling Suction Piping

Description:

This procedure change allowed partially draining the volume between valves 1001-50 and 1001-47 on the Shutdown Cooling Suction Header to provide a volume to accommodate any thermal expansion of the water trapped between the isolation valves without exceeding any allowable piping stresses i

as described in Generic Letter 96-06.

Result: Unreviewed safety question does not exist. By partially draining this volume, an air pocket is created that effectively functions as an accumulator and prevents any significant pressurization of the piping during thermal transients. No new failure modes are created nor does it affect RHR due to operating the plant with shutdown cooling suction header partially drained. The draining between these valves does not adversely affect the response of any plant equipment as described in the UFSAR. This activity does not increase the probability of a LOCA. The 1-1001-50 valve is closed in order to maintain Reactor Coolant pressure boundary. Maintaining this valve closed also maintains primary conteinment integrity as specified in Tech Specs. The 1-1001-47 valve is opened as part of the drain path; however, it would still close automatically if an unexpected decrease in reactor water inventory were to occur during performance of this procedure.

Safety Evaluation Number: SE-97-047 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600418

Title:

U2 Safe Shutdown Make-up Pump Motor Operated Valve Dese 1ption: This change configures the control circuit wiring for SS MOV 2-2901-8 to allow limit switches to remain functional within the circuit should the control circuit experience a hot short and is required to ensure that torque and limit switch devices protect the MOV from over torque conditions postulated to occur from a " hot short" in the event of a fire. Response to NRC Information Notice 92-18. .

Result: Unreviewed safety question does not exist. This change decreases the impact of postulated control circuit hot shorts on the MOV but does not change the operation / logic of the valve. There is no change in system interactions.

Safety Evaluation Number: SE-97-049 Type of Safety Evaluation: Technical Specification Basis Change l Evaluation Reference Number: Surveillance Requirement 4.9.E I

Title:

AC and DC Distribution I

Description:

This change clarifies the purpose of the electrical distribution weekly verification  ;

surveillance requirement as verifying that the electrical power distribution systems have the correct circuit  !'

breaker alignment. Alignments that are correct, but other than normal, are controlled by other Tech Specs or administrative controls.

Attachment A, SVP 98113, Page 136 of 153 I

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l Result: Unreviewed safety question does not exist. Equipment fa;tures and plant operation are unaffected by this change. The change clarifies the purpose of a surveillance requirement as verifying that A.C. and D.C. electrical power distribution systems are functioning properly with the correct circuit breaker alignment, not to verify that the alignment is a normal one. Therefore, no new failures modes are introduced by this change.

Safety Evaluation Number: SE-97-052 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700112

Title:

Unit 2 Bottom Head Drain Line ECP Flange

Description:

This change installed an electrochemical corrosion potential (ECP) flange in the bottom head drain line. The ECP probe and its' associated Data Acquisition System was encoded to determine the correct Hydrogen Water Chemistry of the vessel to mitigate the potential IGSCC in the lower plenum region.

Result: Unreviewed safety question does not exist. The ECP flange does not affect plant operation. The containment penetration through which the cables are routed were tested prior to and after insertion of the two Conax Feed through Modules. This ensures there is not a concern inside or outside containment. The connection into the Rx Building Sample Panel does not affect plant operation. The only other interaction with a system is if the probes, which are internal to the piping flange tee, become loose parts. This is highly unlikely that a part of this size could filter into a 2" drain line. However, ifit did become dislodged, it would become part of the flowpath which discharges into RWCU system which is non-safety-related. It would not affect the operation of RWCU and would not enter the reactor. The loose parts analysis determined that the pieces would end up in the demineralizer bede and stop at that point. This scenario is within UFSAR requirements.

Safety Evaluation Number: SE-97-054 Type of Safety Evaluation: Modification /UFSAR Change Evaluation Reference Number: M04-1-92-0288, UFSAR Change No. 97-7

Title:

Nitrogen Containment Atmospheric Dilution System l

Description:

This safety evaluation was sent in the October 1997 summary report under M04-2-92-028B. 1 The same evaluation applies to Unit 1, and the UFSAR change reflects the installation of the NCAD System for boti. . nits in Section 6.2.

Safety Evaluation Number: SE-97-056 Type of Safety Evaluation: Exempt Change  !

j Evaluation Reference Number: E04-1-95-018

Title:

RVLIS Backfill A-Loop and B-Loop Attachment A, SVP-98-113, Page 137 of 153

Description:

This modification improves RVLIS Backfill System filtration design. The existing single 7 -

micron filter in each loop was replaced with an 8 micron and 2 micron filter in series. This modification installs a parallel filter train and all associated components which allows maintenance to be performed on one train without removing the loop from service. in the past, the task of removing the affected loop out of service was manpower imensive and introduces a potential to induce a SCRAM on the Unit. The relocation

' of the locked metering valves from the existing riser to the parallel filter trains will facilitate maintenance. '

i Result: Unreviewed safety question does not exist. Modification is installed on non-safety-related portion of RVLIS Backfill System upstream of the check valves. Affected portion of RVLIS Backfill System is not 1 considered to initiate any of the design basis accidents discussed in the UFSAR. Therefore, probability of a l i

l design basis accident is not increased. Only function of RVLIS Backfill System is to provide a high quality l water to the RVLIS reference leg condensate pots to prevent buildup of non-condensable gases which is non-safety-related. RVLIS Backfill System is not used to mitigate consequences of any equipment failures.

' Safety Evaluation Number: SE-97-064 Type of Safety Evaluation: Design Change l Evaluation Reference Number: DCP 9700141, DCN 001497E i l- .

Title:

4 kV Circuit Breakers: MERLIN GERIN (MG) Model AMHG l

Description:

This change installs a mounting strap on the internally mounted upper and lower breaker auxiliary switches which functions to mechanically hold the auxiliary switch in place. The existing auxiliary switch and contact block mounting T-bolts will be utilized to position the block, thus maintaining correct mechanical configuration. The existing method of mounting the auxiliary switch within MG models AMHG has been found to be susceptible to cracking, causing it to perform inadequately under mechanical stress of normal breaker operation. Resulting mechanical failure of block mounting has l rendered the breakers auxiliary switch and contacts inoperable due to physical separation of the switch from its mounting. Installing a mounting strap prevents that separation.

Result: Unreviewed safety question does not exist. Operation of the circuit breaker is not changed by the l~ addition of a mounting strap. After addition of a strap, the auxiliary switch and respective breaker are less L susceptible to failure. The breaker will open and close as required to allow accident mitigation.

[.

. Safety Evaluation Number: SE-97-065 Type of Safety Evaluation: Design Change i

L Evaluation Reference Number: DCP 9600052 i i L

Title:

Condenser Low Vacuum SCRAM Switch Replacement l l '

Description:

. This change replaces the currently installed Barksdale pressure switches used for detection j ofloss of condenser vacuum, with new pressure switches manufactured by Static-O-Ring. The new j switches are physically larger than currently installed switches and therefore were mounted at different .i locations on the instrument rack. Associated process tubing and electrical conduit were modified l accordingly. The currently installed switches have a history of excessive setpoint drift. This change  !

satisfies the corrective actions associated with LER 2-95-003.

Attachment A, SVP-98 113, Page 138 of 153 I

Result:

Unreviewed safety question does not exist. This change involves only a replacement of currently installed condenser vacuum pressure switches with new switches by a different manufacturer. The replacement switches are more reliable than the current switches. Functional configuration of the system is unchanged. The new switches improve reliability of the RPS system which would result in a higher confidence level that the SCRAM will occur as designed to mitigate the Loss of Main Condenser Vacuum.

Probability of malfunction of equipment important to safety will decrease due to this change. The new pressure switches are less susceptible to setpoint drift.

Safety Evaluation Number: SE-97-066 Type of Safety Evaluation: Design Change Evaluation Refer (nee Number: DCP 9700006, DCN 001452M, Rev. O and Rev.1

Title:

Control Room Emergency Ventilation System

Description:

This design change provides details to improve the sealing capacity of the Control Room Emergency Zone by reducing air leakage through penetrations and other openings. Sealing requirements, including the type and application of acceptable sealants, are approved by this change. Also, repairs were made to the Train B air handling unit to keep the Emergency Zone from overheating. New sections of tubing are required in the evaporator due to leaks in the existing tubing.

Result: Unreviewed safety question does not exist. Sealing of the emergency zone will allow the System to meet the Technical Specification of providing a 1/8" water gauge differential pressure with respect to all adjacent areas. Repairs to the refrigeration unit of the Train B air handling unit will allow the Train B to cool the air in the Emergency Zone. All design requirements are maintained within plant specifications.

This modification helps to ensure that the calculated Control Room dose analysis is met.

Safety Evaluation Number: SE-97-070 Type of Safety Evaluation: Interim Procedure Evaluation Reference Number: Interim Procedure IP-970059, WR 970044347-01

Title:

2C MSIV Outboard Valve l

==

Description:==

Installs temporaryjumpers under above Interim Procedure to remove a Group 1 isolation that has resulted from removal of temperature switches in the Unit 2 MSIV circuits, which ordinarily detect Main Steam line breaks during umt operation. Thejumpers close the circuits that have been de-energized as a result of the switch removal. Re-energizing the circuits enables the Group I isolation to be removed and allows stroking and testing of the 2-0203-2C MSIV valve.

Result: Unreviewed safety question does not exist. Jumpers were installed temporarily and were removed from the circuits prior to core reload. The MSIV Outboard Valve is not required to be operable during modes 4 and 5 when fuel is removed from the vessel.

Safety Evaluation Number: SE-97-073 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600223 Attachment A, SVP-98113, Page 139 of 153 1

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Title:

Reactor Feed Pump Motor Ventilation Dampert

Description:

This design change replaces the existing Reactor Feed Pump Motor Vent Dampers with l , new dampers capable of withstanding the rough service duty conditions at the RFP Motor fan discharge.

L

' Automatic damper operation is restored. . At present, Operators must manually position the dampers as required prior to starting a vent fan. Existing fan dampers are not strong enough to sustain the air-flow I conditions at the outlet of the vent fan motor. The new dampers are fabricated from a heavier gauge of l steel.

Resmit: Unreviewed safety question does not exist.' Retuming the ventilation control system to automatic l operation will dramatically decrease the system response time when compared to the manual actions required at present to reposition dampers. Manual operation is still possible if automatic controls

malfunction. Failure modes for the new dampers are identical to failure modes for the old dampers.

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Safety Evaluation Number: SE-97-075 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700156, DCN 001505M

Title:

Constant Support

Description:

This design change replaced damaged ELCEN pipe suppon hardware with Grinnell hardware of the appropriate load bearing capacity. This restores the constant support to an appropriate design configuration. The existing threaded eye bolt was broken while attempting to adjust the spring. The..

. spring rebounded, resulting in damage to other support components. The original vendor for this support (ELCEN) is no longer in business. Therefore, a like-for-like replacement was not possible.

L Reemit: Unreviewed safety iuestion l does not exist. The design function of the constant support is not altered and there are no new interactions created with other systems, structures, or components. It has sufficient capacity to accommodate the original design loadings. Likelihood of a failure of the changed .

support is no more likely than a failure of the existing support.

Safety Evaluation Number: SE-97-076 : Type of Safety Evaluation: ' Design Change

Evaluation Reference Number: DCP 9700059, DCP 9700060, DCP 9700061

Title:

Unit 1/2 SSMP Throttle Test Valve, Unit 1(2) SSMP System Injection Valves i

Description:

This design change replaces the current spring pack assembly which has a maximum torque rating of 56 fi-lb with one that can generate a higher torque maximum of 90 ft-lb for the above listed valves, j

' This change is to allow the valves to close under worst case design dP conditions for the Safe Shutdown l l

} Makeup Pump System while not exceeding torque ratings of the motor generator or actuator. When l

- Limiting Component Analysis reports were received from various vendors, it had been identified that torque l

' to close the above listed valves against design conditions was greater than the maximum torque value for the motor operator.

i Attachment A, SVP-98-113, Page 140 of 153 l'

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} Result: Unreviewed safety question does not exist. The additional torque being applied as a result of the spring pack replacement will not increase the consequences of a malfunction of equipment important to l safety because it is within the allowable torque of the motor operator and does not exceed the capability of I the motor gear or actuator. Other characteristics of the valve are not affected.

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l Safety Evaluation Number: SE-97-080 Type of Safety Evaluation: Design Change l

{ Evaluation Reference Number: DCP 9700178, NWR No. 970052973,970053035 l

Title:

CRD Flow Control Valves and Valve Actuators

Description:

CRD Flow Control Valves and their actuators were replaced with new ones that provide for better control of flow in the Control Rod Drive Hydraulic System. New regulators were also installed for each actuator. Existing valves were not working properly and replacement parts were not available. New l

valves provide increased control of the hydraulic system and reduce the supply air pressure from 85 psig to 45 psig in order to accommodate the c. w air diaphragm actuators.

Result: Unreviewed safety question does not exist. Function of these valves is to control flow. They do not interact with any other equipment important to safety. Failure modes remain unchanged. Plant operation is not affected because the CRD pumps will continue to provide water to the HCUs. This design change may be used during all operating modes. No other plant equipment is affected by the change in valves and actuators. Operation of the hydraulic system will be better after the design change is installed because the valves have been evaluated to provide adequate flow control of the hydraulic system.

Safety Evaluation Number: SE-97-083 Type of Safety Evaluation: Interim Procedure Evaluation Reference Number: Interim Procedure # 97-0070

Title:

Reactor Core Isolation Cooling (RCIC) Steam Line Leak Check of 1-1301-16 Valve l Description; This interim procedure allows for a leak check of the reactor core isolation cooling inboard primary containment isolation valve (1-1301-16). The leak check will be performed by closing the 1-1301-16 valve and deenergizing its' breaker. The volume between the 1-1301-16 valve and the 1-1301-61 valve was then isolated and its' pressure monitored to determine the feasibility of performing an LLRT on the 1-1301-17 valve which needs to be performed after planned repairs are made to the seal ring of the valve.

Result: Unreviewed safety question does not exist. The normally open 1-1301-16 valve is closed and deenergized, therefore, RCIC system will not automatically initiate upon a low-low reactor level signal during this procedure. Plant operation with RCIC inoperable is allowed by Tech Specs. He system can still be manually initiated after the 1-1301-16 valve is renergized. Additionally, the HPCI subsystem will be l operable while this procedure is executed and provide backup for RCIC.

Safety Evaluation Number: SE-97-084 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700173, ECN 0015131 i

l Attachment A, SVP-98-113, Page 141 of 153 1

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Title:

Standby Liquid Control Tank Thermistor

Description:

The thermistor for the Standby Liquid Control Tank temperature controller is being replaced with a new model thermistor. This thermistor is installed in a thermowell in the side of the tank.

Currently, the thermistor and temperature controller are not properly matched which results in an incorrect i controller output which at 95 degrees Fahrenheit, reads 8 degrees low.

l

- Result: Unreviewed safety question does not exist. The replacement thermistor is from the same manufacturer as the original and is better suited for use with the installed temperature controller. It

! functions in the same manner as the currently installed thermistor. No new failure mechanisms are l introduced by this replacement. The thermistor and the temperature control loop of which it is a part serve

! ~no safety function.

i:

Safety Evaluation Number: SE-97-088 Type of Safety Evaluation: Design Change l'

Evaluation Reference Number: DCP %00346

Title:

HPCI and RCIC Drain and Vent Lines inside Unit 2 Drywell

Description:

. This evaluation addresses only Addendum 4 to this modification, whichjustifies leaving three hot, small-bore lines uninsulated in the Unit 2 Drywell. The additional heat load on the Unit 2 Drywell will be small, and the change will have a small effect on the Drywell temperature (less than 1

, degree Fahrenheit at normal operation). This additional heat load is acceptable and the Drywell temperature -

L should still be maintained at acceptable levels.

l Result: Unreviewed safety question does not exist. This change will not increase the chance of the failure L of these lines or any other piece of equipment inside the Unit 2 Drywell.' The drain and vent lines inside the drywell are not degraded by this design change. The removal ofinsulation from these lines do not adversely affect the integrity of these lines. A malfunction or failure of any ' quipment e have the same effects as before insulation was removed. Calcium silicate insulation will be added as needed to prevent local temperature effects. No other equipment is affected by not insulating these lines.

l Safety Evaluation Number: SE-97-090 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QGA 101, QGA 200, QGA 500-4

~

Title:

Standby Liquid Control, Primary Containment, Reactor Core

Description:

his change revises Emergency Operating Procedures to support the loading of Siemens

. fuel. The changed values provide the amount of boron that must be injected for cold and hot shutdown weights. These changes encompass both GE and Siemens fuel and the values given are conservative for both fuel types. These changes, therefore, apply to both units with either fuel type present.

Result: Unreviewed safety question does not exist. The changes to the EOP values do not affect probability of a malfunction of equipment important to safety. The values are provided to protect equipment when accident conditions and equipment failures are in excess of the design basis. The new values were obtained by using the physical and nuclear characteristics of the new Siemens fuel.' These new values cannot create i

L Attachment A, SVP-98-113, Page 142 of 153

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an accident, they merely provide the values at which actions should be taken once an accident beyond the design basis has already occurred.

l Safety Evaluation Number: SE-97-092 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700200

Title:

Fire Protection Systems, Turbine Deck Floor, MG Sets, Process Radiation System, Station Heating System ,

Description:

The scope of work being evaluated for this DCP consists of the following:

f STRUCTURAL CHANGES I

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l. ' A 3-hour rated masonry fire wall installed on the turbine building operating floor elevation 639 which replaces an existing water curtain. ' A qualified fire rated seal used for each penetration.

MECHANICAL CHANGES

1. Remove water curtain between the Unit I and Unit 2 MG Sets to facilitate installation of new wall.

2. Fire Protuilon Supply lines require relocation of a Victualic coupling away from wall penetration seal area.

l 3. Trim the W-6 beam on seismic support / restraint to ensure embedment in the penetration seal detail.

4. Shonen dead ended piping and recap to avoid any interference with new wall.

5. Add two fire extinguishers to have one on each side of the new wall.

6. Reroute Reactor Building Vent Separate Particulate Iodine and Noble Gas (SPING) monitor sample inlet line to pass through new fire wall with rigid pipe utilizing a qualified fire rated penetration seal.

' 7. Modify piping configuration and suppod bracket for 2 station heating unit heaters to avoid interference with wall. .

ELECTRICAL CHANGES

1. Relocate SPING monitoring panel and associated Communication Line Isolators.

2. Modify existing general lighting fixtures in area of new wall.

3. Modify turbine building elevation 639 emergency lighting.

l

4. Modify logic circuits associated with abandoned MG Set Water Curtain Suppression System Isolation Valve and Deluge Valve. These valves will be abandoned in place.

i Original Appendix R exemption request for the 3-hour barrier requirement in this location stated that the water curtain along with the deluge system on the MG Set and the ceiling wet pipe system is the equivalent i of a 3-hour fire barrier. There were no hydraulic calculations available to provide justification for the  :

1 exemption request. Additionally, the Safe Shutdown Analysis took credit for the 3-hour barrier so that J manual actions could take place on the opposite side of the water curtain. The water curtain will not prevent I

radiant heat from an MG Set fire so that the ability for the operator to complete the manual actions cannot be assured. Therefore, installation of a 3-hour barrier will improve on the 3-hour barrier system and also j assure that the manual actions will not be impeded. l l

Attachment A, SVP-98113, Page 143 of 153

Result: Unreviewed safety question does not exist. The removal of the active water curtain system improves the hydraulic characteristics of the remaining water based fire suppression systems. The wall provides a more positive assurance that a design basis fire will not affect redundant or alternate shutdown equipment, and provides literal conformance to Appendix R without the need of the exemption. The ability to achieve and maintain safe shutdown is therefore enhanced. Change does not introduce additional fire hazards which would require a different form of fire detection or fire suppression system or alter the operating requirements of fire detection systems. Operating capabilities of fire suppression system are changed in a conservative direction. Change does not introduce failure modes for fire detection or fire suppression systems that are not already analyzed in the SAR.

Safety Evaluation Number: SE-97-093 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700200

Title:

Fire Protection Systems, Turbine Deck Floor, MG Sets, Process Radiation System, Station Heating System

Description:

This safety evaluation addresses the electrical portion of this DCP.

1. Relocate SPING monitoring panel and associated Communication Line Isolators.

2. ~ Modify existing general lighting fixtures in area of new wall.

3. Modify turbine building elevation 639 emergency lighting.

4. Modify logic circuits associated with abandoned MG Set Water Curtain Suppression System Isolation Valve and Deluge Valve. These valves will be abandoned in place.

The specific electrical activities being evaluated here comprise only a portion of the overall scope of this DCP. (see SE-97-092). The electrical portion of this design merely removes interferences associated with installing the new 3-hour barrier wall.

Result: Unreviewed safety question does not exist. The removal of the active water curtain system improves the hydraulic characteristics of the remaining water based fire suppression systems. The wall

_ provides a more' positive assurance that a design basis fire will not affect redundant or alternne shutdown equipment, and provides literal conformance to Appendix R without the need of the exemption. The a'vility to achieve and maintain safe shutdown is therefore enhanced. Change does not introduce additional fire hazards which would require a different form of fire detection or fire suppression system or alter the operating requirements of fire detection systems. Operating capabilities of fire suppression system are changed in a conservative direction. Change does not introduce failure modes for fire detection or fire suppression systems that are not already analyzed in the SAR.

Safety Evaluation Number: SE-97-094 Type of Safety Evaluation: Design Change )

Evaluation Reference Number: DCP 9700219

Title:

Excess Flow Check Pipe Unions

Description:

This change replaced 1" socket welded pipe unions that allow removal of excess flow check valves with socket welded couplings. Seal welds were applied to the threaded pipe connections on both Attachment A, SVP-98-113, Page 144 of 153 o

r sides of the excess flow check valves. These valves are seal welded on one side. The 1" socket we!ded pipe unions have leaked in the past at the threaded connections for the excess flow check valves and have been seal welded. The unions were replaced with socket welded couplings to prevent possible leakage points.

l The threaded pipe connections to the excess flow check valves were seal welded on each side of the valve to prevent the possibility of future leakage at these connection points.

l _ Result: Unreviewed safety question does not exist. The socket welded connections and seal welds provide l a better leakage barrier than the threaded connections and provide a better connection to ensure leak j tightness than the existing configuration. This change will not create any new failure modes nor will the j change have any impact on the plant operating modes and applicable accident conditions.

Safety Evaluation Number: SE-97-095 - Type of Safety Evaluation: Work Request l Evaluation Reference Number: WR 970023062 l

Title:

Low Flow Feedwater Regulating Valve (FRV), Closure Solenoid Valve

Description:

This work request temporarily installs an electricaljumper to the electrical circuit for the l solenoid valve that supplies air to the low flow feedwater regulating valve. Installation of thisjumper energizes the solenoid causing the low flow FRV to fail in the closed direction. Purpose of the test is to l verify that the low flow FRV closes during runout flow control (ROFC) mode to ensure that the runout flow control circuitry is controlling flowrate of feedwater into reactor vessel by controlling the position of the main feedwater regulating valves.

Result: Unreviewed safety question does not exist. The low flow FRV will not be in-service during this test ,

i and is isolated from the feedwater system. Herefore, mis-positioning of the low flow FRV would not effect flow of feedwater into the reactor vessel. Flow of water into the reactor vessel is maintained by a combination of the CRD and RWCU systems. If the ROFC circuit were to become energized by an unexpected event, the control room operators will be capable of exiting ROFC mode of operation by ,

depressing the reset pushbutton. The low level reactor scram signal and low-low reactor water level signals ,;

for high and low pressure ECCS systems are not altered by this functional test and are capable of mitigating l a total loss of feedwater transient.

Safety Evaluation Number: SE-97-096 Type of Safety Evaluation: Design Change (see also SE-97-066 for this DCP)

Evaluation Reference Number: DCP 9700006, Mod Addendum Letter No. 3, NWR 970004211

Title:

Control Room Emergency Ventilation System (CREVS), HVAC Trains A and B

Description:

DCP 9700006 will be closed. Work completed under Addendum No. I and ER9605933 are reviewed by this evaluation to provide a better description of the work that was performed on emergency zone penetrations, ventilation dampers, the 'B' Train refrigeration control unit, and shaft seals on the booster fan. The repairs to CREVS are complete and the system has produced satisfactory inleakage and differential pressure results.

Attachment A, SVP-98-113, Page 145 of 153 l

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i Result: Unreviewed safety question does not exist. The Control Room Emergency Zone has been sealed to minimize inleakage and increase the ability to produce a difTerential pressure with respect to adjacent areas.

Failures are not any different than before this work was completed. Sealants were chosen that would minimize air leakage, but would not adversely impact surrounding equipment during any operating mode.

The booster fan shaft seals reduce inleakage to the Emergency Zone and therefore reduce consequences.

The refrigeration unit still has the same amount of available heat transfer area, so its performance will not l cause equipment to fail. The work that has been performed ensures that the CREVS can perform its' design function. No new functions or system interactions were created. l Safety Evaluation Nur'ober: SE-97 097 Type of Safety Evaluation: Design Change  !

(see also SE-97-092) '

Evaluation Reference Number: DCP 9700200 i

Title:

Fire Protection Systems, Turbine Deck Floor, MG Sets, Process Radiation System, Station Heatmg  !

System 1

Description:

This safety evaluation covers the structural / civil scope of work for DCP 9700200 which includes installing a 3-hour rated masonry fire wall on the turbine building operating floor elevation 639' l

which replaces an existing water curtain. The original Appendix R exemption request, for the required 3- I hour barrier in this location, stated that the water curtain along with the deluge system on the MG Set and the ceiling wet pipe system is the equivalent of a 3-hour fire barrier. There were no hydraulic calculations available to providejustification for the exemption request. Additionally, Safe Shutdown Analysis took credit for the 3-hour barrier so that manual actions could take place on the opposite side of the water curtain.

)

Subsequent evaluation determined that the water curtain will not prevent radiant heat from an MG Set fire.

This will inhibit the operator from being able to complete the required manual actions. Therefore, I installation of a 3-hour masonry barrier will replace the water curtain system and also assure that the manual actions will not be impeded.

Result: Unreviewed safety questior does not exist. The removal of the active water curtain system improves the hydraulic characteristics of the remaining water based fire suppression systems. The wall provides a more positive assurance that a design basis fire will not affed redundant or alternate shutdown equipment, and provides literal conformance to Appendix R without the need of the exemption. The ability to achieve and maintain safe shutdown is therefore enhanced. Change does not introduce additional fire hazards which would require a different form of fire detection or fire suppression system or alter the operating requirements of fire detection systems. Operating capabilities of fire suppression system are changed in a conservative direction. Change does not introduce failure modes for fire detection or fire suppression systems that are not already analyzed in the SAR.

Safety Evaluation Number: SE-97-100 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QCTS 0600-06

Title:

Main Steam Line Drain Valve Local Leak Rate Test Attachment A, SVP-98-113, Page 146 of 153

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Description:

This 50.59 evaluates the administrative limits for the leakage rates for the 2-0220,-01 and 2-220-02 volume. During a Local Leak Rate test during Q2R14, after the 2-0220-2 valve's seal ring and stem were replaced, leakage was cecordeo N a2.0 scfh for the combined volume of 2-0220-01 and 02, inboard and outboard valves. QC15 0600 05 su that the combined total must be less than 10.0 scfh or less than 10.0 scfh per valve. Individual tr tan vere , ,t determined because this required the Main Steam Lines to be flooded. QCTS 0130-01 " Leak k Tm Nogram" allows for increasing leakage limit for 1 operating

[ cycle provided a safety evaluaticu cou.:%. vre is no si ;nificantl safety impact and the "As Left" LLRT Maximum Pathway is less than 0.6La. The ei.ci, a procedure change was put in place to temporarily increase (until Q2R15) the 2-220-01 and 2V% 4. valmne to 12.0 scfh for one operating cycle.

Result: Unreviewed safety question does act c '.st. There have been no physical che ses made to this volume. The maximum path leakage is 12.0 sufh attribut<d all to either the 2-0220-01 or 2-0220-02 valve.

In this case the other valves leakage would be zero .iid cantainment integrity is maintained. Minimum path leakage splits the leakage evenly between the two valves (6.0 scfh each) which means that even if one of the valves fails to isolate, the other valve would maintain containment integrity with an acceptable leakag3 of

,- 6.0 scfh. Therefore, allowing the valves' combined leakage to operate one cycle does not increase the consequences of an accident.

r Safety Evaluation Number: SE-97-101 Type of Safety Evaluation: Problem Identification Evaluation Reference Number: PIF 97-2233

Title:

Operable but Degraded 2A Core Spray Room Cooler at Unit 2 Startup

Description:

The 2A Core Spray Room Cooler was found to have a flow of 67 gpm through it during a j recent surveillance, the required design flow rate is 68 gpm. The cooler is known to be operable because calculations have shown that the minimum required flow rate through the cooler is 30 gpm assuming a 95 degree F river water temperature. This evaluationjustifies Unit 2 startup activities with the 2A Core Spray Room Cooler in operable but degraded status.

Result: Unreviewed safety question does not exist. The 2A Core Spray Room Cooler is operable but degraded due to the flow below 68 gpm design but above the 30 gpm min required. Therefore, it will still perform its function to keep Core Spray equipment below equipment qualification temperatures and will not affect normal plant operation.

Safety Evaluation Number: SE-97-102 Type of Safety Evaluation: Problem Identification Evaluation Reference Number: PIF 97-1982 l

Title:

Relief Valve Discharge Piping Routing Code Noncompliance I

Description:

This PIF identified 5 relief valves in the Off Gas System with discharge piping configuration not in accordance with design codes. Specifically, the discharge pipe from these valves has an ' initial vertical riser section, which is not consistent with code requirements for it to be able to drain and not accumulate liquid. This condition has been evaluated and determined to be technically acceptable to operate with the code noncompliance for an additional cycle.

Attachment A, SVP-98-113, Page 147 of 153

o Result: Unreviewed safety question does not exist. An Off Gas System overpressurization event involving one or more of the identified relief valves lifting would result from either an Off Gas System transient that would ultimately result in a loss of condenser vacuum transient regardless of relicf valve discharge piping integrity, or misalignment of SJAE valves. In the first case, the analyzed transient is already in progress and thus does not involve an increase in the probability of occurrence. In the second case existing step-wise procedural controls on valve manipulations reduce the probability of occurrence to a negligible level.

Probability of occurrence remains unchanged.

Safety Evaluation Number: SE-97-121 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700278

Title:

Bus 31 Cubicle 1, Feed to Safe Shutdown Make-up Pump (SSMP)

Description:

This change provides the design to replace the existing limit switches in Bus 31 cubicle 1 breaker elevator circuit with a GE approved alternate switch. It also provides a replacement cover plate catalog number for use in facilitating installation of the limit switch in the " Front Enclosure". The original switches used in the switchgear have failed and like-for-like replacement switches are no longer available from GE.

Result: Unreviewed safety question does not exist. Function of the new switches is identical to the old switches. Equipment failures are identical so there is no new failure mode. The limit switches and coverplate addressed in this review only contribute to the vertical racking of the breaker and cannot affect actual operation of the breaker or the designed function of the SSMP system. The breaker is not relied upon to perform any function to support plant operation during the insertion or removal of the breaker.

Probability of any accident is not increased.

Safety Evaluation Number: SE-97-122 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700283,9703943, DCN 001557M, NWR 970078180

Title:

IB RHR Pump Motor- Upper Bearing

Description:

This change modified the configuration of the shaft nut assembly for the 1B RHR Pump Motor from a lock nut, lock washer, jam nut configuration to a lock washer and and locknut configuration.

This is a change to the motor design to reflect "as-left" maintenance conditions. The motor in question was originally purchased from Monticello to support RHR motor replacement at Quad Cities Station. The pump loads and seismic design of the Monticello plant are different that Quad Cities Station. The Monticello L pump / motor design required an additionaljam nut to support downward thrust requirements of the motor at ll that plant. The design also requried a longer threaded region along the shaft. While installing the shaft nut, the threads were damaged to the point where it was difYicult to reinstall thejam nut. A PIF was written to address this damage. This design change has been issued to accept the as-left condition of the shaft nut assembly. The As-Left condition is consistent with the original design of the RHR motors.

Result: Unreviewed safety question does not exist. This change does not affect RHR system in any manner. Since this is a sub-component of the IB RHR motor, there are no interactions with other structures, systems, or components. The new shaft nut assembly has sufficient capacity to accommodate the original Attachment A, SVP-98-113, Page 148 of 153 i

design load. Therefore, the likelihood of a failure of the changed assembly is no more likely than a failure of the existing assembly. The new shaft nut assembly is functionally equivalent to the assembly that it replaces, therefore there are no new failure modes introduced.

Safety Evaluation Number: SE-97-123 Type of Safety Evaluation: Procedure Change Evaluation Reference Number: QOP 3200-02, QOP 3200-03, QOP 3200-04, QCOP 3200-05

Title:

2C Reactor Feedwater Pump (RFP) Minimum Flow Manual Isolation Valve

Description:

This change allows operation of the 2C Reactor Feedwater Pump (RFP) with its associated manual isolation valve (at main condenser) for the pump minimum flowpath,2-3213-C, in the closed position. With 2-3213-C in the closed position, there is not a minimum flow path for 2C RFP. FSAR Section 10.4.7 states that a minimum flow of 900 gpm for each RFP is required. A revision to Operating )

Procedures QOP 3200-02, QOP 3200-03, QOP 3200-04, and QCOP 3200-05 is required to allow operation

. of the 2C RFP while the 2-3213C manual valve is in the closed position. Procedure changes will allow operation of the 2C RFP until plant power level is within capacity of the available RFPs and then the 2C '

RFP should be removed from service, and to require process flowrate of 450,000 lb/hr through the 2C RFP be verified when the pump is operating to ensure adequate cooling.

q Result: Unreviewed safety question does not exist. Lack of a minimum flow path for the 2C RFP does not have any potential to create an accident or malfunction of a type different from those evaluated in the UFSAR. Revisia to Operating Procedures provide direction for operation of the 2C RFP with no minimum flow path available. The accident analysis for the malfunction of a reactor feedwater pump has been previously evaluated and bounds the consequences of a malfunction of a loss of a RFP due to damage caused by mis-operation of a minimum flow valve.

Safety Evaluation Number: SE-97-125 Type of Safety Evaluation: Problem Identification Evaluation Reference Number: PIF #Q1997-03027 l

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Title:

Operable but Degraded 2B RHR Room Cooler at Unit 2 Startup

Description:

This evaluates the continued operation of Unit 2, with the 2B RHR Room Cooler operable l but degraded due to a decrease in the cooler flowrate. During monthly performance of QCOS 5750-09, ECCS Room and DGCWP Cubicle Cooler DP Test, the flow through the 2B RHR Room Cooler was

- recorded at 113 gpm. Design flow rate for the cooler is 114 gpm. Calcul_ation NED-H-MSD-26 Rev 0 shows that the min required flow through the 2B RHR Room Cooler is 50 gpm. This 50 gpm will assure that the 2B RHR Room Cooler will remove the necessary heat load in the room in the event of an accident.

Because we are currently below the design flow rate but above the min required flow rate as shown in the l calculation above, then we must conservatively declare this cooler operable but degraded.

- Result: Unreviewed safety question does not exist. The RHR Room Cooler removes necessary heat load to keep the RHR System below equipment qualification so that it will be able to perform its design function of providing water to the core during an accident situation. The reduced flow rate will not affect the coolers

' design function, therefore the 2B RHR System will be able to operate to reduce the consequences of the accident. There are no new failure modes introduced from the reduced flow rate of the room cooler.

Attachment A, SVP-98-ll3, Page 149 of 153

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' Safety Evaluation Number: SE-97-126 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9600163 l

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Title:

Setpoint Change on the Reserve Auxiliary Transformer (RAT) Cooling Equipment l l

Description:

The setpoint for the forced air cooling fan temperature switch on the RAT will be lowered from 60 C to 50 C. The setpoint for the forced oil cooling pumps temperature switch on the RAT will be lowered from 65 C to 60 C. These setpoint changes are required to provide improved heat dissipation on i the RAT durmg penods of high ambient temperature conditions.

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Result: t Unreviewed safety question does not exist. Changing the setpoints does not change how the pumps j' and fans operate, except that it will cause them to run niore often. Increased operation of the pumps and fans may lead to a small increase in the probability'of a failure of the pumps and fans. However, this is

!: i outweighed by the increased reliability of the RAT. There are no new interactions created with other SSCs.

There are no new failure modes created by lowering these setpoints.~

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l Safety Evaluation Number: SE-97-127 Type ofSafety Evaluation: Addition of Equipment Evaluation Reference Number: QCFHP 0400-26

Title:

Operation of the ABB Telescope Fuel Sipping System i

Description:

This evaluates the installation, testing, operation and removal of ABB Telescope Fuel l Sipping System to detect the presence of failed fuel rods. Most of the equipment will be residing on the  !

. refueling bridge, with pick-up tubes clamped onto the grapple head.' Use of ABB's Telescope Sipping j System will allow for detection ofleaking fuel assemblies. This is an altemate method to using GE's  !

sipping equipment and methodology. 1 l

Result: Unreviewed safety question does not exist; The additional five pounds added by equipment installation are bounded by the design of the mast and grapple assembly. The mast and grapple assembly are maneuvered using a single failure proof hoist rated at 1600 pounds Joad, with an expected maximum  ;

load on the hoist of 1310 pounds (per the GEK 97086B). The addition of the ABB sipping equipment will l

not exceed the load rating of the hoist, nor will it create any type ofinterference with hoist operation. Per i

L the new sipping procedure, operational and visual checks will be performed to ensure that no part of the sipping apparatus interferes with the functionality of the grapple, mast or refueling platform. Assumptions

, and initiating conditions in Section 15.7 of the UFSAR remain unchanged.

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, Safety Evaluation Number: SE-97-129 Type of Safety Evaluation: Fire Protection Report

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-Change. .

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' Evaluation Reference Number:- Fire Protection Report Change Request Number 97-07

Title:

Safe Shutdown Path C1 Attachment A, SVP-98-113, Page 150 of 153 f

Description:

This evaluation supports the change to the Fire Protection Reports to remove the reliance on Unit 2 480 VAC power for Safe Shutdown Path Cl. During a self-assessment of Quad Cities IPEEE analysis it was found that feeder cable 20865, from Bus 23-1 to Bus 28, could be damaged during a fire in fire area TB-I QARP 1000-01 (Safe Shutdown Path Cl) requires Bus 28 to supply power to MCC 28-1 A, MCC 28/29-5, and MCC 28-1B for a fire in area TB-1. Therefore, essential shutdown equipment powered

- from these MCC's must receive power from an alternate source (MCC 18-1 A and MCC 18/19-5).

l Result: Unreviewed safety question does not exist. This activity has been previously analyzed for Shutdown Path C2 (QARP 1100-01) with regards to bus loading and emergency diesel generator loading.

An evaluation (NDIT No. QDC-97-086) was performed to evaluate the ability to use the alternate power feeds for safe shutdown in the event of a fire in Fire Area TB-1. The evaluation indicated that the fire would not have an adverse effect on the ability to use these cables for safe shutdown. Therefore no new failure modes are introduced. This change to the method for performing a shutdown using Safe Sht.tdown Path Cl does not place additional burden on the operator in the form of manual actions. The shutdown l methodology of using unit one power for critical unit two shutdown loads has been previously analyzed for

! the loads in question (Path C2). Therefore, there is not a negative impact on systems or functions.

Safety Evaluation Number: SE-97-130 Type of Safety Evaluation: Temporary Alteration Evaluation Reference Number: Temporary Alteration 97-1-31; DCP 9700313

Title:

Temporary Alteration 97-1-31 to remove the pendant and bypass the reactor building overhead crane (RBOC) pendant control from the bridge control.

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Description:

This change will wire the restricted mode / normal mode key selector switch into the control I circuit at the bridge and trolley restricted area panel to allow for operation in both modes using the bridge i

!' controls. There is a problem with the pendant that is isolating the control power to the bridge and rendering i the bridge contrels inoperable. The controls will be bypassed to allow crane operatior, with the pendant removed.

Result: Unreviewed safety question does not exist. Normal operation presently allows for controls from i either the bridge of the crane or the pendant. The TA does not bypass any of the protective features of the  !

crane except for the pendant high limit switch. Since the pendant will be removed, the switch is not needed -  !

and is bypassed to simulate the raised condition. Removing the pendant and installing thejumpers does not i add any new interactions to the bridge control circuit. Therefore, no new failure modes have been introduced. The probability of an accident has not increased because the procedural requirements for operation of the crane.with the pendant removed are required to be made in conjunction with the TA.

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I Safety Evaluation Number: SE-97-131 Type of Safety Evaluation: Operability Determination 1

L Evaluation Reference Number: Temporary Alterations 97-1-28 and 97-1-29  ;

L Operability Determination for PIF Qi997-03327 i QCOA 4100-11 and QCOP 4100-03

Title:

1/2A and 1/2B Diesel Fire Pumps i

i Attachment A, SVP-98-ll3, Page 151 of 153

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Description:

This evaluation assesses the acceptability of the operability determination which has been performed for PIF Q1997-03327 and temporary alterations 97-1-28 and 97-1-29 which disabled the remote i

and local alarms on the 1/2B and 1/2A Diesel Fire Pumps, respectively. This operability determination

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addresses the applicability to the 1/2B fire diesel of recent equipment problems on the 1/2A fire diesel.

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These equipment problems are due to trip of the fire diesel due to an overspeed signal following removal of a l

an alarm condition. In order to provide assurance that the problem would not cause an inadvertent trip and i

inoperability of the fire diesels, a temporary alteration was installed which disconnected a lube oil pressure <

switch and water temperature switch. These instruments actuate a series of relays which cause an alarm. '

The local alarm bell and remote alarm relay have also been disabled.

I l Result: Unreviewed safety question does not exist. The function oflocal and remote alarms is to provide

!- notification to operations that a mechanical failure has occurred or is possible. During non-emergency diesel runs, an operator will monitor the conditions at the diesel fire pump locally, and will be available to intervene if degraded conditions occur as directed in station procedures. During a design basis fire, the i alarm functions bypassed by these temporary alterations are not relied upon. The elimination of these alarms therefore does not create the possibility for a different accident or malfunction.

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Safety Evaluation Number: SE-97-133 Type of Safety Evaluation: Design Change Evaluation Reference Number: DCP 9700329 l

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Title:

Revises Comed Specification R-4411; Instrument and Control Air Piping and Tubing

Description:

The use of polytubing in certain applications will be inserted into Comed Specification R- l 4411. Currently, this spec does not allow the use of polytubing and therefore requires the use of a design I i change when this tubing is needed in the plant. This revision allows the use of plastic or polytubing in l instrument-and-control piping to supply control air in non-safety-related applications in which normal

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operating conditions are below a temperature of 125 degrees F and 40 psig internal pressure with a 1/4" i O.D. Polytubing may only be used on instruments that were designed to connect to polytubing and can j handle the temperature and pressure. R-4411 only allows copper tubing, but the plastic tubing is required l

for some applications where copper tubing will not work.  !

Result: Unreviewed safety question does not exist. The polytubing will be subjected to design temperatures less than 125 degrees F and pressures less than 40 psig which is the same as the copper tubing that is being replaced. The function of the new tubing is identical to that of the copper tubing. The affected instruments will not increase the consequences of a malfunction of equipment important to safety because .

the instruments are not assumed to function during or after an accident. None of the applications included  !

in this DCP performs a safety function. l l

Safety Evaluation Number: SE-97-140 Type of Safety Evaluation: Procedure Approval

! Evaluation Reference Number: GE Procedure GENE-VSP-96-004, Rev.1, April 1997

Title:

Spent Fuel Storage Pool, Vacuum Sipping

Description:

This GE procedure GENE-VSP-96-004, Revision 1 is being approved as a vendor procedure in accordance with QCAP 1100-15 to provide instruction for setup and operation of the GE l Attachment A, SVP-98-ll3, Page 152 of 153 l

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Vacuum Sipping System to detect the presence of failed fuel rods. Use of GE's Vacuum Sipping System allows for detection ofleaking fuel assemblies. Use of the vendor procedure is required since the existing station procedure (QCCP-1000-4) is applicable to an older model vacuum sipping system. This is an alternate method to using ABB's sipping equipment and methodology (QCFHP 0400-26).

Result: Unreviewed safety question does not exist. The worst case analyzed accident is a single fuel bundle drop. This accident is analyzed in the UFSAR and is bounding. Installation and removal of the GE <

sipping can is well within the ~ capability of the crane and the refuel bridge hoists as specified in procedure. l l Approximate empty weight'of the GE sipping can is approximately 25% of a fuel bundle. The GE sipping )

l can includes temperature measurement to prevent high water temperatures in the can, is designed to accept our fuel assembly types, and is designed to sit in the CRD blade storage / defective fuel storage location.

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Attachment A, SVP-98-113, Page 153 of 153 L .