ML20080T529
| ML20080T529 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/06/1995 |
| From: | Aitken P COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PCA-95-019, PCA-95-19, NUDOCS 9503130372 | |
| Download: ML20080T529 (27) | |
Text
4 W ' Osmmonwealth IMiwsn Gnnpany
.l
. Quad Oties Generating htation
. 22710 2(6th Avenue North Girdova,11.612424)740 ,
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309f54224I
,J, w: py hhk PCA-95-019 l
March 6, 1994- )
U. S. Nuclear Regulatory Commission ,
ATTN: Document Control Desk l Washington, D.C. 20555 l
SUBJECT:
Quad Cities Nuclear Station Units 1 and 2 Changes, Tests, and Experiments Completed i NRC Docket Nos. 50-254 and 50-265 Enclosed please find'a listing of those facility and procedure changes, tests, and experiments requiring safety
,- evaluations completed during the month of January, 1995, for Quad-Cities Station Units 1 and 2, DPR-29 and DPR-30. A summary of the safety evaluations are being reported in
", compliance with 10CFR50.59 and 10CFR50.71(e).
'1 Respectfully, Comed Quad-Cities Nuclear Power Station t t i k-Paul C. Aitken ;
System Engineering Supervisor PCA/dak .
Enclosure l
cc: J. Martin, Regional Administrator C. Miller, Senior Resident Inspector ;
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9503130372 950306 f PDR ADOCK 05000254 i R PDR / I A Unicom company Il
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DCR-4-93-185 DESCRIPTION: ,
This change provided new and revised P& ids for the Main Turbine Lube. Oil Piping System based on the "as-built" configuration as determined by system walkdown. The piping configuration changes are for the Main Turbinc Oil Reservoir
, (MTOR) and associated components, the MTOR Console Instrumentation, turbine bearing lubrication, the lift pump enclosure's equipment and configuration and the hydrogen seal oil / turbine shaft sealing subsystem._ The P& ids affected by this DCR are: M-2021 and M-48 Sheets 1 through
- 9. General Arrangement drawings M-3, M-4, M-5 and M-10 are also affected.
Note that the new and revised P& ids for the Recirculation Pump Motor-Generator (MG) Set' Oil Piping System were addressed by DCR 4-93-204 and that High Pressure Coolant Injection (HPCI) oil transfer pump removal was addressed via DCRs 4-91-168 and 4-92-248.
The following changes are purely editorial due to being traceable to an original design document.
Revised Unit 2 Turbine exhaust hood ord2r on General Arrangement drawing.
Reviewed turbine lube oil, turbine bearing lift pump, turbine shaft sesling, turbine oil storage, transfer and reservoir piping configuration.
Provided piping details for turbine centrifuge and vapor extractor.
Provided piping / valve configuration details for: level indicators and flow indicating controllers for the clean and dirty oil tanks, pressure indicators for the turbine oil filters and filter pumps, turbine oil reservoir level transmitters, temperature indicators, and console instrumentation, and turbine lube oil.
The following changes are not traceable to a unique design document but meet the original design intent. These changes are primarily due to presentation of original design information not previously shown on the P&ID, rather than documentation of unanalyzed changes to the Main Turbine Lube Oil System:
Revised EPN from system code 5600 to 5100 for: bearing oil lift pumps, hydrogen seal oil units, turbine oil tanks, coolers, vapor extractors, filters, and filter pumps.
Piping configuration and line number changes on the abandoned FN'CI turbine oil transfer piping.
Presentation of drain valves on capped lines as being normally closed.
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., : DCR 4-93-185 CONTD
- Designation of= valves'toLbe administratively controlled
'in:the closed position with S-locks.
- Addition of. capped vent and drain < lines with normally.
3* closed valves for component maintenance / testing.;
- Addition of normally:open root valves and reducers, associated with pressure sensing components.
Use of 1" valve (1/2-5199-174).on a non-seismic 3/4" line. .
- Substitution of globe valves in place of gate valves 1 for various vents,~ drains and oil transfer lines and a gate valve in place of a globe-valve for the MG set oil-transfer crosstie line.
- Presentation of valve 1/2-5199-53 as normally open in'a dead headed' process line.
- Revised valve line-up configuration to dxsignate the-1-51218. filter instead of the 1-5121A filner as the normally used-filter for the bearing oil supply.
- Addition of normally closed valve 1(2)-5199-246 between the hydrogen seal cooler level switch and the oil separator drain in lieu of a pipe plug. This valve ,
permits proper operation of the liquid detector while !
allowing ease of maintenance by providing a L controllable drainage path.
.. SAFETY EVALUATION
SUMMARY
- 1. - The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR whers any of the following is true:
- The change alters the initial conditions used in the d UFSAR analysis.
- The changed structure, system or component is ,'
explicitly or implicitly assumed to functior during or after the accident.
- Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
Anticipated operational UFSAR Section 15.0.2.1 Occurrences Turbine Trip Without Bypass UFSAR Section 15.2.3.1 Turbine Trip With Bypass UFSAR Section 15.2.3.2 Turbine-Generator Trip / Load UFSAR Section 15.8.4 Rejection i
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l DCR 4-93-185 CONTD j i
- For each of these accidents, it has been determined that the change described above will not increase the probability'of l an occurrence or the consequence of,the accident, or ;
malfunction of equipment important to safety.as previously i evaluated in the UFSAR. 2 C !
- 2. The possibility-for an accident or malfunction of a.
different type than any.previously evaluated in the UFSAR is~ .i not created because the consequence of a. failure of.the system is to trip the turbine and, consequently, the ;
reactor, which has been evaluated. This change is a !
, documentation enhancement which depicts the system in more !
detail. It does not affect the turbine oil' system original" design functions, or the consequences of malfunction of this i equipment. 'l
- 3. The margin of safety, is not defined in-the basis for any i Technical Specification, therefore, the safety margin is not- !
teduced. [
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., .DCR 4-94-160 l DESCRIPTION 8 -l This:DCR revised P&ID'M-78 to remove the "L.C.":(locked ,
closed)' designation- for- UnitL 2 Core SprayL Valve' 2-1402-328.
1 This change matches the existing _ plant conditions ~for. valve I
'2-1402-32B..
i This DCR revised schematic' diagram 4E-1508B Sht.~2 to.show' ;
the correct terminal-point designation at 250LVDC Motor l control Center (McC) 18, Compartment T02.for Residual Heat j
' Removal (RNR) Discharge to Radwaste Valve NO 1-1001-21. ,
This change matches the existing plant conditions and1 !
associated-wiring diagrams for valve MO. 1-1001-21. ;j This DCR revised key diagram 4E-2317 to show the correct !
breaker rating-("40A") for;the Unit 2 High Pressure' Coolant j Injection (HPCI) Turbine Emergency Bearing Oil Pump.- This !
change will match the existing plant conditions and f associated wiring diagrams for this breaker located at 250 i VDC MCC 2A, Compartment C01. -l This DCR revises wiring diagrams 4E-1822 Land 4E-2822 to }
correct equipment piece numbers'(EPNs) for the Sample Pump '
and Bypass Pump associated with the Unit 1 and 2 Primary.
Containment oxygen Analyzer System. This change will match-the existing plant conditions-and the' associated P& ids.
e This DCR revises wiring diagram 4E-1629 and wiring tabulation 4E-1878-to show existing connections for the'
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Turbine Electro-Hydraulic Control.(EHC) System pressure switches. This change matches the existing plant conditions j;
for the Unit 1 Turbine EHC System. <
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SAFETY EVALUATION SU30tARY: j i
- 1. The change described above has been analyzed to determine !
each accident or anticipated transient described in the :I UFSAR where any of the following is.true: !
The change alters the initia1Jconditions used in the !
UFSAR analysis. ;
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The changed structure, system or component is l explicitly or implicitly assumed to function during or ?
after the accident.
Operation or failure of the changed structure, system, I or component could lead to the accident. '
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- ' The accidents which meet these! criteria'are listed below:. -]
Decrease in Heat' Removal )
By The Reactor Coolant System UFSAR'Section'15.2 Increase in Reactor Coolant ~ j Inventory-- UFEAR Section'15.5 -j Decrease in Reactor Coolant Inventory-- UFSAR Section'15.6- j; o Sequence of Events and. Systems . !
Operation UFSAR Section 15.6.5.2 i L Anticipated Transients Without Scram UFSAR Section.15.8
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For each of these' accidents, it'has been determined that the i change described above will not increase the probability of -:
an occurrence or the consequence of the accident, or malfunction of equipment important to safety.as previously evaluated in the UFSAR. [
- 2. The possibility for an accident or malfunction of a !
different type than any previously evaluated in the UFSAR :is not created because revising P&ID M-78 to remove the "L.C." l (locked closed) designation for Unit 2 Core Spray ~ valve 2-1402-32B, revising schematic diagram 4E-15088 Sht. 2 to show the correct terminal point designation associated with RHR j Discharge to Radwasta valve MO 1-1001-21, revising key- l diagram 4E-2317 tn show the correct breaker rating ,
associated with the Unit 2 HPCI Turbine. Emergency Bearing ;
Oil Pump, revising the EPNs for Sample Pumps. 1(2)-8741-8 and if Bypass Pumps 1(2)-8741-9' associated with the: Unit 1 and 2- ;
Primary Containment Oxygen Analyzer and revising wiring ;
drawings 4E-1629 and 4E-1878 to show the correct wiring i connections associated with the Unit 1 Turbine EHC System -i i
will provide better assistance to operations and ~
maintenance, help clarify the design'and will not add any ,
new accident scenarios. !
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- 3. The margin of safety, is not defined in the basis for any !
Technical Specification, therefore, the safety margin is not i reduced. }
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SE-95-01 Alternate Limits' Position use for Control Rods !
DESCRIPTION:
Added a step to QCTS 930-1 and QCTP 930-4 procedures to allow substitute rod positions to be used with BPWS. .GE ,
evaluation MDE-42-0386 allows the following:
- 1. Up to 3 one notch changes (On 3 different rods) from BPWS-specified bank positions. ,
- 2. Position 02 substituted for position 00. ,
SAFETY EVALUATION
SUMMARY
- 1. The change described above has been analyzed to determine each accident or anticipated transient described in the l UFSAR where any of the following is true:
~
The change alters the initial conditions used in the UFSAR analysis. .
The changed structure, system or component is explicitly or implicitly assumed to function during or ;
after the accident.
Operation or failure of the changed structure, system, ,
or component could lead to the accident.
The accidents which meet these criteria are listed below:
Rod Drop Accident UFSAR SECTION 15.4.10 i For each of these accidents, it has been determined that the change described above will not increase the probability of ,
an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously :
evaluated in the UFSAR. !
- 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the CRD System is not changed in any way by adding the results of the GE evaluation to the rod sequence procedures. In addition, no other systems are to ,
be changed. The only changes to be made are the allowances of 3 one-notch errors on 3 separate rods, and to substitute 02 for 00 of the normal BPWS bank positions. GE has ,
analyzed the rod drop accident with these exceptions to BPWS '
and has concluded that the peak enthalpy deposited from a rod drop accident will be less the 280 cal /g (current analysis limit).
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< Because.the'CRD. system operation'is not; changed,_there is no- :
newfaccidentforeated.u The exceptions-(stated:in step.1) to j
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BPWS have been'avaluated by GE.in'MDE-42-0386, and have been
'found to;be within the licensing _ limit lof 280-cal /g.
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- 3. The margin'of safety, as defined in:the basis for any- . Bd Technical Specification, is not reduced'because statistical- '
analysis of .the results of past CRDA ' analysis for plants (
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using-SPWS was performed'by General. Electric (
Reference:
6). . i Most of the relevant rod worths were under 14 Ak.and all',. .l including a statistically determined 95/95- !
probability / confidence: level were wall under 1.5% Ak.: .All- 1 rod worths used resulted in peak enthalpies well under the. )
280 cal /gm licensing limit using the General Electric (NRC
> approved) calculation method. On the kasis of the data jl gathered and the analyses performed it ium statistical -.-
study, the highest peak fuel enthalpy calculated for past '
j CRDA analyses was 158 cal /ga.
For the postulated CRDA with thraa one notch insert / withdraw ,
errors on three different rods the highest incremental rod -
worth was bounded between 0.94% Ak and 0.95%Ak. This rod- 1 worth is below 14 Ak and, therefore, will result in a peak ll enthalpy well under the 280 c41/gm design limit.- j i
The BPWS analysis assumes-that the core will not be critical !
until after the first control rod of the second group to be Li withdrawn is pulled. When notch' position.02 is substituted l for notch position 00 criticality is reached before the l first rod of the second. group to-be withdrawn has been !
completely pull. However,-the highest incremental rod worth j beyond criticality was 0.1% Ak which is of no consequence- j with respect to the CRDA. Further, for the nostulated CRDA i where notch position 02~is substituted for r>tch position 00 !
with 3 one notch withdraw errors on three different rods, :
the highest rod worth was 0.95% Ak. This rod worth is below :
1% Ak and, therefore will result in a peak enthalpy well !
under the 280 cal /gm licensing limit. j i
To reasonably bound all fuel cycles for Ceco plants using l General Electric BWR fuel the results from the notch error !
cases were compared to the generic.BPWS results (Reference -l 6). Based on this statistical analysis for BPWS plants the !
statistically determined 95/95 probability / confidence level i for hot standby condition was 1.11% Ak. This 95/95 rod worth, however, does not include' increases in rod worth caused by notch errors. To approximate the increase in rod
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worth caused by notch errors and notch substitution the ]
largest incremental increase in rod worth caused by these +
errors for both full in to full out rod drops and for rods dropped to intermediate axial notch positions were added to the 95/95 rod worth. ,
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- ' , The greatest incremental increase in rod worth _ for. a: rod' dropped full in to full out with notch. errors and notch substitution was-shown to be 0.02% Ak:which when added to.
-the 95/95_ rod worth gives 1.13% Ak.
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The greatest' incremental. increase in rod worth for_a rod dropped'to an intermediate notch position:with notch. errors and notch substitution was.shown to be 0.33% Ak which when' added to the 95/95 rod worth gives 1.44% Ak.
The conservative rod drop. accident results presented lin.
Reference 2, which were based'on-bounding accident shape functions, scram shape-functions, and Doppler coefficients,:
indicate that for a rod dropped full in to full out a-1.42%
Ak rod worth will result in a peak'enthalpy of'280 cal /gn.
For a rod dropped from full in-to full out the bounding notch error rod worth was 1.13% Ak. From the-same conservative rod drop accident result presented in Reference 2, 1 rod worth of 1.13% Ak will result in a peak fuel enthalpy of 212 cal /gn.
Similarly, for a rod dropped to an intermediate notch position the bounding notch error rod worth was 1.44% Ak which will. result in a peak fuel enthalpy of 284 cal /ga.
However, the conservative analysis presented in Reference 2 does not take credit for-the improved scram shape. function which is obtained when a rod is dropped to an intermediate notch position with BPWS. The scram shape functions that were previously generated (References 1,.2, and 3) were with initial control rod configurations that positioned the control rod either fully inserted or. fully withdrawn.
However, in Reference 4 it is shown that when a group of-control rods is at an axial bank position (i.e.. notch 08) the scram reactivity shape is changed. More negative scram reactivity is added to the core in the early part of'the scram stroke which is'due to the group of control rods being deeply positioned in the core. The increased' negative reactivity insertion in'the early portion of the scram stroke helps to reduce the peak enthalpy of the fuel. It was shown in Reference 4 that when a control rod group is banked at notch position 08 the resultant peak fuel enthalpy; was 25% lower than when all of the control rods are either fully inserted or fully withdrawn. When credit is taken for the improved scram function the peak fuel enthalpy corresponding to the 1.44% Ak rod becomes 213 cal /gm which is well below the 280 cal /gm licensing limit.
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'In. addition, CRDA analysis' performed by Brookhaven National i Laboratory (BNL) (Reference 6)'for the NRC, using. .!
appropriate thermal-hydraulic feedback, confirmed that.the ,
General' Electric calculation procedure.is'very'conaarvative: :;
when the initial _ thermal-hydraulic state is at or near-saturation. The'BNL' calculations of the CRDA using .
appropriate thermal-hydraulic feedback and. conservative-assumptions on the initial thermal hydraulic state resulted in peak enthalpies well under 150 cal /gm-for a 1.5% Ak rod i and.well under 200 cal /gm for a 2.04.Ak rod. .In Comparison, !
I using General Electric's adiabatic model which is based on.
bounding scram shape functions,. accident shape functions, _'
and Doppler coefficients, yields-a peak fuel enthalpy of 280 !
cal /gm for a 1.42% Ak rod. ;
t It'is concluded that the consequences of a CRDA with three- l insert / withdraw errors of one notch on three different rods zi are below the 280 cal /gm licensing limit. - Furthermore, the :
I consequences of a CRDA when notch position 02 substitutes-.
notch position 00 in the BPWS criteria and 3 insert / withdraw ,
notch errors are allowed (notch errors are not allowed on ' i notch substitution rods) are also below the 280 cal /gm j licensing limit. However,lwhen notch substitution is :
allowed it is-possible that criticality will be reached ;
2aile the first rod of the second group withdrawn-is being 1 pulled completely out. j ll
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J TBCHOP33APETY\953AN.Rrr
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. M04-2-92-020 ~
Permanent Repair to the Reactor Vessel Shroud Access Hole Cover ,
1 DESCRIPTION:
This modification involved repairing the cracked AHCs. ,
There are two types of cracks this repair eliminated. ,
Circumferential cracks when developed along the weld affected area and radial cracks which could propagate to the vessel or shroud wall. The original plates and the weld -[
affected area was Electric Discharged Machined out. The new !
cover plates were designed in accordance with ASME Section III, Subsection NG.
SAFETY EVALUATION
SUMMARY
- 1. The change described above has been analyzed to determine !'
each accident or anticipated transient described in the UFSAR where any of the following is true:
The change alters the initial conditions used in the !
UFSAR analysis.
The changed structure, system or component is explicitly or implicitly assumed to function during or !
after the accident. l Operation or failure of the changed structure, system, !
or component could lead to the accident The accidents which meet these criteria are listed below:
LOCA UFSAR SECTION 14.2 FSAR SECTION 15.6.5 For each of these accidents, it has been determined that the ,
change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR. '
- 2. The possibility for an accident or malfunction of a ,
different type than any previously evaluated in the UFSAR is '
not created because this repair will not create an accident :
or malfunction different than evaluated in the SAR. ASME !
Section III Subsection NG was used'to assure reliability and adequate margins of safety in the design. The materials of '
construction are compatible with the vessel internals for a 40 year life. There has been no new malfunctions that have !
been associated with this repair. UFSAR SECTION 6.3.3.1.2.2. will be revised to reflect the designed leakage rate of 78 gpm. This will be added to the original 807 gpm i total due to the jet pump slip joints and bolted joints. i nrnonurmmawr t
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The new; total is 885 gpa. The::LPCI system' capacity was sized to accommodate 3000 gym. leakage at:these~ locations.
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.j Therefore,ithis does.not create.the. possibility of an' ;
accident'or malfunction different than previously evaluatedL i
-in the FSAR. 'f a
- 3. The margin of safety, as defined in the basis for any- :j Technical 1 Specification, is not reduced because the current- !
MCPR Safety Limit will remain valid and'the basis'for the. {
Technical' Specifications will not be affected as long as no !
more than one double: tap per recirculation' loop and;two l
' single tap jet pump flow-instrumentation are.out of service. !
1 This repair will not effect any accident or. transient safety l analysis which foras'the basis for the Technical Specifications. The only limit potentially impacted is the' !
MCPR safety. limit. .The MCPR safety limit considers.the '
effect of core flow uncertainty. The amount of bypass flow is negligible during steady stateLconditions. The total core flow uncertainty therefore remains-below.the'value used to generate the safety-limits. 'See The General Electric ]j Safety Evaluation for Quad Cities Unit Two AHC repair.' . .
Therefore, the current MCPR safety limit is valid and thel I basis.for the Technical Specifications will not be effected-provided no more than one double tap per recirculation loop.
and two single tap jet pump flow instrumentation are out of service.
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. E04-2-93-286 l 2A-3904, RBCCW TCV Replacement with Anti Cavitation Trim 1
' DESCRIPTION: )
i RBCCW TCV 2-3904A and existing operator was replaced with a 12" Copas Vulcan, Class 125,-Series D-600 valve with hush style anti-cavitation trim, and a Model D-600-16-50 actuator. The 18" to 12" reducers on each side of the valve was replaced with reducers coated with Belzona. A new pipe support was added to the 3" Instrument Air supply header.
One hanger on the 18" to 12" reducer was modified to accommodate the increased load of the replacement valve. ,
The new valve with anti-cavitation trim allows for multi- !
stage pressure reductions across the valve which minimizes )
the occurrence of cavitation. The existing TCV was replaced due to cavitation induced erosion and vibration.
SAFETY EVALUATION
SUMMARY
- 1. The change described above has been analyzed to determine each accident or anticipated transient described in the UFSAR where any of the following is true:
l The change alters the initial conditions used in the UFSAR analysis.
- The changed structure, system or component is explicitly or implicitly assumed to function during or after the accident. +
Operation or failure of the changed structure, system, or component could lead to the accident.
The accidents which meet these criteria are listed below:
None For each of these accidents, it has been determined that the change described above will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
- 2. The possibility for an accident or malfunction of a '
different type than any previously evaluated in the UFSAR is not created because the replacement of the RBCCW TCV is a like for like replacement in physical configuration and function. The new valve weighs 500 lbs more than the old valve. The piping due to the increase in valve weight has been evaluated per original design requirements, Power Piping Code B31.1 1967 edition. See Bechtel Calculation QC-429-P-054, Rev. 1. The support on the 18" to 12" reducer TIOl0Ps\1Afr!Y\95JAN.ItPT
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-will-be modified to~ accommodate the increase in load.- The
' existing pipe ~ stanchion supporting the 3904B valve.has been _
-shown to be acceptable.- .!
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-The'ev of~the new valve (880) is lower;than-the old valve, 's
-(1460). However, the lower'Cv of the new valve is still .j greater than the required CV of the system (564) See Bechtel ' i Calculation QC-429-N-010, Rev. O. The flow characteristics !
oof the new valve are adequate'and will throttle properly._ l All original design basis requirements are met for the valve. ;
replacement, j
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This is a non-safety related system whose failure will not j create a accident'of a different: type that would adversely l affect the health and safety of the public. It has been l demonstrated that replacement of the TCV 3904A valve meets -!
the original design requirements. !
- 3. The margin of safety, is not defined in.the basis for any Technical Specification, therefore, the safety margin is not reduced.
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-Exempt Change;E04-1-93-334
'1D Condensate Pump ~ j 4
i DESCRIPTIONt-
- The scope of work for Exempt Change E04-1-93-334 was to i finstall seven different design changes to the 1D Condensate- .:
Pump and the condensate Booster Pump (1-3302D.and 1-3401D "
_respectively). These: changes _are listed'as: ,
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1.. Reduction'in the gap between.the pump' casing' cut-water- !
and.the pump impeller. This' modification helpedireduce: 1 the pump vibration seen-at vane-passing frequency;on. -l the Condensate Pump and the condensate Booster Pump'. j
- 2. Changed out of the inboard and the outboard bearings ;
with that of aldifferent type.- This work ~was' completed i on both the Condensate and Condensate Booster Pump'to- ;
increase the life of these bearings, thus enhancing 1 pump reliability. .
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- 3. Relocation of the inboard bearing thermocouples (two.in !
all) on both-the Condensate and condensate Booster ;
Pump. This scope of work was required to accommodate j the new style bearing so that temperature'is properly j monitored. l 4
- 4. Drilling of_a new internal passage for the inboard and i outboard bearings oil sump. This was completed to both !
the Condensate Pump and Condensate ~ Booster pump. . The ;
new internal passage improved oil flow through;the l bearings and allow complete drainage of all oil in the 4 bearing housing.
- 5. Installation of oversize pump casing. wear rings and an d 0-Ring between the pump casing and casing ring nut ion.
both the_ Condensate Pump and Condensate Booster Pump.
These wear rings help minimize any pump casing erosion. ;
- 6. Chamfering of the pump shaft sleeve on the inboard and outboard side of the Condensate Pump ONLY, to' accept the installation of a new O-Ring. This O-ring is to stop leakage that is seen along the pump shaft.
- 7. Installation of a check valve on.the seal cooling line on the condensate Pump ONLY. This check valve'was' original design on the pump and must be installed to return pump to its original configuration.
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"The pump.manutsi .ure,.Ingersoll-Dresser reviewed and q
', M ' approves all work that,was performed under the' direction:of -
.thisiExempt: Change 1(E04-1-93-334); which'isibeing: reviewed. j?
.TheLdesign chhngesLlisted abovesimproved.the. reliability of '
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both the condensate and condensate Booster Pump. The work 4 completed-was classified asinon-Safety Related.;'There are
^'.f .No Unreviewed' Safety Questions associated =with this. Exempt Change.- Nor'are any: revisions required'to the UFSAR or-the q m Technical Specifications as a result of this work. j
^
4 SAFETY ETALUATION SUIDEARY:
1.. .The. change described above has been' analyzed to determina j
- Leach accident or' anticipated transient described in tho' j UFSAR where any of the'following is true:- l
- 'The change alters the initial conditions used in the !
UFSAR analysis. j
, 'j
- The changed structure, systensor. component is 1 explicitly or implicitly assumed to' function during or j after the accident. )
- operation.or failure of the changed structure, system, 1 or component could lead to the accident. j The accidents which meet these criteria are listed below:
' None a For each of these accidents, it has been' determined that the :l
- . change described above will not increase thelprobability of-
an occurrence or the consequence of the accident, or l
- malfunction of equipment important to safety'as previously ' ~
i'
- evaluated in the UFSAR.
i' i
- 2. The possibility for an accident or malfunction of a jl
- different type than any previously evaluated in'the~UFSAR is
- 'not created because this Exempt Change provides for several- q
! design changes to the condensate / condensate Booster Pump. :
The worst possible event this work'could cause'is'a sudden
.. failure on the Condensate / Condensate Boosv.er pump, with a j remote chance that this would trip tha Reactor Feedwater q Pumps. Although this is considered an incredible' event, the ,i UFSAR has analyzed the transient of the los:s of all-three j Reactor Feedwater Pumps and found it acceptable to loose all !
three simultaneously. Since the condensate pump and the j Condensate Booster Pump are insignificant to plant safety, l no Accident Analysis was required to be performed on their i malfunction. Therefore, it can be stated that there is NO ;
other malfunction of accident of a different type which l could occur as a result of this Exempt Change.
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- 3. The margin of. safety, is not defined in the' basis-for any-Technical Specification, therefore, the safety margin is not -
reduced.
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. E04-1-94-208 EDG HVAC Supply Fan Control Circuit Modification -;
DESCRIPTION
( A fire in the TB-II area could cause failure of the AR1 relay located in panel 2212-47 and associated conductors and disable'the U1-EDG HVAC. Consequently, the Unit 1 EDG would be disabled by a loss of HVAC.
The addition of a bypass switch at the U1 EDG reacte start I panel to bypass and isolate the AR1 relay in the HVAC supply >
fan control circuit in the event of a TB-II area fire !
corrects this situation and satisfy the requirements of the !
Quad Cities Appendix R program. ,
[ SAFETY EVALUATION
SUMMARY
- 1. The change described above has been analyzed to determine i each accident or anticipated transient described in the UFSAR where any of the following is true: ,
The change alters the initial conditions used in the .
UFSAR analysis. l The changed structure, system or component is explicitly or implicitly assumed to function'during or -
after the accident.
Operation or failure of the changed structure, system, :
or component could lead to the accident.
The accidents which meet these criteria are listed below: ,
Loss of Offsite AC Power SAR SECTION 15.2.6 For each of these accidents, it has been detetuined that change described above will not increase the probabilit, ,
an occurrence or the consequence of the accident, or !
malfunction of equipment important to safety as previour l evaluated in the UFSAR. .
- 2. The possibility for an accident or malfunction of a i different type than any previously evaluated in the UFSAR is ;
not created because the addition of the bypass switch to the ;
U1 EDG HVAC logic does not create a new failure mode. As !
described previously, it increases reliability of the U1 EDG I HVAC. Therefore, an accident or malfunction of a type ,
different from those previously evaluated is not created.
- 3. The margin of safety, is not defined in the basis for any }
Technical Specification, therefore, the safety margin is not reduced. {
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- ADRR 95-002,gSESRL4-2358-q I'
1 DESCRIPTION-
- . )
Installed temporaryilead4 shielding on the: ERV/Targetrock1"T- 1 Quenchers"'in the Unit.2 Pressure Suppression = Chamber. . . :l
.(Torus). ' Loading of the."T-Quenchers" is not to exceed 160; .l lbs/ft on the; pipe. All lead shielding-was3 installed >after j the torus was drained:and removed prior to filling the - l torus. ll ,
SAFETY EVALUATION SUIDEARY:
- 1. The change described above has been. analyzed to determineL
?-
The. change alters the initial conditions used in the. j UFSAR analysis. j 1
The changed structure,4 system or component is -j explicitly or implicitly assumed to function during or- .
)
after'the accident. Ll
'i Operation or failure of the' changed structure, system, or component could' lead to the accident.
j g
The accidents which meet these criteria are-listed below: .;:l l
None '
For each of these accidents, it has been determined that the change described above'will not increase-the--probability of an occurrence or the consequence of1the accident, or: 1 malfunction of equipment important.to safety.as previously evaluated in the UFSAR. ,
] )
')
- 2. The possibility for an accident or malfunction of a~ i different type than any.previously evaluated in the UFSAR is j not created because Site Engineering has evaluated the l ERV/Targetrock "T-Quenchers," and has determined that the 1 allowable additional lead shielding load is~160lb/ft. This i application is restricted to the time.that Primary containment is not required and the torus-is drained. Under these conditions, the torus would not be required to mitigate the consequences of an accident. In the unlikely event of a failure as a result of the lead shielding; installation, this failure would be contained within the torus. Safety-related' equipment outside the torus would not be affected by this failure. Therefore, the lead shielding installation does not adversely impact systems or functions that would create the possibility of an accident or malfunction of a type different from those in the UFSAR.
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SE-95-002 CONTD
- 13. - -The margin of safety, is not defined in the basis for~any Technical specification, therefore, the~ safety margin is not
- ., ' reduced.
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-Set'oint1 Change No's:95-001E, 95-002E,'and 95-003E p -l i
EDESCRIPTIOWs.
Installed new overload heaters ~in MCC'18-4, cubicles'1C,f2B,- "'
i and 2D. These cubicles correspondcto.the B Control-Roon
, l HVAC 1/2A'.and 1/2B booster-fans and.the Air-' Handling Unit. .j fan motors, respectively. -The new overload heaters are- . 1 models FH42 and FH85Lfrom Westinghouse.and replaced models: !
FH41-and PH83. Once overloads were replaced, the. relay dial. -!
, setting was placed at 100 percent.
-SAFETY-EVALUATION
SUMMARY
- 1. The change-described above:has been analyzed to determines ;
each accident or' anticipated transient described in the ,
UFSAR where:any of the following is true: I
~!
The change alters the initial conditions used in the i UFSAR analysis. ;
The changed structure, system or component is .
explicitly or implicitly assumed to function during or' :
after~the accident. ,j i
Operation or failure of the: changed structure, system, i or component could lead to the accident. j i
The accidents which~ meet these criteria are listed below: i LOCA UFSAR SECTION 15.6 j i
For each of these accidents, it has been determined that the j change described above will not. increase 1the. probability of j an occurrence or the consequenceLof the accident, Tor !
mal. function of equipment important to safety.as.previously j evaluated in the UFSAR.- '
- 2. The possibility for an accident or malfunction of aL .
a different type than any previously evaluated in the UFSARLis j
-not created because replacement of.the' overload heaters for i the booster fans and the air handling unit fan' motors for !
the B Control Room HVAC system cannot create an' accident or: .j malfunction that is different from those already evaluated.- 1 A failure of a booster fan is already assumed and so.2 fans j were installed. Additionally, the A Control Room HVAC j
-system is assumed to operate with the B Control Room HVAC l acting as the redundant backup to it. The new' overloads i will not affect normal operation and will protect the fans from unplanned trips or failures'during a degraded voltage j condition. No other components are affected by the a replacement. Therefore, no new interactions or functions are created that could cause a malfunction or accident different from those already evaluated.
11DIOP3\1AFETYi951AN.RPT I
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- 3. The margin of. safety,.'is not defined in'the basis for.~any
~
Technical Specification, therefore,1the safety margin is.not reduced.
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TDCHOl'34AITIY\95JAN.RPT
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'SE-95-004 !
HPCI Roon Cooler Thermostat Satpoint Change
-DBSCRIPTION s
Changed the thermostat setpoint to.the HPCI Room Cooler to 100*F. ;
SAFETY EVALUATION SUIDEARY ;
1
- 1. The change described above has been analyzed to determine l each accident or anticipated transient described in'the. !
UFSAR where'any of the following.is true: {
t
- The change' alters the initial conditions used in the l UFSAR analysis, j The changed structure, system or component is -
explicitly or implicitly assumed to function during'or j after the accident. )
Operation or failure of the changed structure, system,. ,
or component could lead to the accident. 1 l
The accidents which meet these criteria are listed below: l
.Small Line Break
- UFSAR SECTION 15.6 For each of these accidents, it has been determined that the ~
l change described above will not increase-the probability of j an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously evaluated in the UFSAR.
- 2. The possibility for an accident or malfunction of a .;
different type than any previously evaluated in the UFSAR is '
not created because this setpoint. change to 100*F on the HPCI Room Cooler thermostat will ensure that HPCI Room is maintained at the EQ normal temperature of 120*F and during an accident scenario when the HPCI System is running the room temperature.will be maintained at EQ emergency temperature of 150*F. This change will'not impact the operation of the HPCI system or any other system. This change will not create the possibility of a'new accident.
- 3. The margin of safety, is not defined in the basis for any Technical Specification, therefore, the safety margin is not reduced.
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..SE-95-006- . . .
j 3ESR'4-2199; DCR'4-95-007L
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-DBSCRIPTION: [
- i Changed the safety. classification of>the Steam Jet Air- l Ejector;(SJAE), Suction Isolation Valves, AO1(2)-5401A/BEand' i 1(2)-5402A/B , and the.Offgas:to ChimneyJIsolation Valve,~AO' !
~1(2)-5406 fron' safety related to non-safety related. The.
document change involved deleting safety flags and j references as they appear in: ,
j UFSAR ~ - - Section -10.4. 2.1 Design Bases - Main Condenser :
Evacuation. System.
P&ID's - M-42 Sheet 1 & M-84 Sheet 1
. Master Equipment List,' Tab 5400 ~j SAFETY EVALUATIOM
SUMMARY
.l t
- 1. The change described above has been analyzed to determine- ,
each accident _or anticipated transient. described in the' l UFSAR where any of the following is true:- j The change alters the initial conditions used in the :
UFSAR analysis. l
- t The changed structure, system or component is i explicitly.or implicitly assumed to function during or- 1 after the accident. j ?
Operation or failure of the changed structure, system, !
or component could lead to the accident. ;
1
- The accidents which meet these criteria are listed below:
}
Control Rod Drop Accident (CRDA) -UFSAR-SECTION 15.4.10 j l
For each of these accidents, it has been determined that the !
change described above will not increase the probability of !
an occurrence or the consequencelof the accident, or '
i malfunction of equipment important to safety as previously evaluated in the-UFSAR. }
- 2. The possibility for an accident or malfunction:of a f different type than any previously evaluated in the UFSAR is
~
i not created because this change does-not cause a functional i change in the offgas system or its interaction with other !
systems. Reclassification of these valves does not alter j any physical parameters or process variables of.the plant. I There are no new failure modes introduced to the offgas !
system, because there~are no new components added to the- l system. !
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'3. The margin of safetyi,is not defined in the basis for.any. .
Technical Specification, therefore, the safety' margin is noti reduced.
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_ Temp Alt-for Unit 2 Reactor Building. Sample Panel DBSCRIPTION:
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^
Routed sample water.fron'the outlet'of the conductivitvfCell for the Reactor Water Clean-Up-Inlet Sample-Point.to the- )
inlet side of-the Conductivity. Cell forithe Reactor: Waters '
Recirc Sample Pointito provide the Control Room with better monitoring of Reactor Water Quality.:
SAFETY EVALUATION
SUMMARY
~
- 1. The change described above has been analyzed to determine each accident or anticipated transient described'in the-UFSAR where any of the following is true: i 1
~
- The change alters the initial conditions used in the-UFSAR analysis. 1
~
- The changed structure, system or component is' i explicitly or implicitly assumed to function during or !
after the accident. !
'l
- Operation or failure of thatchanged structure, system,- l or component could lead to the accidents l 1
The accidents which meet these criteria are listed below:. ;
l Break in Reactor Coolant Pressure Boundary Instrument Line l Outside Containment UFSAR SECTION. 15.6.2 'l.:
For each of these-accidents, it has'been determined'that the change. described above'will not increase the probability of an occurrence or the consequence of the accident, or malfunction of equipment important to safety as previously ;
evaluated in the UFSAR. :
- 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because the tubing used for the Temp-Alt shall ;
be rated for 2 100 psig; which is greater than the rating ;
for the existing conductivity cell module. 1The sample water ~
is routed through a pressure reducing: valve prior to going ;
to the conductivity cell and subsequently through the tubing ;
installed by this Temp Alt. The pressure reducing valve ;
will maintain a pressure of approximately 20-40 psig, which ]
is well below the pressure rating of the tubing installed by I this Temp Alt.
TIrHOP34AFLTn95/AN.RIT
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SE-95-007 CONTD ,
I The pressure at. the ' inlet to the conduce ' 'ity Cell for the '
= Reactor-Water.Recirc-Sample Point is~expouted to;be zero psig; due to valves 2-220-44 and'2-220-45 being closed. ~ A- - ,
pressure reducing valve upstream of..the' conductivity cell '
would maintain pressure < 40 psig even if valve leak-by.-
occurs for valves 2-220-44 and 2-220-45.
- 3. The margin-of safety, is-not defined in the basis for any. .
Technical Specification, therefore,.the safety margin is not i
. reduced. --l s
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