ML20081B766
| ML20081B766 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/31/1994 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20081B753 | List: |
| References | |
| NUDOCS 9503160334 | |
| Download: ML20081B766 (237) | |
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I Attachment 2 l PY-CEI/NRR-1917L -
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i PERRY NUCLEAR POWER PLANT l
l SAFETY EVALUATION SUlefARY ..,
i PURSUANT TO ;
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10 CPR 50.59(B)(2) i I
1994 l i
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9503160334 950313 PDR ADDCK 05000440 R PDR -
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-i SE No.: 93-150 l Source Document: TAP-1916, Rev. 4, TC-3 l Description of Change .l This change to Plant Administrative Procedure (PAP) 1916 incorporates f specific direction for the fire watch when a situation occurs which could !
cause a delay in the timely completion of the fire watch round.
i Summary 7 I. No. The changes are administrative in nature and are found to be !
consistent with the fire protection requirements of the USAR. The i changes only provide direction to the person performing the fire i watch tour and remains consistent with previously evaluated code j requirements. Therefore, the probability of occurrence or the '
consequences of an accident or malfunction of equipment previously evaluated in the USAR is not increased.
II. No. These changes are administrative in nature and are consistent with l the fire protection requirements of the USAR. The changes do not !
impact any plant system or component. Therefore, the possibility of creating an accident or malfunction different from any previously evaluated in the USAR does not exist. ;
f III. No. The changes are administrative in nature and do not impact any ;
activity described in the Technical Specifications. Therefore, no i margin of safety will be reduced. '
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SE No.: 93-157 Source Doctunent: NR 93-S-200, Rev. O Description of Change Superseded by SE 94-122.
SE No.: 93-159 Source Document: DCP 92-0213A, Rev. O Description of Change ,
This design change replaces the Residual Heat Removal (RHR) system Loop 'A' return to suppression pool valve 1E12-F024A. The new alve is a motor operated globe valve versus the existing motor operated gate valve.
The new valve provides better closing performance than the old valve which has exhibited valve factors greater than the manufacturer's predicted values. This valve replacement results in a minor reduction in valve stroke time and a minor increase in flow resistance compared to the existing design. The change includes new larger power fuses to accommodate the larger motor operator and changes to a maintenance platform proximal to the valve.
Summary I. No. This change maintains original system performance requirements. The design change adheres to established codes and standards.
Furthermore, this change does not reduce the redundancy or independence within the RHR system or the onsite power supply. This change does not alter system test frequencies or compromise the availability of the RHR system. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. This modification does not introduce any new equipment types or new systems. Original RHR design bases are maintained. The replacement valve maintains the reliability of the RHR system as evidenced by :
the successful use of this valve in an identical application onsite. l Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated is not created.
III. No. This design change maintains the requirements of the original !
equipment design and construction codes, the original design bases and the equipment qualification requirements. Thus, there are no reductions in the margins of safety.
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Source Document: NR 93-N-0218, Rev. 0 Description of Change ;
This nonconformance report analyzes the rework / repair disposition on the Direct Contact (DC) feedwater heater wall. UT examination on the center i section indicates wall thickness readings below the minimum wall allowables as described by the National Board Inspection Code (NB1C) and ,
the ASME code. .
i Summary I. No. This disposition restores the walls to the code required wall thicknesses. The repair does not alter the design, material, and construction standards applicable to the DC heater. Welding will be ;
performed in accordance with the NBIC. The related USAR accidents !
are: loss of feedwater flow, feedwater line break outside of ;
containment, and loss of feedwater heating. None of the accidents !
are affected by this disposition. PT or MT examinations will be performed after the repairs to ensure structural integrity of the pressure boundary. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipnent important to i safety has not been increased.
II. No. The overall design of the DC heater is unchanged. All repairs will be made in accordance with the applicable codes. The USAR eccidents {
of interest are not impacted. The original design for materials and construction has not been changed. Therefore, the possibility of an accident or a malfunction of equipment of a type different than previously evaluated is not created.
III. No. The condensate system will still function as designed. Repairs will be performed in accordance with the applicable codes. Accident analysis has not been impacted. Therefore, no margin of safety has been reduced.
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Source Document: PSTG, Rev. 2, TC-5 Description of Change j These changes to the Perry Specific Technical-Guidelines (PSTG) correct- !
typographical and. grammatical errors, and deletes an unnecessary line on the Heat Capacity Limit (HCL) curve. ,
Summary I. No. These PSTG changes do not affect the various accident initiators previously evaluated in the USAR. The same operator actions are ,
being taken at the same point in the scenario for accidents already -!
evaluated in the USAR. Therefore, these changes do not increase '
the probability of occurrence or the consequences of an accident or i a nelfunction of equipment important to safety previously evaluated ,
in the USAR. ;
II. No. These changes do net directly affect any actions taken in the Plant l Emergency Instructions (PEIs). Therefore, the possibility of an i accident or a malfunction of equipment important to safety of a type different than previously evaluated is not created. l III. No. These changes do not directly affect any actions taken in the PEIs.
The changes st!!i meet the intent of the analyzed actions described in the Emergency Procedure Guidelines (the basis of the PSTG). ;
Therefore, no margins of safety are impacted. !
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Source Document: DCP 93-0166, Rev. O j Description of Change ;
This design change eliminates the cross-tie between the Auxiliary Steam I system and Direct Contact Heater (Condensate / Extraction Steam). The cross-tie.is utilized for Direct Contact Heater steam blanketing. The change was implemented to eliminate potential radiological cross :
contamination between both systems due to possible valve leakage. ;
Summary j i
I. No. The steam blanketing function for the Direct Contact Heater is ;
utilized during long system outages to prevent internal corrosion to i the heater. During normal plant operation, the steam blanketing i function hence the cross-tie is not utilized. The modification is :
constructed in accordance with ASME/ ANSI B31.1. The cross-tie !
was/is nonsafety and non-seismic and serves no safety function. l This modification did not change the location of "high energy" lines :
as discussed in the USAR. The design function of the Direct Contact Heater, the Auxiliary Steam system, Condensate system, and the !
Extraction Steam system did not change. Therefore, the probability i of occurrence or the consequences of a previously analyzed accident l or malfunction of equipment will not be increased. .
II . No. There will be no adverse operational impact on the Direct Contact Heater, the Condensate system, Auxiliary Steam system or the j Extraction Steam system. These systems will remain as reliable as- t with the cross-tie functional. The potential for radiological cross contamination will be eliminated by this change. Therefore, this change will not create the possibility for an accident or ,
malfunction of a different type thsn any previously evaluated.
III. No. This change will not alter the function or the overall operation of ;
any of the systems involved. Therefore, this change will not reduce any margin of safety. t r
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Source Document: DCN 4193, Rev. O Description of Change l
This drawing change makes various changes to the B-022 series drawings.
The B-022 drawings represent the environmental conditions of the plant.
The changes result from a review of the guidance provided in Regulatory i Guide 1.89, Revision 1 on the subject of equipment qualification !
radiation dose. The revised methodology used in calculating the ,
radiation environment assumes 100% of the core noble gas activity, 50% of the iodine activity, and 1% of the remaining ' solids' inventory is released. The original calculations assumed an additional 50% cesium l inventory. 4 Summary i 1
I. No. This change reflects the radiation environments expected for various !
zones of the drywell and the containment based on the guidance as !
setforth in Regulatory Guide 1.89, Revision 1. This change had no ;
effect upon previously qualified equipment since the previously used !
radiation qualification environments were more conservatively f calculated. No physical changes occurred to the plant as a result of this evaluation. This change has no effect upon any previously. !
defined accident or on equipment important to safety. Therefore, !
the probability of occurrence or the consequences of an accident or l malfunction of equipment has not increased. ;
II. No. This change reflects the radiation qualification environments [
expected based on the guidance as setforth in Regulatory Guide 1.89, l Revision 1. No physical changes to the plant have occurred. !
Therefore, creating a new accident or malfunction of equipment that l has not been previously evaluated is not possible. j e
III. No. The methodology approved by the NRC was applied and was used as the .
basis for these changes. The Technical Specifications do not '
address post LOCA radiation environments inside containment. No physical changes occurred to the plant. Ba. sed on the above, no ;
margin of safety has been reduced.
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Source Document: SOI-G41 (FPCC), Rev. 7, TC-10 3 Description of Change !
" i This change to System Operating Instruction'(SOI) G41 allows the Fuel !
' Pool Cooling and Cleanup (FPCC) system to be shutdown for maintenance or ;
modification when fuel is stored in the spent fuel pools. i Summary i i
I. No. The FPCC system is not analyzed as a possible initiator of any accident / transient in the USAR. The instruction has provisions to )
ensure a fuel handling accident outside the containment does not !
occur during the period of time that the FPCC system is shutdown. ;
Limits were established to provide adequate margin to the assumed l normal temperatures for the spent fuel pool and upper containment '
pool should fuel be in the pools during the period the system is i shutdown. Therefore, the probability of occurrence or the l consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR is not changed. !
II. No. Provisions in the instruction ensure that the pool heat-up time ,
assumptions will remain valid when the FPCC system shutdown is !
performed. Testing demonstrated that adequate margin exists should !
the FPCC system be shutdown. Therefore, an accident or a i malfunction of equipment important *o safety of a different type than previously evaluated is not cr. r _.
III. No. Testing demonstrated that adequate nargin relative to pool heat-up '
exists should the FPCC system be shutdown. Technical l Specification 3/4.6.3.4 is not impacted by this change. Therefore, !
no margin of safety is impacted.
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.SE No.: 93-169 Source Document: DCN 4202, Rev. 0 l
Description of Change )
This drawing change revises P&ID D-912-603, Drywell Cooling System, by l adding a note which permits removal of the roughing filters from the -
l Drywell Air Handling Units (M13) during normal operating periods. :
i Summa q I. No. This change removes the roughing filters from the M13 air handling ,
units during normal operations. The basis for this is that during I past filter change-outs, dirt loading on filters after an operating cycle is very limited to non-existent. The only time the filters collect dust and dirt is during outage periods whe.,the drywell is i open to an outside environment and work is being performed. The M13 j system is nonsafety/non-seismic and is not required to mitigate the consequences of an accident. Therefore, the probability of I occurrence or the consequences of an accident or malfunction of equipment has not increased. j II. No. System function will not be affected by the removal of roughing '
filters for normal operation. The system will continue to meet all operational and design parameters. Therefore, creating a new .
accident or malfunction of equipment that has not been previously evaluated is not possible. ,
III. No. System function will remain unchanged. Operation of the system will remain within design limits and in compliance with Technical Specification 3/4.6.2.6. Therefore, no margin of safety has been reduced.
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SE No.: 93-170 Source Document: DCP 92-7178, Rev. O Description of Change The Meteo.rological Tower Wind Direction Sensors and Processors are obsolete and are no longer supported by the vendor. This change installs vendor approved replacement Wind Direction Sensors and Processors.
Summary I. No. The new Wind Direction Sensors and Processors are direct replacements for existing equipment. The new Sensor and Processor, as a pair, will provide the same voltage input to the Meteorological Tower Computer and Recorders as the existing equipment. The installation of the new Wind Direction Sensors and Processors does not alter the plant operation for normal, abnormal, or safe shutdown conditions. There is no interface to plant systems that are safety-related or important to safety. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The existing Wind Direction Sensors and Processors have a high failure rate and are no longer supported by the vendor. If a failure of a sensor should occur, it would not impair any plant safety system or system that is important to safety. Weather data is available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day from the Cleveland-Hopkins branch of the National Weather Service. The data can be manually inputted into the emergency dose assessment computer program. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The new Wind Direction Sensors and Processors are an equivalent
< version of the existing units. There is no interface to any plant safety system or system that is important to safety. The replacement Sensor and Processor is a vendor approved replacement part and does not change the Meteorological Tower Computer or its programming. Therefore, no margin of safety has been reduced.
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t SE No.: 93-171 Source Document: SCN 00404-ISS-2151, Rev. 0' i
Description of Change j This evaluation analyzes a revision to the testing standards'for baseplate and equipment grout.- The standards originally referenced were {
very specific, including specific ASTM standards and the year of i revision. The grout supplier has since updated testing standards. -
e Summary '!
'l I. No. This evaluation performed a comparison review of the updated 1 standards versus the original standards. The revised standards were found to be technically equivalent to the original standards. -
Baseplate and equipaent grout supplied per the updated standards .
will perform the same as grout supplied per the original standards. .
Therefore, the probability of occurrence or consequences of an !
accident or malfunction of equipment important to safety previously j evaluated is not increased.
II. No. Since the grout supplied per the updated standards will perform the same function to the same degree as the grout supplied per the ,
original standards, the possibility of an accident or malfunction of ;
equipment important to safety not previously evaluated is not. !
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III. No. The function of the grout supplied per the updated standards is !
equivalent to that of grout supplied per the original standards, and .
therefore will not have any impact on the ability of plant ;
components to function as required. Thus, the margin of safety is not reduced.
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1 SE No.: 93-172 Source Document: -DCP 90-0140B, Rev. O Description of Change This design change installs projectors and screens to view process /ERIS displays in the Control Room. The change will'be an enhancement to the system and a aid for Control Room personnel. The 3 process computer screens on 1H13-P680 and the 1 on the Unit Supervisor's desk will be unchanged by this design change. The modification was reviewed for any.
potential seismic interactions and no new seismic interactions were found. This equipment will be diesel backed and thus will be operable-during a Loss of Offsite Power (LOOP). The addition of the projector equipment adds approximately 16,000 BTU /HR total heat load to the Control Room. Also, there will be 155W (529 BTU /HR) heat load from the batteries used to support projector equipment. The impact to Control Room habitability was evaluated for Station Blackout, LOOP and LOCA scenarios.
For Station Blackout the addition of 155W to Unit 1 Control Room heat load has an insignificant effect on calculated steady state Unit 1 Control Room ambient air temperature. For the LOOP and LOCA situations, the accident of biggest concern would be a LOOP with the loss of the Control Complex Chillers (P47). The addition of approximately 16,000 BTU /HR to the calculated Control Room heat load of 1,963,654 BTU /HR will still be less than that which was assumed in the Control Room temperature analysis (i.e., 2.0 MBTU/HR). Therefore, the calculated Control Room ambient temperature during this event is not changed. The addition of this 16,000 BTU /HR will not impact the Control Room Emergency Recirculation system. The Control Room HVAC system will be affected only in the smoke clear mode of operation. The maximum temperature is increased by approximately 10F. This increase in maximum temperature change has been evaluated to be insignificant to the control Room environment.
Summary I. No. Currently to access the ERIS and Process Computers, operators have their backs to the primary control board P680. The new design allows visual access to both the projector screens (displays both ERIS and Process Computer information) and the P680 concurrently. .
The increase in Control Room temperature has been evaluated as having a negligible impact to existing accident analysis. There is no affect on diesel loading. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR will not increase.
II. No. Loss of HVAC has been previously evaluated with regards to Control Room impact. The increase ia temperature caused by the projector does not affect the existing analysis. Therefore, the possibility of an accident or malfunction of equipment different than those previously evaluated is not created.
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' SE No.: .cs-172 (Cont.)
Sunnary (Cont.)
III..No. Technical Specifications do not describe the availability or presentation of ERIS or process computer data. The Control Room -
-temperature analysis is not impacted. Diesel: loading is not affected by this change. Hence, no margin of safety is reduced.
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Source Document: TXI-0174, Rev. 0 l
-Description of Change This temporary test instruction detail's operating the Residual Heat
-Removal (RHR) system in shutdown-cooling during Operational Condition 4, :
when the position of the'feedwater shutoff valve is unknown. For the purposes of this evaluation, the valve is assumed to be open and the nonsafety, non-seismic piping from the last feedwater heater outlet '
isolation valves is considered to be within the reactor coolant boundary.
Summary I. No. For this instruction, the two events of importance are feedwater i piping failure and loss of shutdown cooling. Feedwater piping )
failure is already analyzed in the USAR and the use of this instruction is bounded by this analysis. Further, the use of this .
instruction should not increase the potential of a feedwater piping !
failure. If failure of the feedwater piping should occur and i shutdown cooling is lost, the plant has the ability to restart l shutdown cooling using an alternate return path. Hence, the l probability of a loss of shutdown cooling as described in the USAR 't is not increased. Therefore, the probability of occurrence or'the l consequences of an accident or malfunction of equipment important to )
safety previously evaluated in the USAR is not changed. ;
II. No. The instruction contains a requirement to limit steady state makeup ;
flow to the reactor pressure vessel to 70 gpm to ensure that for any l leakage greater than 70 gpm an RPV Level 3 isolation signal will '
prevent operation of this mode of shutdown cooling. Should this path of shutdown cooling not be available, shutdown cooling using alternate paths is possible. These alternate paths are not impaired I by this instruction. Therefore, creating an accident or a malfunction of equipment important to safety of a different type than previously evaluated is not possible.
III. No. The actions described in this instruction do not directly impact any Technical Specification. The RHR system still satisfies all Technical Specification 3/4.9.2 requirements. Therefore, no margin of safety has been reduced.
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Source Document,: DCP 92-0202, Rev. 0 i Description of Change i
This design change installs a cross-tie between the Division II and Division III electrical systems. The cross-tie will~be used in the event '
of a Station Blackout (SBO) condition to allow isolation valve and l suppression pool makeup system valve manipulations. A power cable is run ;
between Division II and Division III Motor Control Centers (MCCs).
Independence is maintained by opening the disconnect switches and :
removing the fuses at both ends of the cross-tie. Operation of the ;
cross-tie is administrative 1y controlled. ;
l Summary i I. No. The design has been approved by the NRC per letter PY-NRR/CEI-0604L.
Load change under the SB0 scenario is within the capability of the Div. III diesel as shown in calculation PRLV-0059 and is within the capability of the MCCs. Electrical independence is maintained for non-SB0 conditions by open disconnect switches and removed fuses, eliminating the possibility of a single failure affecting two i divisions. There is no impact on safe shutdown capability circuits. ;
Accident analysis as described in the USAR is not changed. j Therefore, the probability of occurrence or the consequences of an i accident or malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. Division II-and III diesel-generator functions are unchanged for ,
non-SB0 conditions. The systems remain' separate through open disconnect switches and removed fuses under normal operation. A malfunction of the crosa-tie cable does not affect any divisional equipment. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR is not created.
III. No. The operability of both diesel systems is unchar:- :lectrical divisional separation is maintained. There is no t on Emergency Core Cooling System (ECCS) functions c; Afe shutdown capability. Therefore, no margin of safety is ci.1 .
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SE No.: 93-176 Source Document: NR 93-S-200, Rev. 2 i Description of Change Superseded by SE 94-122.
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Source Document: ' Emergency Plan, Rev. 11, TC-4 9
Description of Change ,
This change to the Emergency-Plan completely revises the automated and-manual offsite dose assessment methodologies and protective action logic ;
to address EPA-400-R-92-001. The change implements NUREG-1228 source l term development. !
Summary I. No. This change does not direct or impact the operation or design of any plant structure, system or component. Only the methodology used to assess the magnitude of an offsite radiological release and the !
recommendation of protective action recommendations to offsite agencies under the Emergency Plan is being revised. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously analyzed will not be increased.
II. No. This change does not alter the design of the plant; the type, ,
frequency or consequences of an accident; or direct plant mitigating ,
actions. Therefore, this change will not create the possibility for ;
an accident or malfunction of a different type than previously i evaluated. ;
I III. No. This change does not adversely affect any equipment or operation !
relied upon by the Technical Specifications. Therefore, it will not i reduce any margin of safety. ;
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Source Document: ONI-Nll, Rev. 6 Description of Change .
This change to Off-Normal Instruction (ONI) Nll incorporates the Emergency Procedure Guidelines (EPGs) on Secondary Containment Control. ;
Some of the Secondary Containment Control actions can result in different ;
operator responses (than that assumed in the USAR) for certain potential high and low energy pipe breaks. j Summary I. No. The actions taken to shutdown, cooldown and depressurize the reactor are within the bounds _of USAR analyses. The EPGs have been reviewed / approved for incorporation by the NRC. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. Actions to mitigate leak events by isolation of the leak source and, if necessary, elimination of the driving force for the leak are ,
within the range of analyzed actions in the USAR. The EPGs are l reviewed / approved by the NRC. Therefore, creating an accident or malfunction of equipment important to safety of a type different ,
than previously evaluated is not possible. j III. No. Leak mitigation actions are bounded by the USAR analyses. Technical i Specification 3/4,6,1 requires reanalysis prior to repressurization.
Therefore, no margin of safety has been reduced.
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I SE No.: 93-179 Source Document: ISTP, Rev. 2, 70-2 PDB-G0001, Rev. O, TC-4 SVI-E21-T2004, Rev. 5, TC-9 Description of Change .
1 These procedure / doc 2 ment changes reflect implementation of a design change which modifies the Limitorque gearing of the Low Pressure Core Spray (LPCS) inject!.on valve lE21-F005 to increase the thrust / torque capability of tha valve. The increased thrust / torque provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10.
Summary ;
I. No. These changes reflect the design change which maintains original system performance requirements in that the revised valve can provide the required flow to the reactor vessel within the established response time. The design change adheres to established codes and standards such that the pressure boundary integrity is not compromised. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment will not be increased.
II. No. These procedure / document changes do not reduce safety system redundancy or independence since the changes do not create any new-or altered interactions between the Residual Heat Removal and LPCS.
systems. Original system design margins and functions are maintained. In addition, no new permanent equipment types or new systems are introduced and original LPCS design functions are maintained. Therefore, these changes do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Since the safety functions of this valve are maintained by these procedure / document changes, there is no reduction in the margins of safety.
SE No.: 93-180 Source Document: NR 93-S-284, Rev. O NR 93-S-285, Rev. O NR 93-S-286, Rev. O Description of Change These nonconformance reports evaluate the temporary use of the Emergency Service Water keepfill check valves. The nonconformance reports were generated to document that the acceptable amount of leakage past the check valve seats was exceeded.
Summary l
l I. No. Service Water pressure will be monitored in the Control Room by an i alarm available to the operators. In the event a low Service Water header pressure alarm is received, all three Emergency Service Water pumps will be started thus negating the need for the keepfill function provided by the Service Water system. The monitoring of the Service Water system will ensure that the reliability of the Emergency Service Water system is not compromised. The Emergency - )
Service Water system will maintain its safety function as described i within the USAR. Therefore, the probability of occurrence cr the ;
consequences of an accident or malfunction of equipment has not l increased.
II. No. The Service Water system pressure will be monitored to ensure the adequate Emergency Service Water inventory is maintained. This ,
action provides for the equivalence to the existing and previously I accepted conditions for the Emergency Service Water system. i Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. Technical Specification 3/4.7.1 is not affected. The Emergency I Service Water system and the Service Water system are not impacted. I Therefore, no margin of safety has been reduced.
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' Source Document: LLJED 1-93-068 Description of Change This Lifted Lead and Jumper / Electrical Device (LLJED) involves swapping the controls for the Hot Surge Tank Alternate Level Control Valve, i 1N21-F220, with the controls for the Hot Surge Tank Normal Level Control l Valve, 1N21-F230, such that F220 will be controlled by the Hot Surge Tank !
Level controller, 1N21-R475. F230 will then be controlled by the manual I pot, IN21-R0708. The basis for the change is that F230 has developed a l sticking problem and no longer operates smoothly. A large change in i controller output is required before F230 will change position. This is I causing instabilities in Hot Surge Tank level control. Switching level control to the alternate valve F220 will restore stable level control. i The F230 valve will still be available using the manual pot as a backup to the F220.
Summary I. No. This changes uses currently installed equipment (swapping wires).
No new equipment is added. This change does not impact how the l reactor operator responds to condensate transients (F220 is now considered the normal valve and F230 is now considered the alternate i valve). This change does not impact and is bounded by the Loss of i Feedwater Accident Analysis contained in Chapter 15 of the USAR.
Therefore, the probability of occurrence or the consequences of an I accident or malfunction of equipment has not increased.
II. No. This modification uses currently installed equipment and does not j deviate from the Feedwater/ Condensate system design requirements. '
The change is bounded by the loss of feedwater accident. Thus, creating an accident or malfunction of equipment important to safety of a type different than any previously described in the USAR is not possible.
III. No. Hot Surge Tank level control and the F230/F220 valves are not covered by Technical Specifications or the Operating License, nor ,
does any Technical Specification rely on the proper operation of the 1 level control system. The modification restores stable Hot Surge Tank level control. Only currently installed equipment is used. No i new materials are added. Thus, the margin of safety is not reduced. l I
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" Source Document: DCP 92-0159, Rev. O Description of Change This design change adds twenty-six, nonsafety-related RTD sensors to the Residual Heat Removal (RHR) system heat exchanger piping in order to support performance testing of the heat exchangers. The RTDs are being-added as permanent plant equipment in order to reduce manrem during performance testing.
Summary I. No. The RTDs installed by this change are used for monitoring purposes only. The replacement RTDs are designed to fit in the same thermowells and produce the same type output as the original devices. Some of the replacement RTDs will be strapped on to piping. The strap-on RTDs have been analyzed to not impact the E12 or P45 piping due to the added weight. The additional weight on the conduit supports caused by the additional wiring has also been evaluated to have negligible effects on the supports. No piping boundaries are breached due to the installation or use of these RTDs. The RTD outputs have no interface with equipment important to safety. The routing of the RTD wires and the installation of the RTDs on the safety conduit and piping has been analyzed to have no impact en the conduit or piping. The monitored equipment will maintain intended design function. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not been increased.
II. No. The RTDs have no active function and will not impact the piping they are monitoring. These RTDs have been analyzed to have no affect upon any safety equipment. Therefore, the possibility of an accident or a malfunction of equipment of a type different than previously evaluated is not created.
III. No. The RTDs have no active function and will not impact the piping they are monitoring. The RTDs have no interface with any equipment important to safety. Therefore, no margin of safety has been reduced.
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. SE No.: 93-185 Source Document: NR 93-N-331, Rev. O Description of Change This nonconformance report provides a ' repair and use-as-is' disposition for use of a leak sealant device and sealant on a IB Moisture Separator Reheater (MSR) shell pocket drain. The clamp and sealant will plug a pin hole leak on a 1-1/2" diameter pipe.
Summary I. No. This disposition has not affect on the loss of vacuum event described in the USAR. The design and manufacture of the leak sealant device is to approved industry codes and standards. The postulated separation of the 1-1/2" pipe is estimated to generate a 4 GPM leak of condensed steam which would be relatively insignificant from a flooding concern. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The affected piping is located in the Turbine Building which is nonsafety, non-seismic. No safety related equipment is in the vicinity of this line or sealant clamp. Flooding cannot possibly affect any equipment important to safety. Therefore, creating an accident or malfunction of equipment of a type different than has been previously evaluated is not possible.
III. No. The leak sealant device installed is designed and manufactured in accordance with ASME Code Section VIII, Division I criteria and ASTM standards. The clamp will act as a secondary boundary around the pipe and will therefore restore the drain line integrity. The clamp and piping are not expected to fail during normal plant operations, so system and plant operation remain unaltered. Therefore, no margin of safety has been reduced.
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_ Source Document: ' NR 93-N-333, Rev. 0
' Description of Change This nonconformance report evaluates a crLex in piping upstream of valve j IN21-F0787. This is a Condensate system 3/4' vent line which attaches
_ vertically to a larger 30' pipe. The crack is about one inch in-circumference and is leaking a small plume of steam. This steam leak '
appeared to be coming from the heat affected zone' of a fillet weld' adjoining a one half coupling and a pipe nipple. Since it was not possible to isolate the steam leak, a temporary leak sealant enclosure was chosen to be used. The installation of the enclosure serves as a temporary repair. The enclosure will be used until the Condensate system can be shutdown and the weld permanently repaired.
Summary I. No. The operation of the Condensate system is unaffected by this temporary repair. Flood protection is not' required since any possible flooding caused by.the complete failure of the 3/4" vent line is bounded by a "Feedwater Line Break - Outside Containment" ,
accident. The loss of the nonsafety-related 3/4' vent line does not impact the accidents analyzed within Chapter 15 of the USAR. i Therefore, the probability of occurrence or the consequences of-an J accident or malfunction of equipment has not increased.
II. No. The worst case postulated results of the failure of the 3/4" Condensate system vent line would not affect any equipment important to safety. Any plant flooding that would occur as a result of a complete failure of this vent line is bounded by the "Feedwater Line Break - Outside Containment Event.' Therefore, creating a new '
accident or malfunction of equipment.that has not been previously evaluated is not possible.
III. No. Failure of this Condensate system weld will not impact plant operation with respect to the Technical Specifications. Therefore, no margin of safety has been reduced.
SE No.: 93-188 Source Document: DCP 90-0280, Rev. O Description of Change
' This design change installs a Closed Circuit Television (CCTV) camera system on the refueling platform mast to facilitate r:. actor refueling operations and vessel internal component inspections. A CCTV camera is mounted internally in a new fuel grapple head at the end of the refuel mast. Camera cable is routed and supported within the mast. Local controls and a monitor are mounted on the mast. The mast is modified to accommodate this installation. A camera control unit with video tape recording capability and a monitor are located on a tray attached to the trolley cab handrail. A new cable reel is mounted on the second level of the trolley car. Power for the system comes from an existing 120VAC receptacle in the trolley cab.
Summary I. No. This design change does not affect the structural integrity of l either the refueling platform or the trolley. The function of the !
CCTV system is independent of fuel hendling equipment functions.
Fuel handling functions are not affected by this design change. A failure of the CCTV system does not contribute to the likelihood of 1 a fuel bundle accident as described in USAR Chapter 15. Therefore, :
the design change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The structural integrity of the refueling platform and trolley is not compromised by this design change. The CCTV function is 1 independent of fuel handling equipment funculons. Failure of the '
CCTV system has no impact on these functions. Therefore, the change does not create the possibility of an accident or malfunction of a different type than any evaluated in the USAR.
III. No. All operating requirements associated with the refueling platform and fuel handling equipment as defined in the Technical Specifications are unchanged. The design change has no impact on the original design basis of the refueling system. Therefore. the margin of safety is not changed.
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69urce Document: DCN 4052, Rev. O Desctiption of Change Superseded by SE 94-135.
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SE No.: 93-190 Source Document: Emergency Plan, Rev. 11, TC-5 Description of Change This change to the Emergency Plan revises the Emergency Action Level (EAL) initiating conditions and indications for an abnormal radiological effluent release. The change implements the nomenclature .
and methodology outlined i1 EPA-400-R-92-001.
Summary I. No. This change does not direct or impact the operation or design of any plant structure, system or component. Only the classification of accidents under the Emergency Plan is being altered, not the accident initiators. Therefore, neither the probability of occurrence nor the consequences of an accident or malfunction of equipment previously analyzed will be increased.
II. No. This change does not alter the design of the plant; the type, frequency or consequences of an accident; or direct plant mitigating actions. Therefore, creating an accident or malfunction of a different type than previously evaluated is not possible.
III. No. This change does not adversely affect any equipment or operation relied upon by the Technical Specifications. Therefore, no margin of safety has been reduced.
SE No.: 93-191 Source Document: DCP 92-0082, Rev. O Description of Change This design change replaces the Instrument Air system check valve IP52-F550. The valve is safety-related and its function is to provide containment isolation. The original valve installed was a hard seated 2* Dresser valve. The replacement valve is a soft seated 1.5" Edwards lift check valve. The change simply replaces the original valve with a functionally identical one that has increased flow capacity and improved seat leakage characteristics.
Summary I. No. The portion of the Instrument Air system piping between the containment isolation valves is safety-related, class 2.
Valve 1PS2-F550 is part of penetration P306 and provides the inboard isolation barrier. The replacement valve selected meets the same design code (ASME III, class 2) as the original. The improved valve design enhances seating capability which reduces existing seat leakage rates at this particular location. Fission product transport to the environment involves pathway leakage to the environment by several differant mechanisms (containment leakage being pertinent to this change). The improved valve design enhances seating capability which in turn reduces overall containment leakage which ultimately serves to reduce any potential radiological impact.
Therefore, the consequences of previously evaluated USAR accidents are not impacted. Hence, the probability of occurrence or the consequences of an accident or malfunction of equipment has not ,
increased.
II. No. The implementation of this change provides an improved safety-related check valve which improves system reliability.
System / plant function and operation remain unaltered by this change.
As a result, the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR is not created.
III. No. The replacement valve will improve tha plant's capability of meeting the associated Technical Specification requirement for combined secondary containment bypass path leakage. Therefore, the margin of safety is not reduced. ,
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SE No.: 93-192 Source Document: DCP 92-0156C, Rev. O Description of Change This design change provides improvements to the integral !
nonsafety-related, compressed air systems for the fuel handling and !
refueling bridge platforms. The changes include the replacement of the existing compressor with dual compressors (switch selected), replacement of a carbon steel tank with a larger stainless steel tank, replacement of rubber supply lines with stainless tubing, improvements in filtration, and installation of branch isolation valves to ease component i maintenance.
Summary I. No. This change is limited to the fuel handling and re'uel platform integral air systems and has no affect on the acciaents or transients evaluated in USAR Chapters 15.7 and 9.i for fuel handling related activities either inside or outside the containment. The structural integrity and seismic qualification of each platform is maintained by this modification. The new design of the compressed air systems is equivalent with respect to general fit and function ,
of the existing systems except for improvements which increase system reliability. This change does not alter functions or operation of any air supplied fuel handling tools and the design ensures there is no increase in the probability of equipment fall down during a seismic event. Therefore, the probability of occurrence or the consequences of a previously analyzed accident or '
malfunction of equipment will not be increased.
II. No. The new tanks, while larger in size, conform to the ASME Section VIII pressure vessel requirements. The new compressors, j lines, filters and valves are equivalent to the existing design 1 requirements. Thus, the pressure integrity of the air systems is maintained. In addition, this design change maintains the structural integrity of the bridges. Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The changes do not alter the load capacity or limit the capability of platforms L handle fuel or control rods since the weight ;
increase is negligible and the installation has been analyzed to i maintain the structural integrity of the seismically designed platforms. Therefore, the margin of safety is not reduced.
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SE No.: 93-193 Source Document: PAP-1912, Rev. 4, TC-2 '
Description of Change ,
P This change to Plant Administrative Procedure (PAP) 1912 implements [
controls on hot' work activities in the Radwaste Interim Storage Facility and the Waste Abatement / Reclamation Facility buildings, adds guidelines i on the use of 'Siltemp" as protective covering, and clarifies sections on the work forms.
Summary [
I. No. All changes are administrative in nature and are found to be consistent with the fire protection requirements of the USAR. i Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the !
USAR is not increased. !
II. No. These changes are consistent with the fire protection requirements ;
of the USAR and do not impact any plant system or component.
Therefore, the possibility of creating an accident or malfunction different from any previously evaluated in the USAR does not exist.
III. No. The changes are administrative in nature and do not impact any activity described in the Technical Specifications. Therefore, no margin of safety will be reduced.
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2 SE No.: 93-194 Source Document: . PAP-1919, Rev. 1, TC-4 [
Description of Change l This change to Plant Administrative Procedure (PAP) 1919 changes the l location of Fire Brigade Stations #1 and #3 to the 620' level of the Service Building. This change also adds Equipment Storage Area #3, ,
located in the PACP HVAC room, t Summary j i
I. No. The changes are administrative in nature and are consistent with the l fire protection requirements of the USAR. The changes identify new
- locations for the storage and staging of the fire brigade equipment. :
The changes remain within previously evaluated code requirements.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the 1 USAR is not increased.
II. No. These changes address the re-location of the fire brigade stations are consistent with the fire protection requirements of the USAR.
The changes do not impact any plant system or component. Therefore, the possibility of creating an accident or malfunction different from any previously evaluated in the USAR doeu not exist.
III. No. This change is administrative in nature and does not impact any activity described in the Technical Specifications. Therefore, no margin of safety will be reduced. i
1 SE No.: 93-195 i Source Document: DCP 90-0140D, Rev. O i Description of Change This design change will upgrade the work stations in the Unit #1 Control '
Room to better accommodate the tasks of the operating crews. The upgraded work stations will eliminate potential personnel safety hazards, ,
improve storage of required materials, and provide a method of controlling traffic and access through the Control Room. The P895 panel :
will be relocated to provided better access and a more stable mounting of !
the panel. The controls on this panel will not be altered. The work j stations are not directly tied to plant control equipment. '
Summary I. No. The work stations involved in this change are not directly tied to plant control equipment and do not involve any modification to equipment important to safety as evaluated in the US?.R. Therefore, '
the probability of occurrence or the consequences of an accident or 3 malfunction of equipment has not increased.
II. No. The new work stations and displays do not directly impact the i operation or control of safety-related systems. One of the_ work stations has been tested and validated in the simulator over an extended period of time prior to implementation in the plant Control i Room. The new work stations increase visual access to critical indicators for enhanced awareness and provide access control through i the Control Room. Therefore, creating a new accident or malfunction of equipment that-has not been previously evaluated is not possible, j III. No. The work stations included in this change do not impact any :
Technical Specifications. The work stations do not contain controls '
or displays that directly impact plant control equipment.
Therefore, no margin of safety has been reduced. '
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SE No.: 93-196 Source Document: PAP-0803, Rev. 6 Description of Change This change to Plant Administrative Procedure (PAP) 0803 incorporates various changes which will make the Chemical Control Program more consistent with other existing plant procedures.
Summary I. No. These changes are administrative in nature. The changes do not modify the plant or change the process on how storage approval for chemicals in the plant is obtained. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These changes do not alter the operation or the configuration of the plant. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated is not created.
III. No. These changes are administrative in nature and do not affect the Technical Specifications. Plant operation is not affected. ,
Therefore, no margin of safety has been reduced. !
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' Source Document: DCP 89-0216, Rev.-2 Description of Change This design change removes the Monicore Vax 11/780 from parallel !
operation with the.new VAX 7000/610, 3D Monicore system, leaving the j
'3D Monicore system as.a discreet system. [
Summary l I. No. There is no impact upon the plant by the removal of the Mon 1 core Computer from the parallel configuration'with the new 3D Monicore l' system. The Monicore Computer and the 3D Monicore system are nonsafety-related and have no control interface with safety-related i systems; therefore, this removal does not alter the plant operation i for normal, abnonnal or safe shutdown conditions. The plant computers are not relied upon for any accident analysis. Therefore, ,
the probability of occurrence or the consequences of an accident or }
nelfunction of equipment has not increased. l t
II. No. The 3D Monicore system utilizes state-of-the-art hardware and l software which will functionally operate in the same manner as the >
Monicore system. The 3D Monicore system includes a fast monitor ,
function, which is more accurate than the previous Monicore LPRM i alarm function it replaces. The removal of the parallel operation; i between Monicore and the 3D Monicore does not change plant i operations, failure modes or reliability. Therefore, the t possibility of creating an accident or malfunction of a type ,
different than previously evaluated does not exist. '!
t III. No. The PANACEA software is utilized by the 3D Monicore System to verify compliance with the same Technical Specifications criteria as the Monicore system. The PANACEA software is approved by the NRC for ,
core design and thermal limit calculations and was used by General !
Electric for Perry core design. Thus, no margin of safety is {
reduced. :
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SE No.: 93-198 Source Document: DCP 87-0725, Rev. O i Description of Change i
This design change replaces safety-related analog Riley temperature modules associated with the Leak Detection system with' General Electric NUMAC digital leak detection monitors. The modification removes 52.of.
the 60 Riley temperature modules in the Leak Detection System (LDS).
These temperature modules provide divisional alarms, and when necessary, '!
isolation signals which close either inboard or outboard containment isolation valves for a specific system, when high ambient or high differential temperature is sensed. This evaluation resulted in a i determination that NRC review is necessary prior to implementation. A request for NRC review has been submitted under separate cover.
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SE No.: 93-199 Source Document: DCP 93-0017, Rev. O Description of Change Superseded by SE 94-108.
SE No.:- 93-200 l Source Document: DCP 90-0098, Rev. 2- I I
DCP 90-0098A, Rev. O DCP 90-0098B, Rev. O DCP 90-0098C, Rev. 2 'i DCP 90-0098D, Rev. 2 l
. Description of Change l This design change removes piping support / snubbers on the Main Steam I lines inside the drywell, on the 10' Reactor Core Isolation Cooling -l piping inside the drywell, and on all associated branch lines as part of- i the Perry Snubber' Optimization Program. {
Summary j 5
I. No. General Electric (GE) and Cleveland Electric Illuminating. I!
Company (CEI) has performed a piping stress re-analysis using ASME i Code Case N-411. Per this analysis, GE and CEI have found that all i loadings and stresses are within ASME Code allowables for the ;
affected piping systems. The piping system performance is not changed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment'important to r safety previously evaluated in the USAR is not increased. (
e II. No. All equipment interfacing with the piping system were reviewed to l ensure compliance with their applicable design specifications. !
There are no changes to the configuration or function of the piping '
system, nor any adverse effect on interfacing equipment.- Therefore, the possibility of creating an accident or malfunction of a -
l different type than any evaluated previously in the USAR does not l exist.
III. No. The pipe stress analysis and pipe support design calculations meet the allowable stresses of the ASME Code. Interfacing equipment has similarly been checked to ensure compliance with pertinent ASME code ,
requirements as specified by applicable design specification. Also, f the configuration and function of the piping has not been changed. l Therefore, the margin of safety will not be reduced. '!
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. Source Document: DCP 93-0198, Rev. 1 l
Description of Change ,
i l l This design change provides architectural, electrical and communications l changes to support the creation of a new Radiologically Restricted )
Area (RRA) entrance on the 599' level of the control Complex.
l Summary I. No. All changes were done in accordance with the original design criteria. All affected components / systems / structures will continue ;
to function as originally designed. Therefore, the probability of i occurrence or the consequences of an accident or malfunction of equipment has not increased.
1 l II. No. All changes were done in accordance with the original design l l criteria. All components / systems / structures will continue to '
l function as originally designed. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. Since all affected components / systems / structures will continue to
! function as originally designed, the changes will not have any affect on any margins of safety.
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l l SE No.: 94-002, 94-178 Source Document: DCP 92-0060, Rev. O Description of Change This design change converts valves OP42-F0315A, OP42-F0315B, and OP42-F0315C from Motor Operated Valves (MOV) to manually operated valves.
Valves, actuators and motors will remain installed to reduce work scope.
Circuits are lifted to disable both the power and control functions of the MOVs. The cables and raceways are spared in place and the valves will be operated manually by their handwheels. This change essentially removes the automatic control feature for the MOVs. Manual operation is currently in place and will remain unaffected.
Summary I. No. The accident relevant to this change is a loss of cooling to the Control Complex Chillers thus resulting in a loss of cooling to the i Main Control Room. A failure analysis was performed to evaluate the '
failure associated with converting motor operated valves OP42-F315A, B and C to manually operated. Once these valves are positioned to maintain 1200 gpm through the Control Complex Chillers, adequate i cooling will be maintained. The electrical power has been removed i from these valves. Since the desired position for safe shutdown is normally open, no closure is required during loss of offsite power or LOCA. These motor operated valves are no longer an active component. Hence, probability of an active failure has been eliminated. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Prior to this change, valves OP42-F315A, B, and C were operated in one of two modes, auto or hand. In auto, the valves were electrically controlled, and in hand, they were manually positioned to provide 1200 gpm through the Control Complex Chillers. This change eliminated the auto mode of operation. No changes were made to the hand mode of operation. This change is such that the failure of OP42-F315A, B, or C will not prevent the system from performing its safety function. Therefore, the change does not create the possibility of an accident or malfunction of a different type than any evaluated in the USAR.
III. No. This change does not affect the previously evaluated hand / manual mode of operation. Flow through the Control Complex Chiller condenser will be controlled by existing plant procedures requiring approximately 1200 gpm to the chiller. There is no impact on the operation of safety-related equipment, specifically the Emergency Service Water, the Emergency Closed Cooling Water, and the Control Complex Chilled Water systems. Therfore, no margin of safety has been reduced.
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Source Document: DCP 93-0054, Rev. O SCRs 1-93-1193 and 1-93-1194 Description of Change This design change provides for the addition of Control Room indication for the Residual Heat Removal (RHR) and the Low Pressure Core Spray (LPCS) systems' pump suction pressure. Four local pressure indicators (1E12-R0002A-C, IE21-R0001) will be removed and replaced with Rosemount pressure transmitters (lE12-N0084A-C, 1E21-N0071). Trip units (1E12-N0684A-C, 1E21-N0671) will be added a the Control Room to provide ERIS indication for the RHR A/B/C and LPcS systems' pump suction pressure, and to provide annunciation on 1H12-P601 Sections 20A and 17A for RHR A and B low pump suction pressure respectively.
Summary I. No. This change provides Control Room indication rather than local indication to enhance the ability of the Control Room to monitor the RHR and LPCS pump suction pressures and indirectly to assess the condition of their associated suppression pool strainers. The annunciator setpoint will occur before pump cavitation during normal and accident conditions, such that the Control Room is provided the earliest opportunity to take remedial actions. This change does not increase the probability of a radioactive release siace the probability of an ECCS system pipe break or component malfunction is not increased. All components used for this modification were procured Level 1 which will assure that any consequences of failure due to a malfunction of equipment are not greater than those previously analyzed for similar equipment types. Also, the ECCS instrument sensing line modifications will meet the design / installation requirements of the original installation, and the transmitters were procured to meet the same pressure boundary design requirements as the existing gauges. As such, there will be no increase in post accident ECCS liquid line leakage as the result of this change. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This design change maintains the requirements of the original equipment design and construction codes as well as the equipment qualification requirements, such that this change does not introduce a new potential for a malfunction of equipment. Rosemount transmitters and trip units and Agastat normally de-energized relays are currently used throughout the plant and their failure modes have been previously addressed and accepted. Hence, this change does not create the possibility of an accident or malfunction of equipment of a different type than any previously evaluated in the USAR.
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Summary (Cont.)
III. No. This modification will increase the visibility of strainer fouling for RHR/LPCS by adding ERIS points (RHR A/B/C and LPCS) and annunciation (RHR A'and B)-for. pump suction pressure. The changes maintains the requirements of the original equipment design bases and construction codes, and the equipment qualification-requirements. Technical Specification Sections 3/4.3.3, 3/4.4.9, 3/4.5.1 and 3/4.5.2 are not impacted by increasing the visibility of low suction pressure on the RHR/LPCS pumps. Therefore, no safety margins are reduced as the result of this change.
1 SE No.: 94-005 Source Document: TXI-0170, Rev. O Description of Change This temporary test instruction covers the temporary operation of the Emergency Closed Cooling (P42) system during the dynamic diagnostic testing of P42-F150A, P42-F150B, P42-F300A, P42-F300B, P42-F330A and P42-F330B. The performance of these valves will be evaluated to satisfy the testing requirements under NRC Generic Letter 89-10, " Safety-Related Motor Operated Valve Testing and Surveillance." The ECC pumps will be momentarily operated below minimum flow requirements during this testing.
Summary I. No. All automatic functions and interlocks for the Emergency Closed Cooling (ECC) system will be operable during this testing. This testing will not inhibit the system from initiating or operating as needed in an accident. The system will perform as designed in the event of an accident. Operating the ECC pumps below minimum flow requirements momentarily (1 minute) will not degrade pump nor system performance as determined by engineering evaluation. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction will be increased by performance of this instruction.
II. No. All automatic functions and interlocks for the components operated under this instruction will remain operable. These changes will not inhibit the system's ability to function as designed in an accident.
Additionally, administrative controls are provided in this instruction to address operating the ECC pumps below minimum flow requirements momentarily (1 minute). Operation of the pumps in this manner will not degrade pump nor system performance as determined by engineering evaluation. Therefore, performance of this instruction will not create the possibility for an accident or malfunction of a
- different type than previously evaluated.
1 III. No. System actuation setpoints required by Technical Specification are not affected by this instruction. Operation of the system in this instruction will not degrade performance or reliability. Therefore,
! no reduction in the margin of safety will occur.
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SE No.: 94-006 Source Document: DCP 92-0156D, Rev. O Description of Change This design change eliminates the Refueling and Fuel Handling Bridge Radiation Monitors 1D21-K0091, 1D21-N0090, OD21-K0291, and OD21-N0390, and their associated supports. The purpose of the radiation monitor on the bridges is to provide an indication of increasing dose-rates which is now accomplished through the Merlin-Gerin (M-G) dosimetry system.
Summary I. No. The PNPP refueling and fuel handling bridges were designed in accordance with General Design Criteria 61 (10CFR50 Appendix A),
" Fuel Storage and Handling and Radioactivity Control", and Standard Review Plan 12.3-12.4, " Radiation Protection Design Features". The dose limiting requirements specified in 10CFR20.101 are fulfilled l without permanent radiation monitors on the bridges. Permanent I
monitors are not required to maintain compliance with regulatcry documents. Credit is not taken for the Refueling and Fuel Handling Bridge Radiation Monitors in regard to criticality accident monitors. The brackets used to support the monitors do not enhance the structural integrity of the bridges. The monitors are used for personnel protection only, and they have no interface with equipment important to safety. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The Refueling and Fuel Handling Bridge Radiation Monitors are not i integrated into the operation of the bridges, procedurally or through interfaces. Therefore, there is no additional risk associated with their removal. The radiation monitors located on the Refueling and Fuel Handling Bridges are not designed to mitigate or prevent accident scenarios of any kind. Therefore, the possibility of an accident or raalfunction of equipment of a different type being created does not exist.
III. No. Technical Specifications Section 3/4.9.6 contains several load limits and cutoffs for the refueling and fuel handling bridges. The radiation monitors currently installed on the bridges are not integrated in the logic controller for these limits (interlocks).
The removal of the permanently installed monitors will not reduce any margin of safety.
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SE No.: 94-007 Source Document: MFI 2-94-0001 Description of Change This Mechanical Foreign Item (MFI) evaluates the repair of a flanged connection at Unit 2 Instrument Air flow element, 2P52-N0090, via Work Order 94-336. This flanged connection is in the common cross-tie between the Unit 1 and Unit 2 Instrument Air (PS2) distribution systems. The purpose of the MFI was to install three (3) air hoses between Unit 1 PS2 J-headers and Unit 2 PS2 J-headers. This allowed the two distribution systems to remain cross-tied via alternate means while the normal cross-tie was isolated to implement repairs on the leaking flanged connection.
Summary I. No. The installation of this MFI will pose no risk to the safe operation of the plant. Specifically, the hoses to be utilized in the MFI will be rated equal to or greater than the rating of the Instrument-Air (PS2) system. Also, should any or all three of the temporary hoses rupture / break, the leak would be isolated by closing the appropriate connection valve on the respective J-header, or by isolating the J-header itself. Should the leak not be isolated, failure of the hosing would be bounded by accident analysis provided in the USAR regarding the loss of instrument air. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. As previously stated, this MFI installs three (3) temporary hoses to ensure that the Unit 1 Instrument Air (lPS2) system remains cross-tied to the Unit 2 PS2 distribution system. Should any or all three of the temporary hoses rupture / break, the following actions could be taken: 1) close the specific isolation valve at the J-header supplying the hose (s) - this could result in the loss of instrument air to a limited number of plant systems; 2) isolate the entire J-header (s) from the leak - this would involve isolating a larger number of systems from their instrument air supplies; or
- 3) take no action at all. No matter which action is taken, they are all bounded by the accident analysis (Loss of Instrument Air) provided in USAR Chapter 15.2.10. Therefore, the installation of this MFI will not create the possibility of an accident or malfunction of equipment of a different type tnan previously evaluated.
III. No. As previously stated, this MFI installs three (3) temporary hoses to ensure that the Unit 1 Instrument Air system remains cross-tied to the Unit 2 P52 distribution system. Should any or all three of the temporary hoses rupture / break, the scenario would be bounded try the accident analysis (Loss of Instrument Air) provided in USAR Chapter 15.2.10. Therefore, the installation of this MFI does not reduce any margin of safety.
SE No.: 94-008 Source Document: Emergency Plan, Rev.11, 'IC-6 Description of Change This change to the Emergency Plan revises the Emergency Action Level (EAL) indications for a contaminated injured victim. Additional guidance added allows health physics judgment to be used in determining whether or not a victim should be considered contaminated, if due to medical considerations the exact radiological status of individual cannot be determined.
Summary I. No. This change does not direct or impact the operation or design of any plant structure, system or component. Only the classification of accidents under the Emergency Plan is being altered, not the accident initiators. Therefore, neither the probability of occurrence nor the consequences of an accident or malfunction of equipment previously analyzed will be increased.
II. No. This change does not alter the design of the plant; the type, frequency or consequences of an accident; or direct plant mitigating actions. Therefore, creating an accident or malfunction of a different type than previously evaluated is not possible.
III. No. This change does not adversely affect any equipment or operation relied upon by Technical Specifications. Therefore, no margin of safety has been reduced.
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SE No.: 94-009 Source Document: DCP 92-0006, Rev. O Description of Change This design change provides access manways on both the supply and return Circulating Water (N71) system 36' carbon steel lines associated with the Auxiliary Condensers. The manways will provide a means of system access for inspections and associated work activities inside the N71 pipe. The lines are located in the plant yard. The design of each manway will ,
consist of welded pipe fittings and a 24' flanged cover that meet the design requirements of the carbon steel nonsafety line specification L1-4. The 36' lines in question are designed in accordance with line specification L1-4 and therefore the modification is consistent with the original design.
Summary I. No. The manways have no affect on the operation, function or integrity of the Circulating Water system. All flowpaths remain unaltered since the manways only represent additional locations for pipe entry during system shutdown. Circulating Water yard piping failure is evaluated in USAR Sections 10.4.5.3.2 and 2.4.13.5.2. Both of these sections evaluate the postulated break of a 12-foot diameter N71 line. This modification has no bearing on those evaluated accidents or any other accidents previously evaluated in the USAR. The N71 system serves no safety function. There is no equipment important to safety affected by this change. Therefore, the modification does not increase the probability of occurrence or the consequences of an accident or a malfunction of any equipment important to safety previously evaluated in the USAR.
II. No. This change meets the applicable design codes, hence the design i basis of the system remains unaltered. Overall system integrity l remains equivalent to that which currently exists. As stated in the !
USAR, N71 serves no safety function and yard piping failures will not jeopardize safe plant shutdown or adversely affect operation of safe shutdown systems. Therefore, there is no accident or )
malfunction of equipment important to safety of a different type l than previously evaluated in the USAR created by this change.
III. No. System and plant operation remain unaffected. As a result, the Technical Specifications remain completely unaffected. Therefore, this change does not reduce any margin of safety. I I
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SE No.: 94-010 Source Document: TXI-0181, Rev. O Description of Change This temporary test instruction details how the corrosion product film adhered to the internal surface of the reactor recirculation pumps and piping can be dissolved. The chemical process utilized is LOMI, an acronym for Low Oxidation State Metal Ion. Chemical process flow will be through the suction and discharge decontamination flanges of the recirculation pumps. Ion exchange resin is used to remove all process chemicals from solution following the decontamination. Continuous monitoring will be maintained on reactor water conductivity, both reactor pressure vessel and line pressure, and the in-leakage rate of water.
Summary I. No. The activities that will be performed during this chemical decontamination will utilize design provisions of the Reactor Recirculation (B33) system and utilize approved chemicals to reduce personnel dose attributed to the B33 pumps and piping. The plant will be in a normal shutdown condition with the Reactor Recirculation system in a normal shutdown (isolated) status. The waste treatment activities related to the chemical decontamination will be of the same scope as those currently performed in radwaste processing and will have the same controls. The estimated offsite doses as presented in Section 5.2.4 of the Environmental Report will remain the same and thus will not exceed the 10CFR50 Appendix I design objective. This chemical decontamination activity is bounded by the analysis of radioactive release from subsystems and components in Chapter 15.7. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The decon activities fall within the envelope of previously analyzed events. The plant will be returned to the normal design condition, the decon equipment will be removed, and any residual decon chemical will not adversely affect plant equipment /naterial once the B33 system is placed back into service. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The decon activities do not affect any equipment required by Technical Specifications while the plant is in a shutdown condition.
The decon chemical will not adversely affect the equipment it will come in contact with. Therefore, no margin of safety will be reduced.
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Source Document: MFI 1-94-003 l FCR 018679 l Description of Change This Mechanical Foreign Item (MFI) installs a freeze seal necessary to support valve repairs on 1G33-F101 and 1G33-F103. The freeze c.eal is located on the Reactor Water Cleanup (RWCU-G33) system bottom head drain line from the Reactor Recirculation system (B33) and the reactor pressure vessel. The freeze seal will be placed on an unisolated pipe directly off the reactor. However, no nuclear fuel will be inside the reactor pressure vessel during the freeze seal operation.
i Summary I. No. The accident of interest is a LOCA within the Reactor Coolant Pressure Boundary (RCPB), more specifically, an unisolatable bottom head drain (B33/G33) line break or loss of freeze seal. The freeze seal will be placed on the RWCU piping while the reactor vessel is being maintained between 700F and 1400F with only static head pressure (approximately 50 psid) during refueling (Mode 5). All nuclear fuel will be removed from the reactor during the freeze seal evolution. These pressure and temperature conditions are not those defined in the USAR as prerequisites for a LOCA inside drywell.
Therefore, the accident evaluated in the USAR is not possible under these conditions. Nondestructive examinations (PT or MT) for indications, variations, and outside diameter differences performed both before and after the freeze seal will ensure that the pressure integrity of the reactor pressure boundary. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. With the plant in Mode 5 during the freeze seal duration and with the nuclear fuel removed from the reactor, the RCPB is no longer necessary and can be declared inoperable. LOCA events analyzed in the USAR assume the events are at rated reactor power. With the vessel depressurized, fuel removed, RPV head off, and upper pools flooded, there are little similarities in accident types between a B33/G33 pipe break during the freeze seal duration and a design basis LOCA. Although the prerequisites for a LOCA inside drywell appears to be different (reactor at power verses the reactor defueled) the design basis LOCA bounds the freeze seal failure.
Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
Summary (Cont.)
III. No.
The freeze seal cycle on the 2' pipe is an industry proven method of isolation of water system that have limited isolation capability.
Industry tests have proven that deviations below the transition temperature and back-to-normal temperature do not change the crystal structure characteristics of ferritic material. However, strength.
characteristics do change (higher yield, lower elongation, low toughness) while the material is below the transition temperature.
Nondestructive examinations (PT or MT) for indications, variations, changes and outside diameter differences are performed both before and after the freeze seal to ensure that the pressure integrity of the reactor pressure boundary is maintained. Therefore, no margin of safety has been reduced.
SE No.: 94-013 Source Document: SCRs 1-93-1141 and 1-93-1142 Description of Change Superseded by SE 94-159.
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SE No.: 94-015 Source Document: TXI-0182, Rev. O Description of Change This temporary test instruction evaluates the intentional synchronization of the turbine-generator to the grid 5 to 10 electrical degrees out of phase. The test will be performed to achieve a sudden change in electrical torque equivalent to 50% of rated turbine-generator torque capacity in order to measure and validate the sub-synchronous modes of '
vibration. The test will also determine the level of response of super-synchronous modes of vibration due to normal transient operations.
Summary I. No. Performance of the out of phase synchronization would not increase the probability of a generator load rejection or turbine trip because, during the performance of this ac. kity, reactor power will be less than 20%. Therefore, transient forces on the turbine-generator will be limited to approximately 50% of the equipment rating. Additionally, in the unlikely event that a load rejection or turbine trip would occur, the consequences would actually be less severe than described in the USAR due to the low power level at which the test will be conducted. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. No plant equipment modifications are required in order to perform this test. The only interface of test equipment with plant equipment were connections to spare turbine-generator speed sensors I and to torque collars on the turbine whose installation was previously evaluated under MFI l-93-067. As stated above, should a turbine trip or load rejection occur, the consequences are bounded by the existing USAR analyses. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. Technical Specifications require at least one turbine overspeed protection system to be operable in conjunction with the turbine steam bypass system. Performance of this test does not require any modification to these two systems. Therefore, no margin of safety has been reduced.
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SE No.: 94-016 Source Document: DCP 93-0017, Rev. 1 Description of Change This design change replaces a blind flange on the 6" High Pressure Ccre Spray (HPCS-E22) system flush line with a coupling connection. This configuration will enable quick alignment of a connection to a fire hydrant on the Fire Protection (P54) system using 5" fire hoses. This will allow for the use of the fire protection water supply from the diesel driven fire pump as a back-up water source for the reactor vessel.
The piping in the flush line is safety-related, designed as ASME Class 2.
The line is normally empty, at ambient temperature and pressure. The new fitting changed this classification to nonsafety and the pressure rating of this section of pipe changed from 900 psig to 175 psig at the flange connection. There is a normally closed valve and a check valve between the pressurized piping of the E22 system line and the part of the line modified by the new connection. This double isolation meets the ASME Section III requirements for separation of safety classes and is adequate isolation between high pressure and low pressure pipe classifications.
The additional weight added by the new connection has been analyzed and will not adversely impact the support of these valves for seismic events.
This modification will not adversely affect the operation of the HPCS system. The use of the back-up water supply for HPCS will not impact available water for fire fighting as is evaluated in the USAR.
Summarv, I. No. The separation of the high and low pressure piping with two independent means of isolation will withstand expected operating and accident initiated pressures. The increased weight due to the modification will not adversely impact the support of these isolation valves to withstand the safe shutdown earthquake. The '
modification does not reduce the ability of the HPCS system to function in a manner required for safe shutdown. Therefore, the probability of occurrence or consequence of any accident or a malfunction of equipment important to safety as evaluated in the USAR is not increased.
II. No. This modification involves utilizing existing plant systems in their design configurations, operating within existing design limits. The required operating modes and functions of the HPCS system involved in this change as they relate to safe shutdown are not different than previously evaluated in the USAR. Therefore, the modification i
does not create an accident or malfunction of equipment important to safety of a type different than previously evaluated.
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SE No.: j 94-016 (Cont.)
. Summary (Cont.)
III. No. The connection modified by this design change is not'part of the ;
t operation of the HPCS. system as required for safe shutdown. .The !
I flushing function is not required for system' operability as defined h
in. Technical Specifications. The separation of the high and low pressure' piping with two independent means of isolation will prevent ;
L. any. interaction between the modified portion of piping and the parts :i of the HPCS required to function to achieve safe shutdown. 'Ite use !
'of the back-up water supply for HPCS will not impact available water :l for fire fighting as evaluated in the USAR. Therefore,.the change j does not reduce any margins of safety.
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Source Document: DCP 88-0382, Rev. O SCRs 1-93-1243 through 1-93-1252 Description of Change i 1
This design change modifies the design of Hydrogen Analyzers, 1H51-P0022A/B, to improve the performance and reliability of the j monitors. I Summary I. No. The hydrogen analyzers are used by plant personnel for indications-as to when to take manual actions to reduce hydrogen concentration in the drywell and containment. The analyzers are seismic, Class lE. The modifications are consistent with this criteria.
Code requirements related to tubing, fittings, valve installation have been maintained. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The modifications are limited to the enhancement of the hydrogen analyzer operation / reliability. The design criteria are being maintained. No new interfaces to other systems or equipment as a result of these modifications are generated. These modifications will not create new accident scenarios. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The modifications are limited to the enhancement of the hydrogen analyzer operation / reliability. 'The design criteria are being naintained. Therefore, no margin of safety has been reduced.
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l Source Document: DCP 93-0051, Rev. O MFI l-94-001 TXI-0192, Rev. O Description of Change This design change will provide a temporary hookup of the plant Fire Protection system in place of the Service Water system to interface with the Nuclear Closed Cooling (NCC) system. This will provide a cooling water heat sink for the nornal plant heat loads during the refuel mode of operation. The modification includes addition of permanent fittings and temporary piping from fire lines for direct connection to the service water side of the NCC heat exchanger. j i
Summary I. No. The systems associated with decay heat removal are not impacted in that a qualified cooling system and a reliable backup cooling system will always be available. Thus, the Regulatory Guide 1.13 requirements are maintained. Capabilities of the temporary system meet all potential anticipated demands, during the time frame of operation, including concurrent occurrence of plant fires. Plant personnel response to potential events such as pipe breaks and earthquakes has not changed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of aquipment has not increased.
II. No. This change does not functionally impact systems classified as important to safety. The requirements of Regulatory Guide 3.13 are maintained. Therefore, creating a new accident or malfunction of equipment is not possible.
1 III. No. The margin of safety is not reduced since provisions for alternate decay heat removal should a qualified primary method is lost are maintained.
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.SE No.: 94-020 Sourcs Document: DCN 4044, Rev, O Description of Change Superseded by SE 94-176.
f' SE No.: 94-021 Source Document: DCN 4437, Rev. O Description of Change This drawing change revises P&ID D-302-621, Emergency Closed Cooling System, to clarify the safety /nonsafety interface between the Emergency Closed Cooling system and the Nuclear Closed Cooling system.
Summary I. No. The changes provide clarification associated with the original design basis of the Emergency Closed Cooling system and the Nuclear Closed Cooling system and seismic qualification. These changes do not involve any physical changes to the plant. The changes further ensure that the isolation function between the two systems remains as designed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. These changes reflect analyses that were conducted on the isolation valve stroke times and the respective system designs. The changes also reflect the results of an analysis which was conducted to seismically qualify sections of a Nuclear Closed Cooling system nonsafety/nonseismic line. These changes were concluded not to result in a safety-related concern. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The changes ensure the designed safety functions of the Emergency Closed Cooling system are maintained. Therefore, no margin of safety has been reduced.
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SE No.: 94-022 Source Document: DCP 93-0075, Rev. 4 Description of Change ,
This design change revision is based on the concerns addressed in NRC '
Information Notice No. 93-89. That is,'the eight isolation. valves -
associated with the operability of the Reactor Vessel Level Reference Leg l Purge Control system be'placed in a " locked open" position. !
Summary I. No. This change physically places all eight isolation valves in a
' locked open' position. It is concluded that by taking this course ;
of action, closure of the reference leg manual isolation valves due ,
to operator error, which could result in an over-pressurization of l the reference leg, has been eliminated. Therefore, the probability of occurrence or the consequences of an accident or malfunction of .
equipment previously evaluated has not changed.
II. No. Operation of-the Reactor Vessel Level Reference Leg Purge Control !
system will remain as designed._ The design basis remains unaffected. Therefore, this change will not create the possibility ,
for an accident or malfunction of a different type than any '
previously evaluated. i i
III. No. The Reactor Vessel Level Reference Leg Purge Control system will=
continue to operate as designed. Isolation valves which have been placed in a
- locked open' position have no interaction with the plant shutdown capability. Therefore, this change will not reduce any margin of safety. l l
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l SE No.: 94-023 Source Document: SCRs 1-93-1133, 1-93-1134, 1-93-1143 through 1-93-1165 l
Description of Change l
l These setpoint changes revise the allowable values for the M:in Steam i l Line (MSL) Low Pressure (1B21-N676A-D), Main Steam Line High Flow (E31-N686A-D, 1E31-N687A-D, 1E31-N688A-D, and 1E31-N689A-D), ADS Timer.
(1B21-07005A/B), and Residual Heat Removal-Low Pressure Core Injection A/B/C Low Pressure Permissive (1E12-N658A/B/C) trip functions. ]
The setpoints are being revised in the conservative direction (to current i Technical Specification values) based on the results of applying a new methodology to associated setpoint calculations. The basis for applying the new methodology was to address Perry Licensing Commitment 14 in USAR l Appendix 1B. In addition, setpoint Leave-As-Is Zone (LAIZ) values are being revised for the MSL High Flow, ADS Timer, and LPCI Low Pressure j Permissive (reset also) trip functions in the conservative direction.
The LAIZ and reset values are consistent with those reflected in the i instrument calibration SVIs.
Summary 1
I. No. The net effect of implementing these changes will be to establish ,
more conservative acceptance criteria in current surveillance ;
instructions with respect to licensing operability requirements l (allowable values) and Leave-As-Is-Zone criteria. Actual I performance criteria for the instruments to actuate at its current setpoint remains unchanged. No field changes are resultant from l these SCRs. Revision to the LAIZ for the referenced components is in the conservative direction and well within the capability of the components as identified in actual SVI test results and the current l
Calibration Trip Data Sheets. Therefore, the probability of occurrence or consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the USAR has not increased.
l II. No. Revising the allowable values is conservative with respect to current settings in Tables 3.3.3-2 and 3.3.2-2 of the Technical Specifications. This change allows the implementation of the new GE Setpoint Methodology and provides more conservative allowable values for the referenced components. There will be no physical changes to the plant as a result of this change. Therefore, creating an accident or a malfunction of equipment important to safety different than any previously evaluated is not possible.
, III. No. Revising the allowable values for the referenced components increases the margin of safety as defined in Tables 3.3.3-2 and 3.3.2-2 of Technical Specifications. This change allows for the implementation of the new setpoint methodology and is more conservative than the allowable values previously calculated.
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SE No.: 94-024 Source Document: USAR Change Request 94-002 Description of Change This USAR change clarifies the use of the offsite radioactive waste reprocessing facilities, the storage of radioactive waste at the site's Onsite Storage Container (OSSC) Area, and the use of a vehicle type other than a tractor trailer to transport radioactive waste.
Summary I. No. The USAR change is simply a clarification of existing plant activities. There is no impact upon the site's compliance with 10CFR71 or Federal Department of Transportation regulations regarding transport of radioactive materials. Therefore, the probability of occurrence or consequences of an accident or s malfunction of equipment important to safety previously evaluated in the USAR is not changed.
II. No. The USAR change is simply a clarification of existing plant activities. There is no impact upon the site's compliance with 10CFR71 or Federal Department of Transportation regulations regarding transport of radioactive materials. Therefore, creating an accident or malfunction of equipment of a different type than previously evaluated is not possible.
III. No. This change does not alter the requirements to comply with federal shipping, transportation, and disposal site requirements.
Therefore, this change does not impact any margin of safety.
L SE No.: 94-025 Source Documer.t: PTI-E12-P0002, Rev. 2 PTI-E12-P0003, Rev. 2 Description of Change These revisions to the Periodic Test Instructions (PTI) for Residual Heat Removal (RHR)' loops A and B heat exchanger performance testing include incorporation of the ability to throttle Emergency Service Water (P45) system flow and the option to utilize additional M&TE not explicitly specified in the instructions.
Summary I. No. These instruction changes provide for data collection to support performance analysis of the RHR loops A and B heat exchangers per the requirements of Generic Letter 89-13. During performance of these tests, all of the design safety interlocks are maintained and system control logic as described in USAR Chapter 7 is not impacted.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment will not be increased.
II. No. These instruction changes maintain original plant performance capability and do not compromise established equipment design nor equipment qualification. The redundancy / independence of the Emergency Core Cooling Systems (ECCS) are not reduced. In addition, required system test frequencies are not increased by these instructions. Therefore, these changes do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. These instruction changes do not compromise original equipment design bases. The instructions direct testing to be performed within operating limits prescribed by existing, approved system operating procedures which assure system and equipment reliability are not degraded. Hence, no reductions in the margins of safety are created by these instruction changes.
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SE No.: 94-026 i Source Document: DCP 92-0015,-Rev. O !
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Description of Change i t
This design change modifies the Reactor Recirculation pumps, IB33-C0001A/B,'to eliminate susceptibility to shaft' cracking.--The new' j design includes a new pump shaft design, a new hydrostatic. style bearing -l design, a new pump seal heater / cooler and changes to the shaft / motor- -l coupling spacer lengths. In addition, this modification. includes changes ;
to the seal piping and instrument connections, modifies seal purge flow regulators and adds flow measurement improvements. j Summary I. No. This design change maintains the original system performance !
requirements and confornance to governing design codes such that !
original accident' analyses are not affected and the system's i pressure retaining capability is maintained. The modifications ;
improve the reliability of the pumps by reducing the potential for !
shaft cracking. The equipment qualification of the pumps and piping systems is maintained by these changes. Therefore, neither the j probability of occurrence nor the consequences of a previously i analyzed accident or malfunction of equipment will be increased. l II. No. These modifications maintain the pressure retaining capability of I the pumps and the new heater / cooler design maintains design code ,
requirements such that the probability of an inter-system LOCA is l not affected. Furthet, heavy load lifts associated with !
implementation are controlled via existing procedures and are ;
identified to the Outage Shutdown Risk Assessment Team to assure no increased accident risks during implementation. Therefore, these ,
modifications do not create the possibility of an accident or !
malfunction of a different type than any previously evaluated. j i
III. No. As evaluated, the design change does not compromise the USAR !
accident analyses. The modification will maintain the ability to i cool and monitor the reactor as defined in Section 3/4.4 of the i Technical Specifications. Therefore, the nargin of safety as l previously evaluated is not compromised by this design change.
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SE No.: 94-027 Source Document: DCP 93-0020A, Rev. O Description of Change Superseded by SE 94-183.
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SE No.: 94-028 Source Document: DCP 92-0084, Rev. O Description of Change This design change modifies Reactor Recirculation system discharge isolation gate valves, 1B33-F067A/B. The changes include internal modifications of the disc assemblies to eliminate potential flow induced vibration as identified in General Electric Service Information Letter (SIL) #528. In addition, this design change makes improvements in the valves' packing design to include the use of live load packing and changes the body / bonnet gasket from an asbestos base to a graphite base flexitallic.
Summag I. No. The subject valves fulfill a passive safety function only as a pressure retaining boundary and are used for maintenance of system components. The modifications inprove the mechanical reliability of the valves and limit packing leakage while maintaining their pressure retaining function. The equipment qualification of the valves is maintained by these design changes. These changes do not alter or create any radiological release paths. Therefore, neither the probability of occurrence nor.the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications maintain the pressure retaining capability of the valves and maintain conformance to the specified design codes and equipment qualification. Further, these changes will be implemented with the reactor core off-loaded and represents no added safe shutdown risk. Valve reliability is improved and all original design specifications are met by these changes. Therefore, these modifications do not create the possibility of an accident or malfunction of equipment of a different type than any previously evaluated.
III. No. The modified recirculation system discharge gate valves meet all original design specification requirements. Therefore, the margin i of safety cannot be reduced.
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y SE No.: 94-029 Source Document: DCP 94-0008, Rev. 0
. Description of Change This design change makes the following modifications to the Division 1 and 2 Standby Diesel Generators: adds two plugged boroscope inspection.
ports in the gearcase, lube oil pump adapter plate machining for clearance from the jacket water pump adapter plate (Division 2 only), and jacket water pump dowel pin changes. These modifications are consistent with the diesel manufacturer's recommendations as delineated in Service Information Memo (SIM) No. 388, Revision 1. In addition, the jacket water pump opening on each diesel will be enlarged and the gear carrier modified as recommended in the above referenced SIM. These latter changes will facilitate the installation and removal of the jacket water pump drive gear.
Summary I. No. This change will not adversely affect the reliability or availability of the onsite power supplies and therefore will not contribute to the potential of a station blackout. The redundancy and independence of the onsite power supplies is maintained.
Further, the standby diesel generators are used to mitigate Chapter 15 accidents / transients and do not prevent the occurrence of any initiating event. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The reliability and margin of safety of the original design is not compromised. Since this change is limited to the standby diesels and system design functions are not compromised, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. These modifications are fanufacturer recommended changes which do not adversely affect the reliability of the onsite power' supplies.
Therefore, the margin of safety as defined in the bases of Technical Specification Section 3/4-8 is not reduced.
SE No.: 94-031 Source Document: DCP 93-0113, Rev. O Description of Change This design change replaces the motor of the Limitorque actuator of the - )
Reactor Core Isolation Cooling (RCIC) system injection valve, lE51-F013, '
to increase the thrust / torque capability of the valve. The increased thrust / torque provides further assurance of proper valve function, accounting for various-uncertainties with respect to valve operational loading as identified in Generic Letter 89-10.
Summary I. No. This change maintains original system design requirements such that the original accident analyses are not affected. The design change adheres to established codes such that the pressure boundary integrity of the valve is not compromised. This design change improves the capability of the valve to perform its design safety functions as identified in Generic Letter 89-10. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. This modification maintains the original design requirements by providing adequate torque / thrust for the valve to perform its design functions. In addition, since this change does not affect the control or operation of the RCIC system and meets the requirements of the original construction codes, this change does not introduce a new potential for a malfunction of equipment important to safety.
Therefore, the modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the ASME code is maintained by the change. With this modification, the RCIC system will continue to provide the required reactor core cooling in the event of an isolation without requiring actuation of ECCS equipment. Technical Specifications 3/4.7.3, Table 3.3.7.4-1 and associated bases are not affected by the change. In addition, bases for Technical Specifications 3/4.5.1 and 3/4.5.2 for ECCS are not affected. Since the safety functions of the valve are maintained by the design change and the associated Technical Specifications and bases are unaffected, there is no reduction in the margins of safety.
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SE No.: 94-032 Source Document: LLJED 1-94-020 Description of Change ;
This Lifted Lead and Jumper / Electrical Device (LLJED) installs temporary power to portions of the non-Class 1E 480V power system to maintain power to non-essential.120V distribution buses during the refueling outage. A power feed is run from 480V panel h25-S006 to 120V panel R25-S122 via transformer R25-S128. A power feed is run from 480V panel R25-S006 to 120V panel R25-S012 via transformer R25-S024. A power feed is run from 480V panel 2R25-S158 to 120V panel 1R25-S017 via transformer 1R25-S023. i A power feed is run from 480V panel 2R25-S158 to 120V panel 1R25-S011 via transformer 1R25-S030. These panels provide power to Control Room panels, diesel panels and miscellaneous loads.
Summary I. No. This change provides temporary power to non-Class 1E panels during outage work activities while the normal supply is out of service for maintenance. The plant enters Off-Normal Instruction (ONI) R25-2 and R23-2 in response to loss of a non-essential 480V and 120V power during this time period. The operability of this nonsafety equipment does not affect the availability or operability of I equipment important to safety. Therefore, the probability of I occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is ,
not changed.
II. No. This temporary configuration does not c! oge the normal functions and operability / operator responses regarding equipment important to safety. The plant enters ONI R25-2 and R23-2 during this time period. Equipment important to safety remains operational and unaffected by this activity. Therefore, the possibility for an accident or malfunction of a different type than any evaluated in the USAR is not created.
III. No. This temporary power line-up supports operation in a manner equivalent to the normal power sources. Operability of the i remaining portions of the plant power system is unchanged.
Equipment and functions are unchanged. Margin of safety is not changed.
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SE No.: 94-033 Source Document: LLJED 1-94-013 Description of Change This Lifted Lead and Jumper / Electrical Device (LLJED) installs temporary power.to portions of the non-Class 1E 480V power system to maintain power to a non-essential 120V distribution buses during the refueling outage.
A power feed is run from 480V panel 1R25-S008 to 120V panel 1R25-S114 via transformer 1R25-S111. The panel provides power for the operation of the Public Address (PA) system.
Summary I. No. This change provides temporary power to a non-Class lE panel during outage work activities while the normal supply is out of service for maintenance. The plant enters Off-Normal Instruction (ONI) R25-2 and R23-2 in response to loss of a non-essential 480V and 120V power during this time period. The operability of this nonsafety equipment does not affect the availability or operability of equipment important to safety. Therefore, the probability of occurrence or the censequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not changed.
II. No. This temporary configuration does not change the normal functions and operability / operator responses regarding equipment important to safety. The plant enters ONI R25-2 and R23-2 during this time period. Equipment important to safety remains operational and unaffected by this activity. Therefore, the possibility for an accident or malfunction of a different type than any evaluated in the USAR is not created.
III. No. This temporary power configuration supports operation in a manner equivalent to the normal power source with the exception that the power source is not backed up by the diesel. A loss of this power system is similar to the analyzed condition of a LOOP /LOCA loss of the plant PA system. There is no other impact on PA system operation. Therefore, the margin of safety is not changed.
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SE No.: 94-034 Source Document: LIa7ED l-94-026-Description of Change This Lifted Lead and Jumper / Electrical Device (LIa7ED) installs temporary power to a portion of the non-Class lE 480V power system to maintain-power to a non-essential 120V distribution bus, 1R25-S060, during the refueling outage.
Summary I. No. This change provides temporary power to a non-Class lE panel during outage work activities while normal supply is out of service for maintenance. The plant enters Off-Normal Instruction (ONI) R25-2 and R23-2 in response to loss of a non-essential 480 and 120V power during this time period. The operability of this nonsafety
- equipment does not affect the availability or operability of equipment important to safety. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not changed.
II. No. This temporary line-up does not change the normal functions and operability / operator responses regarding equipment important to safety. The plant enters ONI R25-2 and R23-2 during this time period. Equipment important to safety remains operational and unaffected by this activity. .Therefore, the possibility for an accident of malfunction of a different type than any evaluated in the USAR is not created.
III. No. This temporary power configuration supports operation in a manner equivalent to the normal power source. Operability of the remaining portions of the plant power system is unchanged. Equipment and functions are unchanged. Therefore, the margin of safety is not changed.
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'SE No.: 94-035 Source Document: PTI-E12-P0002, Rev. 2, TC-1 PTI-E12-P0003, Rev. 2, TC-1 Description of Change These Periodic-Test Instructions (PTI) provide for-Residual Heat Removal (RHR) system Loop A and B heat exchanger performance monitoring.
These changes provide for heat exchanger testing with the Emergency Service Water (P45-ESW) system flow throttled. There is no impact on P45 system operability, since the associated ESW pump discharge valve is designed to fully open when the system receives an auto start signal.
Summary I. No. This activity collects data to enable an analysis to be performed on the capacity of the RHR Loop A and B heat exchangers. The RHR system is functional with all of the designed safety interlocks still active during the data collection process. The changes do not impact system control logic as described in USAR Chapter 7. These changes do not compromise the independence or redundancy of the onsite power supply or the ECCS systems. The probability of a multi-system maloperation is not increased as the reliability of the affected systems is maintained. Operational radiological conditions are not affected by the methods implemented by these modified instructions. These changes do not compromise the original equipment design and construction codes, or the equipment qualification. The changes will not compromise the ability of the ECCS systems to mitigate accidents. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment to safety previously evaluated in the USAR is not changed.
II. No. Testing performed according to these instructions will not exceed the current operating limits prescribed by approved system operating instructions. The instructions create no new systems, introduce no new equipment types, and maintain originally evaluated plant performance / capability. The instructions do not compromise the requirements of the original equipment design and construction codes, or the equipment qualification. Therefore, these instructions cannot create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
III. No. These instructions direct testing to be performed within the operating limits prescribed by existing, approved system operating instructions. The instructions do not compromia the requirements of the original equipment design and construction codes, or the equipment qualification. Therefore, no margin of safety has been reduced.
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SE No.: 94-036 Source Document: DCP 93-0116, Rev. O Description of Change _
Superseded by SE 94-192.
Source Document: DCP 90-0233, Rev. O Description of Change This design change involves the addition of a three-inch check valve to the Radwaste Building alternate return line to the Condensate Storage ,
Tank. This line is part of the Condensate Transfer and Storage system that transfers processed liquid to the Condensate Storage Tank for eventual plant re-use. This valve was added to this line to eliminate.
the effects of pressure surging and pipe movement that occurs when running the upstream High Pressure Core Spray pump for surveillance '
testing. ;
i Summary I. No. This valve meets or exceeds the requirements within the ANSI B31.1 Pcwer Piping Code, and meets or exceeds the design pressure and temperature conditions for the Condensate Transfer and Storage system. The involved system is only affected by the installation of this valve by slightly increasing the flow restriction through the return line. This has been evaluated to not be significant enough to adversely impact the Condensate Transfer and Storage system or those systems important to safety. The installation of this valve i is in compliance with all design, material and construction standards and practices. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The Condensate Storage and Transfer system serves no safety function. Failure or malfunction of the new check valve would not )
jeopardize the safe shutdown of the plant or adversely impact those !
systems required for safe shutdown. The proposed check valve meets i the applicable design codes and standards. Therefore, creating a !
new accident or malfunction of equipment is not possible. ,
III. No. The addition of this check valve is consistent with the current l piping design and does not affect system or plant operation. '
Therefore, this design change does not reduce any margin of safety.
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SE No.: 94-038 Source Document: DCP 93-0189, Rev. O Description of Change i
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.This design change installs three fiberglass access-manways and two j internal blanks in buried Service Water (P41) system yard piping. The ,
manways provide additional points of access into the system for j inspections and repairs. .The blanks isolate Unit 2 lines that are dead' 1 legs and no longer needed since Unit 2 construction has been terminated. !
The blanks serve to eliminate the potential of system / plant shutdown that might have been caused by a failure in either of these fiberglass lines. .,
The fiberglass piping in these lines was of indeterminate condition. l l
Summary I. No. The current operation and performance of the nonsafety-related. I Service Water system is not altered by this change. The blanked off lines increase system reliability. The P41 system is a nonsafety-related system that is not required for safe shutdown of the reactor. The fiberglass manway additions and blanks meet or exceed the requirements of P41 Line Specification R16-7 and the i plant specification to which the fiberglass piping components were. >
originally purchased. This change does not increase the probability of occurrence or the consequences of an accident or malfunction'of any equipment important to safety previously evaluated in the USAR.
II. No. All Unit 1 system flow paths and required flows will remain unaltered. System and plant operation remain unaltered. .As a result, there are no adverse effects possible to the P41 system or-to those systems which interface with it. Therefore, the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR is not created.
III. No. The operation of the P41 system is not addressed in Technical Specifications nor does the system affect the basis of any Technical Specification. The changes are consistent with the current piping design and do not affect system or plant operation in any way.
Therefore, this modification does not reduce any margin of safety.
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SE No.: 94-039 Source Document: DCP 93-0083, Rev. O DCP 93-0083A, Rev. O DCP 93-0083B, Rev. 0 Description of Change These design changes modify the Limitorque actuators of the Residual Heat Removal (RHR) pumps minimum flow valves, IE12-F064A/B/C, to increase the thrust / torque capability of the valves. The changes include gear, motor and power fuse size increases. The increased thrust / torque provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10. These modifications also result in stroke times changes.
Summary I. No. These changes naintain original system performance requirements such that the original accident analyses are not affected. The design changes adhere to established codes and standards such that the pressure boundary integrity is not compromised. These changes are an equipment improvement to assure the valves will perform the required safety function and do not reduce the reliability of the RHR pumps. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of ,
equipment will be increased.
II. No. These modifications do not reduce safety systems redundancy or independence since the changes does not create any new or altered interactions with other Emergency Core Cooling Systems (ECCS) or within trains of RHR. Postulated failures of the modified valve / operator assemblies will not compromise pressure boundary integrity or the safety function of the valves. In addition, no new permanent equipment types or new systems are introduced and original ,
RHR design functions are maintained. Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the ASME is maintained by these design changes. Although the stroke times of the RHR pump minimum flow valves have changed, these changes have no adverse affect on the USAR LOCA accident analyses since the RHR systems will provide the require ECCS flows. Since the required safety functions of these valves are maintained, these design changes will not result in i a reduction of any safety margins.
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SE No.: 94-040 Source Document: DCP 93-0100, Rev. O i Description of Change !
This design change replaces the existing Control Rod Drive (CRD) to ';
reactor vessel CRD housing cap screws and washers in order to reduce cap '
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screw stress corrosion cracking as identified in General Electric Service !
Information Letter (SIL) No. 483, Rev. 2. The new cap screws are of a [
higher strength material and the new washers are designed to reduce thes -
potential to trap water around the cap screws. This design change ,
provides for replacing at least four of the eight cap screws and washers .!
on each CRD.
Summary :
I. No. The new material meets all ASME Sections III and XI requirements for ,
material, manufacturing, quality and testing. The new design .
reduces the potential for cap screw cracking and thus, reduces the ;
potential for loss of reactor coolant pressure boundary due to-fastener failure. The postulated simultaneous failure of all' cap screws in USAR Section 4.6.2.3.2.2.3 remains unchanged by this i design change. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of ;
equipment will be increased.
a II. No. Since analysis has shown that two good cap screws can retain joint integrity, replacement of a minimum of four cap screws per flange ;
ensures structural integrity of the bolted joints. Thus, the !
reactor coolant pressure boundary integrity is maintained. The postulated failure scenarios considered in the USAR are not altered by this design change. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The application of redesigned cap screws and washers does not change any of the functional characteristics of the CRDs. This change represents a design improvement to prolong equipment life without affecting any safety margins or operational limits. Thus, there is no reduction in the margin of safety.
SE No.: 94-041 Source Document: MFI l-94-017 1 1
Description of Change j i
This Mechanical Foreign Item (MFI) installs floor drain plugs in floor drains around the Steam Bypass and Pressure Regulating (C85) system l l
hydraulic reservoir. This was done to preclude hydraulic fluid from entering the Turbine Building Lube Oil Area Sump in the event fluid was spilled during reservoir maintenance activities.
Summary l
1 I. No. The Turbine Building Lube Oil Area Sump does not automatically pump l down. The operation of the Floor Drain system and this sump are not I credited in the USAR analysis for Turbine Building flooding.
In the event the C85 reservoir catastrophically failed, the USAR flooding analysis remains bounding. Therefore, the probability of l occurrence or the consequences of an accident or malfunction of l equipment has not increased. l l
II. No. The USAR Turbine Building flooding analysis bounds the installation of these floor drain plugs. The USAR analysis allows for flooding i
up to five feet. The volume contained in the C85 reservoir is i l insufficient to cause this degree of flooding. Therefore, the l possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The operation of the Floor Drain system is not covered by the l Technical Specifications. The USAR Turbine Building flooding i l analysis bounds the installation of the drain plugs and the catastrophic failure of the C85 reservoir. Therefore, no margin of safety has been reduced. .
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SE No.: 94-042 Source Document: DCP 93-0089, Rev. O DCP 93-0089A, Rev. O Description of Change These design changes modify the gearing and motors of the Limitorque actuators of the containment spray second isolation valves, 1E12-F537A/B, to increase the thrust / torque capability of the valves. The increased, thrust / torque provides further assurance.of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10. Also, included in these changes is a decrease in electrical power fuse size to provide improved motor protection. These modifications also result in stroke time changes.
Summary I. No. These changes maintain original system design requirements such that the original accident analyses are not affected. The design change adheres to established codes such that the pressure boundary integrity of the valves is not compromised. These design changes improve the capability of the valves to perform their design safety functions as identified in Generic Letter 89-10. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications maintain the original design requirements by providing adequate torque / thrust for the valves to perform their safety functions. In addition, no new permanent equipment types or changes in valve functions are introduced. Thus, these changes do not introduce a new potential for a malfunction of equipment important to safety. Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the ASME code is maintained by these changes. Although the valve stroke times are changed, these changes are consistent with existing design basis stroke time requirements and do not adversely impact the valves' containment isolation function. Since the safety functions of the valves are maintained by the design changes, there is no reduction in the margin of safety.
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Source Document: -LLJED l-94-029 j Description of Change -I This Lifted Lead and Jumper / Electrical Device-(LLJED) installs temporary i power to a portion of:the non-Class lE-480V power system to maintain ;
power to a non-essential 120V distribution bus during the refueling e outage. A power feed is run from 480V panel 2R25-S009 to'the Refueling' Equipment Platform, 1F15-E0003.
Summary l
I. No. This change provides temporary power to a non-Class lE panel during i outage work activities while the normal supply is out of service for i maintenance. The plant enters Off-Normal Instruction-(ONI) R25-2. j and R23-2 in response'to loss of a non essential 480V and 120V power '
during this time period. No fuel movement occurs during this time period. The operability of this nonsafety equipment does not affect the availability or operability of equipment.important to safety.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously.
evaluated in the USAR is not changed.
II ~No. This temporary line-up does not change the normal-functions'and operability / operator responses regarding equipment important to.
safety. The plant enters ONI R25-2'and R23-2 during this time period. No fuel movement will occur during this time period.
Equipment important to safety remains operational and unaffected by.
this activity. Therefore, the possibility for an accident or i
malfunction of a different type than any evaluated in the'USAR is not created. I III. No. This temporary power configuration supports operation in a manner equivalent to the normal power source with the exception that the power is not backed by the diesel. A loss of this power system is !
similar to the analyzed loss of a stub bus during a LOOP /LOCA condition. There is no other impact on Fuel Handling system operation. The margin of safety is not changed. _
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SE No.: 94-044 Source Document: MFI 1-94-027 Description of Change This Mechanical Foreign Item (MFI) installs temporary air hose jumpers to maintain the Service Air system in service to the drywell and containment during testing of containment penetration P308.
Summary I. No. This MFI allows for the continued operation of the Service Air system to the containment and drywell during testing of containment penetration P308. The jumpers installed will cross-connect the Service and Instrument Air systems. The quality of air provided to the Service Air system will be better than the air normally supplied by the Service Air system. The complete loss of instrument air is analyzed in USAR Section 15.2.10. A loss of air due to a failure of these jumpers will not inhibit safety-related equipment from performing their safety function. Therefore, this MFI does not increase the probability nor the consequences of an accident or malfunction of equipment previously analyzed in the USAR.
II. No. The analysis of USAR Sections 9.3.1.3 and 15.2.10 analyze the loss of Service and Instrument Air systems. The installation of these jumpers remains bounded by the analysis presented in the USAR. All aspects of system operation remain unchanged, except for the supply path to the containment. Since the worst case failure is bounded by the USAR analysis, an accident or a malfunction of equipment important to safety of a different type is not created.
III. No. The Service and Instrument Air systems are not addressed by the Technical Specifications. The loss of these systems will not inhibit the ability of their loads from performing their intended safety functions. Therefore, the margin of safety is not reduced.
l SE No.: 94-045 Source Document: Physical Security Plan, Rev. 18 Description of Change This evaluation analyzes changes made to the Physical Security Plan (PSP). The changes have been evaluated to ensure that the effectiveness of the Perry Nuclear Power Plant Security Plan has not been reduced and to ensure that the requirements of 10CFR73, Physical Protection of Plants and Materials, are met. Site Protection must be contacted for further details since this is considered ' SAFEGUARDS' information.
Summary I. No. The PSP describes the comprehensive Physical Security Program and does not direct the operation of plant systems or equipment.
Therefore, the PSP changes do not affect the occurrence or consequences of an accident or malfunction of equipment.
II. No. The PSP does not direct the operation of plant systems or equipment and, therefore, does not create the possibility for an accident or malfunction.
III. No. The PSP changes do not reduce any margin of safety.
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Source Document: DCP 92-0128A, Rev. 0 '
Description of Change This design change adds a manual throttle valve in the 12' cross-tie-line between the Main and Auxiliary Condensers. This is the flow path presently used to pull a vacuum on the Main Condenser during startup.
This valve will provide throttle capability to decrease flow to the Steam Jet Air Ejectors. This will allow condenser pressure to increase which will raise condenser temperature. Raising condenser pressure during startup is needed to facilitate turbine rotor prewarming which has been recommended by General Electric. Prewarming the turbine rotors during plant startup reduces the likelihood of generating and or propagating material flaws that could lead to wheel burst or missile generation from the turbine. By performing this activity, the respective probabilities are decreased and turbine operation between required inspections can be extended.
Summary I. No. Use of the installed valve will allow plant operators to control condenser vacuum as desired to facilitate turbine rotor prewarming.
Operation of the valve is needed only during plant startup. Once the prewarming requirements have been met, the valve will be completely opened and remain in that position until the next plant startup. As a result, continued system and plant operation through power ascension up to full power will remain completely unaltered.
Evaluation has concluded that the use of this valve will not increase the probability of occurrence of applicable USAR evaluated accidents (i.e., Failure of Main Turbine Steam Jet Air Ejector Lines, Loss of Condenser Vacuum and Turbine Trip). Closure or over-throttling of the valve after turbine roll could cause a loss of condenser vacuum, but no more likely than the misuse of any other already existing system valve. If misuse of the valve is assumed, the worst case result would be a slow loss of condenser vacuum and potential turbine trip, if uncorrected. The associated consequences of these events would not be increased. All systems impacted by this change are nonsafety-related and are located in the nonsafety-related Turbine Building. This change does not increase the probability of occurrence or the consequences of an accident or nelfunction of any equipment important to safety previously evaluated in the USAR.
II. No. This modification provides a positive effect on the turbine since it will help decrease the probability of wheel burst / missile generation. The above referenced USAR evaluated accidents are the only potential events applicable to this change. As stated, the ;
systems affected by the use of this valve are all nonsafety which create no potential impact on equipment important to safety.
Therefore, no accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR is created by this change.
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SE No.: 4-046 (Cont.)
Summary (Cont.)
III. No. Effluents are still processed through the Offgas system since the flow path from the condenser to the Offgas system remains unaltered.
Safe hydrogen concentration in the Offgas system is not impacted since steam flow to the Steam Jet Air Ejectors is not altered.
Technical Specifications 3/4.11.2.6 and 3/4.11.2.7 remain unaffected. Therefore, no margin of safety has been reduced.
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i SE No.: 94-048 l Source Document: DCN 4451, Rev. 0 l Description of Change This' drawing change' revises P&ID D-302-353, Standby Diesel Generator Lube Oil System, to relabel the Division 1 Diesel Generator Oil Duplex Filter j; as R47D005A. i Summary I. No. This change re-identifies a piece of plant equipment. The change does not. alter any equipment or plant operating practice. l Therefore, neither the probability of occurrence nor the i consequences of a previously analyzed accident or malfunction of .
equipment will be increased by this change. l II. No. This change re-identifies a piece of plant equipment. The' change does not affect the operation of any equipment required for safe :
shutdown. Therefore, this change will not create the possibility of. l an accident or malfunction of a different type than any previously- !
evaluated. ;
III. No. This change re-identifies a piece of plant equipment. The change !
does not affect any equipment or operating practica relied upon by- l the Technical Specifications. Therefore, this change will not !
reduce any margin of safety. ;
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SE No.: 94-050 Source Document: DCP 93-0020, Rev. 1 Description of Change This design change provides testable flanged spacers / testable restricting orifices to replace the existing orifices on the Residual Heat Removal (RHR) system test return lines to the suppression pool. The new spacer / orifice design is equipped with a test connection to allow local leak rate testing of the flanged joint. This new arrangement eliminates the need for an external test box to perform the testing following joint assembly. l Sumraary I. No. This change adheres to existing design requirements and established codes and standards (ASME Section III Subsection NC) for the RHR ,
system. Excluding minor design differences necessary to allow l in-place testing, the new design is identical to the existing i design. The new design provides an improved means of ensuring the leak tightness of the mechanical joint to maintain containment integrity without degrading the capability of the RHR system to perform its design function. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The design change enhances containment integrity by providing an improved means of testing joint leak tightness and maintains the RHR test return line pressure boundary. The new design meets all original design criteria. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The new design ensures the maintenance of containment integrity under post LOCA conditions and maintains RHR system design standards. Thus, there is no reduction in the margin of safety.
SE No.: 94-051 Source Document: DCP-0084, Rev. O DCP-0084A, Rev. 0 Description of Change Superseded by SE 94-077.
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l SE No.: 94-052, 94-053 Source Document: DCP 93-0112, Rev. O !
DCP 93-Oll2A, Rev. 0 s
Description of Change I' This design change modifies the Limitorque actuator of service water to l cooling tower make-up line isolation valve,.OP41-F420, to increase the: !
torque cepability of the valve. The increased torque provides further j assurance of proper valve function, accounting for various uncertainties l with respect to valve operational loading as identified in Generic l Letter 89-10. The change is limited to the replacement of the motor '
l pinion and worm shaft gears. This modification results in a stroke time i change. l Summary I. No. This change is being implemented to increase the torque capability l for the valve actuator. The increased torque capability will ensure
! that the isolation valve can perform its design basis function to i close and limit flooding of the Turbine Building for postulated line 1 breaks. This change maintains original system performance !
requirements in that the revised valve can close in the required i response time of the transient analyses. The design change adheres to established codes and standards such that the pressure boundary ;
integrity is not compromised. Therefore, neither the probability of I l
occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
( II. No. This modification does not affect the control or operation of the
! P41 system and meets the original requirements of the construction l codes and equipment design. Therefore, this change does not create-the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Performance criteria for this valve are not included in the Technical Specifications or Operating License. The margin of safety assured by compliance to the ASME code requirements is maintained.
This design change will result in the valve operating with adequate torque margin during a design basis event. Thus, there is no I
reduction in the margin of safety.
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SE No.: 94-054 Source Document: PAP-1916, Rev. 4, TC-4 Description of Change This change to Plant Administrative Procedure (PAP) 1916 enhances the performance of the fire watch tours.
Summary I. No. The change is administrative in nature and is found to be consistent with the fire protection requirements of the USAR. The change adds additional documentation to ensure all fire watched locations are viewed. The PAP remains consistent with previously evaluated requirements. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously '
evaluated in the USAR is not increased. .
II. No. This change addresses the documentation of the fire watch rounds and is consistent with the fire protection requirements of the USAR.
The change does not impact any plant system or component.
Therefore, the possibility of creating an accident or malfunction different from any previously evaluated in the USAR does not exist.
III. No. This change is administrative in nature and does not impact any activity described in the Technical Specifications. Therefore, no margin of safety will be reduced.
SE No.: 94-055 Source Document: DCP 92-0049, Rev. O Description of Change This design change redesigns the High Pressure' Core Spray (HPCS): system standby Diesel Generator (DG) instrumentation tubing and supports.
I Summary I. No. This change replaces existing instrumentation tubing on the HPCS DG and starting air skid with seismically qualified instrumentation tubing and associated supports, and isolation valves to be used for instrumentation calibration activities. The qualified instrumentation tubing will be designed and installed in accordance with the appropriate specifications to ensure that the operation of the HPCS DG is not affected. Therefore, neither the probability of l
occurrence nor the consequences of a previously analyzed accident or j malfunction of equipment will be increased.
II. No. The qualification, reliability, availability, and operation of the HPCS DG remains unchanged by the implementation of this design change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than previously evaluated.
III. No. The installation of qualified instrumentation tubing, associated supports, and isolation valves on the HPCS DG does not affect the operation of the HPCS DG nor the HPCS system with regards to the requirements of the Technical Specifications. Therefore, this change will not reduce any margin of safety.
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- 94-056, 94-071,94-185 Source Document: DCP 88-0382, Rev. 1 DCP 88-0382, Rev. 3 SCRs 1-94-1138 and 1-94-1139 Description of Change This design change modifies the design of Hydrogen Analyzers, 1H51-P0022A/B, to improve the performance and reliability of the monitors.
Summary I. No. The hydrogen analyzers are used by plant personnel for indications as to when to take manual actions to reduce hydrogen concentration in the drywell and containment. The analyzers are seismic, Class 1E. The modifications are consistent with this criteria.
6 Code requirements related to tubing, fittings, valve installation have been maintained. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The modifications are limited to the enhancement of the hydrogen analyzer operation / reliability. The design criteria are being maintained. No new interfaces to other systems or equipment as a result of these modifications are generated. These modifications will not create new accident scenarios. Therefore, creating a new accident or malfunction of equipment is not possible.
'III. No. The modifications are limited to the enhancement of the hydrogen I analyzer operation / reliability. The design criteria are being maintained. Therefore, no margin of safety has been reduced.
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SE No.: 94-057 Source Document: TXI-0178, Rev. O Description of Change This temporary test instruction will operate the Residual Heat Removal (RHR) system in the shutdown cooling mode to evaluate the performance of the shutdown cooling to feedwater shutoff valves. The testing will meet the requirements of Generic Letter 89-10,
" Safety-Related Motor Operated Valve Testing and Surveillance". To facilitate this test, the RHR pump minimum flow valves will be electrically disabled to prevent transferring reactor inventory to the suppression pool when the feedwater shutoi' valves are closed. ;
i Summary I. No. Administrative controls are provided in this instruction to address !
the disabled features. No adverse affects wid occur as a result of l disabling the RHR pump minimum flow valve. The reliability or !
performance of the system will not be degraded as a result of performing this instruction. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction will be increased by this instruction.
II. No. All automatic functions and interlocks for the components operated under this instruction will remain operable with the exception of the minimum flow valve. These changes will not inhibit the system's ability to function as designed in an accident. Additionally, administrative controls are provided in this instruction to address the disabled features. Therefore, this instruction will not create.
the possibility for an accident or malfunction of a different type than previously evaluated.
III. No. System actuation setpoints required by Technical Specification are j not affected by this instruction. Operation of the system in this j instruction will not degrade performance or reliability. Therefore, i no reduction in the margin of safety will occur. i
SE No.: 94-058 Source Document: TXI-0195, Rev. O Description of Change This temporary test instruction will perform differential pressure (flow) testing for certain Emergency Service Water (ESW) system valves. The ESW Loop A will be started and the Residual Heat Removal (RHR) system heat exchanger inlet and outlet valves will be closed and then opened under l full flow conditions. The automatic opening of the heat exchangers inlet and outlet valves when ESW Pump A starts, will be defeated, so that test data can be obtained under controlled conditions in both the open and close directions. By defeating this interlock, the valves will not automatically open on a Reactor Core Isolation Cooling (RCIC) initiation or Division 1 diesel generator start signal. However, neither RCIC nor the Division 1 diesel generator require these valves to be open for operability.
Summary I. No. Automatic opening of the RHR heat exchanger inlet and outlet valves upon the receipt of a LOCA signal will not be defeated. If an <
accident occurs, the ESW system will operate as designed. Although I the heat exchanger inlet and outlet valves will not automatically open on a RCIC initiation nor Division 1 diesel generator start, flow through the heat exchanger is not required to support those l systems. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
l II. No. All equipment has been designed to operate under the conditions j established by this instruction. Adequate ESW flow will be verified before any interlocks are bypassed. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. All required Technical Specification actions will be met if needed.
No Technical Specification setpoints will be exceeded. Automatic I opening of the RHR heat exchanger inlet and outlet valves upon !
receipt of a LOCA signal will not be defeated. Therefore, no margin !
of safety has been reduced.
SE No.: 94-059 Source Document: DCP 93-0110, Rev. O DCP 93-0110A, Rev. O DCP 93-0110B, Rev. O I Description of Change l
I These design changes modify the Limitorque actuators of the Low Pressure l Coolant Injection (LPCI) valves, 1E12-F042A/B/C, to increase the I
thrust / torque capability of the valves. The increased thrust / torque provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter (GL) 89-10. These modification will result in stroke time changes. In addition, this modification provides larger power cables for 1E12-F042A/B to improve the valves' degraded voltage performance per the requirements of GL 89-10.
Summary I. No. These changes maintain original system performance requirements in that the revised valves can provide the required flow to the reactor vessel within the established response time of the accident analyses. The design changes adhere to established codes and standards such that the pressure boundary integrity is not compromised. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications do not reduce safety systems redundancy or independence since the changes does not create any new or altered interactions with other Emergency Core Cooling Systems (ECCS) or within trains of LPCI. Postulated failures of the modified valve / operator assemblies will not compromise pressure boundary integrity. In addition, no new permanent equipment types or new systems are introduced and original LPCI design functions are maintained. Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Although the stroke times of the LPCI injection valves have changed, these changes have no affect on the accident analyses. The LPCI system will still provide the required flow to the reactor within the required ECCS response time of Technical Specification 3/4.3.
This is consistent with Technical Specification Section 1.13 definition of ECCS response time. The containment isolation function as defined in Technical Specification 3/4.6.4 is not affected since this valve does not receive any containment isolation signals. Thus, there are no reductions in the margins of safety.
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i SE No.: .94-060,94-065 i Source Document: DCP 94-0037, Rev. 0 '
DCP 94-0037, Rev. 1 ;
Description of Change This design change modifies the underground fire protection'(P54) yard main piping to provide mechanical. joint pipe fittings at two locations.
The types of materials used in this design change will change the piping ,
line specification. The new fittings provide flexibility for pipe 3 movement / alignment in place of the current run of Yoloy. welded piping to !
each valve. The capacity of the underground mains are unchanged. The material meets the requirements of the design parameters of this fire protection piping and NFPA Standard 24, " Installation of Private Fire i Service Mains and Their Appurtenances'.
Summary !
I. No. This design change does not impact the design requirements of the !
Fire Protection system as defined within the USAR. Since this design did not change.the fire hazard in any plant area, the probability or consequences of a fire are not increased and the severity or nature of any fire related challenge to any systems important to safety or other plant feature is not affected.
Therefore, in the event of a fire, the ability to perform safe shutdown functions and to minimize radioactive releases to the environment will be maintained by the Fire Protection system.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
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II. No. The design change only impacts the Fire Protection system and is not l functionally related to any known accident mode or plant feature ,
important to safety. The Fire Protection system will maintain its ;
current design rating as defined within the USAR so the consequences ;
of a fire are unaffected. Therefore, the implen.entation of this design change cannot create any new accident or malfunction beyond those previously postulated within the USAR.
III. No. The consequences of a fire are unaffected by this change since the l Fire Protection system maintains its design rating as defined within ;
the USAR. Therefore, the applicable margins of safety regarding the '
plant and specifically concerning the Fire Protection system are not i affected. j
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SE No.: 94-061 Source Document: PTI-E12-P0005, Rev. O Description of Change
~ This Periodic Test Instruction ~ (PTI) will operate the Residual Heat Removal (RHR) system in the suppression pool cooling or the test return-mode to evaluate the performance of several RHR system valves. The control. circuitry of the RHR A/B pump minimum flow valves will be altered during portions of this PTI to allow testing of the valves near design basis, differential pressure. This testing will meet requirements of' Generic Letter 89-10.
Summary I. No. Operation of the RHR system in this instruction is nearly identical to the RHR system operation instruction. The only difference is that the automatic cycling of the minimum flow valve, IE12-F064A/B, based on low flow or high flow, is disabled. The instruction contains administrative controls to prevent adverse affects on the -
RHR system. Other than the minimum flow valve interlock, all actions specified in this instruction fall within the norr.nl scope of system operation. Therefore, neither the probability ot occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. The instruction contains administrative controls to prevent adverse affects on the RHR system. Other than the minimum flow valve interlock, all actions specified in this instruction fall within the normal scope of system operation. Therefore,-this change will not create the possibility for an accident or malfunction of a different-type than previously evaluated.
III. No. This instruction does not alter any RHR system characteristics, setpoints or functions. Administrative controls address disabling the minimum flow-valve interlock. The performance of this PTI will not degrade RHR system pump / components or their reliability. Since this system will continue to perform its design basis functions,.the-margins of safety will not be reduced.
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SE No.: 94-062 Source Document: DCP 93-0157, Rev. O Description of Change This design change modifies the Inclined Fuel Transfer System (IFTS) containment boundary bellows, 1F42-G001, by installing.a Bellows Test-Apparatus (BTA). The BTA allows isolation of the annular region between the Inclined Fuel Transfer Tube (IFTT) and the containment vessel penetration sleeve for test pressurization.
Summary I. No. Stresses of the IFTT and the BTA have been analyzed and found to remain within allowables of the GE and Pacific Nuclear stress reports which in turn meet ASME Section III, ANSI /ASME B3131 and AISC Code design allowables. Should a leak of the bellows occur, the Annulus Exhaust Gas Treatment System (AEGTS) will process the leakage by normal means. The IFTS, AEGTS and containment normal or emergency operations will be unaffected by this modification. This change will not alter or degrade the normal flow path of any system.
No accident evaluated in the USAR will develop more severe radiological consequences as a direct result of this change.
Therefore, the probability of occurrence or the consequences of any accident or malfunction of equipment important to safety previously evaluated in the USAR will not increase.
II. No. This modification does not change any equipment or equipment operation. Primary containment integrity is maintained. The BTA will not require additional interaction with other systems during normal or emergency operations. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated in the USAR will not be created.
III. No. The function and the configuration of the IFTS and the containment remain essentially the same. Primary and secondary containment pressure and structural integrity will be maintained. As such, no margin of safety will be reduced.
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SE No.: 94-063 Source Document: DCP 93-0088, Rev. O ;
DCP 93-0088A, Rev. O !
Description of Change i
These design changes replace the Limitorque actuators of the Main Steam
' Isolation Valve before seat drain line's inboard and outboard containment isolation valves, 1B21-F016-and F019, to increase the thrust / torque L capability of.the valves. Also, included in the changes are power fuse size upgrades consistent with the new operators. The increased thrust / torque provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10.
Summary I. No. These changes maintain original system performance requirements such .
that the original accident analyses are not affected. The design ,
changes adhere to established design codes such that the pressure !
boundary integrity of the valves is not compromised. These changes !
maintain equipment reliability and qualification and provide further '
assurance of containment integrity. Therefore, neither the l probability of occurrence nor the consequences of a previously l analyzed accident or malfunction of equipment will be increased.
II. No. These modifications create no new permanent equipment types or new !
systems, and maintain original plant design bases. Since these :
changes meet the requirements of the construction codes and original ;
equipment design, no new potential for a malfunction of equipment - ;
important to safety is introduced. Therefore, these modifications i do not create the possibility of an accident or malfunction of a i different type than any previously evaluated.
III. No. The margin of safety assured by the ASME code is maintained by these ;
design changes. Although the stroke times of the valves are increased slightly, these changes have no adverse affect on the )
containment isolation function of the valves as described in ;
Technical Specifications 3/4.3.2 and 3/4.6.4. Since the required. l safety functions of these valves are maintained, these design l changes will not result in a reduction of any safety margin. '
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L SE No.: 94-064 Source Document: DCP 93-0085A, Rev. O DCP 93-0085A, Rev. 1
' Description of Change This design change modifies the Limitorque gearing of the Low Pressure Core-Spray (LPCS) injection valve, lE21-F005, to increase the thrust / torque capability of the valve. The increased thrust / torque provides further assurance of proper valve function,_ accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10. The gear change results in a stroke time change. In addition, this change increases the power cable size for the valve electrical feed to obtain improved degraded voltage performance per Generic Letter 89-10 requirements.
Sumary -
I. No. This change maintains original system performance requirements in that the revised valve can provide the required flow to the reactor vessel within the established response time. The design change adheres to established codes and standards such that the pressure boundary integrity is not compromised. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. This modification does not reduce safety systems redundancy or independence since the change does not create any new or altered interactions with other Emergency Core Cooling Systems (ECCS).
Postulated' failures of the modified valve / operator assembly will not compromise pressure boundary integrity. In addition, no new permanent equipment types or new systems are introduced and original LPCS design functions are maintained. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Although the stroke time of the LPCS injection valve has changed, this change has no affect on the accident analyses. The system will still provide the required flow to the reactor within the required ECCS response time of Technical Specification 3/4.3. This is consistent with Technical Specification Section 1.13 definition of ECCS response time. The containment isolation function as defined in Technical Specification 3/4.6.4 is not affected since this valve does not receive any containment isolation signals. Thus, there are no reductions in the margins of safety.
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l SE No.: 94-066 Source Document: NR 94-S-069, Rev. O Description of Change This nonconformance report analyzes the temporary "use-as-is' disposition for the Emergency Service Water (P45) system having a pipe hanger support member that has stresses in excess of the ASME code allowables.
Summary I. No. A reanalysis of the P45 piping was performed assuming the hangers associated with the nonconforming support member would no longer function. The reanalysis showed that the D45 piping still complies with the ASME code. The reanalysis also showed that adjacent piping supports, though having increased loading, still satisfy their design intent. Overall, the P45 system's safety functions are not impacted. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Analysis shows that the P45 piping and adjacent piping supports still function as designed. The function of the P45 system has not been impacted. Therefore, the possibility of an accident or malfunction of equipment of a type different than previously evaluated is not created.
III. No. Analysis shows that the P45 piping and adjacent piping supports are still able to function under all design loading conditions. The integrity and functionality of the P45 system has not been compromised. Therefore, no margin of safety has been reduced.
SE No.: 94-068 Source Document: DCP 93-0097, Rev. 0 ,
Description of Change This design change adds orifice flanges and a restricting orifice-into +
the~ suction line and branch connections on the discharge-and suction '
piping of the High Pressure Core Spray (HPCS) Waterleg Pump,'1E22-C003.
This will provide the capability to accurately measure the differential .:
pressure across the pump.
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I. No. The overall HPCS system design configuration, component selection, !
system interaction, and operability will be unaffected. Therefore, '
neither the probability of occurrence nor the consequences of a . '
previously analyzed accident or malfunction of equipment will be -
increased by this change. ;
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II. No. The design configuration, component selection, fabrication, and installation meets or exceeds all design and installation standards established for this installation. Therefore, this change will not create the possibility for an accident or malfunction of a different ;
type than any previously evaluated. !
III. No. This change does not affect any equipment or operating practice I relied upon by the Technical Specifications. The HPCS system will l continue to operate as designed. Therefore, this change will not l reduce any margin of safety. '
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SE No.: 94-069 Source Document: DCP 94-5035, Rev. O Description of Change This change change installs connections to the Two-Bed Demineralizer and Distribution (P21) system to accommodate chemical backflush cleaning of each of the Residual Heat Removal (RHR) heat exchanger banks (1E12-B001A/C and 1E12-B001B/D). Added for each heat exchanger train are a block valve, a check valve (to ensure no introduction of the potentia 21y contaminated chemical cleaning fluid into the main process lines), a chut-off valve and associated piping.
Summary I. No. Implementation of this change will in no way impact the overall performan,'e of any system. No adverse system interactions will occur and 63 malfunctions of equipment will be introduced. No accident evaluated in the USAR will develop more severe radiological consequences as a direct result of installation of this change. The design simply adds connection paths for a chemical cleaning system but does not alter or degrade the normal flow paths of any system.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. This modification does not alter the normal flow path of the P21 system. The new valves / pipe / fittings to be installed will not interact with other systems during normal or emergency operations.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR will not be created.
III. No. Technical Specifications do not address the P21 system. The implementation of this modification in no way alters the normal or emergency functions of any system. As such, no margin of safety will be reduced.
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. SE No.:= '94-070 1
' Source Document: ' NR 93-S-200, Rev. 3-
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l SE No.: 94-072 i Source Document: DCP 93-0087, Rev. O DCP 93-0087A, Rev. O Description of Change These design changes replace the Limitorque actuators of the inboard and I outboard Reactor Core Isolation Cooling (RCIC) steam supply containment isolation valves, 1E51-F063 and F064, to increase the thrust / torque capability of the valves. Also, included in the changes are cable size upgrades consistent with the new larger operators. The increased thrust / torque provides further assurance of proper valve function, I accounting for various uncertainties with respect to valve operational j loading as identified in Generic Letter 89-10. 1 i
Summary I. No. These changes maintain original system performance requirements such that the original accident analyses are not affected. The design ,
changes adhere to governing design codes such that the pressure l boundary integrity of the valves is not compromised. These changes )
maintain equipment reliability and qualification. Therefore, I neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be !
increased. j II. No. These modifications create no new permanent equipment types or new I systems, and maintain original evaluated control functions and I operation of the RCIC system. Since these changes maintain original l equipment design requirements, no new potential for a malfunction of I equipment important to safety is introduced. Therefore, these l modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the ASME code is maintained by these ;
l design changes. These changes have no adverse affect on the safety l functions of the valves including containment isolation. Since the required safety functions of these valves are maintained, these .
design changes will not result in a reduction of any safety margin. l l
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SE No.: 94-073 Source Document: TXI-0184, Rev. O Description of Change This temporary test instruction details withdrawing a control rod using.
the electrical override method with a temporary solenoid. This method, which can be utilized to move a control rod out of the '00* position when other alternate control methods are unsuccessful, is based in part upon the " Electrical Override Method" as delineated in GEK-75598B.
Summary I. No. The performance of this instruction is limited to Operational Modes 4 or 5 with Technical Specifications 3/4.9.10, 3/4.1.1, 3/4.1.5, 3/4.9.1, and 3/4.9.2 being applicable depending upon actual plant conditions. USAR Section 15.4 discusses postulated accidents associated with the control rods. The only applicable accident evaluated in the USAR is the " Control Rod Removal Error During Refueling" and is described in Section 15.4.1.1. To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being located into the core. This requirement is backed up by refueling interlocks on rod withdrawal and movement of the refueling bridge. The electrical override method for control rod withdrawal does not operate the control rod in a configuration outside its design capability as described in USAR Sections 4.6.1, 3.9.4 and GEK-75598B. The pressures applied over and under the piston do not exceed nornal values. The pressure differential across the piston is limited to 525 psid which is less that the 550 psid setpoint of the in-line relief valve 1Cll-F0040. The in-line relief valve is provided to avoid exposing a drive to pressure that will cause the drive to move at excessive speeds. Additionally, because the EP-120 solenoid valve will be deactivated and as such closed during performance of this instruction, rod withdrawal speed will be slower than normal. Also, the ability of the control rod to scram when required is not affected by this evolution. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The electrical override control rod withdrawal method deactivates and bypasses the hydraulic control unit directional control solenoid valves for EP-120, EP-121, and EP-123. These activities will not i effect the safety system interlocks associated with control rod withdrawal. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
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1 Summary (Cont.) 1 III. No. The Technical Specifications associated with refueling activities
' ensures that all activities limit the possibility-of an inadvertent core criticality and that adequate cooling is maintained for the I reactor fuel. The reactor core is designed to remain subcritical j with the withdrawal of the most reactive control rod. This instruction does not alter the method in which the control rod is retracted from the core. The control rod electrical override method ;
is only performed on one control rod at a time. Performance of the instruction does not impact any system safety interlocks. l Therefore, no margin of safety has been reduced. !
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SE No.: 94-074 Source Document: MFI l-94-028 Description of Change This Mechanical Foreign Item (MFI) installs floor drain plugs in the -
Offgas Building. These floor drain plugs will be used to prevent Glycol from entering the Floor Drain system. The potential for Glycol leakage was created when maintenance was scheduled to be performed on the Offgas Cooler Condenser and on one of the Offgas system isolation valves.
Summary I. No. The Offgas area of the plant is not considered in the analysis for flooding in USAR Section 3.6.2.3.5.8 which states "no flood protection is required for any of these areas (Offgas)".
Additionally, the Offgas Building floor drain sumps are not considered in the USAR flooding analysis. Per USAR Table 3.6-3 neither the Offgas system nor the Floor Drain system are required to achieve safe cold shutdown of the plant. Therefore, the probability of occurrence or the consequences of an accident or a malfunction of equipment important to safety is not increased.
II. No. The installation of floor drain plugs represents a flooding hazard which is bounded by the flooding analysis in USAR Section 3.6.2.3.5.
Therefore, there are no new accidents or malfunctions created by this MFI. j III. No. The operation of the Floor Drain system is not covered by the l Technical Specifications. The USAR takes no credit for the Floor i Drain system to be in operation. Potential flooding is bounded by the flooding analysis contained in the USAR. Therefore, no margin j of safety is reduced. l
SE No.: 94-075 Source Document: GEI-0133, Rev. O Description of Change This generic electrical instruction provides temporary power in a plant outage for containment isolation dampers during a divisional power shutdown. The instruction provides specific prerequisites for the performance of the instruction, including containment integrity not required per Technical Specifications for the duration of performance.
Temporary jumpers are also installed to complete the logic interruption caused by out of service divisional isolation logic.
Summary I. No. This instruction will permit operation of the Containment Vessel and Drywell Purge (M14) system for the ventilation of the containment and drywell during outage periods which include a divisional power shutdown. Containment integrity is verified not to be required and the system is declared inoperable prior to performance of the instruction. No handling of irradiated fuel in the primary containment, no core alterations, and no operations with a potential to drain the reactor vessel are permitted to take place during the period of time in which this instruction is active. The opposite divisional power supply is still available and the associated valves are still fully functional for containment isolation purposes. The dropped fuel rod event cannot occur since fuel movement is prohibited. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The use of temporary power / logic for operation of one train of isolation dampers for M14 will occur at a time when containment integrity is not required. The opposite divisional power supply is still available and the associated valves are still fully functional for containment isolation purposes. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. For the duration of the performance of this instruction, containment integrity, as specified in the Technical Specifications, is not required. The opposite divisional power supply is still available and the associated valves are still fully functional for containment isolation purposes. With the supporting prerequisites fulfilled, the margin of safety as provided in the Technical Specifications and supporting documents remains unaffected.
SE No.: 94-076 Source Document: DCN 4404, Rev. 0 .
Description of Change This drawing change updates P&ID D-302-605, Nuclear' Boiler System, to reflect the ' locked open" position of the reference leg isolation valves associated with Reactor Vessel Level Purge Control system.
Sumary I. No. Implementation of this change updates P&ID D-302-605 only. The ,
change is editorial. Operability and potential effects on safety-related structures, systems, and components have been ,
evaluated separately by Safety Evaluations93-166 and 94-022. l Therefore, neither the probability of occurrence nor the l consequences of a previously analyzed accident or malfunction of equipuent will be increased by this change. ;
I II. No. Implementation of this change updates P&ID D-302-605 only. This a change is an editorial drawing change and does not affect the j operation of any equipment required for safe shutdown. Therefore, l this change will not create the possibility or an accident or :
malfunction of a different type than any previously evaluated. I III. No. This change is only an editorial drawing change. Operability and any potential effects on the basis for any Technical Specification has been evaluated by Safety Evaluations93-166 and 94-022.
Therefore, this change will not reduce any margin of safety.
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SE No.: 94-077 Source Document: DCP 93-0084, Rev. O DCP 93-0084A, Rev. O Description of Change These design changes modify the gearing of the Limitorque actuators of the containment spray isolation valves, 1E12-F028A/B, to increase the thrust / torque capability of the valves. The increased thrust / torque provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10. Also, included in these changes is a reduced electrical power fuse size to provide improved motor protection.
These modifications result in stroke time changes.
Summary I. No. These changes maintain original system design requirements such that the original accident analyses are not affected. The design change edheres to established codes and standards such that the pressure boundary integrity of the valves are not compromised. These design changes improve the capability of the valves to perform their design safety functions as identified in Generic Letter 89-10. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications maintain the original design requirements by providing adequate torque / thrust for the valves to perform their safety functions to open and close. In addition, no new permanent equipment types, new systems or new system functions are introduced.
Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the USAR LOCA accident analyses is not compromised by these changes. Although the valve stroke times have changed, these changes are consistent with existing design basis stroke time requirements and do not adversely impact the valves' containment isolation functions. Since the safety functions of the valves are maintained, there is no reduction in any margin of safety.
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' Source Document: MFI 1-94-030 Description of Change This Mechanical Foreign Item (MFI) installs temporary air hose jumpers to maintain the Instrument Air syrtem in-service to the drywell and containment during testing of containment penetration P306.
Summary I. No. This MFI allows for the continued operation of the Instrument Air system to the containment and drywell during testing of containment penetration P306. The air quality provided is unchanged from normal operating parameters. Additionally, the complete loss of instrument air is analyzed in the USAR Section 15.2.10. A loss of air due to a failure of these jumpers will not inhibit safety-related equipment from performing their safety function. Therefore, this MFI does not increase the probability or the consequences of an accident or malfunction of equipment previously analyzed in the USAR.
II. No. The analysis of USAR Section 15.2.10 is the bunding accident for the Instrument Air system. All aspects of system operation remain unchanged, except for the supply path to the containment. Since the-worst case failure is bounded by the USAR analysis and the quality of air provided is unchanged, an accident or a malfunction of equipment important to safety of a different type is not created.
III. No. The Instrument Air system is not addressed by the Technical Specifications. The end users of the air supplied (which may be Technical Specification related) will not see a difference-in the quality of air supplied. Therefore, no reduction in the margin of safety exists.
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SE No.: 94-079 Source Document: NR 94-N-099, Rev. 0
-Description of Change
~This nonconformance report provides the-following configuration changes.
to the Circulating Water (N7A) system: removes and scraps flow element 1N71-bG05, installs a pipe plug in the process pipe, and removes and scraps all associated instrument hardware.
Summary I, No. There is no safety function associated with flow element 1N71-N205 or its associated instrumentation. The Circulating Water system is designed nonsafety/non-seismic and previous system analysis has shown that a failure of this system will not compromise any safety-related system or prevent safe shutdown. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. Operetion of the Circulating Water system will remain as designed.
This change will not compromise any safety-related system or prevent safe shutdown. Therefore, this change will not create the possibility of an accident or malfunction of equipment of a different type than previously evaluated.
III. No. Operation of the Circulating Water system is not reliant upon the removed flow instrumentation. Therefore, this change will not reduce any margin of safety.
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Source Document: 'DCP 94-0036, Rev. O !
i Description of Change l t
'This design change adds four spectacle flanges, two in each loop, to the '
Emergency Service Water (ESW) system. The spectacle flanges will be located in the 3-inch bypass lines around manual valves IP45-F541A/B on the ESW outlets from the Emergency Closed Cooling Water (ECCW) heat i exchangers. During normal operations the spectacle flanges will be ,
installed ~in the open position. The flanges will be blanked for- l maintenance activities. ,
Summary i i
I. No. This design change maintains original system performance requirements and conforms to governing design codes. Seismic '
qualification of equipment and systems is naintained. The ch;nge does not provide a new pathway for radioactive releases. Accident analyses of the USAR are not affected by this change since original flow capability and pressure integrity of the ESW system are not i altered. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or salfunction of equipment will be increased. q II. No. These modifications maintain the pressure retaining capability of l the ESW system by conformance to design code requirements. This j change does not introduce any moving parts that could malfunction or !
fail. Operations of other safe shutdown systems are not altered by ;
this change and safe plant shutdown is not jeopardized. The :
spectacle flanges only provide a neans for isolation for future !
work. Therefore, this modification does not create the possibility I of an accident or malfunction of a different type than any previously evaluated.
III. No. There is no inpact upon Technical Specifications 3/4.7.1 and 3/4.7.2 as a result of this change. There is no affect on safety-related systens which interface with the ESW system. Therefore, the modification does not reduce any margin of safety.
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SE No.: 94-083 Source Document: NR 94-N-103, Rev. O NR 94-N-108, Rev. 0 l Description of Change These nonconformance reports evaluate replacing valve IN33-F0120 with the 1 same model valve (Rockwell model K3626M) due to seat leakage. There are i no differences between the two valves. Additionally, drawing errors were l identified which~ indicated the installed valve was a~1-1/2" valve when in "
t fact it was a 2' valve.
Summary j I. No. The two activities (valve replacement and accepting the drawing l error) are being performed on the Steam Seal (N33) Evaporator. !
make-up water supply bypass system. The replacement of the damaged l valve with the same valve will be performed in accordance with !
existing code requirements. The new valve will meet all pressure i and temperature requirement of N33. This installation and I acceptance of the drawing error are not related to any accidents or ;
transients addressed in the USAR. Radiological consequences are not l altered. Welding will be conducted in accordance with code !
requirements. The associated Steam Seal (N33). system is not [
considered " equipment important to safety". Therefore, the i probability of occurrence or the consequences of an accident or j malfunction of equipment has not increased. ;
II. No. The installation of this replacement nonsafety-related globe motor ,
operated valve is per the requirement of ASME/ ANSI B31.1. No new l pipe breaks will be introduced by this valve replacement. The i associated Steam Seal (N33) system is not considered ' equipment I important to safety". Replacement meets or exceeds the required ,
pressure and temperature. Therefore, no new accidents or l malfunctions of a different type than previously evaluated will be ;
created. !
III. No. The replacement valve is the same in design and will be installed in j accordance with applicable code requirements. Design rating of the i valve will meet those requirements set forth in the original' design.
Therefore, no margin of safety has been reduced.
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SE No.: 94-084 Source Document: DCP 92-0097, Rev. O Description of Change Superseded by SE 94-187.
SE No.: 94-085 Source Document: DCN 4496, Rev. O Description of Change This drawing change modifies P&ID D-302-242,- Service Air Distribution.
The item addressed in this change is the drawing symbol which depicted a specific valve a gate valve instead of a globe valve. The valve was confirmed to be a globe valve and design specifications confirmed the design.
Summary I. No. This drawing change is an editorial change to the P&ID for the Service Air system. The change corrects an error in the valve symbol for a valve. This change does not affect the Chapter 15 analysis, physically modify the plant, or affect the plant'r operation. Therefore, the probability of occurrence or the consequences of an accident or a malfunction of equipment important to safety previously evaluated has not changed.
II. No. This activity is an editorial change to documentation to correct the valve symbol on a P&ID drawing. The change will not affect the Chapter 15 accident analysis, modify the plant or alter its operation. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. This activity is an editorial change to system documentation. The change will not affect the Chapter 15 accident analysis, modify the plant or alter its operation. Therefore, no margin of safety has been reduced.
SE No.: 94-086 Source Document: TXI-0188, Rev. O Description of Change This temporary test instruction performs functional verification of the- ,
High Pressure Core Spray (HPCS) system as part of the System Operation and Test Review Program. The functional verification of HPCS is an integrated test using the methodologies of the preoperational test, the system operating instructions, and the existing surveillance tests.
Summary I. No. The HPCS system will be isolated from the reactor vessel during the initiation portions of this TXI by a tagged close isolation valve to prevent an injection into the vessel. During valve logic testing, the HPCS pump breaker will be racked out into the test position to prevent a start of the HPCS pump. Temporary modifications will be controlled in accordance with plant administrative procedures. At all other times, the HPCS system will be in a normal line-up as it would be according to the applicable operating instruction. The plant will be in a refueling condition and HPCS will not be required to be operable per Technical Specifications during performance of this testing. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not ;
increased. '
II. No. The HPCS system will be operated within its design basis throughout the performance of this test. The plant will be either in the refueling or 'at all times' mode. In addition, HPCS system will be isolated from the reactor vessel during system initiations and the HPCS pump will be racked into the test position during valve logic testing. Temporary modifications will be controlled in accordance with plant administrative procedures. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The plant will be in the refueling or "at all times" mode where the HPCS system is not required to be operable. All Technical Specifications will be adhered to during the performance of this testing. The HPCS system will be operated within its design basis. l Therefore, the margin of safety will not be reduced. -
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SE No.: 94-087 Source Document: DCP 93-0205, Rev. O Description of Change I
This design change replaces processing tubing and supports on the High l Pressure Core Spray (HPCS) Diesel Generator skid to ensure seismic -
Category I qualification of the affected lines.
Summary L I. No. This design change replaces skid mounted process tubing on the HPCS Diesel Generator with seismically qualified tubing and supports. l The revised tubing will maintain the design requirements of the l affected equipment and systems, and will meet the original material l and construction requirements of ANSI B31.1. The fundamental design and function of the HPCS Diesel Generator are not altered by this change. The performance and reliability of the diesel generator is not degraded by this change. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. This modification will ensure the skid mounted process tubing satisfies the design requirements. Compliance with established design and construction standards will be maintained. The reliability and performance requirements of the HPCS Diesel Generator are not impacted. Therefore, this modification does not create the possibility of a new accident or malfunction of a different type than any previously evaluated.
III. No. This modification maintains the performance and reliability of the HPCS Diesel Generator as an onsite power source as defined in Technical Specification 3/4.8. Therefore, the margin of safety cannot be reduced.
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94-089 Source Document: USAR Change Request 94-042 Description of Change This USAR change alters the Technical Staff and Managers (TSM) training.
program to satisfy the Engineering Support Personnel (ESP) Training.
Program. This change was driven by a new INPO Training Guideline, ACAD 91-17, which replaced ACAD 82-022 and ACAD 88-007. This change is consistent with USAR Section 13.2.1.c, which states that Perry's training programs " meet or exceed current INPO training guidelines." The USAR change also adds a section entitled " Management Supervisory Training
- to incorporate the provisions for the supervisory training that was in the original program but not in the new program.
Summary I. No. This change does not eliminate any training requirements, it just reassigns the requirements into different training programs. NRC training requirements contained in USAR Section 13 continue to be satisfied. Therefore, the probability of occurrence or the consequences or an accident or malfunction of equipment has not increased.
II. No. This change does not eliminate any training requirements, it just reassigns the requirements into different training programs. NRC training requirements contained in USAR Section 13 continue to be satisfied. Therefore, the probability of a new accident or malfunction of equipment of a type different than previously evaluated is not created.
III. No. This change does not eliminate any training requirements, it just reassigns the requirements into different training programs. NRC training requirements contained in USAR Section 13 continue to be satisfied. Therefore, no margin of safety has been reduced.
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l SE No.: 94-090 . l Source Document: MFI l-94-034 l Description of Change
! mis Mechanical Foreign Item-(MFI) allows'the operation of the Offgas. -
Building Exhaust Train'A without the charcoal adsorber installed. - W e j
operation of this exhaust train is necessary to support the surveillance !
testing of the Division I diesel generator. -
Summary ,
l I. No. The operation of the Offgas Building Exhaust Train A without'- I charcoal will be the same as with the charcoal except that j radioactive iodines will not be removed. This is not of concern -
because the plant is-in a shutdown condition and any radioactive .
iodine that may have been. generated during plant operation has since !
decayed away. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident cr :nalfunction of -
equipment will be increased.
II. No. The operation of the Offgas Building Exhaust Train A without charcoal while in a refueling outage and without a source term will not create the possibility for an accident or. malfunction of a different type than previously evaluated.
III. No. This MFI will not change the way the Offgas Building Exhaust Train A will be operated during plant operations. Since there is no source-term,-this change will not reduce any margin of safety.
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l SE No.: 94-091 Source Document: ONI-R10, Rev. 4 I Description of Change This change.to Off-Normal Instruction (ONI) R10 incorporates responses to 10CFR50.63, the station blackout rule. This instruction provides guidance to overriding the High Pressure Core Spray (HPCS) pump suction transfer to the suppression pool on high suppression pool level while maintaining the HPCS pump suction transfer to the suppression pool on low Condensate Storage Tank (CST) level in automatic.
Summary I. No. ONI-R10 is used in conjunction with the Plant Emergency Instructions to maintain RPV level within the USAR analyzed band, to maintain suppression pool temperature less than 185 F, and to restore the plant electrical system. For all analyzed transients, actions performed prevent any barrier (fuel, pressure vessel, or containment) from exceeding any of the design criteria. The analyses submitted demonstrate that no malfunctions of equipment important to safety are expected due to environmental conditions or '
equipment usage. Therefore, this instruction does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR.
II. No. The actions taken preclude the possibility of containment failure since RPV pressure is maintained less than the analyzed limit.
Diesel generators are operated within their design bases.
Therefore, the possibility of an accident or a nelfunction of equipment of a different type than evaluated in the USAR is not created.
III. No. During a loss of AC power, a number of Technical Specification parameter limits may be exceeded. This is due to the loss of various systems due to the loss of electrical power. In all cases, the actions taken will restore Technical Specification parameters and complete the required Technical Specification actions as soon as power availability permits. Both the Emergency Procedure Guidelines and Perry's Station Blackout submittal have been reviewed by the NRC and found acceptable. Therefore, since the appropriate Technical Specification actions are taken, the margins of safety are maintained.
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SE No.: 94-092 Source Document: DCP 93-0116, Rev. 1 Description of Change Superseded by SE 94-192.
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Source Document: DCP 94-0019, Rev. O Description of Change This design change installs a digital strip chart recorder and analog signal conditioners on the Generator Recorder Panel in the Main control Room. The recorder monitors generator negative sequence current and generator output frequency. An alarm is generated if values are outside desired operating limits.
Summary I. No. The equipment is used for monitoring purposes only and provid's no control functions. The equipment is installed in accordance with electrical separation requirements and adds little heat load to the panel. Potential failure modes were examined including short circuits, open circuits, operator errors, software failures and failures due to Electromagnetic Interference (EMI). None of these failures will cause a generator load rejection or any other accident analyzed in the USAR. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. The equipment is used for monitoring purposas only, not control functions. Equipment failures including those due to EMI have been ,
analyzed for impact on the generator system. Operator errors have been addressed. None of these failures will cause an accident or Halfunction beyond those discussed in the USAR. There is no possibility'for creating an accident or malfunction of a different type than any evaluated previously in the USAR.
III. No. The equipment is used for monitoring purposes only and does not affect systems monitored per Technical Specifications. The parameters monitored are not described in Technical Specifications.
No Technical Specification actions are required and generator operation is not degraded. The margin of safety is not changed.
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SE No.: 94-094 Source Document: LLJED l-94-036 Description of Change This Lifted Lead and Jumper / Electrical Device (LLJED) will permit operation of the Containment Vessel and Drywell Purge (M14) system in the refuel mode with the 18-inch isolation valves in the supply path closed.
Ventilation air will still be provided by the 42-inch supply path while-actuator maintenance is conducted on the 18-inch valve operators.
Summary -
I. No. This LLJED simulates the open position of the 18-inch supply' path valves to the supply fans in order to permit fan operation. The permissives are provided to prevent operation of the fans without the appropriate valve line-up. The LLJED does not create any change in the flows external to containment (i.e., no change in the filtered release path). Containment integrity is relaxed during the installation and use of the LLJED. The 42-inch supply path valves are fully functional for isolation purposes. Additionally, the valves affected are in the safe position (closed) for any event.for which the M14 system is required to mitigate the consequences. . No temperature changes are expected to occur as a result of this modification. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The M14 valves are required to close to prevent / minimize the release to the environment in case of a radiological event. The 42-inch supply path valves remain fully functional for isolation purposes.
The 18-inch supply valves will be maintained in a closed position.
Therefore, there is no impact upon any event for which the Mid i system is required to mitigate. The 42-inch isolation valves are l rated for closure against full flow and are fully functional.
Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. For the duration of this LLJED, the M14 system is inoperable, yet is j in a configuration such ti 2 the system can fully perform its i intended isolation function if required. Therefore, no margin of i safety has been reduced.
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SE No.: 94-095 Source Document: NR 94-N-177, Rev. 1 NR 94-N-234, Rev. O Description of Change These nonconformance reports disposition cracks on three extraction .
steam bellows located in the upper compartments of the main condenser.
This "use-as-is" disposition will be used until the next main condenser inspection.
Summary I. No. The Extraction Steam (N36) system is nonsafety-related and is not required for safe shutdown. The cracks on the bellows have been inspected and found to pose no current plant operational problems.
The cracks will be re-inspected during a mandatory main condenser inspection (next refueling outage) to ensure they did not propagate.
Complete failure of the bellows will not adversely affect the overall operation of this system, nor will it adversely impact any other plant system. The Loss of Condenser Vacuum (USAR Section 15.2.5) and Loss of Feedwater Heating (USAR Section 15.1.1) I transients are not impacted. Therefore, neither the probability of '
occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The bellows and associated piping / equipment will continue to perform the design function with the identified cracks. The potential steam leakage from a complete failure of the bellows will be contained by the main condenser and will not affect any safety-related equipment.
Therefore, this disposition will not create the possibility for an accident or nalfunction of a different type than any previously evaluated.
III. No. The Extraction Steam system is nonsafety and is not addressed within the Technical Specifications. This system and the main condenser will continue to function as required. No Technical Specification related equipment will be adversely affected by this disposition.
Therefore, no margin of safety will be reduced.
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-t O j SE No.: 94-096 Source Document: DCP 93-0082, Rev. O j t
Description of Change l I
This design change modifies the Limitorque gearing of the High Pressure !
Core Spray (HPCS) injection valve, 1E22-F004, to increase the thrust / torque capability of the valve. The increased thrust / torque ;
'provides further assurance of proper valve function, accounting for i various uncertainties with respect to valve operational loading as )
identified in Generic Letter 89-10. This modification results in a i stroke time change. l.
Summary I. No. This change maintains original system performance requirements in that the modified valve can provide the required flow to the reactor !
vessel within the established response time of the accident ;
analyses. The design change adheres to established codes and .
standards such that the pressure boundary integrity is not !
compromised. Therefore, neither the probability of occurrence nor l the consequences of a previously analyzed accident or malfunction of i equipment will be increased. !
II. No. This modification does not reduce safety system redundancy or l independence since the change does not create any new or altered l interactions with other Emergency Core Cooling Systems (ECCS). l Postulated failures of the modified valve / operator assembly will not j compromise pressure boundary integrity. -In addition, no new i permanent equipment types or new systems are introduced, and l original HPCS design functions are maintained. Therefore, this ;
modification does not create the possibility of an accident or- 1 malfunction of a different type than any previously evaluated.
III. No. Although the stroke time of the HPCS injection valve has changed, this change has no affect on the accident analyses. The system will still provide the required flow to the reactor within the required ECCS response time of Technical Specification 3/4.3. This is consistent with Technical Specification Section 1.13 definition of ECCS response time. The containment isolation function as defined in Technical Specification 3/4.6.4 is not affected since this valve does not receive any containment isolation signals. Thus, there are no reductions in the margin of safety.
SE No.: 94-097 i Source Document: DCP 93-0082A, Rev. O Description of Change This design change modifies the Limitorque actuator of the High Pressure Core Spray (HPCS) minimum flow bypass valve, lE22-F012, to increase the thrust / torque capability of the valve. The increased thrust / torque i provides further assurance of proper valve function, accounting for i various uncertainties with respect to valve operational loading as '
l identified in Generic Letter 89-10. This modification results in a stroke time change.
Sumirary l I. No. This change maintains original system performance requirements such that the original accident analyses are not affected. The design change adheres to established codes and standards such that the pressure boundary integrity of the valve is not compromised.
Equipment reliability and qualification are maintained by this design change. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
l II. No. This modification does not reduce safety system redundancy or independence of the Emergency Core Cooling Systems (ECCS). Original -
equipment design margins and system functions are maintained such that a new potential for a malfunction of equipment important to safety is not introduced. In addition, no new permanent equipment types or new systems are introduced. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the USAR LOCA accident analysis is not compromised by this change. Although the valve stroke time specified in the USAR has changed, this change does not adversely impact the valve's containment isolation function as defined in Technical Specification 3/4.6.4. Since the safety functions of this valve are maintained by this design change, there is no reduction in the margin of safety.
l SE No.: 94-098 i Source Document: DCP 93-0086, Rev. 0 l DCP 93-0086A, Rev. O i I
Description of Change l 7
These design changes' replace the Limitorque actuators of the inboard and i outboard Reactor Water Cleanup system containment isolation valves, l 1G33-F001 and F004, to increase the thrust / torque capability of the ;
valves. Also, included in the changes are power fuse and cable size ;
upgrades consistent with the new larger operators. The increased (
thrust / torque provides further assurance of proper valve function, :
accounting for various uncertainties with respect to valve operational ,
loading as identified in Generic Letter 89-10. These modification result in stroke time changes. j i
Summary i
I. No. These changes maintain original system performance requirements such !
that the original accident analyses are not affected. The design changes adhere to established codes and standards such that the pressure boundary integrity of the valves is not compromised. These changes naintain equipment reliability and qualification.
Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications create no new permanent equipment types or new systems and maintain original plant design bases. In addition, these changes meet the requirements of the construction codes.
Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the ASME code is maintained by these design changes. Although the stroke times of the valves are increased, these changes have no adverse affect on the containment ;
isolation function of the valves as described in Technical i Specifications 3/4.3.2 and 3/4.6.4. Since the required safety l functions of these valves are maintained, these design changes will j not result in a reduction in any safety margin. l l
SE No.: 94-099 Source Document: NR 94-N-230, Rev. O NR 94-N-301, Rev. O Description of Change These nonconformance reports evaluate the repair of several ASME test connections that were discovered broken off from where they were originally attached. These instrumentation lines were socket welded to the inner shells of the high, intermediate and low pressure condensers to aid in determining turbine efficiency. The broken connections will be plugged and seal welded closed to prevent any unnecessary air in-leakage to the condensers.
Summary I. No. The purpose of the ASME test lines is for performing ASME turbine i efficiency test and they have no function during normal or abnormal plant conditions. The repair of the existing connections will not alter the operation of the condenser or Extraction Steam system as described in the USAR. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The design integrity of the condenser or extraction steam system is not adversely impacted with tne repair specified within the associated nonconformance reports. Safety-related equipment is completely unaffected by this disposition since the condensers are not required to support the safe shutdown of the reactor.
Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. Since the nonconformance reports disposition will not affect plant operation or affect safety-related equipment, there is no affect on any safety margin.
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.SE No.: 94-100 Source Document: DCP 93-0065, Rev. 0 >
Description of Change l This design change modifies the electrical power circuit of the Residual
- Heat Renoval (RHR) -suppression pool test -return-valve, lE12-F024A, to , ,
improve the degraded voltage capability of the valve. The changes l consist of increasing effective size of the power feed conductor. The ;
increased conductor size provides further assurance of proper valve i function at degraded voltage conditions as identified in Generic i Letter 89-10.
Summary i I. No. This change maintains original system performance requirements such !
that the original accident analyses are not affected. The !
modification improves the valve's ability to operate under accident ,
conditions and reduces the probability of the valve failing to close ,
under degraded voltage conditions. The increase in conductor size ;
does not alter the protection provided to ensure the valve will open to support safe shutdown in the event of a fire. In addition,.the ;
change maintains conformance to the appropriate design codes, i Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of l equipment will be increased.
II. No. This modification does not reduce the redundancy or independence of the RHR system since the changes maintain original RHR design functions and equipment design margins for power circuits. In -
l addition, no new permanent equipment types or new systems are introduced. Therefore, this modification does not create the l possibility of an accident or malfunction of a different type than i any previously evaluated.
III. No. This modification maintains automatic closure of the valve under LOCA accident conditions and does not alter valve stroke times. The change does not adversely affect the ability of RHR to provide suppression pool cooling as required by Technical Specifications.
Since the required safety functions of this valve are maintained, the design change will not result in a reduction of any safety margin.
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SE No.: 94-101 Source Document: TXI-0198, Rev. O t Description of Change This temporary test instruction provides for the chemical backflush cleaning of the Residual Heat Removal (RHR) heat exchanger A train. The process will apply an environmentally safe corrosion inhibitor. A 12%
formic acid solution will then be introduced and circulated through the heat exchanger train for 16-24 hours. This process will dissolve calcium carbonate scale deposits.
Summary I. No. Chemical cleaning of the heat exchangers will enhance overall RHR system performance. No adverse system interactions will occur.
Therefore, the probability of occurrence or the consequences of an accident or nalfunction of equipment important to safety previously evaluated in the USAR will not be increased.
II. No. This instruction involves a chemical cleaning process of the heat exchangers only, and once performed all systems will be restored.
The instruction will not alter the normal operation of any system.
No additional interaction with other systems during normal or emergency operations will result due to execution of this cleaning process. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated in the USAR will not be created.
III. No. Performance of this cleaning process will not impact the Technical Specification requirements of any system. No change to Technical Specifications related to the chemical cleaning of the RHR heat exchangers is required. As such, no margin of safety will be reduced.
SE No.: 94-102 Source Document: MFI 1-94-037 Description of Change This Mechanical Foreign Item (MFI) provides for the temporary installation of a nonsafety, manual three-way valve in the air supply line to the Control Room Inboard Isolation Damper B, OM25-F020B. This valve will replace the installed solenoid valve in order to permit operation of the OM25-F020B damper during the divisional power outage on the Division 2 power supply.
Summary I. No. This temporary modification will be installed with the plant in Mode 5, all fuel removed from the reactor vessel. The requirements of Technical Specification 3/4.7.2.b.2 apply for the duration of this installation. The Control Room Ventilation (M25/26) system is inoperable due to the divisional outage on the Division 2 power supply. However, in the event that an automatic initiation of the M25/26 system should occur, isolation is still possible due to the availability of the Division 1 logic. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not been increased.
II. No. The MFI is installed with the M25/26 system declared inoperable.
The system isolation function can still occur due to the availability of the Division 1 logic. The activity is being performed with the plant in Mode 5, all fuel removed from the reactor vessel. Control Room habitability is not affected.
Therefore, the possibility of an accident or malfunction of equipment of a type different than evaluated is not created.
III. No. The M25/26 system ensures the radiological dose to the Control Room operators is maintained within the requirements of 10 CFR 50, Appendix A, GDC 19. This temporary modification does not adversely impact the ability of the M25/26 system to accomplish this. System isolation capability is maintained. Therefore, the margin of safety is not reduced.
SE No.: 94-103 Source Document: NR 94-S-291, Rev. O Description of Change l
l This nonconformance report evaluates the 'use-as-is' disposition of a l l bent hydraulic line to the 1B33-F060B valve actuator. ;
l' Summary I. No. The 'use-as-is' disposition of the hydraulic piping can at worst case lead to the accumulation of air in the hydraulic system and cause the valve to operate with delay or operate erratically. The accidents addressed in USAR Chapter 15 are concerned with rapid valve movement. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. Although the response time of the valve will be slower, the hydraulic piping and associated equipment will function as designed.
Therefore, this change will not create the possibility for an accident or malfunction of a different type than previously evaluated.
III. No. The nonconformance report has evaluated the condition and concluded that the condition would not affect the valve's ability to. comply with Technical Specification requirements. Therefore, this disposition will not reduce any margin of safety.
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SE No.: 94-104 Source Document: DCP 94-0072, Rev. O Description of Change i i
Superseded by SE 94-168. 1 l
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SE No.: 94-105 Source Document: NR 94-S-252, Rev. 1 NR 94-S-253, Rev. 1 Description of Change These nonconformance reports evaluate the disposition of check valves 1Pll-F545 and 1P43-F721. The disposition provides for the replacement of the valves with equivalent check valves of an improved design.
l Summary I. No. The disposition of valve replacement maintains original system design and performance requiremerd.s such that contsinment integrity and containment isolation capability are not altered. Plant accident analyses are not affected. The replacement valves conform to ASME Section III, subsection NC requirements and thus ensure pressure boundary integrity. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The disposition maintains original system design margins and original system functions. In addition, these valve changes create no new systems and introduce no new permanent equipment types.
Thus, the disposition does not introduce any new potential for a malfunction of equipment / systems. Therefore, these changes do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The replacement valves conform to original design code requirements and thus, maintain the safety function of the affected valves.
Therfore, no reduction in the nargin of safety exists.
Source Document: NR 94-N-324, Rev. O Description of Change ]
t This nonconformance report evaluates the repair of an ASME test i connection that was discovered broken off from where it was originally .
attached. This instrumentation line was socket welded to the inner shell of the intermediate pressure condenser to aid in determining turbine .
I efficiency. The broken connection will be plugged and seal welded closed to prevent any unnecessary air in-leakage to the condenser. 1
-Summary j I. No. The purpose of the ASME test line is for performing the ASME turbine efficiency test. The line has no function during normal or abnormal ;
plant conditions. The repair of the existing connection will not ;
alter the operation of the condenser or the Extraction Steam system l as described in the USAR. Therefore, the probability of occurrence l or the consequences of an accident or nalfunction of equipment has not increased.
II. No. The design integrity of the condenser and the Extraction Steam system is not adversely impacted with the repair specified within the associated nonconformance report. Safety-related equipment is completely unaffected by the proposed disposition since the condensers are not required to support the safe shutdown of the reactor. Therefore, creating a new accident or malfunction of I equipment that has not been previously evaluated is not possible.
III. No. Since the proposed nonconformance report disposition will not affect plant operation, there is no affect on any safety margin.
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SE No.: 94-107 Source Document: TXI-0201, Rev. O Description of Change !
This temporary test. instruction provides the directions necessary to ,
perform dynamic' flow testing of the High Pressure Core Spray (HPCS) l injection valve, lE22-F004, in accordance with Generic Letter 89-10. The testing will evaluate the performance of IE22-F004. The test will be. ';
performed with the reactor flooded, therefore the auto closure of the valve on high reactor level (level 8) will be defeated to perform the test. Also, to preserve the Condensate Storage Tank (CST) as the pump suction source during testing, the automatic suction transfer to the suppression pool will be defeated.
Summary I. No. HPCS operation via this instruction will be similar to normal and emergency operation. The HPCS injection valve, lE22-F004, will not auto close on a level 8 signal, however. This interlock is to be defeated to allow stroking of the injection valve. Also, an automatic suction source transfer from the CST to suppression pool ;
will not occur since the HPCS suppression pool suction valve, f lE22-F015, is overridden closed. All other automatic functions of HPCS will remain operational during the performance of this test. I Administrative controls are provided in the test to ensure no adverse affects occur as a result of the defeated interlocks.
Therefore, neither the probability of occurrence nor the '
consequences of a previously analyzed accident or malfunction of ,
equipment will be increased by this change.
II. No. This instruction provides administrative controls to ensure that no adverse affects occur as a result of the defeated interlocks. All 4 automatic functions and interlocks for the components operated under this instruction will remain operable with the exception of the ,
injection valve level 8 auto closure and the automatic suction source transfer. Therefore, creating a different accident type or -
malfunction of equipment previously evaluated is not possible.
III. No. System actuation setpoints required by Technical Specifications are not directly effected by this instruction. The HPCS system will be inoperable during the performance of this test which satisfies the requirements of Technical Specification 3/4.3.3, Action b (Action 34).. The system will function as required by Technical Specifications, with the exception of the level 8 auto close interlock and the automatic suction source transfer. The operation of the HPCS system as previously discussed will not result in degraded system reliability. Therefore, no reduction in the margin of safety will occur.
SE No.: 94-108 Source Document: DCP 93-0017, Rev. O DCP 93-0017, Rev. 1 Description of Change This design change replaces a blind flange on the 6' High Pressure Core Spray (HPCS-E22) system flush line with a coupling connection. This configuration will enable quick alignment of a connection to a fire hydrant on the Fire Protection (P54) system using 5' fire hoses. This will allow for the use of the fire protection water supply from the diesel driven fire pump as a back-up water source for the reactor vessel.
The piping in the flush line is safety-related, designed as ASME Class 2.
The line is normally empty, at ambient temperature and pressure. The new fitting changed this classification to nonsafety and the pressure rating of this section of pipe changed from 900 psig to 175 psig at the flange connection. There is a normally closed valve and a check valve between the pressurized piping of the E22 system line and the part of the line modified by the new connection. This double isolation meets the ASME Section III requirements for separation of safety classes and is adequate isolation between high pressure and low pressure pipe classifications.
The additional weight added by the new connection has been analyzed and will not adversely impact the support of these valves for seismic events.
This modification will not adversely affect the operation of the HPCS system. The use of the back-up water supply for HPCS will not impact available water for fire fighting as is evaluated in the USAR.
Summary I. No. The separation of the high and low pressure piping with two independent means of isolation will withstand expected operating and accident initiated pressures. The increased weight due to the modification will not adversely impact the support of these isolation valves to withstand the safe shutdown earthquake. The modification does not reduce the ability of the HPCS system to function in a manner required for safe shutdown. Therefore, the probability of occurrence or consequence of any accident or a malfunction of equipment important to safety as evaluated in the USAR is not increased.
II. No. This modification involves utilizing existing plant systems in their design configurations, operating within existing design limits. The required operating modes and functions of the HPCS system involved in this change as they relate to safe shutdown are not different than previously evaluated in the USAR. Therefore, the modification does not create an accident or malfunction of equipment important to safety of a type different than previously evaluated.
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Summary (Cont.)
, III. No. The connection modified by this design change is not:part of the l operation-of the HPCS system as required for safe shutdown. The.. ;
flushing function is not required for system operability as defined in Technical Specifications. The separation of the high and low t pressure piping with two independent means of isolation will prevent :
any interaction between the modified portion of piping and the parts of the HPCS required to function to achieve safe shutdown. The use of the back-up water supply for HPCS will not impact available water for fire fighting as evaluated in the USAR. Therefore, the change does not reduce any margins of safety.
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SE No.: 94-109 Source Document: USAR Change Request 94-025 r
Description of Change [
This USAR' change adds a valve control, acronym to two valves on a table. l The control of the two valves has not changed. j Summary .
1 I. No. The plant configuration has not changed. This change is editorial i in nature. It documents the current design and the as-installed ;
configuration. No field work is required as.a result of this change l which brings the USAF in line with the existing design. Therefore, ;
the probability of ot;currence or the consequences of an accident or i a malfunction of equipment important to safety is not increased. i II. No. This change is editorial in nature. The valves in question still !i satisfy all design requirements. No change to the plant has occurred. Therefore, the possibility of an accident or a malfunction of a different type than any previously evaluated in the USAR is not created. l III. No. This change is consistent with existing Technical Specifications and l the current design bases. Therefore, the margin of safety will not l be reduced. l i
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i SE No.: 94-110 Source Document: TXI-0199,-Rev. 0 '
Description of Change ,
This temporary test instruction perforns injection into the reactor vessel-to evaluate the performance of the Low Pressure Coolant l Injection (LPCI) valve, lE12-F042B, under flow conditions. The testing will meet requirements of Generic Letter 89-10. To facilitate this test, the RER pump B minimum flow valve will be electrically disabled to prevent transferring reactor inventory to the suppression pool when the ;
injection. valve is closed. The instruction will also verify system flow l rates eith the LPCI B injection valve throttled. To accomplish this the a
" seal-in' logic 'or the valve will be disabled. :
Summary i
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I. No. Injection into the reactor vessel in this instruction is acceptable based on engineering evaluation. No adverse affects will occur re14tive to reactor internal components. Administrative controls Are facluded to address disabling the RHR B pump minimum flow valve.
th. reliability or performance of the valve will not be degraded as a result of the throttling application. Therefore, neither the probability of occurrence nor the consequences of a previously ,
analyzed accident or malfunction will be increased by this '
instruction.
II. No. All automatic functions and interlocks for the components operated under this instruction will remain operable with the exception of th! minimum flow valve and the injection valve seal-in logic. These j changes will not inhibit the system's ability to function as designed in an accident. Additionally, administrative controls are provided in this instruction to address disabled features.
Therefore, this instruction will not create the possibility for an accident or malfunction of a different type than previously !
evaluated.
III. No. System actuation setpoints required by Technical Specification are not affected by this instruction. Operation of the system in this instruction will not degrade performance nor reliability. I Therefore, no reduction in a margin of safety will occur.
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l SE No.: 94-111 Source Document: DCP 93-0051, Rev. O DCP 93-0051A, Rev. O
_ Description of Change These design changes modify the Limitorque actuators of the Residual Heat '
Removal (RHR) suppression pool test return valves, 1E12-F024A/B, to increase the thrust / torque capability of the valves. The modifications consist of gear changes to increase actuator torque / thrust. The increased torque / thrust provides further assurance of proper valve function, accounting for various uncertainties with respect to valve operational loading as identified in Generic Letter 89-10. These modifications result in stroke time changes.
Summary ,
I. No. These changes maintain original system performance requirements such ;
that the original accident analyses are not affected. The design changes adhere to governing design codes such that the pressure boundary integrity of the valves is not compromised. These changes are an equipment improvement which assures the valves will perform the required safety function while not reducing the reliability of the valves. In addition, the equipment qualification of the valves is maintained. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. These modifications do not reduce the redundancy or independence of the RHR system since the changes maintain original RHR design functions and equipment design margins. In addition, no new permanent equipment types or new systems are introduced. Therefore, these modifications do not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety assured by the USAR LOCA accident analyses is maintained by these design changes. Although the stroke times specified in the USAR have changed, these changes have no adverse affect on the USAR LOCA accident analyses or suppression pool cooling evolution (Technical Specification 3/4.6.3.3). Containment isolation function, Technical Specification 3/4.6.4, is not adversely affected, since these changes provide further assurance of the valves performing their safety functions. Since the required safety functions of these valves are maintained, these design changes will not result in a reduction of any safety margin.
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Source Document: DCP 93-0065A, Rev. 0 Description of Change This design change modifies the electrical power circuit of the Residual l Heat Removal-(RHR) suppression pool-test return valve,-lE12-F024A, to ~ ,
improve the degraded voltage capability of the valve. The changes !
consist of increasing the effective size of the power feed conductor and :
reducing the power fuse size. The increased conductor size provides. .
further assurance of proper valve function at degraded voltage conditions ,
as identified in Generic Letter 89-10. The smaller fuse size provides !
better circuit and equipment protection.
Summary I. No. This change maintains original system performance requirements such >
that the original accident analyses are not affected. The i modification improves the valve's ability to operate under accident i conditions and reduces the probability of the valve failing to close ;
under degraded voltage conditions. The increase in conductor size I does not alter the protection provided to ensure the valve will open j to support safe shutdown in the event of a fire. In addition, the r change maintains conformance to the appropriate design codes. i Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of ;
equipment will be increased. :
II. No. This modification does not reduce the redundancy or independence of f the RHR system, since the changes maintain original RHR design j functions and equipment design margins for power circuits and -
circuit routing requirements. In addition, no new permanent ,
equipment types or new systems are introduced. Therefore, this r modification does not create the possibility of an accident or [
malfunction of a different type than any previously evaluated. l III. No. This modification maintains automatic closure of the valve under i LOCA accident conditions and does not alter valve stroke times so that Technical Specification 3/4.6.4 in not affected. The change !
does not adversely affect the ability of RHR to provide suppression '
pool cooling as required by Technical Specification 3/4.6.3.3. !
Since the required safety functions of this valve are maintained, !
the design change will not result in a reduction of any safety l margin.
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SE No.: 94-113 Source Document: MFI 1-94-039 Description of Change This Mechanical Foreign Item (MFI) evaluates installation of a freeze seal to support valve testing. This freeze seal is located on the Reactor Water Cleanup (G33) system bottom head drain line from the Reactor Recirculation (B33) system between the reactor vessel and downstream closed valves 1G33-F0101 and F0103.
Summary I. No. The accident of interest is a LOCA within the Reactor Coolant Pressure Boundary (RCPB), more specifically, an unisolatable bottom head drain (B33/G33) line break or loss of freeze seal. The freeze seal will be placed on the G33 piping while the reactor vessel is being maintained between 700F and 1400F with only static head pressure (approximately 50 psid) during refueling. These pressure and temperature conditions are not those defined in the USAR as prerequisites for a LOCA inside drywell. Hence, the accident evaluated in the USAR is not possible under these conditions.
Nondestructive examinations (PT or MT) for indications, variations, and outside diameter differences will be performed both before and after the freeze seal. This ensures that the pressure integrity of the reactor pressure boundary will be maintained before that section of piping is placed into service. If the freeze seal was to fail (pipe failure), it is estimated less than 30 gpm will leak into the drywell. Any Emergency Core Cooling System (ECCS) loop will provide sufficient makeup water to keep the fuel, control rod blades and other vessel internals covered. Therefore, the probability of i occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. With the plant in Mode 5 during the freeze seal duration and with the nuclear fuel and vessel isolated from the pipe to be frozen, the RCPB (pipe) in the area of the freece is no longer necessary and can be declared inoperable. LOCA events analyzed in the USAR assume the events are at rated reactor power. With the vessel depressurized, Reactor Pressure Vessel (RPV) head off, and upper pools flooded, there are little similarities in accident types between a B33/G33 pipe break during the freeze seal duration and a design basis LOCA.
Although the prerequisites for a LOCA inside drywell appears to be different (reactor at power verses the reactor depressurized and fuel isolated by valves) the design basis LOCA bounds the freeze seal failure. Therefore, creating a new accident or malfunction of i equipment that has not been previously evaluated is not pose _ble.
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i Summary (Cont.) i III. No. The freeze seal cycle on the G33 pipe is an industry proven method of isolating water systens that have limited isolation. Industry ,
tests proved that deviation below the transition temperature and !
back-to-normal temperature does not change the crystal structure or l characteristics of.ferritic material. However, strength characteristics do change (higher yield, lower elongation, low toughness) while the material is below the transition temperature. ;
Nondestructive examinations (PT or MT) for indications, variations, '
changes and outside diameter differences performed both before and after the freeze seal will ensure that the pressure integrity of the .
reactor pressure boundary is maintained. Therefore, no margin of safety has been reduced. !
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-SE No.: 94-114 t Source Document: DCN 4539, Rev. 0 l I
Description of Change This drawing change adds a note to P&ID D-302-102, Condensate Transfer !
and Storage System, to identify the Condensate Storage Tank'(CST)
IP11-A0001 and valve 1P11-F0518 as Augmented Quality Master Parts ,
List (MPL) numbers. l Surmary !
I. No. This drawing change makes no functional changes to the CST or to l 1P11-F0518. The equipment will continue to operate as intended by i original design. The change is the result of the CST being
- identified as nonsafety equipment that would be used in the event of :
a Station Blackout (SBO). Therefore, it must be identified as l Augmented Quality. The probability of occurrence or the j consequences of an accident or malfunction of equipment important to l safety previously evaluated in the USAR is not increased.
II. No. Plant equipment is not changed. System functions are not changed.
No procedures, instructions, instrumentation,. controls or operator actions are affected. Therefore, no possibility for an accident or ;
malfunction of a different type than any evaluated previously in the !
USAR is created. [
III. No. This drawing change will not change the lowest functional capability i or performance levels of equipment required for safe operation of l the plant. Surveillance requirements will not be changed for any !
system. The quality of systems and components will be maintained to ;
support operation within safety limits. Therefore, the margin of. !
safety is not changed. l 1
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SE No.: 94-115 Source Document: DCP 94-0079, Rev. O Description of Change Superseded by SE 94-148.
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SE No.: 94-116 Source Document: SVI-T23-T0394, Rev. 4 Description of Change This surveillance instruction revision analyzes a change to the Containment Integrated Leak Rate Test (CILRT) valve lineup and the use of temporary Measuring and Test Equipment (M&TE) for data collection.
Summary I. No. The CILRT surveillance instruction establishes pretest alignments and system controls to ensure safe operation of the plant. The temporary M&TE selected for data collection is in compliance with the original specification codes and standards ensuring test results are within required accuracies. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment is not increased.
II. No. The CILRT and M&TE are controlled by plant operations and procedures. These processes do not involve initiators or failures not previously considered. Therefore, the possibility of creating an accident or malfunction of a different type than any evaluated .
previously does not exist. I III. No. The containment leakage rate limit ensures that total containment leakage will not exceed the value assumed in the accident analyses at peak accidcat pressure. The component alignment and M&TE meet the intent of 10CFR50 Appendix J and the Technical i Specifications. Therefore, no margin of safety will be reduced. l I
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SE No.: 94-117 Source Document: PTI-E22-P0009, Rev. O j Description of Change :
This Periodic Test Instruction (PTI) provides the directions necessary to' l perform dynamic flow testing of the High Pressure Core Spray (HPCS) pump minimum flow valve, lE22-F012, and the HPCS suppression pool suction valve, lE22-F015, in accordance with Generic Letter 89-10
- Safety Related-Motor Operated Valve Testing and Surveillance". ;
i Summary i
I. No. The HPCS system will be operated within the bounds of its design ;
- . sis parameters. Administrative controls, including adherence to !
appropriate Technical Specifications, ensures that other Emergency- !
Core Cooling' Systems (ECCS), the Reactor Core Isolation Cooling (RCIC) system, and the suppression pool are maintained in an -
operable condition, if required by the applicable Operational l Condition. Therefore, the probability of occurrence or the consequences associated with an accident or nelfunction of equipment important to safety has not been increased.
II. No. The HPCS system will be operated within the bounds of its design basis parameters. Administrative controls, including adherence to appropriate Technical Specifications, ensures that other ECCS systems, the RCIC system, and the suppression pool are maintained in an operable condition, if required by the applicable Operational Condition. Technical Specifications also define the required actions to ensure plant safety is not compromised. Therefore, the possibility of an accident or malfunction of equipment of a different type than previously evaluated is not created.
III. No. Technical Specifications, as well as other administrative controls, referenced within the body of th instruction, ensures that adequate compensatory measures are taken to address the unavailability of-HPCS. Therefore, no margin of safety is reduced as a result of performing this instruction.
JE No.: 94-118 Source Document: Component /Part Equivalence Review Form (CERF) 1003, Rev. O Description of Change This CERF evaluates the replacement of a 100HP, 460V, 3 phase, 60Hz motor manufactured by Westinghouse with a motor of the same rating manufactured by Reliance Electric. These motors are used on the Control Room Emergenc" Recirculation Fans, OM26C0001A and B.
Summary I. No. The design,. material and construction standards for the replacement motors are in accordance with IEEE 323, 334 and 344, for Class lE applications. The new motors meet safety class and seismic class requirements for this application. The new motors are wired to the power supply in accordance with divisional separction requirements.
The reliability of the Control Room Heating, Ventilation and Air Conditioning (HVAC) system is not changed. The response of the Control Room HVAC to an accident is not affected. Control Room habitability, in response to an accident, is not changed.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. Performance level of the Control Room HVAC system has not changed.
The motors are equivalent to each other. The ability of the system to operate in the presence of a Loss Of Coolant Accident (LOCA),
Loss Of Offsite Power (LOOP), or high radiation event is not affected. Control Room habitability is not affected. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR is not created.
III. No. The Control Room air temperature, moisture content and habitability are unchanged. Each motor conforms to the design specifications required for this application. The motors are equivalent to each other. Therefore, the margin of safety has not been reduced.
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1 SE No.: 94-119 Source Document: USAR Change Request 94-015 Description of Change This USAR change clarifies the narrative description of the ,
synchronization capabilities of the Division 1 and 2-Diesel Generators. .
3 Summary i
I. No. This USAR change clarifies wording in the text so that it is j compatible with the USAR control logic diagram. There is no change ;
to diesel operation or function. The Division 1 and 2 Diesel Generators will still automatically start and connect to their buses in response to an emergency signal coincident with a Loss Of Offsite Power (LOOP). Therefore, the probability of occurrence or the l consequences of an accident or malfunction of equipment important to !
safety previously evaluated in the USAR is not increased. !
i II. No. This change ensures the USAR text agrees with the USAR logic l diagram. There is no change to diesel function or operation. i Current design and operation meet the design bases. The diesel generators do not need to be synchronized to any source to mitigate '
the consequences of an accident. The diesel generators cannot by .
themselves cause an accident. Therefore, the possibility for an l accident or malfunction of a different type than any evaluated !
previously in the USAR is not created. i III. No. Current design and operation is'in accordance with Technical ;
Specifications and the requirements of Regulatory Guides 1.9, 1.93 ;
and 1.108. This USAR change does not affect current design or operation. Therefore, no margin of safety is affected. l l
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SE No.: 94-120 Source Document:- NR 94-S-463, Rev. 1 j i
Description of Change 1
-This nonconformance report evaluates the "use-as-is" disposition of a I check / relief valve in the High Pressure Core Spray (HPCS) Diesel I Generator (DG) Lube Oil system. The set pressure for the valve is less than the design value. j Summary l I. No. This NR evaluates a condition such that the set pressure for 1E22-F0560 was found to be less than the design value. This condition results in reduced soakback supply pressure to the :
turbocharger during standby conditions. This condition does not l represent a component failure, and is not considered to be wear :
related. The probability of an eventual failure of the HPCS ;
Soakback system or Lube Oil system attributable to continued i degradation of this condition is not considered to be increased :
above the probability of other failures within the HPCS DG. Hence, ,
i this condition will not decrease the reliability or availability of the HPCS DG, including its lube oil system. Therefore, the probability of occurrence or the consequences of an accident or ;
malfunction of equipment has not increased. j II. No. This condition is limited to the HPCS DG Lube Oil system. The HPCS l DG is used to mitigate USAR accidents / transients which have been !
initiated by other causes, and not to prevent the occurrence of any initiating event. This condition does not represent a camponent i failure, and is not considered to be wear related. The probability of an eventual failure of the HPCS Soakback system or Lube Oil system attributable to continued degradation of this condition is not considered to be increased above the probability of other ,
failures within the HPCS DG, Hence, this condition will not i decrease the reliability or availability of the HPCS DG, including its lube oil system. Therefore, the possibility of an accident or i malfunction of equipment important to safety of a type different ;
than previously evaluated is not created.
III. No. This nonconforming condition is limited to the scope described i above. The reliability of the HPCS DG is not adversely affected as a result of this condition, and no system functions are compromised. '
This condition has no effect on the independence or redundancy of the onsite power supply. Therefore, this change does not reduce any l margin of safety. ;
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Source Document: PAP-0803, Rev. 6, TC-1 i l
Description of Change '
This change to Plant Administrative Procedure (PAP) 0803 incorporates ,
various administrative changes which are designed to ensure better i conformity to other plant procedures.
Summary i I. No. These changes to the Chemical Control Program are administrative. i They will not modify the plant or change the process on how storage approval for chemicals in the plant is obtained. Therefore, neither the probability of occurrence nor the consequences of a previously '
analyzed accident or malfunction of equipment will be increased by ;
this change. i II. No. These changes do not alter the operation or the configuration of the plant. Therefore, these changes will not create the possibility for ;
an accident or malfunction of a different type than any previously
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III. No. These changes are administrative and do not affect the Technical :
Specifications. Therefore, this change will not reduce any margin i of safety. ;
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SE No.: 94-122 Source Document: NR 93-S-200, Rev. O NR 93-S-200, Rev. 2 NR 93-S-200, Rev. 3 NR 93-S-200, Rev. 4 Description of Change This nonconformance report evaluates the use of actual yield strength values in performing an operability review of the High Pressure Core Spray (HPCS) Pump Room structural steel platform members at Elevation 574'-10", located in the Auxiliary Building. During review of revised piping loads due to snubber optimization, several connections were found to have slotted beam connections with ANCO nuts, whereas original design calculations and drawings specified friction type A325-F type connections. The Certified Mill Test Reports (CMTR's) were employed to establish actual yield strength rather than the minimum yield strength specified by the American Institute of Steel Construction (AISC)
Specifications (February 1969 version). This is the version of the specification the USAR references. The use of these actual values allowed the justification for continued operation.
Summary I. No. The use of the actual yield strength, for determination of continued operation, is documented via structural calculations. Based on said calculations, it has been concluded that the platform remains adequate with the revised loading imposed by snubber optimization analyses utilizing normal design concepts and the slotted connections with ANCO nuts. Based on the above, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated will not increase.
II. No. Based on the conclusions formed by structural calculations, the ;
platform design remains within allowable stress limits with the new !
piping loads imposed. Therefore, the possibility of an accident or nalfunction of equipment different than those previously evaluated is not created.
III. No. This evaluation has confirmed the adequacy of the platform for the new piping loads. No functional change is being nede to any plant equipment or system. Therefore, the margin of safety is not reduced.
l SE No.: 94-123 Source Document: DCP 94-0079A, Rev. O Description of Change This design change installs a permanent level indicator (ruler) and ultrasonic flow instrumentation for the OP41 Service Water system west discharge structure.
Summary I. No. This modification does not affect the design basis of the Service Water system and does not impact any other safety-related or nonsafety-related plant system. Therefore, neither the probability-of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment important to safety will be increased by this design change.
II. No. The installation of the ultrasonic level monitoring system and the level indicator is bounded by the evaluation and design relative to the existing Service Water discharge flow instruments and existing flow indicator in the main service discharge structure. Therefore, the installation of the new instrumentation has not increased the possibility of creating a different type of accident or malfunction of equipment different than any previously evaluated.
III. No. The Service Water system will continue to function as required. No Technical Specification related equipment will be adversely affected by this modification. Therefore, the modification will not reduce any margin of safety.
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SE No.: 94-124 Source Document: PDB-F0001, Rev. 2 Description of Change This evaluation analyzes updating the Plant Data Book (PDB) F0001 for various plant limits associated with cycle 5 operation.
Summary I. No. The only change to the plant is the introduction of a new fuel configuration and core design. The fuel design was produced using NRC-approved methods described in the GESTAR II. The potentially l limiting plant transients and accidents have been evaluated using j the GESTAR methodology with the same limits on consequences as previously evaluated in the USAR. The fundamental sequences of accidents and transients have not been altered. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Plant operation remains in conformance with the analyzed envelope of l USAR Chapters 4, 5, and 15. The GESTAR II analysis has been '
accepted by the NRC as comprehensive for ensuring that fuel designs will perform within acceptable bounds. Therefore, creating an i accident or malfunction of equipment or a type different than previously evaluated is not possible.
III. No. The new fuel configuration does not alter the design or function of any plant system, outside of the fuel. The fuel design was produced using NRC-approved methods described in the GESTAR II. The design satisfies the acceptance criteria which are consistent with the fuel-related Technical Specifications. Therefore, no margin of safety has been reduced.
i SE No.: 94-125 Source Document: DCP 91-0215, Rev. O Description of Change This design change removes the MSIV isolation signal and RPS scram function from the Main Steam Line Radiation Monitors (MSLRM)
(lD17-K610A-D). The MSLRMs will still trip the mechanical vacuum pumps and isolate the reactor water sample valves on high radiation. This modification was done in accordance with General Electric NEDO-31400A,
" Safety Evaluation for Eliminating The Railing Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Punction of the Main Steam Line Radiation Monitor", which was approved by the NRC. A Technical Specification change was submitted, reviewed, and approved by the NRC prior to approval and implementation of this modification.
Summary I. No. The accident of consideration is the Control Rod Drop Accident (CRDA) which is the only USAR described accident which takes credit or the MSLRMs. The MSLRMs are designed to mitigate the consequences of the CRDA. The Perry plant has a system called the Rod Pattern Control System (RPCS) which is a part of the Rod Control and Information System (RC&IS) which is required by Technical Specification 3/4.1.4.2, " Rod Pattern Control System'. This system prevents the control rods from being placed in a configuration such that the CRDA would be a valid accident. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The MSLRMs will still be operable, will have alarms, and upon indication of high radiation levels, will isolate the reactor water sample valves and trip the mechanical vacuum pumps. The scram and MSIV isolation were only installed to mitigate the consequences of the hypothetical CRDA. The Perry plant has a system called the Rod Pattern Control System (RPCS) which is a part of the Rod Control and Information System (RC&IS) which is required by Technical Specification 3/4.1.4.2, " Rod Pattern Control System'. This system prevents the control rods from being placed in a configuration such that the CRDA would be a valid accident. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The MSLRMs will still be operable, will have alarms, and upon indication of high radiation levels, will isolate the reactor water sample valves and trip the mechanical vacuum pumps. The scram and MSIV isolation were only installed to mitigate the consequences of the hypothetical CRDA. The Perry plant has a system called the Rod Pattern Control System (RPCS) which is a part of the Rod Control and Information System (RC&IS) which is required by Technical Specification 3/4.1.4.2, " Rod Pattern Control System'. This system prevents the control rods from being placed in a configuration such that the CRDA would be a valid accident. Therefore, no margin of safety has been reduced.
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Source Document: DCN 4549, Rev. O j Description of Change This drawing change will revise layout drawing E-015-044, Refueling ;
Floor, to show a revised weight of the shroud head stud wrench due to ae new socket extension. The socket extension is attached to the shroud I head stud wrench, which is in turn stored on the tooling rack located at- :
0' azimuth, on the Refueling Floor, El. 689'-6', in containment.
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l I. No. This drawing change revises the weight.of a refueling tool stored in :
containment. Due to the relatively negligible ~ load increase of the i new tool, no credible concern exists for the overall design loading ;
conditions for this area, either from a normal loading or seismic j loading condition. Therefore, the probability of occurrence or the ,
consequences of an accident or malfunction of equipment has not l increased. !
II. No. Though a minor weight increase exists, the area under evaluation remains well within pertinent design allowables. Therefore, i creating a new accident or malfunction of equipment is not possible, j l
III. No. No functional change is being made to any plant equipment or system. I The area under evaluation remains within design allowables. '!
Therefore, no margin of safety has been reduced. ]
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i Source Document: TXI-0204, Rev. 0 '
Description of Change This temporary test instruction operates the Residual Heat Removal (RNR) system in the suppression pool cooling or test return mode to evaluate _
the performance of several RHR system valves. 'The control circuitry of !
the RHR A/B pump minimum flow valves will be altered during portions of, this test to allow testing of the valves near their design basis, :
differential pressure. This testing will meet requirements of Generic !
Letter 89-10.
Summary I. No. Operation of the RHR system in this instruction is nearly identical to the RHR system operating instruction. The only difference is that the minimum flow valve, 1E12-F064A/B, will be disabled. The i instruction contains administrative controls to prevent adverse affects on the RHR system. Other than the minimum flow valve ;
interlock, all actions specified in this instruction fall within the normal scope of system operation. Therefore, neither the probability of occurrence nor the consequences of a previously l analyzed accident or malfunction will be increased by this change. ,
II. No. The instruction contains administrative controls to prevent adverse affects on the RHR system. Other than the minimum flow valve -
interlock, all actions specified in this instruction fall within the normal scope of system operation. Therefore, this testing will not create the possibility for an accident or malfunction of a different type than previously evaluated.
t III. No. This test instruction does not alter any RHR system characteristics, '
setpoints or functions required by Technical Specifications, except the automatic function of IE12-F064A/B. Administrative controls l address disabling the minimum flow valve interlock. The performance i of this test will not degrade RHR system pump / components or their j reliability. Since this system and components will continue to i perform their design basis / Technical Specification functions, the l margins of safety will not be reduced.
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.SE No.: 94-129 Source Document: GMI-0131, Rev. 0 ,
Description of Change
-t This general maintenance instruction describes the methodology of j installing and removing a small pump inside the reactor vessel t temporarily after the reactor has been shut down for 90 days. The i purpose of this pump is to help minimize thermal stratification in the :
reactor when installing fuel.
Summary I. No. An evaluation of this instruction showed that there would be no :
increase in the potential or severity of a fuel handling accident. ,
In the event that the pump itself were to fall from its mounting, .:
the result would be bounded by the fuel handling accident. Also, !
for this installation, an evaluation was performed in order to !
assess the radiological consequences of a fuel handling accident with the pump running and concluded that the consequences were still i below those evaluated in the USAR. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased. i II. No. The evaluation showed that the potential of the pump coming in contact with the fuel would be bounded by the present USAR analysis.
Also, the_ installation does not introduce any new mechanisms that !
could serve to initiate a new type of equipment failure. Current instructions provide for guidance in the event that water level 3 decreases below the required level. Therefore, this' change will not j create the possibility for an accident or malfunction of a different ;
type than any previously evaluated. j i
III. No. This installation does not impact the ability of the Fuel Pool Cooling and Cleanup (G41) or Residual Heat Removal (E12) systems to ;
renove decay heat from the upper containment pools. Offsite doses are bounded by the Technical Specifications. Therefore, no margin:
of safety will be reduced. !
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SE No.: 94-130 Source Document: PS7G, Rev. 3 Description of Change
'The Perry Specific Technical Guidelines (PSTG) are-the " licensing basis,
~ documents
- for the Plant. Emergency Instructions (PEIs), and therefore, all changes to the PSTG receive safety evaluations. This revision changes entry conditions to agree with revised containment parameters,-
changes the method used to round-off the pressure for securing containment spray on decreasing containment pressure, incorporates ' fast fire water" as an alternate injection system, incorporates editorial changes, and incorporates the secondary containment control guidelines.
Summary I. No. Containment spray to control containment pressure is analyzed in the USAR both for the case of intentional and unintentional initiation.
For all postulated accidents and transients, the operation of containment spray contained within the PSTG is within the USAR analyzed limits. Therefore, the changes to the PSTG do not affect the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR.
l II. No. Alternate injection is only used to compensate for already existing malfunctions of Reactor Pressure Vessel level control systems. The
' fast fire water' method of alternate injection uses the same- l suction source (Emergency Service Water Forebay) as existing l methods. Therefore, the possibility an accident or malfunction of '
equipment important to safety of a different type than previously i analyzed in the USAR is not created. !
III. No. The PSTG does not' adversely affect systems or components required for safety. Therefore, no margin of safety is impacted. j l
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SE No.: 94-131 Source Document: TXI-0205, Rev. O Description of Change This temporary test instruction verifies the operability of isolations initiated by a manual Low Pressure Coolant Injection (LPCI) B/C initiation to partially meet the requirements of Technical Specification 3/4.2.1-1.1.e. It will also test the LOCA initiation logic of the Annulus Exhaust Gas Treatment System (AEGTS) to meet Technical Specification 3/4.6.6.2. The performance of this test is required to bring current the surveillance for the valves in question, as well as AEGTS.
Summary I. No. During performance of this test the reactor will be in a defueled condition. The systems being removed from service due to the removal of LOCA logic required for this test are not required to be operable in this operational condition. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. During performance of_this test the reactor will be in a defueled condition. The systems being removed from service due to the removal of LOCA logic required for this test are not required-to be operable in this operational condition. Therefore, creating an ,
accident or malfunction of equipment of a type different than has l been previously evaluated is not possible.
III. No. The Emergency Core Cooling Systems are relied upon to prevent exceeding thermal limits such that the integrity of the barriers to release to the environment are maintained. The reactor is defueled, i and decay heat removal of the spent fuel pools will be accomplished by systems unaffected by the testing. Therefore, no margin of safety has been reduced.
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SE No.: 94-132 Source Document: NR 94-S-482, Rev. 1 Description of Change This nonconformance report evaluates the disposition of check valve l 1G41-F522. This valve is located in the Spent Fuel Pool Cooling system: l return line from the upper containment pool. The disposition provides for the replacement of the valve with an equivalent check valve of an improved design. ;
Sumary i I. No. The disposition of valve replacement maintains the original system l design and performance requirements such that containment integrity 'i and containment isolation capability are not altered. Plant accident analyses are not affected. The replacement valve conforms 2 to ASME Section III, Subsection NC (Class 2) requirements and thus 'l ensures pressure boundary integrity. Therefore, neither the !
probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The disposition maintains original system design margins and original system functions. In addition, the valve change creates no new systems and introduces no new permanent equipment types. Thus, potential for a new malfunction of equipment / systems is not introduced. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The replacement valve conforms to original design code requirements.
The safety functions of the affected valve for containment and system pressure boundary integrity are maintained. Thus, there is no reduction in the margin of safety.
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v SE N:t: 94-133-
- Soure1 Document: TXI-0206, Rev. 0 l Desc.jycion of Change This temporary test instruction allows for the Residual Heat .
Removal-(RHR) system operation and the data collection necessary for the post-cleaning requirement of the RHR Loop A under the Generic Letter 89-13 Program.
Summary L. No. This instruction utilizes the system as designed. All of the designed safety system interlocks and functional capabilities will still be available during the performance monitoring of the RHR ,
Loop A. The main purpose of the instruction is to collect '
performance data, related to the system in-line heat exchangers. i Under an administrative function the temperature data and flow data ,
will be collected when the system is operating in the suppression l pool cooling mode. The data collected under this instruction is necessary to ascertain the overall heat transfer coefficient of the >
RHR Loop A heat exchangers. During the performance of the instruction, the system performance, control logic, interlocks, and '
overall system capabilities will not be challenged. Further, this instruction will be utilized with the system declared as inoperable. '
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This instruction utilizes the system as designed for data collection to measure temperature and flow for system operation as it relates ,
to the heat exchanger performance. All of the designed safety ,
system interlocks and functional capabilities will still be available during the performance monitoring of the RHR Loop A. ;
Therefore, creating a new accident or malfunction of equipment that i has not been previously evaluated is not possible. j i
III. No. This instruction does not compromise the original equipment design )
bases, construction codes or equipment qualification. The instruction is only used to collect performance monitoring data.
Therefore, no margin of safety has been reduced. I 1
SE No.: 94-134 Source Document: TXI-0204, Rev. O, TC-2 Description of Change l
This change to this temporary test instruction will operate the Residual Heat Removal (RHR) Loop C in the test return mode to evaluate the l performance of lE12-F064C, the RHR C pump minimum flow valve. The control circuitry of the RHR C minimum flow valve is altered during portions of this test to allow testing of the valve near its design basis, differential pressure.
i Summary I. No. Operation of RHR Loop C per this instruction change is nearly I identical to the operation described in the RHR system operating instruction. The only difference is that automatic cycling of lE12-F064C is disabled. The instruction contains administrative l controls to prevent adverse affects on the RHR system. Other than I the minimum flow valve interlock, all actions specified in this i instruction fall within the normal scope of system operation. l Therefore, neither the probability of occurrence nor the l consequences of a previously analyzed accident or malfunction will l be increased by this change.
l II. No. The instruction contains administrative controls to prevent adverse ,
affects on the RHR system. Other than the minimum flow valve l interlock, all actions specified in this instruction fall within the normal scope of system operation. Therefore, this change will not l create the possibility for an accident or malfunction of a different type than previously evaluated.
III. No. This instruction change does not alter any RHR Loop C characteristics, setpoints or functions required by Technical Specifications, except the automatic function of the RHR C pump minimum flow valve, 1E12-F064C. Administrative controls address disabling the minimum flow valve interlock. The performance of this test will not degrade RHR system pump / components or their reliability. Since this system and components will continue to perform their design basis / Technical Specification functions, the margins of safety will not be reduced.
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.SE No.: 94-135 Source Document: DCN 4052, Rev. 0 .;
i Description of Change This drawing change adds a note to P&ID D-352-791, Emergency Service Water System, to state the hydromotors associated with Unit 2 valves, !
2P45-F040A/B and 2P45-F0160.have been electrically disabled and may be !
-disconnected or removed. These valves have been evaluated as no longer necessary to support Unit One operations. The valves are considered Unit-One/ Unit Two boundary valves which are necessary to remain closed in order to support Unit One operations. The ball valve.' remains installed and is necessary to provide a boundary between' Unit 1 and Unit.2 piping..
Summary I. No. This drawing change applies to Unit 1/ Unit 2 boundary valves. The ASME III, Class 3, ball valves will continue to meet design and construction standards of the Emergency Service Water (ESW-P45) :
system. Overall system performance remains unchanged by the I acceptance of the removed actuator at the Unit 1/ Unit 2 boundary.
The valves will remain in a closed position, providing the necessary isolation to prevent possible draining of the ESW Strainer Blowdown- !
system. This drawing change does not involve any accidents j evaluated in the USAR. Therefore, the probability of occurrence'or l
the consequences of an accident or malfunction of equipment has not i increased.
II. No. The valves remain available as passive components isolating the ESW blowdown from Unit 2. The P45 system remains as designed. Only the ESW blowdown isolation valves mass has changed from the original !
design. Stress analysis remains within'the allowable code limits. I The ball valves continue to provide reliable isolation between Unit 1 ESW blowdown piping and Unit 2 piping. Therefore, creating an accident or malfunction of equipment different than has been. i previously evaluated is not possible.
III. No. This drawing change does not affect the bases of Technical Specification 3/4.7.1. Sufficient cooling capacity. remains available for continued operation of safety-related equipment'during normal and accident conditions. Even though the blowdown header connects all three Unit 1 ESW divisions with Unit 2 divisions 1 (piping not completely installed on the Unit 2 side), a single failure (2P45-F040A/B or F160) within the blowdown system cannot affect the redundant cooling capacity of the ESW system. Therefore, no margin of safety has been reduced.
SE No.: 94-136 Source Document: DCP 92-0097, Rev. 1 Description of Change Superseded by SE 94-187.
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SE No.: 94-137 Source Document: TXI-0204, Rev. 1 Description of Change This temporary test instruction will operate the Residual Heat Removal (RHP) system in the suppression pool cooling or test return mode to evaluate the performance of several RHR system valves. The control circuitry of the RHR A/B pump minimum flow valves will be altered during portions of this test to allow testing of the valve near its design basis, differential pressure. This testing will meet the requirements of Generic Letter 89-10.
Summary I. No. Operation of the RHR system in this instruction is nearly identical to that of the RHR system operating instruction. The only difference is that the automatic cycling of the minimum flow valve, lE12-F064A/B, will be disabled. The instruction contains administrative controls to prevent adverse affects on the RHR system. Other than the minimum flow valve interlock, all actions '
specified in this instruction fall within the normal scope of system operation. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of i equipment will be increased by this change. !
l II. No. The instruction contains administrative controls to prevent adverse affects on the RHR system. Other than the minimum flow valve i interlock, all actions specified in this instruction fall within the normal scope of system operation. Therefore, this change will not create the possibility for an accident or malfunction of a different type than previously evaluated.
i III. No. This instruction does not alter any RHR system characteristics, l setpoints or functions required by Technical Specifications, except the automatic function of the RHR pump minimum flow valve, IE12-F064A/B. Administrative controls address disabling the minimum I flow valve interlock. The performance of this te.st will not degrade I RHR system pump / components or their reliability. Since this system and components will continue to perform their design basis / Technical Specification functions, the margins of safety will not be reduced.
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Source Document: DCP 92-0097A, Rev. 0 l l Description of Change This design change provides electrical control modifications necessary to make the Residual Heat Removal (RHR) heat exchanger discharge valves, j i lE12-F003A/B, open automatically on receipt of a Low Pressure Coolant j l Injection (LPCI) initiation signal. l Summary I. No. Adding an automatic opening signal to the lE12-F003A/B valves will l make them similar to the RHR heat exchanger bypass valves, t 1E12-F0048A/B, in design and function, and ensure that LPCI will !
' fulfill its design intent. There are no detrimental affects to any 1 other operating modes of RHR. Because the valves were not j previously required to be either open or closed in the accident l analysis, reqeiring them to be open will have no affect upon i accident analysis. Therefore, the probability of occurrence or the ;
consequences of a previously analyzed accident or malfunction of !
l equipment will not increase. !
i II. No. This design change creates no new systems, introduces no new equipment types, and maintains the RHR design requirements for accident mitigation. Therefore, this change will not create the ,
possibility for an accident or malfunction of a different type than ;
any previously evaluated, j i
III. No. This design change maintains the requirements of the original equipment design and construction codes, RHR's design bases, and the equipment qualification requirements. Installation of this l modification will ensure LPCI provides the required flow with ,
l adequate margin. Therefore, no safety margins are reduced as the !
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-Source Document: MFI/LLJED l-94-051 ,
I Description of Change -
This Mechanical Foreign Item (MFI)/ Lifted Lead and Jumper /
Electrical (LLJED) will permit operation of the Containment Vessel and - ,
Drywell Purge (1M14) system with the 18-inch exhaust path isolation, valves manually blocked open and the' actuators removed for maintenance.. -;
Containment integrity will be relaxed during the period of time this !
temporary modification is active. The 1M14 exhaust system normally I operates in the refuel mode with the plant in a shutdown condition. This-MFI/LlaTED permits the system to operate while maintenance is being i conducted. j Summary j I. No. The performance of the ma ntenance on the valve actuators requires !
that containment integrity is not necessary. The blocking open of :
the associated valves and the installation of the appropriate i jumpers will permit the operation of the IM14 system for ventilation !
purposes. With the plant shutdown, the event of concern is a !
dropped fuel bundle accident. However, with containment integrity j relaxed, fuel handling is not permitted, thus no increase'in the l probability or consequences of the dropped fuel accident will occur. i The operation of the 42-inch exhaust valves remains unaffected by i the MFI/LLJED and full isolation capability still exists for the i exhaust path should it be needed. Therefore, the probability of i occurrence or the consequences of an accident or malfunction of l equipment has not increased. !
II. No. The USAR analyzes the 1M14 system for support of a dropped fuel bundle in containment during a shutdown condition. Since fuel ;
handling is an excluded process with containment integrity relaxed, j the potential for this event is non-existent. Any event which has ;
the potential to provide a radiological release is excluded by the ;
relaxation of containment integrity. Therefore, no accident or l malfunction of a different type is created by this MFI/LLJED. !
III. No. The reqairements and actions of the Technical Specifications with respect to containment integrity and the 1M14 system are met by the j MFI/LLJED. Therefore, no margin of safety has been reduced. !
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SE No.: 94-140 Source Document: SP-2154, Rev. 1 SP-2135, Rev. 1 Description of Change This evaluation analyzes the revisions to Installation Standard '
Specifications (ISS) SP-2154, ' Technical Requirements for Procurement of Class A Fill Material", and SP-2155, " Technical Requirements for Excavation and Installation of Backfill". The revisions entail the .
addition of alternate test methods to be employed during the backfill process. It also deleted the requirement to achieve a specific dry unit weight at 85% relative density for Class A fill. This is acceptable due to the requirements for certification and in-process testing verifies particle size distribution, uniformity coefficient, specific gravity and friction angle. Since these requirements directly affect the minimum and maximum density values, compaction to 85% relative density is adequate regardless of the actual unit weight.
Summary I. No. The concepts adopted by these specification revisions have been chosen and evaluated such that the design intent of the backfill remains unchanged as compared to that originally specified. The backfill which results will continue to fully perform its intended design function as originally evaluated in the USAR. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Since the backfill in question is designed to perform specific functions and civil / structural calculations have been performed to i ensure those properties which allow those functions to be performed !
are not changed, the backfill remains adequate to perform said functions. Therefore, creating a new accident or malfunction of equipment of a different type than previously evaluated is not possible.
III. No. The calculations referenced, confirm the adequacy of the subject backfill and ensures that all original design functions continue to be met or exceeded. Therefore, no margin of safety has been reduced.
SE No.: 94-141 Source Document: NR 94-S-521, Rev. O Description of Change This nonconformance report analyzes the "use-as-is' disposition of continued operation of the Offgas system with missing Steam Jet Air Ejector (SJAE) bellows pieces. It is believed the missing pieces are located between the SJAE and the Offgas Preheater, IN64-B0001B.
Summary I. No. The Offgas system will continue to meet the required design and construction standards with the exception of the cleanliness "D' requirements. The gaseous radioactive effluent will remain unchanged. There is no affect upon dilution steam, hence no affect upon effluent hydrogen concentration. The proposed activity (operation of the system with debris) may affect the overall system performance. Degraded Offgas system performance would only occur if the "B" SJAE train is placed into service and a preheater tube was to break due to vibration of debris within the system. Over time the Offgas system could exceed its moisture removal design rate and overload downstream equipment. At this point, a controlled plant shut down would be required. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The Offgas system does not perform a safety function in accordance with the USAR and only serves to process and control releases of gaseous radioactive effluents from the main and auxiliary condensers. Allowing the use of the SJAE *B' train with debris in the preheater is not related to an accident initiator or failure considered in the USAR. The existing debris and system vibration is postulated to contribute to a possible increase in tube failures within the preheater. The gasecus radioactive effluent will remain unchanged. There is no affect upon dilution steam, hence no affect upon effluent hydrogen concentration. Due to the system configuration, the extent of postulated tube damage is small.
Simultaneous breaking of many tubes due to a random detonation is not considered feasible. Therefore, creating a new accident or nalfunction of equipment is not possible.
III. No. The gaseous radioactive effluent will remain unchanged and will be kept "as low as is reasonably achievable" with the debris still within the preheater. The explosive mixture (Technical Specification B3/4.11.2.6) is unchanged. Dilution steam is unaffected and will still ensure that the concentration of potential explosive mixture in the Offgas system is maintained below the flammability limits of hydrogen. Therefore, no nargin of safety has been reduced, e
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I SE No.: 94-142 Source Document: Emergency Plan, Rev. 11, TC-7 Description of Change This change to the Emergency Plan revises the Emergency Plan-related training requirements for General Employee Training (GET) to meet the INPO Guidelines for GET, as outlined in ACAD 93-009, dated 11/93. The change also incorporates statements allowing participation in drills and exercises towards meeting the annual Emergency Response Organization (ERO) requalification requirements.
Summary I. No. This change does not direct or impact the operation or design of any plant structure, system or component. Only the GET content and criteria for meeting the annual ERO requalification requirements under Section 8.1.2 of the Plan are altered. Accident initiators are not affected. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change. i II. No. This change does not alter the design of the plant; the type, frequency or consequences of an accident; or direct plant mitigating actions. Therefore, it will not create the possibility for an accident or malfunction of a different type than previously evaluated.
III. No. This change does not adversely affect any equipment or operation relied upon by the Technical Specifications. Therefore, it will not reduce any margin of safety.
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SE No.: 94-143 i Source Document: DCP 94-5112, Rev. 0 l DCP 94-5112, Rev. 1 l Description of Change l l
This design change will eliminate the internal bypass valve of the Offgas !
Condenser, IN64-B0002. A
- stationary' orifice plate will replace this
' dynamic' component and will simulate the required pressure drop across.
the water box and still maintain adequate condensate flow through the tube side of the heat exchanger. The Offgas Condenser is nonsafety-related. j Summary i- I.-No. This design change replaces the bypass valve of the Offgas Condenser .
with an orifice plate. The orifice plate will produce a pressure drop which is slightly higher than the bypass valve. The increased pressure drop will not affect the operation of the Condensate (N21) or Offgas (N64) systems. The function and design intent of the N21 and N64 systems will not be adversely affected. Similarly, the design and construction standards of the Offgas Condenser will not be adversely affected. Also, the change will not hinder the N64 system from performing its function of controlling the release of gaseous effluents from the condensers. Hence, an increase in onsite or offsite dose will not occur due to the design change. Therefore. <
the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
1 II. No. This design change will replace a multiple piece dynamic component . I with a single piece stationary component. Thus, significantly i reducing the possibility of the component partially disassembling ;
during the operation of the system and causing damage. The change .l will not adversely affect the function, operation, or design intent l of the N21 and N64 systems. Therefore, creating a new accident or i malfunction of equipment is not possible. .l III. No. The N64 system will continue to operate in accordance with established operating practices. The design intent of the Offgas Condenser is not adversely affected by the design change. The gaseous radioactive effluent will remain unchanged. The explosive mixture is unchanged (dilution steam is unaffected and will still ensure that the concentration of potential explosive mixture contained in the Offgas system is maintained below the flammable !
limits of hydrogen). Therefore, no margin of safety has been reduced.
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u SE No.: 94-144 Source Document: TXI-0201, Rev. 1 Description of Change This temporary test instruction provides the directions necessary to perform dynamic flow testing of the High Pressure Core Spray (HPCS) injection valve, 1E22-F004, and the HPCS pump minimum flow valve, lE22-F012, if necessary, in accordance with NRC Generic Letter 89-10.
This revision was written to allow this test to be performed in Operational Condition (OP CON) 5, with fuel assemblies in the reactor pressure vessel.
Summary I*
I. No. HPCS operation via this instruction will be similar to normal and emergency operation. The primary difference is that the HPCS injection valve, lE22-F004, will not auto close on a level 8 (high reactor water level) signal. This interlock will be defeated to allow stroking of the injection valve. Also, an automatic suction source transfer from the Condensate Storage Tank (CST) to suppression pool will not occur since the HPCS suppression pool suction valve, 1E22-F015, will be maintained closed. The system will not be operated in a manner which would cause any degradation even though the injection valve will not be operational for auto closure on level 8 during certain portions of this instruction and lE22-F015 will be maintained c20 sed. The system will be operated as designed, and no concerns exist with nuclear instrument damage and/or fuel channel wear / creep. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. All automatic functions and interlocks for the components operated under this instruction will remain operable with the exception of the injection valve level 8 auto closure and the automatic suction source transfer. Administrative controls are in the instruction to ensure that equipment damage is prevented. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The HPCS system will be inoperable during the performance of this test which satisfies the requirements of Technical Specifica-tion 3/4.3.3, Action b (Action 34). The system will function as required by Technical Specifications, with the exception of the level 8 auto close interlock and the automatic suction source transfer. The operation of the HPCS system during this test will not result in degraded system reliability. Therefore, no reduction in the margin of safety will occur.
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i SE No.: 94-145 l Source Document: DCN 3547, Rev. 0 l Description of Change This drawing change revises P&ID D-806-022, Plant Radiation Monitoring !
Automatic Isokinetic Sampling System, for the Turbine / Heater Bay stack.. i The revision is editorial for it just corrects a reference.
Sumary ;
I. No. This change is editorial. The change does not affect the Chapter 15 l accident analysis, physically modify the plant, or its operating l program. Therefore, the probability of occurrence or the !
consequences of an accident or a malfunction of equipment important to safety has not increased. i II. No. This activity is an editorial change to documentation to correct a ;
drawing cross-reference. The change will not create the possibility i of a different type of accident or malfunction of equipment !
important to safety. ,
III. No. This activity is an editorial change to system documentation. The :
plant will not be altered. Therefore, no margin of safety is !
affected.
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SE No.: 94-146 Source Document: PSTG, Rev. 2, TC-6 Description of Change This change to the Perry Specific Technical Guidelines (PSTG) evaluates the effect caused by bypassing the RCIC low steam line pressure isolation signal. This deviation does not change the methodology used to bypass the Reactor Core Isolating Cooling (RCIC) low steam line pressure isolation.
Summary I. No. The methodology employed to bypass the RCIC low steam pressure isolation does not result in a concern if RCIC vacuum breakers are open. No systems or components required for safe shutdown are affected. Accident initiators are not impacted. The same operator actions are taken at the same point for the accidents already evaluated in the USAR. Therefore, neither the probability of occurrence nor consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR is affected.
II. No. The actions to bypass the RCIC low steam pressure isolation are taken post-accident. No systems or components required for safe shutdown are affected. Accident initiators are not impacted. The same operator actions are taken at the same point for the accidents already evaluated in the USAR. Therefore, the possibility of an accident or malfunction of equipment different than previously evaluated is not created.
III. No. No systems or components required for safety are adversely affected.
Therefore, no margin of safety is impacted.
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SE No.: 94-147 Source Document: DCP 93-0020, Rev. 2
- Description of Change This design change modifies the bore size and.provides a more precise tolerance for the thickness of the Residual Heat Removal (RHR) test return line restricting orifice plates, 1E12-D003A/B. 1E12-D003A/B were sized per a calculation which utilized system test data as input. The bore size is designed such that it will limit suppression pool cooling flow below 7800 gallons per minute (gpm) and provide a minimum flow of 7100 gpm.
Summary I. No. The restricting orifice will not alter the operational characteristics of the RHR system because the design change maintains all original system design standards. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This design change meets all original design criteria. Therefore, creating a new accident or malfunction of equipment of a type different than previously evaluated is not possible.
III. No. The design change maintains RHR system design standards. Therefore, there is no reduction in any margin of safety.
l SE No.: 94-148 Source Document: DCP 94-0079, Rev. O DCP 94-0079, P.ev. 1 l
Description of Change This design change abandons approximately 370 feet of defective Service Water (P41) system fiberglass pipe (of various sizes) upstream of the Service Water Weir Structure. The abandoned pipe is replaced with carbon steel piping that reroutes the return side system flows to the lake by a different flow path. The replacement piping is routed to the Discharge Tunnel Entrance Structur< The configuration of the P41 system is ;
altered since the rerouted piping bypasses the Service Water Weir i Structure. '
Summary I. No. This design change abandons and replaces a large portion of P41 piping. Flow requirements to supplied equipment are not adversely effected. Overall system and plant operation remain unchanged. All equipment, material and services associated with this modification meet the applicable nonsafety piping design code requirements.
Flooding concerns inside and outside of plant buildings associated with this change are bounded by existing USAR analyses. Therefore, the probability of occurrence of an accident or malfenction of equipment important to safety consequences previously evaluated in '
the USAR is not increased.
II. No. The P41 system is not required for safe shutdown of the reactor and ;
its impact on safety-related systemc/ components is unchanged by this design change. As a result, the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR is not created.
III. No. The dilution of effluencs is not compromised by this modification.
All equipment, material and services associated with the modification meet the applicable piping codes. Flooding concerns are bounded by existing analyses. Therefore, no margin of safety is reduced.
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SE No.: 94-149 Source Document: USAR Change Request 94-028 Description of Change This USAR change deletes the design requirement to consider jet :
impingement reflected off of obvious reflecting surfaces (i.e., secondary _
reflected jets).
Summary I. No. This change only affects common mode hazards associated with a given !
pipe break. The pipe break is the initiating event. Considering that the probability of breaks occurring is very low, and is decreasing due to the implementation of erosion / corrosion monitoring programs, the consideration of reflected jets result in little if any improvement in plant safety. This is based on the fact that items which could be affected by reflected jets will yield but are i not expected to fail catastrophically due to design margins and ductility. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This change affects common mode hazards associated with a given pipe break. The initiating event is the broken pipe, which in turn creates jets which may impinge on essential systems and components.
Equipment already assumed to malfunction in the pipe break analysis of the USAR may be hit by a reflected jet from the initiating pipe break. Additional equipment failure need not be considered since the initial jet impingement will be the governing condition.
Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The secondary effects of reflected jets is a design load consideration relating to the availability of essential systems and components in areas with high energy pipes. These secondary effects need not be considered based on the fact that initial jet impingement loading will govern. Therefore, this change does not reduce any margin of safety.
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SE No.: 94-150 Source Document: 10I-9, Revision 5 Description of Change This change to Integrated Operating Instruction (I0I) 9 provides guidance for performing reactor disassembly, the manipulation of irradiated fuel assemblies and reactor components, and reactor reassembly without the Refueling Operation Atmospheric Radiation Monitor (D17-K650) and the Unit 1 Portable Drywell Radiation Monitor (lD21-N340) installed.
Summary I. No. These radiation monitors only provide alarm / indication functions.
Actions taken in response to receipt of an alarm are manual in nature. Compensatory measures are provided to ensure these actions are still taken should the plant conditions warrant. Therefore, the probability or the consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the USAR .
has not increased. I II. No. There is no change to the USAR described method of disassembling and reassembling the reactor or of handling irradiated fuel assemblies. l These radiation monitors only provide alarm / indication functions. l Actions taken in response to receipt of an alarm are manual in I nature. Compensatory measures are provided to ensure these actions I are still taken should the plant conditions warrant. Therefore, the possibility of creating an accident or a malfunction of equipment of a type different than previously evaluated is not created.
III. No. The D17-K650 and 1D21-N340 radiation monitors are not described in plant Technical Specifications. These radiation monitors only provide alarm / indication functions. Actions taken in response to receipt of an alarm are manual in nature. Compensatory measures are provided to ensure these actions are still taken should the plant conditions warrant. Therefore, the margin of safety is not l impacted. l l
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SE No.: 94-151 Source Document: ONI-B21-1, Rev. 3, 'IC-4 Description of Change This change to Off-Normal Instruction (ONI) B21-1 directs the operator to lower reactor power prior to attempting to close an inadvertently opened Safety / Relief Valve (SRV).
Sumary I. No. The operator action te reduce reactor power prior to closing the SRV is included within the scope of the USAR analysis for inadvertent safety / relief valve opening. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. The operator action to reduce reactor power prior to closing the SRV is included within the scope of the USAR analysis for inadvertent safety / relief valve opening. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated is not created.
III. No. Compliance with Technical Specification 3/4.4.2.1, Action b, is ensured. The USAR analysis for inadvertent safety / relief valve opening is not affected. Therefore, this instruction does not impact any margin of safety,
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Source Document: TXI-0208,..Rev. O, TC-7 i i
Escription of Change This temporary test instruction change adds a section to perform testing .
of the Emergency. Service' Water (ESW) keepfill portion'of the Service Water system.
Summary I. No. Procedurally established alternate methods of decay heat removal are available. Administrative controls are established to prevent ESW inoperability. Therefore, the probability of occurrence or the consequences of an accident or a malfunction of equipment previously ,
evaluated in the USAR has not changed. l II. No. Procedurally established alternate methods of decay heat removal are available. Administrative controls are established to prevent ESW inoperability. Therefore, the possibility of an accident or a -
malfunction of equipment important to safety of a different type than any previously evaluated in the USAR is not created. !
III. No. Administrative controls are established to prevent ESW-inoperability. Alternate methods of decay heat removal are !
available. Plant conditions at the time of the testing provide a i 9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> window in order to re-establish shutdown cooling. Therefore, the margin of safety has not been impacted. l l
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SE No.: 94-153 Source Document: DCP 92-097, Rev. 3 Description of Change Superseded by SE 94-187.
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- SE No.: 94-154 Source Document: TXI-0200, Rev. O Description of Change This temporary temporary test instruction provides the directions-necessary to perform dynamic flow testing of the Fuel Pool Cooling and Cleanup (FPCC-G41) system valves, G41-F280, F285, F290 and F295 in accordance with NRC Generic Letter 89-10.
Summary I. No. None of the accidents analyzed in Chapter 15 of the USAR take credit for FPCC operation, nor is the malfunction of FPCC a precursor to any of the accidents analyzed in the USAR. All of the actions performed by this instruction will ensure that the G41 system is operated within the bounds of the design basis. The valves will remain capable of performing their design basis function of closing during an Residual Heat Removal (RHR) LOCA event. Administrative controls within the instruction ensure that pool water chemistry remains within the normal operating bounds. Therefore, the probability of occurrence or the consequences of any accident or malfunction of equipment previously evaluated in the USAR will not l be increased by this instruction. j II. No. All automatic functions and interlocks of the G41 system will remain operable throughout the performance of this test. No credit is taken, from a USAR Chapter 15 accident standpoint, for G41 operability. The system will be operated within its normal design and operating parameters. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is j not possible. I III. No. The performance of this instruction does not alter any G41 system characteristics, setpoints or functions required by Technical Specifications. Administrative controls within the instruction will ensure that critical fuel pool parameters (e.g., temperature, flow, etc.) are maintained within required operating limits. Therefore, the margin of safety will not be reduced.
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.SE No.: 94-155, 94-156 Source Document: NR 94-S-600, Rev. O r NR 94-S-601, Rev. 1 Description of Change ,
1 These nonconformance reports evaluate the ' rework /use-as-is' disposition ;
which allows the continued use of'the Division 3 Diesel Generator fuel '
oil differential pressure switch design. The pressure switch has i insufficient range to sense the required differential pressure.
i Summary I. No. The scope of these NRs is limited to the Divisional 3 Diesel Generator and its associated fuel oil system. The disposition of the NRS will not affect the reliability or redundancy of onsite power. Compensating measures have been instituted to provide fuel !
oil filter differential pressure monitoring during diesel operation to assure that the condition cannot affect the operability of the
- Division 3 Diesel Generator. Therefore, the probability of occurrence or the consequences of an accident'or malfunction of i equipment has not increased. i II. No. Division 3 Diesel failure has been postulated as part of the Chapter 15 accident scenarios. As this is the worst case failure which could be associated with the operability of these switches, this failure is bounded by existing analysis. Compensating measures have been instituted to provide fuel oil filter differential pressure monitoring during diesel operation to assure that the condition cannot affect the operability of the Division 3 Diesel Generator. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The requirements of the original design are maintained by use of alternate / redundant methods of the subsystem monitoring. Therefore, no margin of safety has been reduced.
SE No.: 94-157 Source Document: ASI-0003, Rev. O NEI-0102, Rev. O Description of Change This Safety Evaluation evaluates the deletion of NEI-0102, " Nuclear Engineering Department (NED) As Low As is Reasonably Achievable (ALARA)
Review Comittee", in its entirety, and revises ASI-0003, 'ALARA Subcomittee" . The responsibilities, as identified in NEI-0102, were transferred to ASI-0003. ASI-0003 will provide a central focal point for all ALARA matters. In addition, ASI-0003 was revised to include a reduced membership, adding the provision to convene a radiological project team, requires primary and alternate members to discuss issues, and adds reporting requirements to the Central Safety Comittee.
Summary I. No. These instruction changes were determined to be administrative functions and will not affect any system, structure or component important to safety. The proposed activities concern the plant's ability to address issues associated with dose reduction. The proposed activities did not affect the comitment or intent of maintaining exposures ALARA. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The instruction changes are administrative functions and do not affect the physical design or safety functions of the plant.
Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The instruction changes do not adversely affect the Radiation Protection Program. The changes strengthen the ALARA program. The physical design or safety of the plant is not impacted. Therefore, no margin of safety has been reduced.
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i SE No.: 94-158 i Source Document: DCP 94-5063, Rev. 0 ,
Description of Change This design change installs four (4) lugs into the ceiling of the Service Water Valve Pit (SWVP). The SWVP is located in the yard to the east of. l the plant. The lugs will be installed to facilitate rigging and removal i of the Service Water valves located in the pit, P41-F0420 and P41-F0430. '
Summary I. No. The lugs were designed to withstand all applicable loading .
conditions. In addition, the only system located in the SWVP is the Service Water system. During use of the lugs to rig the Service ;
Water valves, the Service Water system must be out of service since !
the two valves in question are installed in series. Therefore, the probability of occurrence or the consequences of an accident or-malfunction of equipment has not increased. :
II. No. The installation of the lugs will not affect any plant system. When j the lugs are in use, there are no in-service systems in the SWVP. ,
Therefore, creating a new accident or malfunction of equipment that i has not been previously evaluated is not possible.
III. No. The addition of the lugs into the SWVP ceiling will not impact any plant system. The lugs have been designed to withstand all applicable loading conditions and when the lugs are in use no system is in service in the SWVP. Therefore, no margin of safety has been reduced. .
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1 SE No.: 94-159 Source Document: SCRs 1-93-1141 and 1-93-1142 Description of Change These setpoint changes increase the setpoints for the Loose Parts-i Monitoring (LPMS) system channels 7 and 8 from 0.5 ft-lbs to a value of 20% above the background noise level of the channels.
Summary I. No. The setpoint changes do not impact LPMS design or system compliance with Regulatory Guide 1.133. Failure of the LPMS does not impact the initiators of any accident, and the system is not required by the USAR for accident mitigation. USAR Section 4.4.6.1.3 " Safety Evaluation" states, "The LPM system is intended to be used for information purposes only by the plant operator. The operator does not rely on the infomation provided by the LPM for the performance of any safety-related action." Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not. increased.
II. No. The purpose of the LPMS system is to detect a loose metallic part internal to the reactor vessel. The operability of the LPMS does not change the probability of the presence of a loose part.
Compliance with Regulatory Guide 1.133 is maintained. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The location of these sensors, on the Reactor Recirculation system discharge risers, makes it impossible to leave the setpoints at 0.5 ft-lbs due simply to the background noise levels generated by the recirculation pumps and the water flow through the piping. Use of the new setpoints does not impact LPMS operability or compliance with Regulatory Guide 1.133. Therefore, no margin of safety has been reduced.
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SE No.: 94-160 Source Documenl: Physical Security Plan, Rev. 9 Description of Change This-evaluation analyzes changes made-to the Physical. Security ,,
Plan (PSP). The changes have been evaluated'to ensure that the i effectiveness of the Perry Nuclear Power Plant Security Plan has not been
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reduced and to ensure that the requirements of 10CFR73, Physical .
Protection of Plants and Materials, are met. Site Protection must be l contacted for further details since this is considered "SAFBGUARDS' information. i Summary j i
I. No. The PSP describes the comprehensive Physical' Security Program and does not direct the operation of plant systems or equipment.
Therefore, the PSP changes do not affect the occurrence or ;
consequences of an accident or malfunction of equipment. l II. !!a. The PSP does not direct the operation of plant systems or-equipment-and, therefore, does not create the possibility for an accident or ,
malfunction. )
III. No. The PSP changes do not reduce any margin of safety.
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SE No.: 94-161 Source Document: DCP 94-0092, Rev. O Description of Change Superseded by SE 94-181, k
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SE No.: 94-162 Source Document: TXI-0191, Rev. O Description of Cnange Due to modifications that were performed on the three Reactor Feedwater-L Pumps (RFP), pump flow characteristics will be changed. This tenporary.
j test instruction will collect data as the Feedwater system is being manipulated during power ascension to generate new pump curves.
Summary I. No. Previously approved plant operating instructions will be used,to direct plant operations and power changes. This instruction will collect feedwater flow and pressure data during power ascension.
The USAR loss of feedwater flow event remains bounding during the performance of this test. The test will not alter the physical plant. With the exception of manipulating the RFP's minimum flow valve during the test, feedwater operations will be consistent with system design and operational capability. Therefore, the probability of occurrence of the consequences of an accident or malfunction of equipment will not be increased. :
II. No. This instruction will collect feedwater flow and pressure data ~j during power ascension. The USAR loss of feedwater flow event !
remains bounding during the performance of this test. The test will i not alter the physical plant. With the exception of manipulating the RFP's minimum flow valve during the test, feedwater operations will be consistent with system design and operational capability.
Therefore, the possibility of an accident or malfunction of equipment of a different type than previously evaluated is not created.
III. No. This instruction does not direct operations of the Feedwater system outside of its design basis. This test cannot direct operation of the Feedwater system such that the 143% Nuclear Boiler Rated (NBR) flow requirement could be exceeded. This instruction in no way effects this value or any other Technical Specification basis.
Therefore, there is no reduction in any margin of safety.
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SE No.: 94-163 Source Document: TXI-0194, Rev. O Description of Change This temporary test instruction test will determine the runout capability of the Reactor Feedwater Pumps (RFP) A and B. The instruction is basically a data gathering instruction. Feedwater pump / turbine flow, pressure, and speed parameters will be recorded as reactor power is increased from 60 to 100%. This data will be extrapolated out to the turbine's high speed stop settings. A maximum feedwater flow will then be determined.
Summary I. No. USAR Section 15.1.2 discusses Feedwater Controller Failure-Maximum Demand. The accident assumes the feedwater flow control fails such that a maximum feed demand is inputted to the Reactor Feedwater Pump Turbine (RFPT). The RFPTs will increase speed until their high speed stop settings are reached. At the high speed stops, feedwater flow is required to be below 143% of Nuclear Boiler Rated (NBR) flow. A runout flow of 143% NBR is acceptable and ensures Minimum Critical Power Ratio (MCPR) safety limits. The methodology of the instruction is based on G.E. Design Specification 386HA567 and G.E.
Startup Test Specification 22A7637. Only one RFP will be placed on the Master Level Controller (MLC) until the results of this test are analyzed. This will ensure if the Feedwater Flow Failure-Maximum Demand Accident were to occur during startup, the 143% NBR flow would not be exceeded. Only the RFP on the MLC would speed up to its high speed stop. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The specific feedwater accidents discussed in the USAR are: loss of feedwater flow, maximum feedwater flow and feedwater line break.
This instruction simply gathers feedwater data as the system is progressed through power ascension. Plant procedures are followed through all Feedwater system manipulations. The main difference from normal Feedwater system operations specified in this test is maintaining only one RFP on the Master Level Controller until completion of the test. This mode of feedwater operation is fully covered in the system operating instructions. The three above mentioned accidents are still bounding. No new Feedwater system operations are introduced by the performance of this test. The purpose of this test, as previously stated, is to verify runout capability and ensure 143% NBR flow is not exceeded. At no time during the perfcrmance of this test will that limit be exceeded.
Operation of the components in the Feedwater system are within their design qualifications. No new Feedwater system component operations are introduced by the performance of this test. Therefore, creating i a new accident or malfunction of equipment is not possible. l 1
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Summary (Cont.). t III. No. This instruction does not direct operations of the Feedwater system outside of its design basis. This instruction cannot direct j operation of the Feedwater system that could exceed the 143% NBR ,
flow requirements. Therefore, there is no reduction in any margin- !
of safety.
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SE No.: 94-164 Source Document: PTI-N27-P0004, Rev. 1 Description of Change ,
This Periodic Test Instruction (PTI) provides for the testing of the.
emergency overspeed governor and tripping mechanism of the Reactor Feedwater Pump Turbine (RFPT). The test verifies the overspeed trip mechanisms will trip the RFPT prior to the occurrence of equipment damage. Performance of the test requires the RFPT to be uncoupled from the Reactor Feed Pump (RFP). Depending on plant conditions, up to three RFPT trips can be bypassed during the performance of this test. The bypassed trips are: high reactor vessel level, loss of all running reactor feedwater booster pumps and RFP suction valve closed.
Summary I. No. A requirement of this PTI is to uncouple the RFPT from the RFP prior to the performance of the test. This renders the RFP out of service and completely independent from its associated RFPT. Testing of the uncoupled RFPT in no way can effect feedwater flows or reactor vessel levels. If this PTI is performed with the plant on-line, Feedwater system operation is permitted with one RFP out of service.
Bypassing several RFPT trips during this PTI has no relevance to feedwater operations or plant safety, since the associated RFP cannot effect feedwater flow. The USAR Chapter 15 accidents are unaffected as a result of the bypassed trips because the RFP is out of service. USAR Section 10.4.7.2, Feedwater system and USAR ,
Section 7.7.1.4, Feedwater Flow Control system are both unaffected 1 since the RFPT and RFP are out of service during the performance of this test. A concern during any turbine overspeed test is the !
possible generation of turbine missiles. The PTI has administrative l controls to manually trip the RFPT if its overspeed setpoint is j exceeded and it did not trip. Therefore, this precludes any chance '
of generating RFPT blade missiles as a result of this PTI.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The PTI simply uncouples a RFP and RFPT and then performs an overspeed test on the RFPT. The bypassed RFPT trips are irrelevant i because the effected RFP is uncoupled. There are no new line-ups or 1 feedwater operations as a result of this PTI. The discussed bypassed trips are only performed for a RFP/RFPT that is uncoupled and out of service. The remaining portion of the Feedwater system is unaffected by the performance of this PTI. Therefore, creating a new accident or malfunction of equipment is not possible.
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r III. No. The reactor vessel high level RFPT trip.is addressed in Technical ;
Specifications Section 3/4.3.9. The purpose of this trip is . !
mitigate.the consequences of a failure of a feedwater controller j under maximum demand. As explained, the high level trip is only ;
bypassed for an uncoupled RFP. With the RFP uncoupled, there is no !
impact upon reactor vessel level. Therefore, no margin of safety j has been reduced.
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, SE No.: 94-165 Source Document: NR 94-N-160, Rev. 3 Description of Change This nonconformance report evaluates the "use-as-is' disposition of eight f- nonsafety, non-seismic small bore piping root valve appendages in.the Condensate system which have undersized welds, lack of fusion and arc strikes.
Summary l
l I. No. Plant operations are unaffected by postulated line breaks caused by l- the assumed root weld failures. -Flooding is adequately handled by the floor drains and sump pump. It is bounded by the USAR flooding analysis. Condensate flow is controlled by flow control valves and the USAR loss of feedwater flow and loss of feedwater heating analyses are not affected. The area does not have safety-related equipment so there is not a jet impingement or a falling appendage concern. These appendages are not needed for safe shutdown.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. do. The flooding from the postulated line breaks is bound by the USAR analysis for a failure of the condenser circulating line. The plant effluents would not be changed. The Condensate system is nonsafety-related and is housed in nonsafety-related buildings. It does not support safe shutdown and does not play a role in the operation of reactor safety features. Therefore, the possibility for an accident or malfunction of a different type than previously-analyzed in the USAR is not created.
III. No. The presence of these defective welds in the Condensate system will not affect the safe operation of the plant. No safety margins are reduced.
.SE No.: 94-166 Source Document: NR 92-S-032. Rev. 2 NR 92-S-232, Rev. 1 l Description of Change These nonconformance reports evaluate the temporary *use-as-is' l disposition of a six-inch relief discharge header.that does not conform I to the ASME code. The NRs require that'the Steam Condensing Mode (SCM) l of the Residual Heat Removal system be declared inoperable.
Summary
) I. No. The SCM is not an initiator or a mitigating function in the.USAR Chapter 15 accident analysis. All other modes of Residual Heat Removal (RHR) system operation are unaffected. Therefore, the probability of occurrence or the consequences of a previously analyzed accident or malfunction of equipment will not be increased.
II. No. The change does not compromise the ability to safely shutdown the reactor. Reliability of the RHR (excluding SCM) is not adversely affected. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The margin of safety as defined in the bases to Technical Specifications Section 3/4.6.3 refers to design parameters for the suppression pool and Section 3/4.6.1.5 refers to the structural integrity of containment. The temporary inoperable status of SCM has no impact on these limits. Reliability of the RHR system (excluding SCM) is not adversely affected. Therefore, this change will not reduce any margin of safety.
SE No.: 94-167, 94-194,94-201 Source Document: TXI-0179, Rev. O TXI-0179, Rev. O, TC-4 TXI-0179, Rev. 1 Description of Change This temporary test instruction tests the turbine generator system to assure that the system is not subject to damage by torsional vibration resonance frequencies near 120 Hz. The test is conducted with the generator isolated from the grid, reactor power at 15% or less and generator load less than 5%. Temporary modifications are made to conduct the test. These modifications are restored after test coupletion.
Summary I. No. Potential failure mechanisms were evaluated and found to have no impact on accident analysis contained in the USAR. The test is conducted at power levels much lower than those assumed in the accidents evaluated in the USAR. The effects of a fire are bounded by existing USAR analysis. Equipment failures have a likelihood of occurrence no greater than that encountered during normal operation.
The test set-up ensures that the test has no impact on other ,
systems. Therefore, the probabilit" of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. The test is conducted with reactor and turbine running within normal parameters at less than 15% power. The test set-up ensures that the test has no impact on other systems. No new failure modes are created. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR is not created.
III. No. The test set-up ensures that the turbine has adequate overspeed protection while the test is being performed. The test set-up ,
ensures that power is available to supply safety-related equipment required for safe shutdown, and mitigation and control of accidents.
Therefore, the margin of safety has not changed. j l
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SE No.: 94-168 Source Document: DCP 04-0072, Rev. O DCP 94-0072, Rev. 1 Description of Change This design change replaces defective Service Water (P41) system fiberglass piping with carbon steel piping. Approximately 200 feet of 24' fiberglass pipe will be abandoned in place. Approximately 175 feet of 24' carbon steel piping is routed to bypass the abandoned fiberglass line. The overall configuration of the Service Water system is not significantly altered. Pressure loss changes associated with the new routing are insignificant. As a result, there is no effect on system pump capacity or overall system performance.
Summary I. No. USAR Section 9.2.7.3 states that the P41 system is nonsafety-related and is not required for safe shutdown of the reactor. This modification removes defective Service Water System fiberglass piping which will increase the reliability of the system. Flow requirements to supplied equipment are not adversely effected.
System and plant operation will remain virtually unchanged. All equipment, naterial and services associated with this modification will meet the applicable nonsafety piping design code requirements.
Flooding concerns inside and outside of plant buildings associated with this modification are bounded by existing USAR analyses.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR will not change.
II. No. This modification involves enhancement of the nonsafety-related Service Water system by providing a more reliable piping flow path.
This system is not required for safe shutdown of the reactor and its impact on safety-related systems / components is unchanged. All equipment, naterial and services associated with this modification will meet the applicable nonsafety piping design code requirements.
Flooding concerns inside and outside of plant buildings associated with this modification are bounded by existing USAR analyses.
Therefore, the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the USAR is not created.
III. No. This design change is consistent with the current system design and does not alter system or plant operation. The operation of the P41 system is not addressed in any Technical Specification. All equipment, material and services associated with this modification will meet the applicable nonsafety piping design code requirements.
Flooding concerns inside and outside of plant buildings associated with this modification are bounded by existing USAR analyses.
Therefore, no margins of safety are reduced.
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.SE No.: 94-169 Source Document: PAP-1912,'Rev. 4, TC-3 Description of Change This change to Plant Administrative Procedure-(PAP) 1912 provides clarification for the protection of combustible materials within 35 feet of a hot' work areas.
Summary l I. No. The changes are administrative in nature and are found to be consistent with the fire protection requirements of the USAR'and with NFPA 51B. Therefore, the probability of an occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR is not increased.
II. No. These changes address the manner in which combustible materials are protected when exposed to hot work activities and are consistent with the fire protection requirements of the USAR. The changes do not impact any plant system or component. Therefore, the possibility of creating an accident or malfunction different from any previously evaluated in the USAR does not exist.
III. No. The changes are administrative in nature and do not impact any activity described in the Technical Specifications. Therefore, no margin of safety will be reduced.
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SE No.: 94-170 Source Document: PS'IU, Rev. 2, TC-7 Description of Change This change to the Perry Specific Technical Guidelines (PSTG) evaluates the addition of a deviation sheet explaining the rational used to allow use of revised containment parameters (approved in License Amendment 57).
Summary I. No. License Amendment 57 changed the values for containment average air temperature, suppression pool average temperature, and suppression pool water level. The Plant Emergency Instructions (PEI) operator actions will not be impacted since the actions are to restore the affected parameter to the normal value. Therefore, the probability ;
of occurrence or consequences of an accident or a malfunction of equipment previously evaluated in the USAR is not changed. 4 II. No. License Amendment 57 changed the values for containment average air temperature, suppression pool average temperature, and suppression pool water level. The Plant Emergency Instructions (PEI) operator j actions will not be impacted since the actions are to restore the .
affected parameter to the normal value. Therefore, the possibility I of an accident or malfunction of equipment important to safety of a different type than previously analyzed in the USAR is not created. l III. No. License Amendment 57 changed the values for containment average air temperature, suppression pool average temperature, and suppression pool water level. The Plant Emergency Instructions (PEI) operator actions will not be impacted since the actions are to restore the affected parameter to the normal value. Therefore, no margin of safety is reduced.
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SE No.: 94-171 Source Document: DCN 4743, Rev. O i
Description of Change This drawing change implements editorial changes to P&ID D-302-606 and the 607 series drawings associated with the Reactor Recirculation (B21) system. The change corrects drawing cross-references and component labeling.
Summary I. No. This drawing change is an editorial revision to B21 system drawings.
These changes do not affect the Chapter 15 accident analysis, physically modify the plant, or its operating program. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This drawing change is an editorial change to correct documentation.
The drawing change will not impact accident analysis as described in Chapter 15 of the USAR. Therefore, creating a new accident or malfunction of equipment is not possible. ;
III. No. This activity is an editorial change to system documentation. The plant or its operation is not affected. Therefore, the margin of safety has not been changed.
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Source Document: DCN 4738, Rev. O I Description of Change !
This drawing change implements an editorial revision to P&ID D-302-872, Control-Rod Drive Hydraulic System. The revision corrects component MPL l numbers.
Summary t I. No. This drawing change is an editorial change to the P&ID for the C11 [
system. The change corrects a MPL number on a valve. This l editorial change does not affect the Chapter 15 accident analysis, l physically modify the plant, or its operating program. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. This is an editorial change to documentation to correct the MPL }
designation of a valve. This editorial change does not affect the ,
Chapter 15 accident analysis, physically modify the plant, or its
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operating program. Therefore, creating a new accident or malfunction of equipment is not possible. t III. No. This activity is an editorial change to system documentation. This editorial change does not affect the Chapter 15 accident analysis, ;
physically modify the plant, or its operating program. Therefore, t no margin of safety has been reduced. i i
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.SE No.: 94-173-Source Document: DCN 4736, Rev. O Description of Change This drawing change implements an editorial revision to P&ID D-302-832, Combustible Gas Control Hydrogen Analysis System. The revision removes-the flow switch symbol at the discharge of each sample pump since the switches do not exist. Additionally, a component MPL number was corrected.
Sunnary I. No. This drawing change is an editorial change to the P&ID for the Combustible Gas Control system. This editorial change does not affect the Chapter 15 accident analysis, physically modify the plant, or its operating program. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The drawing change is an editorial change to correct documentation.
The drawing change will not impact USAR Chapter 15. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The plant is not physically changed. USAR accident analysis is not affected. Therefore, no margin of safety has been reduced.
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SE No.: 94-174 Source Document: PS7G, Rev. 3, TC-1 Description of Change This change to the Perry Specific Technical Guidelines (PSTG) evaluates a change which incorporates comments received during the verification and-validation of the associated Plant Emergency Instructions (PEI) flowcharts.
Summary I. No. This change makes editorial revisions to the PSTG, none of which result in technical changes to the PSTG or the PEIs. The change does not adversely impact the plant design or any analysis used in the design. Systems and components are not affected. Therefore, the probability of occurrence and consequences of an accident or a malfunction of equipment previously evaluated in the USAR is not changed.
II. No. This change makes editorial revisions to the PSTG, none of which result in technical changes to the PSTG or the PEIs. The change does not adversely impact the plant design or any analysis used in the design. Systems and components are not affected. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR is not created.
III. No. No adverse impact upon the plant design or any analysis used in the design is created. No adverse affect upon systems or components required for safety is created. Therefore, the margin of safety has not been impacted.
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L SE No.: 94-175 Source Document: USAR Change Request 94-053-Description of Change This USAR change adds Regulatory Guide 1.147, Rev. 10 to USAR Table 1.8-1. The Regulatory Guide endorses ASME code cases accepted by, the NRC.
Sumary I. No. This change adds a Regulatory Guide to the USAR. Application of specific ASME code cases included within the Regulatory Guide will be accomplished by revisions of appropriate plant procedures. The code cases listed provide alternative rules or requirements to meet the ASME code. The code cases do not alter any design requirements for plant systems. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. This change does not modify any existing plant systems. The change does not increase the potential for common mode / common cause failures. In addition, this change does not create any new equipment types or alter any system design requirements. Therefore, this change will not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The ASME code cases included in Regulatory Guide 1.147 were issued by the ASME Boiler and Pressure Vessel Committee. The application of these code cases through approved plant procedures assures.
conformance to ASME code requirements. Therefore, this change will not reduce any margin of safety.
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SE No.: 94-176 Source Document: DCN 4044, Rev. O Description of Change This drawing change provides for the equivalent replacement of reactor Shroud Head Stud (SHS) bolt retainer assemblies. These assemblies provide a means of transmitting torque to the shroud head studs during reactor pressure vessel assembly and disassembly, and also provide a locking function during normal operation. This change also provides the design criteria for removing worn SHS bolt retainers by determining the number of retainers which are required to ensure that the minimum number shroud head studs remain functional. This change is based on a General )
Electric Perry Unit 1 Shroud Head Stud / Bolt evaluation. l Summary I. No. This change maintains all original design criteria and performance requirements for the shroud head assembly and the SHS bolt retainer assemblies. When installed the replacement assemblies are of equivalent design. The relative potential for increased loose parts from shroud head scuds coming unscrewed after removal of up to sixteen locking mechanisms is negligible compared to the reduced potential of loose parts due to worn retainer / locking mechanisms.
Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. This change maintains compliance with ASME Section III, subsection NG code requirements. Thus, there is no increased potential for shroud head stud or assembly failure. There are no new failure mechanisms introduced by the use of an equivalent retainer or the removal of SHS bolts as provided by this change.
Although this change can result in slightly higher shroud head stud loads, this increase has been evaluated to be insignificant and will not result in increased potential for Intergranular Stress Corrosion Cracking (IGSCC) of the studs. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. This drawing change maintains compliance to original design and performance requirements including design loads and minimization of loose parts and IGSCC. Thus, there is no reduction in any margin of safety.
SE No.: 94-177 Source Document: TXI-0193, Rev. O Description of Change This temporary test instruction provides the directions necessary to perform dynamic flow testing of the following Reactor Core Isolation Cooling (RCIC) system valves: 1E51-F019, 1E51-F022, and 1E51-F059. The testing is in accordance with NRC Generic Letter 89-10. The test methodology for the 1E51-F022 and 1E51-F059 will involve cycling the motor operated valves, one at a time, against RCIC pump total discharge head minus the pressure drop associated with minimum flow valve, 1E51-F019, operation. Data for 1E51-F019 will be obtained when the valve opens / closes automatically due to changes in total system flow rate when 1E51-F022 is manipulated.
Summary I. No. The RCIC system will be operated within normal system operating and design parameters. Administrative controls within the instruction are provided to ensure that: the system is declared inoperable while under test so that appropriate Technical Specification actions can be taken; suppression pool level and temperature are maintained within required Technical Specification limits; and CST level is monitored and maintained greater than 250,000 gallons to ensure adequate Net Positive Suction Head (NPSH) exists for the RCIC pump.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. As stated previously, the RCIC system will be operated within normal and design parameters (e.g., pressure, flow, rpm, etc.). 1 Additionally, administrative controls within the instruction ensure j that equipment damage is prevented and entry into the appropriate l Technical Specification actions ensures availability of other i systems to mitigate the consequences of an accident. Therefore, i creating a new accident or malfunction of equipment is not possible.
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i III. No. System actuation setpoints required by Technical Specifications are l not effected by this instruction. The RCIC system will be capable i of operating to mitigate the consequences of any accident. However, ,
as previously stated, the system will be declared inoperable during this test. Technical Specification action statements will ensure that adequate safety systems are available to mitigate the l consequences of any accident. Therefore, no margin of safety has i been reduced. I
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Source Document: DCP 94-0019, Rev. 1 Description of Change !
t This design change installs.a nonsafety-related digital strip chart :
recorder and analog signal conditioner on the 1H13-P811 Generator '
Recorder Panel". The recorder will be setup to monitor the main !
generator negative sequence current (I2), which is an indicator of three i phase electrical load imbalance, and the generator output frequency. l When the levels of I2 or generator frequency are outside of desired i operating limits the recorder will initiate a common annunciator-on the i 1H13-P680 benchboard.
I Summary i
I. No. Failure of this equipment to perform its intended function causes no i safety concern as the recorder is for indication only and has no j interlocks or other control functions which can malfunction in a :
manner detrimental to safety equipment. The worst case failure !
postulated is a short circuit on the input side of the frequency i transducer which could cause the Potential Transformer (PT) fuse to open. A Control Room annunciator on H13-P680-9A would alert the i operator of this condition. Therefore, this equipment cannot .
increase the probability or the consequences of an accident or j malfunction of equipment previously evaluated. .
II. No. This recorder allows plant operators access to plant operating ,
information. No control functions are provided with this equipment. j Electromagnetic Interference (EMI) data has shown levels well below -
Perry's existing envelope. USAR Chapter 15, Section 15.2.2 :
' Generator Load Rejection
- event bounds this modification and i
therefore, no new accidents are created by the addition of the recorder. Recorder software failures and human errors are not of ,
concern since the recorder provides a nonsafety monitoring function
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only and no control functions are provided. These failures are bounded by the worst case failure causing a PT fuse to blow. t Therefore, creating a new accident or malfunction of equipment is ;
not possible.
III. No. The equipment installed by this activity is for monitoring purposes !
only and does not affect the systems being monitored. No Technical i Specification actions are required based on the lack of protective l voltage generator trips, and generator operation is not degraded.
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Therefore, no margin of safety has been reduced. '
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SE No.: 94-180 Source Document: PAP-ll20, Rev. 2 Description of Change This revision to Plant Administrative Procedure-(PAP) 1120 incorporates.
changes to the containment pressure testing methodology; changes to the suppression pool and the upper pool water levels; changes to the containment average air temperature; and changes to the suppression pool average water temperature. l Summary I. No. This procedure is an administrative program to track and trend the LLRT data collected to ensure compliance with respect to the overall .
Leak Design Basis Accident (LDBA). The methods utilized in the !
collection of the LLRT data will remain unchanged. The new parameters were reviewed and approved for use by the NRC.
Therefore, the probability of occurrence or the consequences of an )
accident or nalfunction of the equipment important to safety .
previously evaluated in the Updated Safety Analysis Report (USAR) is l not increased. l II. No. This procedure is an administrative activity. A containment response analysis performed by General Electric (GE), as well as a structural design and operation impact review performed by Gilbert Associates, Incorporated (GAI) were utilized to verify that the changes would not adversely affect the LDBA analysis. Therefore, the possibility of an accident or malfunction of a different type than previously evaluated in the USAR has not been created.
III. No. This procedure does not compromise the containment response safety analysis. Therefore, the margin of safety has not been reduced.
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5 SE No.: 94-181 Source Document: DCP 94-0092, Rev. O DCP 94-0092, Rev. 1 DCP 94-0092, Rev. 2 Description of Change This design change provides a safety grade source of air to the outboard Main E.eam Isolation Valves (MSIV's). The new backup air source will be used to maintain the valves leak tight in the closed position following a design basis accident. The air source for this new supply is one of the two safety-related Instrument Air system low pressure storage tanks which also provide air for the Automatic Depressurization system safety / relief valves.
Summary I. No. This design change adheres to established design criteria and established codes. It provides for enhancements for minimizing post-accident main steam line penetration leakage without degrading the safety functions of the systems affected. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The design change maintains the redundancy and independence of the affected systems such that susceptibility to common mode or common cause failures is not possible. In addition, the changes met the general design criteria for single failures of the affected equipment. Postulated failures of the new equipment do not introduce the possibility of a new accident. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The new safety grade air supply is constructed and maintained in ,
accordance with the American Society of Mechanical Engineers (ASME) l Boiler and Pressure Vessel Code and thus maintains the margin of ;
safety associated with Technical Specification (T.S.) 4.0.5 for l system pressure boundary integrity. The changes provide added assurance that the outboard MSIV's will limit penetration leakage per T.S. 3/4.6.1.2 without degrading the safety functions of i safety-related Instrument Air or the Automatic Depressurization l systems (T.S. 3/4.5.1). Therefore, there are no reductions in the margins of safety, i
SE No.: 94-182 Source Document: DCP 91-0060A, Rev. O DCP 91-0060A, Rev. 1 Description of Change This design change replaces a 25 kva non-regulating distribution transformer in Motor Control Center (MCC) EF1C07 with a floor-mounted 15 kva regulating transformer. The fused disconnect power supply switch is changed from 100 ampere switch /100 ampere fuse to 60 ampere switch /40 ampere fuse. The regulating transformer provides output voltage of 120VAC + or - 2% over a range of -15% to +10% input voltage to accommodate changes in plant voltage under varying operating conditions.
Summary I. No. Individual accidents and transients relevant to the design change are Generator Load Rejection, Loss Of Offsite Power (LOOP) and Loss Of Coolant Accident (LOCA). A failure modes and effects analysis was performed to examine failures of the regulating transformer and the impact on plant systems and the occurrence of these accidents.
The analysis concluded that a transformer failure does not initiate any of these events. The transformer is qualified and installed in accordance with Class 1E requirements. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. Per analysis, a failure of the regulating transformer cannot initiate a failure of Division I equipment and cannot cause a failure of connected Division II equipment. Analysis shows that the failure effects of the regulating transformer are no more severe than the failure effects of the non-regulating transformer that it replaces. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR is not created.
III. No. Analysis shows that the use of a regulating transformer is equivalent to the non-regulating transformer that it replaces. The installation satisfies the requirements of Regulatory Guide 1.81, IEEE Standard 308, Regulatory Guide 1,32 and General Design l Criteria 18. The design accommodates degraded grid voltage requirements of Branch Technical Position PSB-1. The margin of safety has not been changed.
SE No.: 94-183 Source Document: DCP 93-0020A, Rev. O Description of Change This design change provides a testable flanged spacer on the Reactor Core Isolation Cooling (RCIC) turbine exhaust line to replace the existing flanged spacer. The new spacer design is equipped with a test connection to allow local leak rate testing of the flanged joint. This new arrangement eliminates the need for an external test box to perform the testing.
Summary I. No. This change adheres to estEnlished design criteria and established codes and standards for the RCIC system. Excluding minor design differences necessary to allow in-place testing, the new spacer is identical to the existing spacer. The new design provides an improved means of ensuring the leak tightness of the mechanical joint without degrading the capability of the RCIC system to perform its design function. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The design change affects only the RCIC turbine exhaust piping and does not affect any other system. The structural configuration of the bolted joint remains unchanged. The component change is necessary only to permit in-place testing. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The new testable spacer supports 10CFR50 Appendix J 1eak testing pursuant to Technical Specification 3/4.7.3 for primary containment leakage. In addition, the new spacer meets the design, material and construction standards for the RCIC system and does not alter the operation or performance characteristics of the RCIC system. Thus, )
there is no reduction in the margin of safety.
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SE No.: 94-184, 94-188 Source Document: DCP 94-139, Rev. O DCP 94-139, Rev. 1 Description of Change ine keepfill water for the Emergency Service Water (ESW) system is normally provided by the Service Water (SW) system. This design change relocates the keepfill water to a point upstream of a 9:grvice water heat ,
exchanger (turbine lube oil cooler).
Summary I. No. This modification will not affect the ESW system. There is no impact on SW pump operation or SW system performance. The SW system is nonsafety-related and not required for safe shutdown of the reactor. Fission product barriers are not affected. If the keepfill breaks, the flooding is bounded by existing USAR flood evaluations. The welding will not have an adverse metallurgical effect. The new piping conforms to the design code. Therefore, the probability of occurrence or consequence of an accident or a malfunction of equipment important to safety as analyzed in the USAR is not increased.
II. No. This modification does not affect the performance or operation of the SW or ESW systems. The design code is met. The temperature, 1 pressure or flow conditions in the ESW or SW systems are not !
altered. Potential flooding is bounded by existing USAR analysis.
The SW system is not needed for safe shutdown and its impact on safety-related systems is unchanged. This change will not create the possibility for an accident or malfunction of a different type than evaluated in the USAR.
I III. No. This change complies with ANSI /ASME B31.1. ESW or SW system l operations are not affected. The margin of safety is not reduced.
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_SE No.: 94-186 Source Document: DCP 94-5132, Rev. O !
Description of Change This design change installs a blind coupling in the common 2" header for the outboard Main Steam Isolation Valve (MSIV) before seat drain lines.
The blind coupling will eliminate the drain line as a potential secondary <
containment bypass leakage path. This design change will result in the Inboard MSIV Leakage Control System (MSIV-LCS) being inoperable at low power levels. An NRC approved licensing amendment has provided exception to the provisions of Technical Specification 3.0.4 as it applies to "
Technical Specification 3/4.6.1.4 for Cycle 5 of plant operation. ;
g Summary {
I. No. The coupling will not degrade equipment nor alter the drain piping '
that would make it more susceptible to failure. The change will be consistent with the original standards and installation practices, l ASME Code Section III, Subsection NC. The change will not impose i additional loads, will not degrade support system performance, will ,
not change the frequency of system or equipment operation and will !
not affect protective features. The coupling will not cause nor j create conditions conducive to any adverse environment. Further, the j modification will not change, degrade or prevent actions described '
or assumed in USAR accident scenarios. Therefore, the probability l of occurrence or the consequences of an accident or malfunction of ,
equipment has not increased. j i
II. No. The material, design and installation of the coupling will comply with the applicable codes and standards. The coupling will have no adverse effect on systems or the way they will react to normal and abnormal transients. The coupling will not create or compromise the functioning of any new or existing systems or equipment. The change will not result in any equipment failures and will not affect any accident initiators or contributors. No new initiators or contributors will be created for an event which could be considered a new accident. Therefore, the probability of an accident or malfunction of equipment of a different type will not be created.
III. No. The coupling will be installed in accordance with the ASME code and it will not degrade the capability of systems to mitigate the effects of postulated accidents. The margin of safety will be maintained.
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SE No.: 94-187 Source Document: DCP 92-0097, Rev. O DCP 92-0097, Rev. 1 DCP 92-0097, Rev. 3 DCP 92-0097, Rev. 4 DCP 92-0097B, Rev. O Description of Change These design changes modify the Residual Heat Removal (RHR-E12) system Loops A and B, the Containment Spray (E15) system, and the Low Pressure Core Spray (E21) system. The design changes include:
- modifications to the internals of the RHR heat exchanger outlet and bypass valves 1E12-F003A/B and 1E12-F048A/B,
- modifying the diameters of restricting orifices lE12-D004A/B, 1E15-D003A/B, and 1E21-D004,
- modifying the control circuits for the lE12-F024A/B valve to provide for administrative 1y controlled throttling capability, and
- locks closed the cooling water discharge valve 1M51-F0591A/B to the Combustible Gas Mixing Compressor Aftercooler.
Summary I. No. These design changes do not affect the original plant design basis.
The probability of a E12/E21 system pipe break or component malfunction is not changed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Potential failure mechaniscs and system responses to such failures are bounded by those previously analyzed in the USAR. These design changes create no new systems and maintain the original plant design basis. These changes do not affect any required system test frequencies or compromise the availability of any system. These design changes do not increase the potential for a common mode failure beyond that which existed in the original design. Hence, these changes do not create the possibility of a different type of accident or malfunction of equipment important to safety than any evaluated previously in the USAR.
III. No. Technical Specifications related to ECCS Instrumentation, Remote Shutdown, RHR, ECCS (Operating and Shutdown), Primary Containment, Depressurization Systems, Containment Isolation, Refueling, Electrical Power System, Refueling Operations and associated bases are not affected by these design changes. The availability and reliability of the E12, E15, and E12 systems are not changed. Thus, no margin of safety has been reduced.
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SE No.: 94-190 Source Document: NR 94-N-703,.Rev. 0 l Description of Change This nonconformance report analyzes the ' scrap / repair /use-as-is" disposition of flow instrumentation in the Circulating Water (N71) system. Specifically, flow element 1N71-N230A will be scrapped, and flow j indicator 1N71-R231A will be removed. A blind flange will be installed in the' location where IN71-N230A was removed. ),
Sumary I. No. The instrumentation being removed provides local indication only.
The indication is not used for N71 system operation. There are no safety functions associated with this instrumentation. LThe ;
installation of the blind flange satisfies the requirements of ANSI B31.1. Hence, piping integrity is maintained. This disposition maintains the N71 design basis. Therefore, the probability of occurrence or the consequences of an accident or
- malfunction of equipment important to safety has not been increased.
II. No. This disposition maintains the N71 design basis. The changes are installed in accordance with the applicable code requirements. N71 operation is not affected. Therefore, the possibility of an _
accident or malfunction of equipment important to safety of a type different than previously evaluated is not created. l l
III. No. This disposition maintains the N71 design basis. The changes are installed in accordance with the applicable code requirements. N71 ,
operation is not affected. Therefore, no margin of. safety has been reduced.
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Source Document: USAR Change Request 94-037 Description of Change This USAR change evaluates the application of the Control Room Isolation Status Panel to ensure that it is consistent with the containment l isolation systems design bases, evaluates changes to eliminate l inconsistencies between USAR Table 6.2-32 and the design bases, and .
clarifies information associated with the hydrogen igniter control circuitry. ;
. Summary j
'I. No. The information presented on the indication status panel does not affect the operation of the containment isolation systems, nor does ',
it impact the design bases of the containment isolation systems as described in USAR 6.2.4.2.1. The isolation status is an operator aid for quick reference of isolation valve position status. The clarification of the hydrogen igniter control operation does not change the design bases of the Hydrogen Control system described in USAR Section 6.2.8.1. Therefore, the probability of occurrence or :
the consequences of an accident or malfunction of equipment ;
important to safety previously evaluated in the USAR has not been l changed. j II. 23. The changes are consistent with design bases. Therefore, a-different type of accident or malfunction of equipment than those i previously analyzed is not created. ;
1 III. No. The changes for the containment isolation systems and the Hydrogen Control system will function as required per the USAR and the Technical Specifications. Therefore, no margin of safety has been reduced.
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SE No.: 94-192 Source Document: DCP 93-0116, Rev. O DCP 93-0116, Rev. 1 Description of Change This design change replaces the R-11 refrigerant of the Offgas Vault Refrigeration system with Trichloroethylene, R-1120. Replacement is necessary to ensure EPA regulations regarding use of non-ozone depleting substances are met. This modification also provides a dike for leakage containment and local indication of atmospheric R-1120 concentration with trouble alarm annunciation in the Control Room.
Summary I. No. The R-1120 refrigerant ambient properties differ from R-11 refrigerant in that R-1120 remains a liquid and does not evaporate at room temperature as does R-11. This more desirable feature of R-ll20 refrigerant ensures Control Room habitability cannot be impacted by a refrigerant leak. This modification does not introduce any new system interactions or malfunctions of equipment.
No accident evaluated in the USAR will develop more severe radiological consequences due to implementation of this modification. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. This modification involves refrigerant changeout and does not alter the normal flow path of the Offgas Vault Refrigeration system. No additional interaction with other systems during normal or emergency operations will result due to this change. The dike to be installed will ensure radwaste effluent is unaffected should a leak occur.
The local indication and Control Room annunciation will alert personnel of any leakage. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated in the USAR will not be created.
III. No. Radiological effluents will not be impacted due to a refrigerant release. Control Room habitability will not be impacted. As such, the margin of safety will not be reduced.
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SE No.: 94-193 Source Document: MFI l-94-055 FCR 020312 Description of Change ,
This Material Foreign Item (MFI) installs a freeze seal to support valve testing per SVI-G33-T9131. The freeze seal is located on the Reactor Water Cleanup (RWCU-G33) system bottom head drain line from the Reactor Recirculation (B33) system and the reactor pressure vessol, downstream of closed valves 1G33-F0101 and F0103. These two valves are known to leak slightly past their seats thus requiring the freeze seal to perform a LLRT per SVI-G33-T9131.
Summary I. No. The accident of interest is a LOCA within the Reactor Coolant Pressure Boundary (RCPB), more specifically, an unisolatable bottom head drain (B33/G33) line break or loss of freeze seal. The freeze seal will be placed on the 6' RWCU piping while the reactor vessel is being maintained between 1100F and 1400F with only static head.
pressure (approximately 20 psid). Nuclear fuel will be in the reactor but isolated by valves during the freeze seal evolution.
These pressure and temperature conditions are not those defined in the USAR as prerequisites for a LOCA inside drywell. Therefore the accident evaluated in the USAR is not possible under these conditions. Nondestructive examinations (PT or MT) for indications, variations, and outside diameter differences will be performed both before and after the freeze seal to ensure that the pressure integrity of the reactor pressure boundary is maintained before that section of piping is placed into service. If the freeze seal was to fail (pipe failure), it is estimated less than 15 gpm will leak into the drywell with the valves closed as previously stated. Any Emergency Core Cooling System (ECCS) loop will provide sufficient makeup water to keep the fuel, control rod blades and other vessel internals covered. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. With the plant in Mode 4 during the freeze seal and with the nuclear fuel and vessel isolated from the pipe to be frozen, the RCPB (pipe) in the area of the freeze is no longer necessary and can be declared inoperable. LOCA events analyzed in the USAR assume the events are at rated reactor power. With the vessel depressurized, there are little similarities in accident types between a B33/G33 pipe break during the freeze seal operation and a design basis LOCA. The design basis LOCA bounds the freeze seal failure. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
Summary (Cont.)
III. No. The freeze seal cycle on the 6" pipe is an industry proven method of
' isolating water systems that have limited isolation capability.
Industry tests have proven that deviation below the transition temperature and back-to-normal temperature do not change the crystal structure or characteristics of ferritic material. However, strength characteristics do change (higher yield, lower elongation, low toughness) while the material is below the transition temperature. Nondestructive examinations (PT or MT) for indications, variations, changes and outside diameter differences will be performed both before and after the freeze seal will ensure that the pressure integrity of the reactor pressure boundary is maintained. Therefore, no margin of safety has been reduced.
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t-SE No.: 94-195 Source Document: DCN 4753, Rev. O Description of Change i
This drawing change makes an editorial revision to P&ID D-302-621, !
Emergency Closed Cooling System. The drawing revision updates MPL i numbers for valves with the appropriate unit or common number. l designation. - ,
Summary ,
I. No. This drawing change is an editorial change to the P&ID for the j Emergency Closed Cooling system. The changes make the drawing more ;
consistent. The' editorial activity does not affect Chapter 15 i accident analysis, physically modify the plant, affect the operating i parameters, or increase.the radiological consequences to the plant !
or public. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not ;
increased. j II . ft). This activity is an editorial change to documentation to put unit or common designations before the MPL numbers on the P&ID drawing. The
. change will not impact Chapter 15 analysis. The plant is not j physically modified. Therefore, creating a new accident or !
malfunction of equipment is not possible.
III. No. This activity is an editorial change to system documentation. The ,
editorial activity does not affect Chapter 15 accident analysis, i physically modify the plant, affect the operating parameters, or ;
increase the radiological consequences to the plant or public. !
Therefore, no margin of safety has been reduced. i h
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li SE No.: 94-196 Source Document: DCN 3749, Rev. 0 Description of Change !
This drawing change incorporates changes to various plant drawings. The i changes provide agreement between the annunciator windows, temperature switch labels, recorder labels,-and the field conditions-for the Reactor Water Cleanup (RWCU) system. ,
Summary ;
i I. No. This drawing chanC- i M s with labels for annunciator windows, l temperature switches and room name plates. These RWCU leak detection labels will not cause the operator to take any action that ;
could cause an accident or malfunction, or effect consequences since !
the primary function of the RWCU system is to maintain water quality +
in the reactor. Changes to the labels will have no affect on the !
RWCU leak detection system since there will be no change in system !
operation. Therefore, the probability of occurrence or the 4 consequences of an accident or malfunction of equipment has not increased. ;
II. No. Changes to the labels of annunciator windows, temperature switches i or recorders cannot cause any type of accident or equipment 3 malfunction. The worst case is that there is a high energy pipe !
break outside containment. In this case, the operator would enter ;
PEI-N11 per ARI-H13-P680-1. At'this point, the operator's actions i are prescribed by the PEI. The annunciator wording does not influence his actions. The labels are passive components. Changes to the labeling do not affect equipment operation or system performance. Therefore, creating.a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The alarm, recorder and temperature switch labels do not affect the Technical Specifications. The label wording has no impact on the RWCU leak detection isolation signals. Therefore, no margin'of safety has been reduced.
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- Source Document: DCP 93-0085,'Rev. O I DCP 93-0085, Rev. 1 l DCP 93-0085, Rev. 2 f DCP 93-0085, Rev. 3 DCP 93-0085, Rev. 4 DCP 93-0085, Rev. 5- ,
Description of Change ;
i This design change modifies the Limitorque actuator of the Low Pressure Core Spray (LPCS) pump minimum flow valve lE21-F0ll to increase the ;
thrust / torque capability of the valve. The increased thrust / torque ;
provides further assurance of proper valve function, accounting for :
various uncertainties with respect to valve operational loading as I identified in Generic Letter 89-10. The modification results in a stroke l time change. !
Sumary )
I. No. This modification maintains original system performance requirements !
, such that the original accident analyses are not affected by this j change. The design change adheres to established codes and j standards such that the pressure boundary integrity is not ;
compromised. The probability of a radioactive release is not l increased since the probability of an LPCS system pipe break is not i increased and the reliability of containment integrity is !
maintained. Therefore, neither the probability of occurrence nor )
the consequences of a previously analyzed accident or malfunction of ;
equipment will be increased. '
II. No. This modification does not reduce safety systems redundancy or independence since the change does not create any new or altered interactions with other Emergency Core Cooling Systems (ECCS).
Postulated failures of the modified valve / operator assembly will not compromise pressure boundary integrity. In addition, no new permanent equipment types or new systems are introduced and original LPCS design functions are main;ained. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Although the stroke time of the valve has changed, this change has no affect on the Tecident analyses. Since the safety functions of this valve and the LPCS syste:1. are maintained and the accident analyses are not affected, there is no reduction in the margin of safety.
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SE No.: 94-198 Source Document: USAR Change Request 94-049 Description of Change This USAR change adds additional details to USAR Appendix 15E, the Anticipated Transient Without Scram (AWS) analysis, for operation in the Maximum Extended Operating Domain (EOD) . The USAR MEOD analysis already contained the information.
Summary I. No. Information from the original EOD analysis was simply added to
, USAR Appendix 15E to better detail the stated conclusion. Operation l in MEOD does not impact the AWS analysis. Therefore, the probability of occurrence or the consequences of an accident or.
malfunction of equipment has not increased. j II. No. Information from the original EOD analysis was added to USAR Appendix 15E to better detail the stated conclusion. Operation in EOD does not impact the AWS analysis. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. Information from the original MEOD analysis was added to USAR Appendix 15E to better detail the stated conclusion. Operation in MEOD does not impact the AWS analysis. Therefore, no margin of safety has been reduced.
SE No.: 94-199 Source Document: DCP 88-0212, Rev. O DCP 88-0212, Rev. 1 i Description of Change This design change provides Loss Of Coolant Accident (LOCA) override capability to drywell instrument air isolation valve 1P52-F646. The valve control logic is modified by the addition of a relay seal-in circuit and an amber override indicating light. The change eliminates the need for operators to lift leads and install jumpers during time-critical Plant Emergency Instruction tasks.
Summary I. No. Individual accidents and transients relevant to the design change are Main Steam Isolation Valve Closure, Loss of Instrument Air, and LOCA Inside Containment. A failure mode effects analysis was performed to examine failures of 1P52-F646 control logic and j operation. The analysis concluded that a spurious isolation or a i failure of the valve to isolate is bounded by existing design basis accidents and transients. The analysis examined common mode failures and determined that none are introduced outside of the ,
design basis accidents and transients. Components used in the I modification are installed in accordance with physical independence j criteria and qualified for this application. The logic is the same !
as that utilized in other isolation valve circuits. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.
l II. No. Single failure criteria and common mode failure protection l provisions have been addressed in the design change. The logic is consistent with existing override designs, utilizes the same components, and is designed to the same codes and standards. The design is such that the failure of 1P52-F646 will not prevent the system from performing its safety function. Therefore, the change does not create the possibility of an accident or malfunction of a different type than any evaluated in the USAR.
III. No. This change does not affect the isolation valve safety function. It has no impact on the system design basis. The design is such that the failure of 1P52-F646 will not prevent the system from performing its safety function. Therefore, the margin of safety as described for drywell integrity, drywell bypass leakage, drywell structural integrity, isolation actuation instrumentation and containment isolation valves is not changed.
f SE No.: 94-200 l Source Document: USAR CR 94-054 :
PAP-0101, Rev. 7 l Emergency Plan, Rev. 11, TC-8 I Description of Change- !
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. This USAR change describes a management reorganization involving onsite, j personnel. j I
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, I. No. This change does not alter.the plant in any way. No functions or 4
' activities have been eliminated. The onsite personnel involved continue to meet the ANSI N18.1-1971 and Regulatory Guide 1.8
- qualification requirements for their positions. The change complies with Technical Specifications. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. This change does not alter the plant in any way. No functions or activities have been eliminated. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated will not be created.
III. No. There are no changes being made to the physical plant. The personnel qualifications continue to meet the requirements of Regulatory Guide 1.8, ANSI N18.1-1971 and the Technical Specifications. Therefore, this change will not reduce any a rgin of safety, i
SE No.: 94-202 Source Document: NR 94-S-728, Rev. 1 Description of Change This nonconformance report evaluates the a use-as-is" disposition of valve 1E51-F046. lE51-F046 is a normally closed, motor operated isolation valve on the Reactor Core Isolation Cooling (RCIC) pump lube oil cooling water line which opens upon RCIC initiation. This use-as-is disposition deactivates the motor operated valve and locks it in the open position.
Sumary I. No. This disposition maintains the original RCIC system design and performance requirements. It does not impact operation of any other system. Since, this NR will maintain 1E51-F046 in its required safety position, the RCIC pump will be assured a continuous supply of cooling without dependence on opening of this valve. Thus, RCIC system reliability is improved. Further, this valve has no safety functions to close. Potential failures and effects associated with lE51-F046 and downstream regulating valve lE51-F015 are not increased. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The NR maintains original system functions and does not reduce the redundancy or independence of the RCIC system and other safety systems. Thus, there is no increase in susceptibility to common i mode failures. In addition, this valve change creates no new I systems and-introduces no new permanent equipment types, thus does not introduce any new potential for a malfunction of equipment / systems. Therefore, this change does not create the l possibility of an accident or malfunction of a different type than ;
any previously evaluated.
III. No. Securing this valve in its required safety position maintains system design requirements, and Technical Specifications (T.S.) 3/4.3.5 and 3/4.7.3 continue to be met. With this valve maintained open, RCIC initiation from the Remote Shutdown Panel will not require opening the valve and T.S. 3/4.3.7.4 is not affected. Thus, there is no reduction in the margin of safo /.
SE No.: 94-203 Source Document: PAP-1608, Rev. O Description of Change This evaluation analyzes the initial issue of Plant Administrative Procedure (PAP) 1608. This PAP establishes the responsibilities, requirements, and process controls for identifying, evaluating, correcting, and recording potential issues adverse to quality, safety, the environment, and business. The procedure satisfies 10 CFR 50, Appendix B, criteria XV and XVI. This procedure consolidates four independent corrective action programs.
Sum. nary I. No. Consolidating the four independent corrective action programs into PAP-1608 results in no changes to the plant, no changes to procedures which operate plant equipment, and does not affect overall system / plant performance. There is no impact on accident initiators. The new process does not replace or bypass any of the existing controls placed on changes to plant design that could affect the consequences of accidents previously evaluated in the USAR. Malfunctions of equipment important to safety will receive at least the same level of investigation and corrective action as provided under existing programs. No adverse system interactions will occur and no failure of structures, systems and components to perform functions described in the USAR will be introduced.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously i evaluated in the USAR is not increased.
II. No. The corrective action process is administrative in nature and is used to identify and correct conditions adverse to quality which assures that plant material conditions maintain fidelity with analyzed conditions. PAP-1608 combines four independent corrective action programs into one program. Therefore, the possibility of an an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR is not created.
III. Ne The administrative process change of consolidation does not reduce the effectiveness of the Quality Assurance Plan, and compliance continues under 10CFR 50.54 (a) (1) and 50.54 (a) (3). The corrective action program does not reduce any commitments to the QA program description previously accepted by the NRC. Therefore, the activity does not reduce any margin of safety.
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1 SER No.: 94-204,94-205 ii Source Document: MFI 1-94-0058 1 Description of Change ,
l This Mechanical Foreign Item (MFI) installs a connection from a !
processing trailer to OP21-A001, Two-Bed Storage Tank. This'MFI provides )
a temporary source of water to OP21-A001 during a period of. time when the l Two-Bed Demineralizers are not available due to maintenance activities. l The water quality specifications for the effluent of the processing '
trailer are specified to ensure the water supplied to the Two-Bed Storage i Tank is at least equal to the quality of'the effluent of the Two-Bed I Demineralizers. All of the normal monitoring and alarm functions are l unchanged and would function to notify the Control Room crew of a change in water chemistry or other abnormality.
Sumary I. No. This does not impact any USAR Chapter 15 accident initiator. The bounding event in the USAR is flooding, as described in Chapter 10.4.5. The postulated worst-case failure would be a failure of the processing connection. The analysis provided in the.
USAR for the circulating water break in the yard remains bounding. I The Mixed-Bed and Two-Bed systems are nonsafety, and therefore are not required to provide water to mitigate the consequences of an accident. The equipment serviced by water treatment will not see a negative change in water chemistry. Therefore, the probability of occurrence or the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. The failure of the connection is bounded by the flooding event evaluated in USAR Section 10.4.5.3.2. The equipment. serviced by these systems will not see a negative change in water chemistry.
Therefore, the possibility of an accident or malfunction of equipment of a type different than any previously evaluated in the USAR is not created.
III. No. The limits for reactor coolant chemistry as specified in Technical Specification 3/4.4.4 are not changed. The ability to maintain reactor coolant chemistry within limits will not be reduced.
Therefore, the margin of safety has not been reduced.
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SE No.: 94-206 Source Document: TXI-0221, Rev. O Description of Change This temporary test instruction will provide guidance to identify possible air inleakage sources into the Offgas (N64) system. The instruction will systematically eliminate N64 ties to outside air sources in an attempt to identify if any of the current Offgas process flow rate can be attributed to this air inleakage. This testing will temporarily isolate air to N64 valve stems and check for changes in the Offgas flow rate. Other manipulations will be performed in an effort to identify the .
source of air inleakage so that repairs may be performed to further reduce Offgas system flow rates.
Summary I. No. The affected Offgas valves are supplied with air to the lantern ring leak-off connection between the double-seal packing. USAR Section 11. 3.2.1.7 states the " leakage of radioactive gases from the system is limited by ... using bellows stem seals or equivalent".
The General Electric design basis states that valves with air supplied to the lantern ring leak-off connection are equivalent design alternatives to bellows stem seals. This TXI does not change this design feature. The testing will be performed with the seal air temporarily isolated. The Offgas Building Exhaust system is designed to treat any leakage in the Offgas system and this function will not be compromised. Any minor leakage that may result during this testing is bounded by the Offgas pipe break analysis in USAR Section 15.7.1. Therefore, the probability of occurrence or the l consequences associated with an accident or malfunction of equipment l important to safety has not been increased.
l II. No. The sealing air system will be isolated but not breached. Since I currently there is no flow in the sealing air supply, the packing area will remain pressurized. Should leakage through the bypass line isolation valves occur, radiation monitors in the Offgas exhaust stack will alarm and isolate the bypass line if air is not restored. This minor leakage, should it occur, is bounded by the USAR 11.3.2.1.6.3 analysis. Offgas system pressure in the bypass line is less than 1.0 psig so it is unlikely the valves will leak.
The offgas Building Exhaust system is designed to treat any leakage in the Offgas system and this function will not be compromised. Any minor leakage that may rfsult during this testing is bounded by the Offgas pipe break analysis in USAR Section 15.7.1. Therefore, the possibility of an accident or malfunction of equipment of a different type than previously evaluated is not created.
III. No. Sealing air is not a Technical Specification required component for Offgas system operation, nor is it considered in any Technical Specification basis. Since required Technical Specifications will be met, particularly those related to radiation monitors and their setpoints, the margin of safety related to Offgas and offsite releases will not be reduced.
U SE No.: 94-208 Source Document: PTI-R46-P0001B, Rev. 1 ,
Description of Change j i
This periodic test ~ instruction performs data collection to confirm the !
-overall heat transfer capability of the Division II Diesel Generator Jacket Water Heat Exchanger (Div. II D/G JWHX) through performance monitoring. During the operation of the system, in-plant instrumentation !
will be removed and replaced with more accurate test equipment, i
Summary ;
I. No. This instruction was written as an administrative program to operate i performance monitoring equipment to confirm the overall heat ,
transfer coefficient associated with the Div. II D/G JWHX. The methods utilized in the collection of the performance data will not :
compromise the safety significance of the components being !
monitored. Therefore, the probability of occurrence or the :
consequences of an accident or malfunction of the equipment l important to safety previously evaluated is not increased. ;
II. No. This instruction was written as an administrative function. The collection of the data during the operation of the Emergency Service !
Water (ESW) and the Diesel Generator Jacket Water (R46) systems do i i
not compromise system performance or overall system response. The instruction utilizes external temperature measuring devices. These devices do not i-terfere with any existing control logic or system
- interlocks. Therefore, the possibility of an accident or malfunction of a different type than previously evaluated in the USAR is not created.
III. No. This instruction does not compromise the original equipment design bases, construction codes or equipment qualification. It only used i to collect performance monitoring data as it relates to the !
Division II Diesel Generator Jacket Water Heat Exchanger.
Therefore, no margin of safety has been reduced.
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SE No.: 94-214 Source Document: DCN 4555, Rev. O Description of Change This drawing change updates P&ID D-302-643, Residual Heat Removal System, ,
to add a 'B" suffix to valves 1E12-F549 and 1E12-F587, which was '
inadvertently omitted when this document was updated to the CAD system.
Summary ,
I. No. The configuration of the plant was not altered or changed as a result of updating this drawing. The drawing change did not change the system operation, nor alter the valve line-up. The purpose and operation of the valves remained unchanged. Therefore, the l probability of occurrence or the consequences of an accident or I malfunction of equipment has not increased.
II. No. The plant configuration was not changed or altered as a result of this drawing change. The safety classification and valve operation has not changed. The drawing changes were editorial in nature. '
Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The drawing change was editorial in nature and did not alter the ,
function or manner in which these valves are operated. The valves !
still provide the intended safety function of maintaining the piping i system pressure boundary. Therefore, no margin of safety has been l reduced.
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SE No.: 94-221 Source Document: DCN 4849, Rev. O Description of Change This drawing change corrtets a drafting error on P&ID D-302-341, MSIV Leakage Control System, to represent MPL 1E32-R0601A/E/J/N as millivolt converters.
Summary I. No. The Leakage Control system is a manually initiated system for use during a LOCA inside containment. The system is used for accident mitigation. The equipment associated with the MSIV Leakage Control system has not been changed by clarifying its nomenclature.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. The change is an editorial correction to a P&ID. No physical changes have been made to the Leakage Control system. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. The change is an editorial correction to a P&ID. The Leakage Control system has not changed. Therefore, no margin of safety has been reduced.
SE No.: 94-223 Source Document: US. Change Request 94-114 Description of Change This USAR change ~ request changes the allowable stroke time of lE51-F064, which is a normally open motor operated valve. The lE51-F064 v11ve is the outboard containment isolation valve for the 10' steam line to the Residual Heat Removal (RHR) and Reactor Core Isolation Cooling (RCIC) systems.
Summary I. No. No actual change to lE51-F064 is involved. Environmental conditions are not more severe than previously evaluated. The electrical supply components to the valve are assessed as suitable for the new stroke time. Consequences of a line break outside containment remain bounded by the main steam line break case. Automatic line break detection and isolation signals are not affected. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
t II. No. This change involves a change in allowable stroke time. Safety ]
system operation and containment isolation provisions are not changed. Potential environmental conditions for equipment were previourly evaluated for the longer allowable stroke time. ;
Therefore, creating a new accident or malfunction of equipment that 4 has not been previously evaluated is not possible.
III. No. Isolation response time for lE51-F064 remains in conformance with assumptions used in the LOCA analysis and the line break analyses applicable to the line. Therefore, no margin of safety has been i reduced. l
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Source Document: USAR Change Request 94-119 l i
Description of Change l This USAR change describes the existing method used to ensure, through diagnostic testing per NRC Generic Letter (GL) 89-10, that Motor Operated j Valves (MOVs) will operate under normal and design basis conditions.
Summary I. No. Testing MOVs in accordance with GL 89-10 does not affect accidents or malfunctions previously evaluated in the USAR. The GL 89-10 l diagnostic testing program sets the MOVs with a higher, more ,
conservative margin than the original vendor requirements to ensure j the MOVs will operate reliably during their design basis functions.
The diagnostic testing assures the proper functioning of the equipment by quantitatively demonstrating the adequacy of the design. Although the GL 89-10 program increases the testing l requirements on the equipment, the original design specifications, !
qualifications, and criteria will not be violated or exceeded. The j controls and logic for operation of the equipment are not changed as a result of the testing Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not increased.
II. No. Testing MOVs in accordance with GL 89-10 does not exceed design specifications, criteria, and qualifications. The controls and logic for operation of the equipment are not changed as a result of the testing. The diagnostic testing is performed in accordance with approved plant procedures. Therefore, creating a new accident or malfunction of equipment is not possible.
III. No. This testing ensures that the MOVs perform in accordance with design bases and Technical Specification requirements. Each diagnostic test is performed in accordance with approved plant procedures. !
Therefore, no margin of safety has been reduced.
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SE No.: 94-232 l Source Document: PAP-0101, Rev. 7, TC-1 l PAP-0103, Rev. 4, TC-8 PAP-0507, Rev. 10, TC-1 ;
Description of Change i This change to Plant Administrative Procedures (PAP) 0101, 0103, and 0507 l incorporates a site reorganization of personnel, and changes in the 3 Onsite Review Committee memoership and quorum requirements. 1 Summary ;
I. No. These changes are administrative in. nature. Accident analysis is not affected. The level of expertise to perform the activities in l question have not been reduced. Therefore, the' probability of '
occurrence or the consequences of an accident or malfunction of ,
equipnent has not increased. !
II. No. The changes do not alter the design, operation, or function of the plant. Therefore, creating a new accident or malfunction of equipment of a type different than previously evaluated is not possible.
III. No. The changes have no impact on the Technical Specifications. No activities or functions were eliminated, just reassigned. The changes do not alter the design, operation, or function of the plant. Therefore, no nargin of safety has been reduced.
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