ML20128L561
| ML20128L561 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 04/10/1996 |
| From: | CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20128L541 | List: |
| References | |
| NUDOCS 9610150122 | |
| Download: ML20128L561 (280) | |
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Attachment 2 l PY-CEI/NRR-2099L j l
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PERRY NUCLEAR POWER PIJGIT <
SAFETY EVALUATION
SUMMARY
PURSUANT TO 10 CFR 50.59 (B) (2) ,
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1994 - 1996 {
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l 9610150122 961008 l
i SE No.: 94-0207 Source Document: TXI-0213, Rev. O Description of Change This temporary instruction evaluates the use of Calgon Towerbrom 960 as a microbiocide in the Circulating Water (N71) system.
Summary I. No. The use of Towerbrom 960 is an equivalent substitute for the use of sodium hypochlorite as a biocide. The biocide will not affect the !
design or operation of the N71 system. Towerbrom is compatible with l plant materials. Accident analysis, as described in USAR )
Chapter 15, will not be impacted. Therefore, the probability of I occurrence or the consequences associated with an accident or '
malfunction of equipment has not changed, f l
II. No. The use of Towerbrom will not impact the design or operation of the N71 system. The biocide is compatible with plant materials. If a condenser tube leak were to occur, the Condensate Demineralizer ;
system would minimize any impact upon the Condensate system. !
Therefore, the possibility of an accident or malfunction of a i different type than any previously evaluated in the USAR is not i created. j III. No. Towerbrom 960 is compatible with plant materials. The biocide will l not impact the design or operation of the N71 system. Towerbrom 960 ' l is an equivalent substitute for sodium hypochlorite, which is ;
currently in use. Therefore,.no margin of safety has been changed. l t
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. SE No.: 91-0209 Source Document: Physical Security Plan, Rev. 19 l l
pescription of Change This evaluation analyzes changes made to the Physical Security Plan (PSP). The changes have been evaluated to ensure that the l effectiveness of the Perry Nuclear Power Plant Security Plan has not been j 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 4
and, therefore, does not create the possibility for an accident or
- malfunction of a different type.
j III. No. The PSP does not reduce any margin of safety.
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SE No.: 94-0210 Source Document: USAR Change Request 94-075 ;
i Description of Change This change request clarifies the wording in Appendix 1A, Item III.D.l.1, Integrity of Systems Outside Containment Likely to Contain Radioactive Material. The changes simply clarify the means by which systems are aligned and leakage is identified.
Summary:
I. No. The USAR provides guidance for determining leakage from systems outside containment likely to contain radioactive material during a serious transient or accident. This change request clarifies the stated guidance. The clarification conforms to the current methods and conditions used to satisfy the leakage surveillance and preventative maintenance program as described within USAR, Appendix 1A, III.D.1.1 and Technical Specification 6.8.4.a.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
i II. No. The changes to USAR, Appendix 1A, Item III.D.1.1 are administrative I in nature and do not it.terfere with any existing controls, logic, or system interlocks. Therefore, the possibility of an accident or malfunction of a different type than previously evaluated in the USAR has not been created.
III. No. The existing surveillance and preventative maintenance program will ensure the leak tight integrity of USAR stipulated systems. This program minimizes the dose rates received onsite and to the public following a serious transient or accident condition. The change request does not effect leakage limits but clarifies the guidance used to monitor for system leak tightness. Technical Specification 6.8.4.a requirements for primary coolant sources outside containment will be maintained. Therefore, no margin of safety has been changed.
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SE No.: 94-0211 Source Document: PTI-R46-P0001, Rev. 1 Description of Change This periodic test instruction performs data collection to confirm the overall heat transfer capability of the Division I Diesel Generator Jacket Water Heat Exchanger (Div. I 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. This temperature instrumentation is not associated with any control function or interlock.
Summary I. No. This instruction was written to operate performance monitoring equipment to confirm the overall heat transfer coefficient associated with the Div. I 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 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 cperation of the Emergency Service Water (ESW) and the Diesel Generator Jacket Water (R46) systems do not compromise system performance or overall system response. The instruction utilizes external temperature measuring devices. These devices do not interfere 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 is only used to collect performance monitoring data as it relates to the Division I Diesel Generator Jacket Water Heat Exchanger. Therefore, no margin of safety has been reduced.
SE No.: 94-0212 Source Document: PTI-E22-P0007, Rev. 1 Description of Change This periodic test instruction performs data collection to confirm the overall heat transfer capability of the Division III Diesel Generator Jacket Water Heat Exchanger (Div. III D/G JWHX) through performance monitoring. During the operation of the system, in-plant instrumentation will ha removed and replaced with more accurate test equipment. This tempe e are instrumentation is not associated with any control function or interlock.
Summary
. I. No. Thio instruction was written to operate perfornance monitoring 1
ey.ipment to confirm the overall heat transfer coefficient associated with the Div. III D/G JWHX. The nethods 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 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 (E22) systems do not compromise system performance or overall system response. The instruction utilizes external temperature measuring devices. These devices do not interfere 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.
s III. No. This instruction does not compromise the original equipment. design 4
bases, construction codes, or equipment qualification. It is only used to collect performance monitoring data as it relates to the
- Division III Diesel Generator Jacket Water Heat Exchanger.
q Therefore, no margin of safety has been reduced.
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SE No.: 94-0213 Source Document: NR 94-S-572, Rev. 2 Description of Chance This nonconformance report evaluates the "use-as-is' disposition associated with the use of Fuel Pool Cooling and Cleanup (G41) system valves 0G41-F280, 0G41-F285, dG41-F290, and 0G41-F295.
Summary I. No. The use of OG41-F280, 0G41-F285, 0G41-F290, and 0G41-F295, and the ability to manually isolate these valves under accident conditions will not preclude the adequate functional performance of the Fuel Pool Cooling and Cleanup (G41) system. An analysis of the associated consequences related to flooding, shielding, and spent fuel pool heatup will not result in any adverse conditions related !
to safety. Sufficient time is available for manual alignment of the '
G41 system to provide spent fuel pool cooling. Therefore, the l
probability of occurrence or the consequences of an accident or l
malfunction of equipment has not increased. !
II. No. Sufficient time is available for manual alignment of the G41 system to provide the safety-related function related to spent fuel pool cooling. Based on the time available and the analyzed consequences associated with the worst case scenario, no concerns related to safety are created by the disposition. Therefore, creating a new accident or malfunction of equipment that has not been previously evaluated is not possible.
III. No. The safety-related function of the Fuel Pool Cooling and Cleanup system to provide adequate spent fuel pool cooling is not precluded i as a result of this disposition. The requirement to maintain 23 feet of shielding over the spent fuel assemblies as defined in Technical Specification Section 3/4.9.9 is also not violated.
Therefore, no margin of safety is reduced.
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4 SE No.: 94-0215 :
Source Document: SVI-P53-77312, Rev. 5, TC-22 1
Description of Change This surveillance instruction analyzes a change to the expected containment peak accident pressure from 11.31 PSIG to 7.8 PSIG with regards to local leak rate testing.
l Summary I. No. This change to the local leak rate test program has been incorporated based on the analyses performed by NRC Safety Evaluation, dated March 23, 1994. The test methods utilized to ensure Technical Specification compliance with containment leakage rates as described in this instruction will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR.
1 II. No. The containment peak accident pressure analyzed by the NRC is within the design limits of the containment structure. The reduction of local leak rate test pressure to comply with the reduced containment i accident pressure does not create the possibility of an accident or j malfunction of a type different than previously evaluated.
III. No. The change in local leak rate test pressure is a result of the change to contaioaent accident pressure described above. The conduct of containment leak rate testing at analyzed accident pressure will ensure that the containment leakage will not exceed allowable totals. Therefore, the margin of safety as defined in the bases for the Technical Specifications will not be reduced, l
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SE No.: 94-0216 Source Document: SVI-M14-T9313, Rev. 6, TC-2 SVI-M14-T9314, Rev. 6, TC-3
- Description of Change These surveillance instructions analyze a change to the expected containment peak accident pressure from 11.31 PSIG to 7.8 PSIG with l regards to local leak rate testing.
Summary i I. No. This chunge to the local leak rate test program has been !
incorporated based on the analyses performed by NRC Safety Evaluution, dated March 23, 1994. The test methods utilized to ensure Technical Specification compliance with containment leakage 4 rates as described in these instructions will not increase the probability of occurrence or the consequences of an accident or
- malfunction of equipment previously evaluated in the USAR.
II. No. The containment peak accident pressure analyzed by the NRC ic within the design limits of the containment structure. The reduction of l' local leak rate test pressure to comply with the reduced containment accident pressure does not create the possibility of an accident or i malfunction of a type different than previously evaluated.
i III. No. The change in local leak rate test pressure is a result of the !
change to containment accident pressure described above. The conduct of containment leak rate testing at analyzed accident i pressure will ensure that the containment leakage will not exceed ,
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allowable totals. Therefore, the margin of safety as defined in the :
bases for Technical Specifications will not be reduced.
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4 SE No.: 94-0217 Source Document: Emergency Plan, Rev. 12 Description of Chance This revision to the Emergency Plan incorporates various changes affecting event classification, facility staffing, equipment testing, and program naintenance.
Summary i I. No. This revision does not direct or impact the operation or design of any plant structure, system or component. 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.
II. No. This revision 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 equipr< nt or operation relied upon by the Technical Specifications. Therefore, it will not reduce any margin of safety.
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l SE No.: 94-0218 Source Document: SVI-P53-T7305, Rev. 6, TC-4 Description of Change:
1 This surveillance instruction evaluates the test pressure reduction for I Type C Leak Testing of Penetration 312, Equalizing Ball Valves, due to the incorporation of Technical Specification Amendment 57.
j Summary:
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No. The changes made to this instruction do not affect any interlocks, I controls or plant functions. The changes are administrative in nature and comply with the evaluation performed for the determination of the peak containment pressure following a postulated LOCA. The changes made do not change the methods employed by USAR Section 6.2.6 for data collection. The data collected during this instruction reflect the change made to the ,
I peak containment pressure from the existing 11.31 PSIG to the Technical Specification Amendment 57 pressure of 7.8 PSIG.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR) has not changed.
II. No. The analysis for the containment response, structural design and operational impact review along with the NRC Safety Evaluation, dated March 23, 1994 are utilized as the basis of this change.
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 instruction change relies upon an anelysis performed on the containment pressure response following a postulated LOCA. This analysis was approved by the NRC in Amendment 57. Therefore, no margin of safety has been reduced.
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l SE No.: 94-0219 Source Document: DCN 4846 Description of Change This drawing change is an editorial change to various plant drawings to assign Master Parts List (MPL) numbers to the suppression pool, the upper containment pools, and the lower fuel storage pools.
Summary I. No.
The drawing change simply assigns MPL numbers to the suppression, upper containment, and lower fuel storage pools. No physical changes are occurring to the plant or in its operation. The addition of MPL numbers to the suppression pool, upper containment pools, and lower fuel storage pools will help to ensure proper safety classification of work associated with these pools.
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 increased.
II. No. The drawing change simply assigns MPL numbers to the suppression, upper containment, and lower fuel storage pools. No physical changes are occurring to the plant or in its operation. Ther.efore, the possibility of an accident or malfunction of equipment of a type different than previously evaluated is not created.
III. No. The drawing change simply assigns MPL numbers to the suppression, upper containment, and lower fuel storage pools. No physical changes are occurring to the plant or in its operation. Therefore, no margin of safety is reduced. ,
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SE No.: 94-0220 Source Document: DCN 4750 Description of Change This drawing change revises various plant drawings to resclve ambiguities with the system diagram symbols that occurred when the drawings were reformatted onto the Computer Aided Drafting and Design (CADD) system.
Summary
.. No. These changes are editorial. Valve symbols were revised to reflect symbols as they were depicted in previous drawing revisions. Plant operations or control 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. Revising the system diagram symbology on these drawings will not affect plant operations or control in any way. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated is not created.
III. No. These changes do not affect any equipment or operation relied upon in the Technical Specifications. The changes are strictly editorihl changes to the drawings. Therefore, no margin of safety has been reduced.
SE No.: 94-0224 Source Document: DCP 94-5165, Rev. O Description of Change This design change provides minor architectural, electrical, and HVAC changes required to construct a full height walled office on the third floor of the Service Building.
Summary I. No. The architectural features, the HVAC, and the electrical systems l
l involved with this change do not have any direct or indirect connection with any plant safety systems. Therefore, the probability of occurrence or the consequences associated with an accident oi nalfunction of equipment has not changed.
II. No. The changes involved do not have any connection to plant safety l features. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The architectural changes are to nonstructural features. The changes to the electrical and HVAC components do not involve any significant load increases. All involved components will continue to function as designed. There is no impact upon any plant safety system. Therefore, the changes will not affect any margins of safety.
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SE No.: 94-0225
- Source Document: DCP 89-0054, Rev. O !
Description of Change '
This design change disconnects the electrical connections to the Feedwater Turbidity Analyzer (lB21-N0703) located in Panel B21-Z001.
Summary I I. No. This modification removes unused equipment. It does not affect ;
safety-related or nonsafety-related plant systems. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The Feedwater Turbidity Analyzer and associated instrumentation does not control or affect the operation of the Feedwater system. The elimination of this equipment will not increase the possibility of creating a different type of accident or malfunction of a type previously evaluated. ,
III. No. The Feedwater System will continue to function as required.
Therefore, no margin of safety has been changed.
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SE No.: 94-0226
! Source Document: PAP-1914, TC-2 I
l Description of Change:
l This procedure change revises the Fire Protection Program inspection and testing frequencies for fire detection instruments, sprinkler system alarm testing, fire barrier inspection clarification, and fire hose hydrostatic testing.
Summary:
I. No. This change does not introduce any new fire hazards to the plant.
No new or changes to existing safe shutdown circuits or equipment were introduced. The accidents listed in the USAR or associated analyses were not affected. The ability to detect and suppress a fire within a reasonable time frame remains. All fire protection and safe shutdown systems and equipment will operate as previously analyzed. Therefore, this change does not increase probability or the consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The required operating modes and functions of any system important to safety or radiological dose mitigation as they relate to safe shutdown in event of a fire are not changed. Operation of the systems in response to a fire are not affected. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
III. No. Only administrative and audit aspects of fire protection are contained within the Technical Specifications. These consist of review and audit responsibilities, the need for administrative procedures, and the associated reporting requirements. The basis for testing fire protection systems are not in the scope of the Technical Specifications. As described above, there are no changes to the fire protection and safe shutdown systems and equipment.
Changes to the periodic '.esting frequencies and methods will not affect any other system istsined in the Technical Specifications.
Therefore, the margins of safety are not impacted.
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4 f SE No.: 94-0227 Source Document: Offsite Dose Calculation Manual, Rev. 5 l i
Description of Change i
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! This revision to the Offsite Dose Calculation Manual (ODCM) incorporates j the procedural details of Radiological Effluent Technical I
Specifications (RETS) into the ODCM, in accordance with the guidance of l Generic Letter 89-01, and as described in License Amendment Request j
PY-CEI/NRR-1655L. The changes being made are administrative and only involve the relocation and format of the procedural details of the RETS. ;
Summary 1
I. No. This change relocates the procedural requirements of the RETS to the ODCM. All program requirements previously required by RETS will now
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, be required by the ODCM. Since all program requirements will remain the same, this change will not increase the probability or
' occurrence of an accident or malfunction of equipment as described in the USAR, t
II. No. The procedural requirements previously contained in RETS have been relocated to the ODCM. This will not change the current program t
requirements. Therefore, the possibility of an accident or f malfunction of a different type than previously evaluated is not j created.
) III. No. The changes do not involve any actual change in the methodology used i in the control of radioactive effluents or radiological effluent i monitoring. The effluent monitoring program remains in compliance l with the requirements of 10CFR20.106, 40CFR190, 10CFR50.36a, and j 10CFR50 Appendix I. Therefore, no margin of safety has been reduced.
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dE No.: 94-0229 ;
Source Document: DCP 94-5186, Rev 0 Description of Change This design change caps the Potable Water (P71) system line to the Sodium Hypochlorite Generation Building. The line only serves an eyewash / safety shower. ;
i Summary I. No. The P71 system does not have any connection to any plant systems important to safety. It is not relied upon in any accident analysis. Therefore, the probability of occurrence or the <
consequences of an accident or malfunction of equipment has not '
increased.
1 II. No. The change involves capping off a line to an eyewash / safety shower. l The balance of the system is unchanged. The P71 system does not {
have any connection to any plant systems important to safety. It is ;
not relied upon in any accident analysis. Therefore, creating a new !
accident or malfunction of equipment that has not been previously evaluated is not possible. .
l III. No. The P71 system does not have any connection to any plant systems i important to safety. It is not relied upon in any accident analysis. Therefore, no margin of safety has been changed.
SE No.: 94-0230 Source Document: DCN 4594 Description of' Change This drawing change revises various plant drawings by renaming the Hydrogen Ignition (M56) system 120 volt AC distribution panels as
" isolation panels" instead of " distribution p tnels' .
Summary I. No. This drawing change renames the Hydrogen Ignition (M56) system 120 volt AC distribution panels to eliminate confusion. Panel circuit protection is provided in accordance with IEEE 317-1976 as modified by Regulatory Guide 1.63 and as listed in USAR Table 8.1-2.
The change does not alter the design basis of the M56 system. This change will not impact accident events discussed in USAR Chapter 15.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The design basis of the M56 system has not been altered and no field modifications are required. The hydrogen ignitors are for accident mitigation. No new failure modes have been created. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change does not alter the M56 system design basis or make any field modifications. As such, there is no vehicle for reducing the margins of safety established for containment /drywell pressure and temperature design values compared to the maximum calculated accident values.
l SE No.: 94-0231 Source Document: USAR Change Request 94-121 Description of Chance This change request revises USAR Table 9A.3-2 by adding emergency lighting units 1R71-S0229 and 1R71-LO229, and removing unit 1R71-S0248.
Summary I. No. This change request is limited to revising a table of emergency lighting. Compliance of the requirements of Appendix R is maintained. The lighting units cannot initiate a design basis accident or transient. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change request is limited to revising a table of emergency lighting. Compliance of the requirements of Appendix R is maintained. The lighting units cannot initiate a design basis accident or transient. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change request is limited to revising a table of emergency lighting. Compliance of the requirements of Appendix R is maintained. The lighting units cannot initiate a design basis accident or transient. Therefore, no margin of safety has been changed.
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SE No.: 94-0233 Source Document: PORC Meeting 94-174
( Description of Change l
PORC Meeting 94-174 was convened to determine if regulatory relief to Technical Specification Action Statements 3.1.3.2.c.1, 3.1.3.2.c.3, and 3.1.3.2.c.4 should be sought following the fourth failure of a control rod to achieve required insert times to position 43 during scram time testing on 12/11/94. This safety evaluation evaluated the decision to seek relief with respect to the requirements of 10CFR50.59.
Summary l
I. No. The consideration cf all rods failing to insert is the activity' under evaluation. The Anticipated Transient Without Scram (ATWS) analysis in Chapter 15 is bounding. Furthermore, analysis provided by General Electric for this operating cycle showed that even if all rods were slow by an additional 70 ms, the impact upon the limiting accidents was minimal. Based upon this analysis, it was concluded that no increase in the probability or the consequences of an accident or a malfunction of equipment important to safety previously analyzed in the USAR would occur.
II. No. The complete failure to scram properly is already analyzed in USAR Chapter 15 Appendix C, Anticipated Transient Without SCRAM (ATWS).
l The failure of rods to insert to position 43 within the required i time (but acceptably to later notch positions) is bounded by this accident. 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. In order for the safety limits to be exceeded, reactor pressure L
would have to exceed 1325 PSIG and the Critical Power Ratio (CPR) would have to be less than 1.07. Based upon analysis the increase in reactor pressure was just 2 PSIG and the delta CPR increased only 0.01. Therefore, no margins of safety were reduced.
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SE No.: 95-0001, 96-0072 Source Document: DCN 3990 Description of Change This drawing change revises various drawings to reflect the permanent retention of inspection port drain lines of the Residual Heat Removal and High Pressure Core Spray systems. Each port consists of a welded half coupling and a threaded plug. These ports were originally installed to aide in checking for system boundary valve leakage. The affected lines are non-ASME.
Summary I. No. The installed inspection ports have been analyzed to maintain original system performance requirements and maintain conformance with the applicable ANSI B31.1 code stress requirements. The proposed change does not create any new system interactions and thus, does not reduce the redundancy or independence of the affected safety systems. In addition, the functions and operation of the affected systems is not altered by this drawing change. Therefore, neither the probability of occurrence nor the consequencea of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The drawing changes are limited to permanent retention of inspection ports in drain lines which are normally isolated from adjacent safety system piping. The revised piping meets the original design requirements and does not alter any system functions. Thus, the change will not introduce any new potential for common mode / common cause failures. Therefore, this drawing change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The revised piping maintains conformance to original design code requirements such that system pressure boundary integrity is !
assured. Further, the changes will not affect the performance or j operation of the adjacent section of the safety systems. Thus, !
there are no reductions in the margins of safety.
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SE No.: 95-0002 Source Document: DCP 94-5183, Rev. O Descripcion of Change l
The D51 Meteorological Tower Dew Point Sensor (0D51-N0711) and Processor I (0D51-K0711) are obsolete and no longer supported by the vendor for I repair or repair parts. This change installs a vendor approved '
replacement Dew Point Sensor and Processor.
I Summary i I. No. The new Dew Cell Sensor and Processor is a direct replacement for existing equipment. The new Sensor and Processor, as a pair, will 1 provide the same voltage input to the Meteorological Data Processing l System (MDPS) and Recorder as the existing equipment. There are no l software changes necessary to replace the existing equipment. The l installation of the new Dew Point System does not alter plant i operation for normal, abnormal, or safe shutdown conditions. There is no interface to plant systems, structures, or components that are important to safety. Therefore, the probability of occurrence or j the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The existing Dew Point System is no longer supported by the vendor.
If a failure of the Dew Point System 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 National Weather Service. The data can be manually inputted into the emergency dose assessment computer program. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The new Dew Point System is a replacement version of the existing equipment. There is no interface to a plant system, structure, or component that is important to safety. The replacement Sensor and l
Processor is a vendor approved replacement part and does not change the MDPS or its programming. Therefore, no margin of safety has been changed.
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SE No.: 95-0003 Source Document: SCRs 1-94-1949 through 1-94-1951 Description of Change These setpoint changes revise the Upper Allowable Value for instruments 1E12-N0658A/B/C in conformance with the implementation of the GE Setpoint Methodology. A Lower Analytical Limit and associated Allowable Value was also added for the Low Pressure Core Injection (LPCI) Pressure Permissive during the implementation of the methodology. In addition, the Leave-As-Is-Zone (LAIZ) was added to the Master Setpoint List to show a change from the Technical Specification LAIZ of +5/-10 to the field established value of 12.82 PSIG to provide more margin between the Allowable Value and the Setpoint. The Reset value was also added to the Master Setpoint List to reflect implementation of the GE Setpoint Methodology.
Summary I. No. The Allowable values, Leave-As-Is-Zone, and Reset values added to the Master Setpoint List are consistent with or conservative to existing Technical Specification values. This permits implementation of the new GE Setpoint Methodology. The trip setpoints have not been changed. There has been no change to the physical plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These change allow the implementation of the new GE Setpoint Methodology and provides consistent or conservative Allowable Values fer the referenced components. There will be no physical changes to the plant as a result of this change. No trip setpoints have been i changed. Therefore, the possibility of an accident or malfunction of equipment of a different type than any previously evaluated in the USAR cannot be created.
III. No. Revising the Allowable Values for the referenced components increases the margin of safety as defined in Table 3.3.3-2 of Technical Specifications. This change allows for the implementation of the new GE Setpoint Methodology and is equal to or more conservative than the Allowable Values previously calculated.
Therefore, no margins of safety have been reduced.
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SE No.: 95-0004 Source Document: SOI-B33, Rev. 5 Description of Change This system operating instruction describes the process for bypassing the
- recirculation flow control valve runback on a low reactor pressure vessel water level (Level 4) coincident or subsequent to a reactor feed pump 2
trip during single recirculation loop operation.
Summary s I. No. USAR transient / accident analysis does not take credit for the 1
recirculation flow control valve runback. Technical Specification j
(T.S.) 3.4.1.1 limits reactor power to within the capability of a single reactor feed pump. Therefore, this change does not increase probability or the consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the USAR.
i II. No. As stated above, T.S. 3.4.1.1 limits reactor power to within the 5
capability of a single reactor feed pump. USAR accident analysis is not impacted. Therefore, the possibility of an accident or a i malfunction of equipment important to safety of a different type than previously evaluated is not created.
III. No. The recirculation flow control valve runback is not analyzed as
! providing any input into any safety margin for plant systems or i equipment. Therefore, no margins of safety are impacted.
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1 SE No.: 95-0005 4
Source Document: SOI-G41 (FPCC), Rev. 7 SOI-G41 (FPFD), Rev. 5 4
l Description of Change i These system operating instructions describe changes which will allow removing the Fuel Pool Filter Demineralizing subsystem from service while i spent fuel is stored in the fuel pools.
I r Summary i
I. No. These system operating instructions describe changes which will allow removinq the Fuel Pool Filter Demineralizing subsystem from
- . service whi', spent fuel is stored in the fuel pools. The only analyzed accidents involving spent fuel in storage are Fuel Handling Accidents Inside and Outside Containment which are analyzed in USAR Sections 15.7.6 and 15.7.4. Neither accident takes credit
! for the fuel pool filter demineralizers. Therefore, this change i l does not increase probability or the consequences of an accident or '
i of a malfunction of equipment important to safety previously j evaluated in the USAR. ;
II. No. The Fuel Pool Filter Demineralizer subsystem is not related to j
safety. Accident analysis is not impacted by this activity.
I
- Therefore, the possibility of an accident or a malfunction of i equipment important to safety of a different type than previously
} evaluated is not created.
i III. No. The ability to add water to the fuel pools is unaffected by the l shutdown of the Filter Demineralizer subsystem. Accident analysis
- is not impacted by this activity. Therefore, this activity does not a
affect any margin of safety.
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i SE No.: 95-0006
_ Source Document: USAR Change Request 95-007 Description of Change This USAR change clarifies the personnel qualifications necessary to i satisfy the criteria for testing certification under Regulatory Guide 1.58.
Summary
( I. No. The changes made to USAR Table 1.8-2 are a part of the PNPP administrative program (s) for the certification and qualification of ;
power plant personnel. These changes provide the clarification of ;
the Perry Nuclear Power Plant personnel qualifications with regard to Regulatory Guide 1.58. These changes to USAR Table 1.8-2 do not interfere with any existing controls, logic, or system interlocks.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunctio'n of equipment has not changed.
II. No. The qualification of the power plant personnel are being clarified to delineate the certified and non-certified personnel requirements. :
These changes to USAR Table 1.8-2 do not interfere with any existing ;
controls, logic, or system interlocks. Therefore, the possibility '
of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. i III. No. This USAR change does not compromise the original equipment design bases, construction codes or equipment qualification, conformance with Regulatory Guide 1.58 has been maintained. Therefore, no margin of safety has been changed.
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i SE No.: 95-0007 Source Document: PS7G, Rev. 3, TC-2 Description of Change This change to the Perry Specific Technical Guidelines (PSTG) describes !
changes to Contingency #2, Emergency Depressurization, in which the order l of decisions has been slightly modified to ensure that the termination ,
and prevention of injection occurs prior to vessel depressurization.
Summary I. No.
The changes introduced by TC-2 will serve to lessen the consequences '
of some events which are beyond the plant's design basis. In all scenarios in which termination and prevention of injection are necessary, the method for terminating and preventing injection is the same as was previously reviewed in the PSTG. Therefore, this change does not increase probability or the consequences of an accident or of a malfunction of equipment important to safety i previously evaluated in the USAR. ;
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II. No. The changes introduced by TC-2 will serve to lessen the consequences of some events which are beyond the plant's design basis. These changes do not create the possibility of the event. Therefore, . is change does not create the possibility of an accident or a '
malfunction of equipment important to safety of a different typr than any previously evaluated in the USAR. T III. No. TC-2 does not adversely affect systems or components required for i safety. The changes . itroduced by TC-2 will serve to lessen the l consequences of some events which are beyond the plant's design ;
i basis. Therefore, the margin of safety as defined in the bases for '
l the Technical Specifications are not impacted.
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SE No.: 95-0008 Source Document: DCP 91-0060A, Rev. O and 1 Description of Change This design change replaces the non-regulating 25 KVA Class 1E distribution transformer, 1R25-S0035, in Motor Control Center EF1C07 with an external floor mounted 15 KVA regulating transformer. In addition to this transformer replacement, the existing 100A fusible disconnect switch /100A fuse on the transformer primary side is replaced with a 60A fusible disconnect switch /60A fuse for the transformer protection enhancement.
Summary I. No. The change maintains original system performance requirements such l that the original accident analyses are not affected. The 15 KVA l capacity is adequate for the design requirement of 10.235 KVA connected load (including LOCA response loads) on the Division I 120 VAC bus EK-1-B1 as described in calculation PRCV-004, Rev. O.
The new 15 KVA regulating transformer will provide voltage regulation to accommodate variations in loads or voltage fluctuation .
caused by LOCA loading response or by 345 kV system grid '
fluctuations. There is no impact on the system capability and reliability. 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 replaces the non-regulating 25 kVA transformer with a 15 kVA regulating transformer to improve the voltage on the Division I 120 VAC miscellaneous loads. The Unit 1 Division 1 120 VAC system original function and operation remain unchanged.
l The failure of the new regulating transformer cannot initiate a failure in the Division I equipment. The difference in failure rates between the old transformer (25 kVA non-regulating) and the l
new transformer (15 kVA regulating) is essentially negligible. In f addition, the new transformer will enhance the system operability l due to its voltage stability. Therefore, this distribution l transformer replacement does not create the possibility of an I accident or malfunction of a different type than any previously l evaluated.
III. No. The margin of safety as defined in the Technical Specification Bases B3/4.8.1, .2, and .3 refers to the operability of Class 1E distribution systems. Since the new regulating transformer will improve the system operability as addressed above, this change will not result a reduction of any margin of safety.
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SE No.: 95-0009 Source Document: Potential Issue Form (PIF) 95-0274 l
I Description of Change This PIF evaluates the ' repair' disposition of the installation of a Team Inc. leak sealant clamp on a 1-1/2" portion of the the Main, Reheat, l Extraction and Miscellaneous Drains (N22) system. The leak sealant device will serve to reduce the effects of a steam leak coming from i socket weld that connects a 900 elbow to the 1-1/2" pipe. This leak sealant device is designed in accordance with ASME Code Section VIII, Division I criteria and ASTM standards.
Summary l
I. No. The design and manufacture of the leak sealant device is to approved industry codes and standards. The flooding aspect of this evaluation is completely bounded by the previous Turbine Building flooding analysis. The potential loss of condenser vacuum is bounded by the current loss of condenser vacuum discussed in USAR Section 15.2.5. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The worst case postulated results of this disposition cannot affect ,
any equipment important to safety. Flooding and the loss of condenser vacuum is completely bounded by the previously analyzed USAR conditions. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The leak sealant device to be installed via this PIF disposition is designed and manufactured in accordance with ASME Code Section VIII, Division I criteria and ASTM standards. The clamp and elbow are not expected to fail during normal plant operation, so N22 system and plant operation remain unaltered. Therefore, no safety margins are reduced.
SE No.: 95-0010 Source Document: Potential Issue Form (PIF) 95-0241 Description of Change This PIF analyzes the acceptability of a nonconforming condition identified with Reactor Water Cleanup (G33) system motor operated valve 1G33-F101. The valve was found to have an overstress condition on the structural bolts attaching the Limitorque actuator to the valve yoke.
Summary I. No. The continued use of the yoke-to-actuator bolts will have no effect upon the G33 containment isolation function. The Reactor Coolant Pressure Boundary (RCPB) would remain intact, even if the bolts were to fail. The USAR evaluated accidents do not require the G33 system to perform any safety-related functions that would require 1G33-F101 to be operated. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated has not changed.
II. No. Valve 1G33-F101 is passive and does not provide any safety-related functions relating to the mitigation of an accident. The continued ,
use of the valve will not affect the RCPB or the functions of the G33 containment isolation valves. There is no change to the structural integrity of the pressure retaining components of the valve. Therefore, the possibility of an accident or malfunction of equipment important to safety of a type not previously evaluated is not created.
III. No. The continued use of valve 1G33-F101 would not result in the degradation or reduction in the safety margin of any system used to safely shutdown the reactor or mitigate the consequences of an accident. The RCPB and the functions of the G33 containment isolation valves remain unaffected. Thus, the margin of safety is not reduced, i
SE No.: 95-0011 l Source Document: S01-P21, Rev. 8 Description of Change This system operating instruction describes the use of a portable caustic metering pumo to be used to regenerate the two bed demineralizers.
Summary I. No. There are no USAR analyzed accidents which are assumed to be caused by failure of the two bed demineralizer system. Failure of the two bed demineralizers was analyzed in the NRC safety evaluation report to not affect plant safety. Therefore, this change does not increase probability or the consequences of an accident or of a malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The effluent quality from the demineralizers is not adversely affected by the use of a portable metering pump. Accident :.nalysis is not affected. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a different type i than previously evaluated is not created.
III. No. No Technical Specification is directly related to the two bcd demineralizers. The effluent quality from the demineralizers is not adversely affected. Accident analysis is not affected. This activity does not affect any margin of safety.
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SE No.: 95-0012 Source Document: SMRF 95-5010 Description of Change This modification adds a clean out lateral to a vertical run on a drain line of the Floor and Equipment Drains (P68) system.
Summary I. No. This change will aid in the maintenance of the drain piping, thus increasing its reliability. The clean out connection will be capped closed, except when being used to clean the drain piping. Thus, the flow paths of the P68 system,.as intended by its original design, will not change. The clean out lateral will comply with the original design code of the P68 system (ANSI B31.1). Thus, the function and reliability of the drainage piping will not change.
Since the design intent, function, and reliability of the floor drain piping is not changed by this change, the floor dreins wit.
continue to perform their design function. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The clean out lateral which will be added by this change will be designed and tested in accordance with the original design code of the P68 system (ANSI B31.1). As discussed in Question I, the floor drains will continue to perform their design function. There are no changes to the previous flooding evaluations in the USAR.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. As stated in Question I, the floor drains will continue to perform their design function without a decrease in reliability. Thus, no margin of safety is affected.
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l SE No.: 95-0013 '
Source Document: DCN 4918 I
Description of Change This drawing change revises various plant drawings by indicating an increase in the maximum allowable stroke time of the following ;
containment isolation valves: 1P11-F0060, 1P11-F0090, 1P43-F0055, t 1P43-F0215, 1P50-F0140, and 1P50-F0150. The increased the stroke times for these valves went from 30 to 35 seconds.
Summary
!l I. No. Analysis showed that the increase in valve stroke time would not increase the potential or severity of an accident. Present USAR accident analysis does not explicitly model closure times of containment isolation valves, nor were they required to be modeled I per the guidelines of Standard Review Plan 15.6.5. Therefore, neither the probability of occurrence nor the consequences of a '
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previously analyzed accident or malfunction of equipment will be i increased by this change.
II. No. Analysis showed that no new equipment types or new system I interactions are created. The original plant design basis is l maintained. There are no hardware changes to the plant. l Performance of the affected systems is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. These valves continue to automatically close to provide containment isolation and !
to prevent radiological effects from exceeding guidelines of l 10CFR100. The containment isolation function of these valves as l described in the Technical Specifications are not affected.
Therefore, no reduction of safety margin is caused by this change.
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SE No.:
95-0014 Source Document: USAR Change Request 95-009 Description of Change f
This USAR change describes a management reorganization involving the onsite engineering staff.
1 Summary J
I. No. This change does not alter the plant in any way. No functions or activities have been eliminated. The onsite personnel involved continue to meet the ANSI N18.1-1971 qualification requirements for their positions. The change complies with Technical Specifications.
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. t
! II. No. This change does not alter the plant in any way. No functions or ,
l 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 ANSI N18.1-1971 and the Technical Specifications. Therefore, this change will not reduce any margin of safety.
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i SE No.: 95-0015 Source Document: ONI-R10, Rev. 4 Description of Change This off-nortal instruction evaluates the following items:
- 1) Bypassing Reactor Core Isolation Cooling (RCIC) system leak detection during a Total Loss of All AC (TLAC),
- 2) Not performing a manual pre-lubrication of the Division I and II diesel generator turbochargers as required by DR/QR MP-022/023,
- 3) Overriding the High Pressure Core Spray (HPCS) pump suction transfer to the suppression pool on high suppression pool level,
- 4) Leaving the HPCS pump suction transfer to the suppression pool on low Condensate Storage Tank (CST) level in automatic while the the i HPCS pump suction transfer to the suppression pool on high I suppression pool level is overridden,
- 5) The effect of transferring CST contents to the suppression pool on )
suppression pool temperature, !
- 7) Relationship to Plant Emergency Instructions (PEI) and Perry Si,ecific Technical Guidelines (PSTG).
Summary l
I. No. ONI-R10 provides the actions necessary to recover from a Loss of Offsite Power (LOOP), a SBO, or TLAC. It is used in conjunction with the PEIs to maintain vessel level within the USAR analyzed band, to maintain suppression pool temperature less than 1850F, and to restore the plant electrical system. For the analyzed transients (LOOP and SBO) operation per ONI-R10 does not result in any barrier (fuel, pressure vessel, or containment) exceeding anv of the design criteria. Therefore, ONI-R10 does not alter the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. Overriding the HPCS pump suction transfer to the suppression pool does not result in the possibility of containment failure since the effect on suppression pool temperature is not significant and the PEIs have actions to ensure that vessel pressure is maintained less than the that required by the PS'IU. Additionally, the method employed in overriding the HPCS pump suction transfer to the suppression pool maintains the CST transfer logic in automatic and therefore does not adversely effect HPCS operation. Therefore, ONI-R10 does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
SE No.: 95-0015 (Cont.)
Summary (Cont.)
III. No. ONI-R10 is intended to be used post-event in conjunction with the PEIs to gaintain vessel level within the USAR analyzed band, to I maintain suppression pool temperature less than 1850F, and to I restore the plant electrical system. During a loss of AC power, a l number of Technical Specification parameter limits may be exceeded,-
these include: primary to secondary containment dP, primary containment average air temperature, drywell to primary containment <
dP, drywell average air temperature, suppression pool water level, '
suppression pool temperature, and upper containment pool water I level. This is due to the loss of various systems due to the loss of electrical power. In all cases, the PEIs as supplemented by ONI-R10 will attempt to restore Technical Specification parameters and complete the required Technical Specification actions as soon as power availability permits. Both the Emergency Procedure Guidelines (EPG) (on which the PSTG is based) and Perry's Station Blackout submittal have been reviewed by the NRC and found ;
acceptable. Therefore, the margin of safety as defined in the bases for Technical Specifications are maintained.
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l SE No.: 95-0016 Source Document: DCN 4919 Description of Change This drawing change revises plant drawings by indicating an increase in the maximum allowable stroke time of the Inboard MSIV Before Seat Drain Containment Isolation Valves 1B21-F0016 and 1B21-F0019 from 20 to 25 seconds.
Summary I. No. Analysis showed that the increase in valve stroke time would not increase the potential or severity of an accident. Present USAR accident analysis does not explicitly model closure times of containment isolation valves, nor were they required to be modeled per the guidelines of Standard Review Plan 15.6.5. Environmental and radiological consequences resulting from a nonsafety line break downstream of the 1B21-F0019 valve were shown to be bounded by the design basis Main Steam Line Failure Outside of Containment accident. 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. Analysis showed that no new equipment types or new system interactions are created. The original plant design basis is maintained. There are no hardware changes to the plant.
Performance of the affected system is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. These valves continue to automatically close to provide containment isolation and to prevent radiological effects from exceeding guidelines of 10CFR100. The containment isolation function of these valves as described in the Technical Specifications are not affected.
Therefore, no reduction of safety margin is caused by this change.
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SE No.: 95-0017 Source Document: TXI-0223, Rev. O Description of Change This temporary instruction revises the control circuitry for valve j 1G33-F0101 [ Reactor Water Cleanup (RWCU) System Bottom Head Drain Suction l Valve), to allow the valve to be electrically backseated to stop packing j leakage. Due to its location inside the drywell, the valve cannot be l
, locally backseated with the plant in operation. In order to remotely l l electrically backseat the valve, the open limit switch must be defeated which also results in bypassing the open torque switch.
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Summary I. No. No accident evaluated in tr.e USAR takes credit for 1G33-F0101, or the RWCU system, in mitigating the consequences of an accident. To <
prevent excessive stresses, without crediting the open torque I switch, TXI-0223 backseats the valve under reduced voltage. The reduced voltage prevents the actuator or valve from exceeding any I ttrust/ torque limitations, or any valve ASME Code allowable !
stresses, even during motor stall conditions. Smaller fuses are installed while backseating, to reduce the time that the motor would spend at stall conditions. 1G33-F0101 will be operated within its design basis. However, even if the valve was postulated to fail, the failure would not increase the consequences of an accident previously evaluated in the USAR. USAR Sections 6.2, 6.4.4.1, and 15.6.5 evaluate the most limiting Loss of Coolant Accident.
Failure of 1G33-F0101, or of the bottom head drain line, is bound by these analyses. Similarly, USAR Section 3.6.2.5.1.5 already evaluates a pipe rupture in the bottom head drain line (reference USAR Figure 3.6-73). Therefore, the probability of occurrence or the consequences of an a'ccident or malfunction of equipment important to safety are unchanged.
II. No. Because the RWCU system is operated within its design, the activity cannot create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated.
III. No. Operation of 1G33-F0101 and the RWCU system remains within the plant I design basis and Technical Specification requirements. Therefore, no margin of safety has been changed.
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j SE No.: 95-0019 Source Document: USAR Change Request 95-017
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Description of Change i This change request removes a numerical value from a diesel automatic ;
start permissive description and replaces that value with a word i description of that setpoint.
Summag .
1 I. No. This change request removes a numerical value from a diesel automatic start permissive description and replaces that value with a word description of that setpoint. This activity allows the j original design intent of the automatic start inhibit pressure switch to be fully utilized as described in the USAR. As described in Chapter 8.3 of the USAR, engineered safety feature loads are assigned to three independent load groups designed as Division I, l II, and III. The diesel generators are designed to provide physical and electrical divisional separation, redundance, and load independence. This change does not affect any of these conditions.
If the diesel automatic start inhibit signal is satisfied, then a failure of the diesel to start has occurred. Based on this occurrence, reliance is placed on the redundancy and independence provided by the other divisional diesels. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed. ;
1 II. No. This USAR change removes the numerical figure for the automatic I start inhibit signal and replaces it with a word description of that signal permissive. The actual field setpoint and tolerance value ,
have not been altered and these values now support the function of I the actuating pressure switch as now described in the USAR. The description of the pressure switch as it exists in the inhibit circuitry is acceptable since the design intent is to inhibit automatic engine cranking if starting air receivers decrease to the inhibit pressure level. This change request does not alter the design intent of the pressure switch or other equipment associated with the diesel. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This USAR change is for the removal of the numerical figure for the automatic start inhibit signal and replaces it with a word description of that signal permissive. The actual field setpoint and tolerance value have not been altered. The design intent of the !
pressure switch or other equipment associated with the diesel has l not been affected. Therefore, no margin of safety has been reduced.
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- 1 SE No.: 95-0020 i I
Source Document: WO 95-855 and 95-856 Description of Change j
These Work Orders (WO) will disable the LOCA opening signal to valves 1P53-F035 and 1PS3-F045 when the associated airlock inner door is
! inoperable.
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Summary l
j I. No. Containment air locks are equipped with an inner and an outer door.
j Each door has two inflatable seals. Between the two seals each door has a Leakage Control System (LCS) which is piped to the Shield i
Building annulus. The inner airlock leakage control line is provided with a solenoid operated valve (lP53-F035 for the lower airlock and 1P53-F045 for the upper airlock). Upon a LOCA signal ;
these valves open directing any potential seal leakage to the j annulus to be treated by the Annulus Exhaust Treatment system prior !
to release to the environment. When the inner door seal is J
. inoperable, the design does not provide any provisions to limic the leakage if a LOCA occurred. Thus, the potential leakage can exceed 0.6 La containment leakage value. When either the inner airlock l i
door or the inner seal of the inner door is inoperable maintaining the leakage control system valve in the closed position will prevent excessive containment leakage. These W0s will ensure that this action occurs. The LCS is an accident mitigation system, it cannot i
initiate an accident. Maintaining the valves in closed position will prevent gross leakage, preserving the radiological consequences as analyzed. Operation of the outboard LCS is not affected. Any
, potential. leakage pass the inner airlock door will be processed via j
the outer airlock door LCS. Therefore, the probability of occurrence or the consequences associated with an accident or j
malfunction of equipment has not changed.
- II. No. The disabling of the inboard airlock door LCS does not affect the operability of the redundant outer airlock door LCS. Therefore, the possibility of an accident or malfunction of a different type than j any previously evaluated in the USAR is not created.
III. No. Disabling the containment airlock inner door LCS whenever the inner door is inoperable limits the potential gross leakage from the containment. Outer door LCS will still perform its intended function. Thus, Technical Specifications 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3 and their bases remain unaffected. Therefore, no margin of safety has been changed.
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SE No.: 95-0021 i Source Document: LLJED 1-95-016 '
Description of Change 4
This temporary modification provides for isolation of the Unit 2 Division III DC system from all Unit 1 interconnections to support the removal of the Unit 2 Division III diesel generator and its support systems.
Summary I. No. The temporary modification provides for isolation of the Unit 2 Division III DC system from all Unit 1 interconnections to support the removal of the Unit 2 Division III diesel generator and its support systems. Normal Division III DC system alignments has the Unit 2 DC system isolated from the Unit 1 system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The temporary modification provided for isolation of equipment not required to support Unit 1 operation. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Unit 2 Division III DC system is not required to maintain compliance with any Technical Specification or Operating License requirements for the continued operability of either the Unit 1 Division III DC system or the High Pressure Core Spray system.
Therefore, no margin of safety has been changed.
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! SE No.: 95-0022, 95-0023 l Source Document: P&ID D302-761, Penetration Pressurization and Personnel i
Airlock Leakage Control System
- Description of Change This evaluation analyzes changing the normal operating position of manual
, valves 1P53-F556 and 1P53-F561 from normally opened to normally closed.
Summary i
a I. No. Containment airlocks are equipped with an inner and an outer door.
i Each door has two inflatable seals. Between the two seals, each l door as a Leakage Control System (LCS) which is piped to the Shield
' Building annulus. The inner airlock door leakage control line is provided with a manually operated valve, 1P53-F556 for the lower airlock and 1P53-F561 for the upper airlock. Upon a LOCA signal these valves direct any potential seal leakage to the annulus to be
- treated by the Annulus Exhaust Treatment system prior to release to i
the environment. When the inner door seal is inoperable, the design does not provide any provisions to limit the leakage if a LOCA i occurrs. Thus, the potential leakage can exceed 0.6 La containment leakage value. Maintaining leakage control system valves IP53-F556 4
for the lower airlock door and 1P53-F561 for the upper airlock door, I
in the closed position will prevent excessive containment leakage post-LOCA. Since the LCS is an accident mitigation system, it cannot initiate an accident. Maintaining the valves in closed position will prevent gross leakage post-LOCA, preserving the j radiological consequences as analyzed. Operation of the outboard LCS is not affected. Any potential leakage past the inner airlock door will be processed via the outer airlock door LCS. Therefore, I the probability of occurrence or the consequences associated with an j accident or malfunction of equipment has not changed.
[ II. No. Disabling the airlock inner door LCS cannot cause an accident of a
! different type. The disabling of the LCS does not affect the l operability of the redundant outer airlock door LCS. Therefore, the possibility of an accident or malfunction of a different type than
- any previously evaluated in the USAR is not created.
- III. No. Disabling the containment airlock inner door LCS limits the potential gross leakage from the containment. Outer door LCS will l-still perform its intended function. Thus, Technical Specifications 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3 and their bases remain unaffected. Therefore, no margin of safety has been changed.
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SE No.: 95-0024 Source Document: DCN 4935 Description of Change This drawing change revises plant drawings by indicating an increase in the maximum allowable stroke time for the the Fuel Pool Cooling and Cleanup system containment isolation valves 1G41-F0100, 1G41-F0140, and 1G41-F0145 from 30 to 35 seconds.
Summary I. No. Analysis showed that the increase in valve stroke time would not increase the potential or severity of an accident. Present USAR accident analysis does not explicitly model closure times of containment isolation valves, nor were they required to be modeled per the guidelines of Standard Review Plan 15.6.5. 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. Analysis showed that no new equipment types or new system interactions are created. The original plant design basis is maintaine1. There are no hardware changes to the plant.
Performance of the affected system is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. These valves continue to automatically close to provide containment isolation and to prevent radiological effects from exceeding guidelines of 10CFR100. The containment isolation function of these valves as described in the Technical Specifications are not affected.
Therefore, no reduction of the safety margin is caused by this change.
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SE No.: 95-0025 Source Document: PTI-P42-P0004, Rev. O I
e Description of Change:
This periodic test instruction evaluates momentary operation of the Emergency Closed Cooling (ECC-P42) system pump at a flow rate less than the continuous duty minimum flow value. This will occur when the primary i
flowpath is isolated by closing a motor operated butterfly valve, in 2
order to obtain diagnostic test data on the butterfly valve in accordance t
with NRC Generic Letter 89-10.
, Summary:
! I. No. An engineering evaluation determined that it was acceptable to
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operate the ECC pumps at flow rates below the continuous duty minimum flow value specified by the pump vendor, for a duration as long as one hou.r. This test instruction only intermittently operates the ECC pump below the continuous duty minimum flow value, 3
and for a duration no longer than one minute. Additional pump monitoring will be performed. All other aspects of the system t
operation, automatic functions, and interlocks for the ECC Oystem will remain as described in USAR Section 9.2.2. Therefore, the i probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
l II. No. All automatic functions and interlocks described in USAR Section 9.2.2 for the components operated under this instruction will remain operable. The testing will not affect the way the ECC system functions. Because the operation of the ECC system is a
unchanged, the proposed activity cannot create the possibility of an i accident or malfunction of equipment important to safety of a !
! different type than any previously evaluated in the USAR. l 1
III No. Operation of the ECC system remains within the plant design bases i
and Technical Specification requirements. Therefore, no margin of i l safety as defined in the basis for any Technical Specification is '
affected.
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l SE No.: 95-0026 1 j
Source Document: PAP-0101, Rev. 8 !
l Description of Change This procedure change incorporates various changes to the onsite plant organization.
Summary I. No. The organization changes are administrative in nature. Accident analysis is not affected. The level expertise to perform the activities in question have not been reduced. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment 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 margin of safety has been reduced.
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SE No.: 95-0027 Source Document: DCP 90-0196, Rev. 3 Description of Change This design change replaces the Condensate Filtration (N23) system porous metal precoatable septa with new septa. The porous metal septa are being replaced because they were becoming fouled with iron oxide resulting in short N23 filter run lengths. The shorter run lengths resulted in i excessive radwaste generation due to the use of powered resins as the '
{ precoat media. The installation of new septa will increase run time, i
improve filter efficiency for suspended solids, and reduce radwaste
) generation. The change permits replacement of the filter septa with precoat and non-precoat septa designs. Based on septa performance, septa life, and radwaste economic considerations, the optimum septa design will be chosen.
Summary I. No. The septa are designed to meet or exceed the original performance requirements for the system, and do not adversely affect the function or operation of the system or any other Structure, System, or Component (SSC). This change does not result in an increase in radionuclide concentrations and resultant doses above licensing limits. The use of different septa and process changes are in no way a causal factors in the initiation of an accident (e.g., loss of feedwater, failure of a liquid radwaste tank, or any other accident analysis). 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.
I II. No. The septa are designed to meet or exceed the original performance i requirements for the system and do not affect the design function or operation of the system. Process changes associated with this activity do not change the design functica or degrade the performance of any SSC. Therefore, the change does not create the possibility of an accident or malfunction of a different type than.
1 any evaluated previously in the USAR.
III. No. Technical Specifications 3/4.11.1.1 (Liquid Effluents-Concentration), 3/4.11.1.2 (Liquid Ef fluent'. .e ) ,
3/4.11.1.3 (Liquid Radwaste Treatment System), 3/4.11... 3olid Radwaste), and 3/4.11.4 (Total Dose) are applicable for these changes. The use of different septa and different process operating changes do not reduce the margins of safety with respect to effluent or dose limits. Operability of the Liquid Radwaste Treatment system is not impacted by these changes. These changes do not result in changes in solid radwaste shipping and disposal requirements above licensing limits. The radionuclide concentration, and resultant exposures and total dose as a result of these changes are within limits of 40CFR190; 10CFR20, Appendix B, Table II, Coluna 2; Appendix I, 10CFR50; and 10CFR20.106(e). Therefore, the change does not reduce any margin of safety.
i SE No.: 95-0028 source Document: SOI-C85, Rev. 4 j PTI-C85-P0001, Rev. 1, TC-2 2
Description of Change q
These instruction changes evaluate operation of the Steam Bypass and Pressure Regulating (SB&PR) system in single channel mode.
Summary 1
i I. No. The SB&PR system is described in the USAR as a control system not j required for safety. No new failure modes are introduced. Accident 1
analysis remains unaffected. Therefore, the probabi,ity of occurrence or the consequences associated with an accident or
{ malfunction of equipment has not changed.
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II. No. Placing the SB&PR system in single channel mode does not introduce 1
any new failure mechanism. Accident analysis is not impacted.
Therefore, the possibility of an accident or malfunction of a I
different type than any previously evaluated in the USAR is not created.
l III. No. Technical Specification 3.7.6, Main Turbine Bypass System, is i directly related to the SB&PR system. The bases for this Technical Specification states, 'The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis in FSAR Chapter 15." The SB&PR system uses only one channel to provide combined flow demand, bypass demand and load error signals; therefore, its operation is consistent with the assumptions of the feedwater controller failure analysis when in the single channel mode. Therefore, the margin of safety as defined in the bases for Technical Specifications is maintained.
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j SE No.: 95-0029 Source Document: DCP 94-0073, T.<ev. 0 4 Description of Change This design change establishes an alternate boron injection path to the !
reactor vessel in response to NRC Unresolved Item 91013-05 (Licensing )
Commitment L01408). The Alternate Boron Injection (ABI) system will have l l the capability to transfer the boron solution from the existing Standby l
! Liquid Control (SLC) Transfer system boron transfer mixing tank to the 1
reactor vessel via the High Pressure Core Spray (HPCS) system injection ;
piping flushing connection, by utilizing flexible hoses. '
Summary i I. No. The modifications meet the design, material, and construction
! standards of the affected systems as mandated by CEI Installation i Standard Specification SP-2000. The ABI system does not interface '
- with any existing plant systems during normal plant operation and/or s
Design Basis Accident (DBA). This modification does not affect any 4
existing systems required to mitigate radiological consequences. l The installation of the ABI system does not introduce any adverse ;
i system interactions. Therefore, the probability of occurrence or j the consequences of an accident or malfunction of equipment ;
- important to safety previously evaluated in the USAR cannot be l
! increased. t j II. No. The ABI system will not be connected to the HPCS or the SLC Transfer l l systems during normal plant operation and/or DBA. The ABI system ;
j components meet the design, material, and construction standards of 1
! the affected systens. The changes associated with this activity do not affect the design function / operation or degrade the performance l of any equipment. No new hazards are created due to this i modification. Therefore, the possibility of an accident or ;
! malfunction of a different type than any evaluated previously in the j USAR will not be created.
1 III. No. The ABI system is not addressed in any bases contained in the
- Technical Specifications. The ABI system components meet the
- design, material, and construction standards of the affected
- within limits as described in the Technical Specifications. As such, the margin of safety as defined in the bases for any Technical Specification will not be reduced.
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d SE No.: 95-0030 Source Document: USAR Change Request 95-029 Description of Change i This change request evaluates correction of a note in USAR Tables 6.3-8, 6.3-9, and 6.3-10 to read, " Piping low point drains, high point vents and test connections are provided with dual isolation".
Summary I. No. This change request doe not create any physical changrs to the plant. The changes involve the clarification of USAF, text only.
The changes were analyzed to be consistent with the original FSAR
, analysis assumptions. The design, material, and construction standards applicable to the High Pressure Core Spray (HPCS), Low Pressure Core Spray (LPCS), and Residual Heat Removal (RHR) systems are not modified. The subject drain, vent or test connections have protection features built into the design which will avert any concern for breach of pressure boundary. These protective design features provide inherent reliability of the systems to perform required safety functions. 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. HPCS, LPCS, and LPCI systems comprise the injection network for emergency core cooling and are designed to provide protection against postulated LOCAs. The change request does not make any physical modification to tne plant. The subject drains, vents, and test connections are not part of the main flow stream of the HPCS, LPCS, or LPCI systems. The reliability of the HPCS, LPCS, and LPCI systems to perform required safety-related functions will not be altered. Therefore, the possibility for an accident or malfunction of a different type than previously evaluated in the USAR will not be created.
III. No. Existing Technical Specifications in part provide the basis for acceptability of this USAR change and will be unchanged. As stated above, HPCS, LPCS and RHR system design and operation will remain unaffected. Therefore, the margin of safety remains intact.
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j SE No.: 95-0031 i j Source Document: USAR Change Request 92-028 i
i Description of Change :
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2 This USAR change eliminates postulated breaks RHS1 and RHSlLL in the '
recirculation piping based upon the results of GE Design Report 23A4755, '
4 Rev. 1. No new breaks are added. :
Summary
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- I. No. This change involves the removal of postulated breaks in the ,
recirculation piping based upon updated analysis. No equipment '
changes are being made. Therefore, neither the probability of l
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occurrence nor the consequences of a previously analyzed accident s
will be increased by this change. ;
1 II. No. There is no change to equipment or to plant operation as a result of i this postulated pipe break elimination. Therefore, this change will not create the possibility for an accident or malfunction of a i different type than any previously evaluated.
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! III. No. The Technical Specifications do not address the postulation of rupture locations within any piping system nor do they address the criteria associated with this issue. The configuration and the
} function of the piping remain unaltered. Therefore, this change
- will not reduce the margin of safety as defined in the bases for any i
Technical Specifications, f
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SE No.: 95-0032 Source Document: SOI-C71, Rev. 7 Dascription of Chance This system operating instruction evaluates changes which incorporates
- provisions for bypassing the Reactor Mode Switch Manual Scram.
Summary l
I. No. No accident analyzed in USAR Chapter 15 takes credit for the Reactor Mode Switch Manual Scram. Manual scram capability with the installed scram pushbuttons is not affected by the jumpers being installed. No other functions of the Reactor Mode Switch are
- bypassed when the jumpers are installed. In addition, all control ,
i rods are verified fully inserted prior to installing the jumpers. l 4
Therefore, this action does not alter the probability or 1 consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
- II. No. The instruction includes provisions to ensure all control rods are fully inserted prior to bypassing the Reactor Mode Switch Manual
! Scram and remain inserted throughout the evolution. If a lead
- should be inadvertently grounded during the installation or removal j of the jumpers, it may cause the deenergization of a channel of the a
Reactor Protection (RPS) system; this is an analyzed event.
i Therefore, this action does not create the possibility of an accident or a malfunction of equipment important to safety of a j different type than any previously evaluated in the USAR.
i III. No. The bases for the Reactor Mode Switch Manual Scram is given in Technical Specification'2.2.1.12 as, "The reactor mode switch shutdown position provides additional manual trip capability." The RPS is described in Technical Specification 3.3.1. The instruction ensures that all of the action requirements associated with the Reactor Mode Switen Manual Scram function being inoperable are met.
Therefore, the margin of safety as defined in the bases for Technical Specifications is maintained.
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l SE No.: 95-0033 Source Document: S01-M14, Rev. 11 Description of Change j i
This system operating instruction evaluates the elimination of the l requirement to operate the Containment vessel and Drywell Purge (M14) !
system at the full flow rate prior to personnel entry into the drywell i and for refueling operations.
Summary l
I. No. No changes are being made to the M14 system isolation logic or system operation logic for other ventilation systems which would effect the assumptions made in USAR Section 15.7.6. When the M14 system is not operated prior to personnel entry into the drywell or during refueling operations, other systems will be required to be operated. The USAR states that the Containment Vessel Cooling system can provide all of the cooling capacity required in the containment vessel, and that the Drywell Cooling system is designed to maintain temperatures suitable for equipment operation in the various drywell areas. Therefore, this action does not alter the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The instruction directs compensatory actions to be taken to ensure the design objectives of the M14 system are met. These compensatory actions are all within the normal scope of routine operations.
Therefore, this action does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
III. No. The M14 system is discussed in Technical Specification 3.6.1.8. ;
Operation of the M14 system also affects Technical '
Specifications 3.6.1.1, 3.6.1.2, and 3.6.4. All of these Technical Specifications are concerned with containment isolation and integrity. As discussed above, no changes are being made to the M14 system isolation logic. Therefore, the margin of safety as defined in the bases for Technical Specifications is maintained.
4 SE No.: 95-0034 Source Document: DCN 4951 Description of Change This drawing change provides the necessary alterations to refornat and effect Computer Aided Drafting and Design (CADD) conversion of the Piping and Instrument Diagram (P&ID) lead sheets. No technical changes are being made, the existing information is being presented more clearly, along with direction as to the correct symbology to be used for future changes.
Summary I. No. The change is editorial only. The effect will be to more clearly portray the existing plan configuration on the P& ids. There are no changes to the plant procedures or installed hardware. System performance, pressure boundary integrity and safety system operations are anaffected. The change does not create any new !
system interactions and does not reduce the redundancy or independence of any system important to safety. Therefore, the probability or the consequence of an accident or malfunction of equipment important to safet" previously evaluated in the USAR has not changed.
II. No. The change is editorial only. The change does not involve an initiator or failure not considered in the USAR. The change creates no new systems. There are no physical changes to the plant nor to operating procedures made by the changes. Therefore, change will not create the possibility of an accident nor a malfunction of
, equipment important to safety of a different type than any evaluated
, previously in the USAR.
III. No. There are no physical changes to the plant nor to any operating procedures made by the change. The P& ids are not governed by nor directly referenced in the Technical Specifications. Therefore the change does not reduce any margin of safety.
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SE No.: 95-0035 Source Documen_t_: DCN 4937 Description of Change l
This drawing change revises plant drawings by indicating an increase in !
the maximum allowable stroke time of the following valves: 1G43-F0030A/B l and 1G43-F0040A/B. The change increased the stroke times for these ;
valves from 30 to 35 seconds. I Summary l
I. No. Analysis showed that the increase in valve stroke time would not i increase the potential or severity of an accident. The change does not impact the time requirements for the upper containment pool to drain into the suppression pool. This change does not impact environmental or radiological conditions since the post-LOCA suppression pool vent coverage of at least 2 feet above the top of I i the top row of horizontal drywell vents is maintained for long term J steam condensation. Therefore, neither the probability of I occurrence nor the consequences of a previously analyzed accident or ,
malfunction of equipment will be increased by this change. '
II. No. Analysis showed that no new equipment types or new system interactions are created. The original plant design basis is maintained. There are no hardware changes to the plant.
Performance of the affected system is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of different type than any previously evaluated.
III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. These valves will continue to open to allow a sufficient volume of water to be transferred from the upper containment pools to the suppression pool. The change results in no effect of the results of any LOCA l analysis. Therefore, no reduction of safety margin is caused by j this change.
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SE No.: 95-0036 Source Document: PTI-P42-P0003, Rev. O i
Description of Change !
This periodic test instruction evaluates momentary operation of the !
Emergency Closed Cooling (ECC-P42) system pump at a flow rate less than l the continuous duty minimum flow value. This will occur when the primary i flowpath is isolated by closing a motor operated butterfly valve, in
) order to obtain diagnostic test data on the butterfly valve in accordance {
with NRC Generic Letter 89-10. '
Summary l l
I. No. An engineering evaluation determined that it was acceptable to !
operate the ECC pumps at flow rates below the continuous duty l minimum flow value specified by the pump vendor, for a duration as '
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long as one hour. This test instruction only intermittently i operates the ECC pump operates below the continuous duty minimum l flow value, and for a duration no longer than one minute. ;
Additional pump monitoring.will be performed. All other aspects of the system operation, automatic functions, and interlocks for the !
ECC system will remain as described in USAR Section 9.2.2.
Therefore, the probability of occurrence or the consequences l associated with an accident or malfunction of equipment has not changed.
II. No. All automatic functions and interlocks described in USAR Section 9.2.2 for the components operated under this instruction will remain operable. The testing will not affect the way the ECC system functions. Because the operation of the ECC system is unchanged, the proposed activity 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. Operation of the ECC system remains within the plant design bases and Technical Specification requirements. Therefore, no margin of safety as defined in the bases for any Technical Specification is affected.
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SE No.: 95-0037 !
Source Document: DCP 94-0172, Rev. 0 I l
Description of Change This design change establishes a Vehicle Barrier System (VBS) in response to the August 31, 1994 amendment to 10CFR73, " Protection Against Malevolent Use of Vehicles at Nuclear Power Plants". The purpose of the l
VBS is to ensure that a four-wheel drive vehicle cannot be malevolently used to transport personnel and explosives into direct proximity of plant vital areas.
Summary I. No. No modifications were made to any systems, structures, or components 1 identified as accident initiators or credited with accident '
l mitigating functions in the USAR. All onsite and offsite doses
! remain unaffected by this activity. Installation of the VBS is !
designed such that surface runoff is not impeded, and a new l potential flooding is not created. Therefore, the probability of i occurrence or the consequences of an accident or malfunction of ]
equipment important to safety previously evaluated in the USAR '
cannot be increased.
II. No. The existing equipment modified by this activity is i nonsafety-related and non-seismic. No interactions with .
safety-related equipment are being created by this activity, and, thus, no new types of system interactions are created that could create or contribute to the possibility of a new transient or accident. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the USAR will not be created.
l III. No. The proposed activity is intended to enhance the plant's security by
! modifying the Site Security Plan in response to amended 10CFP.73.
l The VBS will not detract from any existing portions of the site l
Security Plan. Therefore, this activity does not reduce any margin of safety.
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SE No.: 95-0038 Source Document: DCP 94-0151, Rev. O Description of Change This modifica, ion overlays Stainless Steel (SS) cladding on the inside of the 24' diameter, center downcomer between the Direct Contact (DC) Heater and the Hot Surge Tank (HST). The DC Heater and the HST are ASME Section VIII, Division I, U stamped vessels. Since the ANSI /ASME B31.1 Code does not specify repair procedures, the downcomer is being repaired to the requirements of the ASME Section VIII and the National Board Inspection Code (NBIC).
Summary 1
I. No. This modification to the DC Heater downcomer does not alter the ,
design, material and construction standards consistent with the USAR. Code classification is unchanged in that the repair will still comply with ASME Section VIII, Division I. The overall design of the DC Heater, HST and connecting piping is unchanged.
Radiological consequences involving USAR accidents or transients related to feedwater are unchanged. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed. ,
II. No. The SS cladding is not expected to degrade or break free causing flow blockage or pieces to travel into downstream systens, since it meets the original design intent. No new failure mechanisms have been created. The DC Heater will continue to meet the original design for materials and construction. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specifications are not related to the modifications to the DC Heater, the HST or the interconnecting piping. The Condensate (N21) system will still function as designed with no threat to the Feedwater (N27) system, reactor coolant pressure boundary, or reactor water chemistry. Therefore, this modification does not reduce any margins of safety associated with the N21 and N27 systems.
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l SE No.: 95-0039 l Source Document: DCP 94-0139, Rev. 2 1
l Description of Change l
This design change establishes a reliable keep fill water supply to the
' Emergency Service Water (ESW) system by the addition of a new 2' Service i Water (SW) keep fill line. J Summary I. No. The modification meets the design, material, and construction t
standards of the SW and ESW systems (P41 and P45) as mandated by CEI I
' Installation Standard Specification SP-2000 for Line Specifications L1-4 and L2-4. The installation of the new keep fill line performs the same function as was originally designed for the plant. The installation of this new keep fill line does not introduce adverse system interactions, and is limited to ensuring that the ESW system remains filled, thus preventing potential water hammer effects.
I 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 changed.
II. No. This modification meets the design, material, and construction standards of the SW and ESW systems. The changes associated with i
this activity do not change the design function or operation of any equipment and this activity will not affect the function or degrade the performance of any System, Structure, or Component (SSC). The l
modification provides for a reliable keep fill source in accordance with the design basis of the ESW system and maintains the ESW system in a reliable condition. Therefore, the possibility of an accident or malfunction of a different type will not be created.
III. No. This design change provides a reliable source of keep fill water to the ESW system. Maintaining the ESW system filled helps ensure that the ESW is maintained operable in accordance with Technical Specification requirements. Further, this activity in no way alters the normal or emergency functions of any of the affected systems.
As such, no margin of safety will be reduced.
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SE No.: 95-0040 Source Locument: WO 94-3888 ;
RPI-0507, Rev. 2 l Description of Change This work order and instruction change analyzes the acceptability of the Drywell Personnel Airlock Shield Doors being placed in the open position during plant Operational Modes 1, 2 and 3. This action places additional loading onto the 620'-6' platform structure. However, the platform structure remains below the yield stress limits. Dae to the limited time this condition occurs, it is acceptable to leave the doors in the'open position. This safety evaluation permits this activity until the end of Refueling Outage 5.
Summary I. No. The 620'-6' platform continues to perform its intended function, that being support of various plant equipment and piping.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Since the shield doors, supporting hardware and the 620'-6' platform remain functional with the doors in the open position, no new accident initiators or failures are introduced. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The primary function of the 620'-6' platform is support of equipment and piping. With the doors in the open position it has been determined that the platform continues to perform it intended function. As a result, all systems continue to perform their required functions with no degradation in performance. As such, the l overall performance of the equipment which is supported by the j platform remains unchanged. Therefore, the safety margins will l remain bounded by the original acceptance limits and are not reduced.
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l SE No.: 95-0041 I i
Source Document: DCP 87-0725, Rev. 1 l
Description of Change This design change replaces safety-related analog Riley temperature modules associated with the Leak Detection system with General Electric i NUMAC digital leak detection monitors. This modification removes 52 of l the 60 Riley temperature modules. These temperature modules provide i divisional alarms, and when necessary, isolation signals which close ;
either inboard or outboard containment isolation valves for a specific 1 system, when high ambient or high differential temperature is sensed. I This evaluation resulted in a determination that NRC review is necessary prior to implementation. A request for NRC review has been submitted under separate cover. i i
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4 SE No.: 95-0042 l 1
- Source Document: DCP 90-0123, Rev. O l
! Description of Change
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l This design change involves replacing the existing Turbine Sample Panel with a new sample panel manufactured by Sentry Corporation. The panel is i
designed to facilitate on-line sampling of 35 process system streams {
! associated with main feedwater, main steam, condenser hotwell, condensate i i
storage, condensate demineralizers, heater drains, direct contact heater, l condensate hotwell pump discharge header, condensate filter outlet header, two bed distribution system header, mixed bed distribution system j header, and condensate demineralizer system effluent header.
A
! Summary
) I. No. The Turbine Sampling system and interfacing systems are classified
} as nonsafety-related systems. The design change does not alter the design basis for any of the systems, subsystems, or components that i
are the subject of the various accident scenarios outlined in the t
USAR. Radwaste inventories between the old and new sample panel were deemed comparable and as such interactions with the radwaste l process systems were deemed equivalent and within the design basis l of the existing radwaste processing systems. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
l II. No. The design change creates no new systems nor does it make any alterations to existing plant systems. The Turbine Sampling system
} is designated as nonsafety-related. Each system interaction
] required to support operation of the Turbine Sample Panel was evaluated to ensure that each system interaction was bounded by the i
design basis for that system and that the entire scope of the plant
- modification was bounded by the accident scenarios outlined in the I
USAR. There is no change to the operation, function, or logic of the radwaste systems and system interactions with the radwaste systems were deemed equivalent as a result of this plant i i modification. Therefore, the possibility of an accident or j
malfunction of a different type than any previously evaluated in the l
USAR is not created.
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III. No. The Turbine Sampling system is classified as a nonsafety-related system. Process sampling is not discussed in the Technical Specifications. The Technical Specifications do address power systems and liquid effluent systems, but no impact has been identified as a consequence of this change. Therefore, no margin of safety has been changed.
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1 SE No.: 95-0043, 95-0175 Source Document: DCP 91-0085, Rev. O Description of Change ;
This design change replaces obsolete ASCO solenoid valves on the Inboard MSIVs with solenoid valve cluster assemblies manufactured by Ralph Hiller i Co. The existing ASCO valves are solenoid pilot valves that are used to position the existing Norgren 2-way and Norgren-4 way air control valves to establish the open, close, or test position for the Inboard MSIVs.
The new solenoid valve cluster assembly is an engineered replacement for the existing ASCO configuration. :
Summary I. No. The new solenoid valve cluster assemblies satisfy the ASME/ ANSI B31.1, Safety Class 2 requirements currently employed for the existing ASCO configuration. The solenoid valve cluster assemblies are a specifically engineered and qualified assembly that was specifically designed as a direct replacement for the existing obsolete ASCO solenoid valve arrangement currently employed for MSIV -
operation. They are designed to operate the MSIV in the same manner i as the existing ASCO solenoid valves. The design basis or licensing basis for MSIV operation is not altered by this design change. i Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The design basis for operation of the Inboard MSIVs is not altered 4 by this design change. Plant system interfaces are the same as the existing system interfaces required for proper operation of the Inboard MSIVs. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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III. No. The time response requirements outlined in the Technical l
Specifications for MSIV operation are not adversely impacted by this ,
design change. The Controls and Indication requirements outlined in the Technical Specifications for the Inboard MSIVs are not altered ,
by this design change. Therefore, no margin of safety has been '
changed.
.i SE No.: 95-0044 1
Source Document: DCP 93-0125, Rev. O Description of Change This design change modifies the Primary Containment Personnel Airlocks to eliminate chronic shear failures of the airlock ball valve stems and improve airlock reliability. The modifications include:
- 1) disconnecting and removing the existing hydraulic system components and the existing handwheel locking device; 2) converting the existing door linkage-type mechanisms to gear-type mechanism to control the activation of the ball valves; 3) installing modified hatch pins containing springs and a reaction collar; and 4) installing a handwheel locking system.
Summary I. No. The modifications institute measures to improve reliability of the safety-related ball valves and the operation of the airlocks. The new nonsafety-related, nonseismic instrument air tubing runs will not adversely affect any safety-related structures, systems, or components. Pressure boundary components important to safety are designed, installed, and tested in accordance with original criteria for which the plant was licensed. Failure of the modified airlock doors to properly close and seal results in the same consequences as before because the size of airlock opening, the pressure experienced in containment, the operation of the Emergency Core Cooling Systems, ;
and the source term in containment are not changed by this activity.
Furthermore, this activity does not adversely affect the subcomponents that provide the pressure boundary function and does i not impact the pressure boundary function of the airlocks.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR cannot be increased.
II. No. Failure of new components installed by this modification have the same results as failure of the existing components. The personnel airlocks serve tc mitigate the consequences of accidents (by providing containment integrity and pressure boundary), and loss of mitigation function cannot create a new accident. Loss of nonsafety-related circuit or nonsafety-related instrument air have been considered in the original design and remain bounding. Failure of the nonsafety-related circuit or the nonsafety-related instrument air will free the handwheel to operate. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the USAR will not be created.
III. No. The Technical Specifications require only one closed door to maintain the integrity of the containment, and the mechanical interlock system supports this requirement. By adherence to the ASME Code requirements, and testing in accordance with 10CFR50, Appendix J, this modification ensures that the airlocks and the door seals are functional as containment pressure boundaries. Therefore, no margin of safety has been changed.
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i SE No.: 95-0045 i Source Document: DCN 4961 J
j Description of Change 1
This drawing change modifies P&ID D302-739, Liquid Radwaste Sump System, {
- to reflect that the head of pumps 1G61-C0001A/B is 60 feet rather than 65 ;
l feet.
l Summary '
i j I. No. The drywell equipment drain sump pump does not play a role in the i !
accidents evaluated in of the USAR. The sump level indication and i
the rated pump flow (50 GPM) is not affected. The use of a pump
! with 60 feet of head (versus 65 feet of head) will nc* 'nfluence the ;
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radiological aspects of any USAR evaluated accidents, 2nerefore, '
the probability of occurrence or the. consequences associated with an l accident or malfunction of equipment has not changed.
II. No. The sump pump flow remains at 50 GPM. The originally stated pump head of 65 ft. is not utilized as inputs to any engineering calculations. USAR accident analysis is unaffected. Therefore, the l possibility of an accident or malfunction of a different type than i any previously evaluated in the USAR is not created.
l III. No. The sump pump is not described in the Technical Specifications. '
Drywell leakage rates or their identification are not affected by
) the sump pump. The sump pump will remove 50 GPM of water, which is j its intended function. Therefore, no safety margin has been
- reduced.
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I SE No.: 95-0046 Source Document: Potential Issue Form (PIF) 95-0812 Description of Change l
This PIF analyzes the " temporary repair' disposition for the inet-llation of a Team Inc. leak sealant clamp on a 2' line of the High Pressure Heater Drains and Vents (N25) system. This leak sealant device will reduce the effects of a steam leak coming from a coupling socket weld.
The leak sealant device to be installed is designed in accordance with ASME Code Section VIII, Division I criteria, and ASTM standards. The resulting worst case flooding from this steam leak is completely bounded by the previous Turbine Building flooding analysis. The potential failure of the 2' pipe weld would be completely bounded by the postulated loss of feedwater heating accident and a feedwater line break outside of containment accident as discussed in USAR Chapter 15.
Summary I. No. The design and manufacture of the leak sealant device is made to !
approved industry _ codes and standards. The disposition has no clear discernible affect on the Chapter 15 accidents discussed in the USAR. Steam leakage would be handled by the Turbine Building Ventilation system which is monitored for radiological concerns.
There is no possible affect on essential safety-related equipment as a result of this disposition. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Safety systems operability remains unaffected with or without assumed failure of the 2" diameter weld in question. Flooding is completely bounded by the previous flooding accident discussed in USAR Section 10.4.5.3.1, and a complete failure of this weld is bounded by the loss of feedwater accidents as discussed in USAR Chapter 15. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The installed clamp will act as a secondary boundary around the weld and will therefore restore the pipe line integrity. The clamp and weld are not expected to fail during normal plant operation, so the N25 system and plant operation remain unaltered. Therefore, the Technical Specifications, Technical Specification bases and associated safety margins remain completely unaffected.
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l SE No.: 95-0047 Source Document: DCP 95-5032, Rev. O Description of Change i
i Shis design change evaluates the injection of an approved leak sealant into the stuffing box of valve 1G33-F0101 in order to stop a packing leak. Valve 1G33-F0101 is a 3" Borg Warner motor operated gate valve associated with the Reactor Water Cleanup (RWCU) system. The valve is classified as an ASME Section III, Class 1 component and is part of the Reactor Coolant Pressure Boundary (RCPB) . However, the leak sealant act.ivity involves the nonsafety, non-ASME code, non-pressure retaining valve components. The valve will remain open with its motor operator tc.oed out. The change in the positioning of valve 1G33-F0101 from no.a. ally CLOSED to normally OPEN was previously evaluated and found acciotable.
Summary I. No. The leak sealant will have no adverse affect on the pressure retaining capability of the valve or on reactor water chemistry. t Containment isolation capabilities, provided by 1G33-F001 and -F004, remain unaffected by this change. Therefore, the probability of j occurrence or the consequences associated with an accident or '
malfunction of equipment has not changed. l II. No. This design change eliminates the source of an input signal to a
' Control Room identified leakage alarm, but does not alter or effect the containment isolation circuitry. Valve leakage could still be detected and processed as unidentified leakage. Containment isolation capabilities, provided by 1G33-F001 and -F004, remain unaffected by this change. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. l III. No. Technical Specifications 3/4.4.1 and 3/4.4.3.2, and the USAR do not i identify conditions which would prohibit plant operation with i 1G33-F0101 injected with leak sealant. The leak sealant injection of 1G33-F0101 would not result in degradation in any system used to safely shutdown the reactor or mitigate the consequences of an accident. The RCPB and the functions of the RWCU system containment isolation valves remain unaffected. This activity does not reduce the margin of safety as defined in the bases for any Technical Specification.
SE No.: 95-0048, 96-0009 Source Document: DCP 94-0027, Rev. O and Rev. 4 Description of Change This design change installs a bypass line around each of the Emergency Closed Cooling (ECC) heat exchangers (lP42-B0001A and 1P42-B0001B) and utilizes an electro-hydraulic modulating Temperature Control Valve (TCV)
(1P42-F0665A and 1P42-F0665B) to control the flow between the heat exchanger and the bypass line based on the ECC water temperature downstream of the heat exchanger.
Summary I. No. The modifications meet the Usign, material, and construction standards of the ECC (P4:, system as mandated by CEI Installation Standard Specification SP-2000 for Line Specification L1-3. The 4
3-way valve is provided with a fully qualified actuator which is powered by a safety-related lE power supply. Further, the actuators
, are powered from the same Engineered Safety Functions (ESP) division as the components it serves to ensure redundancy is maintained. The installation of the TCV valves and bypass line around the ECC heat exchangers does not introduce adverse system interactions, and is limited to ensuring that the ECC temperature remains within acceptable limits. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR cannot be increased.
II. No. This modification meets the design, material, and construction I standards of the ECC system. The changes associated with this activity do not change the design function or operation of any equipment. This activity will not affect the function or degrade l the performance of any system, structure, or component. The two l TCVs are being installed on the two redundant loops of the ECC system which are powered by separate lE divisions. The response of the ECC system to any accidents is not degraded or changed by this activity. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the USAR will not be created.
III. No, Implementation of this change enhances the performance of the ECC (P42) system. Further, this activity in n.) way alters the normal or emergency functions of any of the affected systems. As such, no margins of safety will be reduced.
i SE No.: 95-0049 j
Source Document: NR 94-S-707, Rev. 2 l
Description of Change:
i This Nonconformance Report (NR) evaluates the "use-as-is" disposition of fourteen Motor Operated Valves (MOV) in the Emergency Closed Cooling (P42) and the Nuclear Closed Cooling (P43) systems that were not analyzed for worst case design conditions.
Summary:
I. No. The valves in question were shown to be capable of performing their j active safety function, to close during a design basis event per the '
requirements of Generic Letter 89-10. Therefore, isolation of P43 from P42, containment isolation, and drywell isolation functions are maintained. The P43 system is not required for safe shutdown of the reactor. The P43 system was originally designed for two unit operation; P43 "A" and 'B' pumps support Unit 1 and common equipment and are powered from Unit 1 busses, P43 "C" pump was to support Unit 2 and common equipment and is powered from a Unit 2 source. It was shown that the continued operation of the P43 'C' pump during and after a LOCA had no adverse impact on the operation of the P43 system or equipment. Therefore, the probability of the occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR have not increased.
II. No. The P43 pumps are not required to function during an accident or-mitigate the consequences of an accident. The valves have adequate capability to isolate against an in-service P43 pump. The operation of the P43 "C" pump during and after a LOCA does not involve an initiator or failure not considered in the USAR. Since the valves have adequate capability to perform their design basis function, no new type of valve malfunction is introduced. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a type different than any previously evaluated in the USAR is not created.
III. No. The MOVs were shown to be capable of performing their design function with the P43 'C' pump in operation. Therefore, Technical Specifications 3/4.6.4 (Containment Isolation Valves) and 3/4.7.1.2 (Emergency Closed Cooling Water) are not impacted. It was also shown that the P43 'C' pump would operate within design parameters with operation post-LOCA. Therefore, no margin of safety has been l
reduced.
SE No.: 95-0050
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Source Document: USAR Change Request 95-030 Description of Change This USAR change clarifies USAR Section 17.2.3.2.k to state that only
" approved" computer programs will be used to perform design activities.
Summary I. No. This change is a clarification to the use of design-related computer programs. Design criteria have not been altered. Computer programs directly related to plant operation have not been affected.
Administrative controls exist to ensure proper use of the programs.
Accident analysis remains unchanged. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change is a clarification to the use of design-related computer i programs. Design criteria have not been altered. Computer programs l directly related to plant operation have not been affected. '
Controls are in place to ensure the programs are reviewed and tested prior to final use. Controls are also in place to ensure program output is correctly used. Accident analysis has not been affected.
Therefore, the possibility of an accident or malfunction of a i different type than any previously evaluated in the USAR is not created. l l'
III. No. This change involves computer programs which are only involved in the design of the plant. The programs are used during the design process to ensure that the design criteria remain satisfied. No design criteria have been altered. Computer programs associated with direct plant operations (e.g., programmable logic controllers, ERIS, and microprocessors) are not affected by this change.
Controls are in place which ensure that the design programs are reviewed and tested prior to final approval and use, and that programming output has been properly used. USAR accident analysis remains unchanged. Therefore, no margin of safety has been reduced.
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j I il SE No.: 95-0051 i DCP 92-0135, Rev. O Source Document:
i l Description of Change i
! This design change is Phase II of a multi-phase program to upgrade the I plant computers. This phase will replace the process computer and its i data acquisition system by integrating the existing nuclear steam supply,
. 3D-Monicore, and balance of plant functions unto a comnon computer
! platform.
J Summary '
i I. No. This design change will integrate various functions unto a single platform. The program functions will not be affected. There is no j plant control interface with the process computer and its programs.
The process computer is not relied upon for accident analysis.
! Therefore, the probability of occurrence or the consequences i associated with an accident or malfunction of equipment has not i changed.
II. No. This design change will integrate various functions unto a single platform. The program functions will not be affected. There is no plant control interface with the process computer and its programs. i The process computer is not relied upon for accident analysis, i 1 Therefore, the possibility of an accident or malfunction of a ,
different type than any previously evaluated in the USAR is not i created. '
III. No. Technical Specification variables supplied by the 3D-Monicare
, software to verify Technical Specification compliance will not be-i affected. There is no plant control interface with the process
! computer and its software. Accident analysis remains unchanged.
Therefore, no margin of safety has been changed.
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SE No.: 95-0052 Source Document: DCP 94-0187, Rev. O Description of Change This design change replaces motor operated valve lE51-F0059 in order to restore the valve to an operable status in accordance with the requirements of Generic Letter 89-10. This valve is the second isolation valve in the test return line for the Reactor Core Isolation Cooling (RCIC) system. Its primary function is to isolate the test return line to the Condensate Storage Tank (CST) in order to assure ;
adequate RCIC flow to the reactor. In addition to replacing the valve, '
the motor actuator capability was increased.
Summary j
l I. No. The modifications meet the design, material, and construction ;
standards of the RCIC system. Since this modification does not i affect the RCIC system functionality, control interlocks, response '
times, system hydraulics, initiation responses, or normal operation of valve 1E51-F0059 or the RCIC system, single failure protection is maintained. Adherence to the ASME Code for the valve and piping ,
design ensures that the probability of a loss of the RCIC system '
will not increase. Restoring valve 1E51-F0059 ensures that the valve performs its safety function as an engineered safety feature !
boundary, thus maintaining the design requirements of RCIC system. !
This ensures that the radiological consequences of postulated I transients and accidents remain below the limits of 10CFR20 and 10CFR100. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR cannot be increased.
II. No. This modification meets the design, material, and construction standards of the RCIC system. The changes associated with this i activity do not change the design function or operation of any plant l component or equipment. Therefore, the possibility of an accident ;
or malfunction of a different type than any evaluated previously in I the USAR will not be created.
III. No. Implementation of this change will restore the capability of the RCIC system to automatically provide isolation cooling if an initiation signal is received with the system in the test mode. By adherence to the ASME Code for the new valve, and the ANSI B31.1 Code for the downstream piping, this change ensures that the margin of safety associated with Technic:1 Specification 4.0.5 for system pressure integrity is maintained. As such, no margin of safety will be reduced.
SE No.: 95-0053 Source Document: DCP 95-0017, Rev. O Description of Change This design change replaces the Reactor Core Isolation Cooling (RCIC) system Turbine Exhaust Valve IE51-F0040, and adds a second 3/4" drain isolation valve downstream of 1E51-F0041. Additionally, this modification changes the classification of the associated penetration, P106, from GDC 57 to GDC 56. ,
Summary 1
I. No. The replacement check valve IE51-F0040 and associated piping are j designed to meet or exceed the design (ASME Section III, Class 2) and performance requirements for the RCIC system including the new ;
function of containment isolation. The replacement of the check !
valve is in no way a casual factor in the initiation of previously evaluated accidents. The specified allowable leakage for the new valve has been determined so that the total containment leakage, I including secondary containment bypass leakage, will not exceed the value assumed in the accident analysis. This activity does not I affect the function or degrade the performance of any Structure, System, or Component (SSC), or cause any SSC to operate its design basis or fail to perform its safety function. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR cannot be increased.
II. No. The replacement of the check valve is in no way a casual factor in the initiation of any accidents. This activity does not affect the function or degrade the performance of any Structure, System, or Component (SSC), or cause any SSC to operate its design basis or fail to perform its safety function. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the USAR will not be created.
III. No. The total containment leakage, including secondary containment bypass leakage, will not exceed the value assumed in the accident analyses which provide the bases for the leakage volume for Technical Specification 3.6.1.2. This activity will assure the availability of the RCIC system, and maintain the containment isolation criteria specified in Technical Specification 3.7.3.
Therefore, the proposed activity does not reduce the margin of safety as defined in the bases for any Technical Specification.
SE No.: 95-0054 Source Document: PAP-0803, Rev. 7 Description of Change This procedure revision will incorporate a screening criteria to be used for the approval of chemicals with regards to a postulated failure in accordance with Regulatory Guide 1.78.
Summary I. No. The procedure changes will ensure the requirements of Regulatory Guide 1.78 for Control Room habitability are met. The basis for the approval of chemicals and chemical storage have not changed. These changes ensures that existing requirements are maintained.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
, II. No. The procedure changes will ensure the requirements of Regulatory Guide 1.78 for Control Room habitability are met. The basis for the approval of chemicals and chemical storage have not changed. These changes ensures that existing requirements are maintained.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change will only clarify responsibilities for chemical storage and inspection and ensure the requirements of Regulatory Guide 1.78 are met. Therefore, no margin of safety has been changed. l 1
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SE No.: 95-0055 i
- 5ource Document
- Potential Issue Form (PIF) 95-0907 Description of Change l I
This PIF analyzes the " temporary repair" disposition for the installation of a Team Inc. leak sealant clamp on a 2' line of the High Pressure '
j Heater Drains and Vents (N25) system. This leak sealant device will i reduce the effects of a steam leak coming from a pipe socket weld. The
- leak sealant device to be installed is designed in accordance with ASME Code Section VIII, Division I criteria, and ASTM standards. The resulting worst case flooding from this steam leak is completely bounded 4
- by the previous Turbine Building flooding analysis. The potential !
) failure of the 2' pipe weld would be completely bounded by the postulated i i
loss of feedwater heating accident and a feedwater line break outside of '
i containment accident as discussed in USAR Chapter 15. ;
Summary b
! I. No. The design and manufacture of the leak sealant device is made to i
1 approved industry codes and standards. The disposition has no clear discernible affect on the Chapter 15 accidents discussed in the 1
USAR. Steam leakage would be handled by the Turbine Building Ventilation system which is monitored for radiological concerns.
] There is no possible affect on essential safety-related equipment as I
a result of this disposition. Therefore, the probability of :
occurrence nr the consequences associated with an accident or !
malfunction of equipment has not changed. !
d i II. No. Safety systems operability remains unaffected with or without I
, assumed failure of the 2* diameter weld in question. Flooding is
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completely bounded by the previous flooding accident discussed in USAR Section 10.4.5.3.1, and a complete failure of this weld is bounded by the loss of feedwater accidents as discussed in USAR Chapter 15. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the i USAR is not created. ;
III. No. The installed clamp will act as a secondary boundary around the weld and will therefore restore the pipe line integrity. The clamp and weld are not expected to fail during normal plant operation, so the N25 system and plant operation remain unaltered. Therefore, the Technical Specifications, Technical Specification bases, and associated safety margins remain completely unaffected.
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i SE No.: 95-0056 Source Document: Potential Issue Form (PIF) 95-0906 l Description of Change This potential issue analyzes the ' temporary repair" of installing a leak :
l sealant device on a 2' portion of the Extraction Steam (N36) system. The ;
l device will minimize the effects of a lerng socket weld. The device to !
be installed is designed per ASME Code ' , ion VIII criteria. The ;
sealant to be used is site approved for .t.is application.
Summary I. No. The installation of a leak sealing device on the N36 piping will )
have no adverse effect on the continued operation of the system. l The worst case scenario for failure of the leak sealing device is complete weld failure of the 2" line. This failure is already bounded by the USAR evaluation of a Circulating Water system expansion joint failure in the Turbine Building, the loss of feedwater heating accident, and a feedwater line break outside containment. Therefore, the leak seal device installation cannot ;
affect the probability or consequences of any accident or malfunction of equipment previously evaluated in the USAR.
II. No. The device will have no detrimental effect on the operation of the N36 system. As detailed above, the results of the worst case failure of the weld are bounded by existing USAR evaluations.
Therefore, no new accident or malfunction types can be created by this installation.
III. No. The installation of this leak seal device will fully maintain the function of the affected N36 piping. If the weld were to fail, the failure would be bounded by existing USAR evaluations. Therefore, no margin of safety has been reduced.
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SE No.: 95-0057 Source Document: USAR Change Request 95-035 Description of Change f
l This change request updates USAR Section 13.1 to reflect the reassignment !
of contract administrator duties from the Maintenance Section to the '
Outage, Projects and Cost Section. No functions or responsibilities as described in the USAR have been eliminated. The background :
qualifications for the section managers r..aains consistent with the :
requirements of ANSI N18.1-1971.
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Summary l
I. No. No functions or activities as described in the USAR have been {
eliminated. The reassignment of functions or activities does not !
impact the design, function or operation of equipment important to '
safety. Accident analysis have not been altered. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The realignment of functions does not alter equipment important to ,
safety. No functions or responsibilities as described in the USAR. I have been eliminated. The background qualifications for the section ;
managers remains consistent with the requirements of l ANSI N18.1-1971. Therefore, the possibility of an accident or !
malfunction of a different type than any previously evaluated in the i USAR is not created. !
III. No. The realignment of functions is consistent with Section 6.0 of the l Technical Specifications. No functions or responsibilities as described in the USAR have been eliminated. The background qualifications for the section managers remains consistent with the requirements of ANSI N18.1-1971. Therefore, no margin of safety has been changed.
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SE No.: 95-0058, 96-0080 Source Document: SOI-E22A, Rev. 5 Description of Change This system operating instruction incorporates the use of High Pressure Core Spray (HPCS) alternate keepfill methods. One of two methods is used: either the HPCS suction is aligned to the condensate Storage Tank (CST), or a hose is connected between IE22-F031 and IP21-F893 and both lE22-F031 and IP21-F893 are opened.
Summary I. No. The source of keepfill (the water leg pump or one of the alternate keepfill methods) will not affect the ability to start the HPCS pump. None of the analyzed USAR accidents / transients in which the HPCS pump starts will be affected. Therefore, alternate keepfill will not alter the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
l II. No. An annunciator is available when using alternate keepfill to ensure the HPCS discharge line remains filled with water. Alternate keepfill will not alter HPCS injection flows, and alternate keepfill pressure is within the normal bounds on HPCS system pressure.
Therefore, alternate keepfill does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
III. No. Technical Specifications require the HPCS system to be maintained filled with water to ensure the minimum possible injection time and to minimize the potential for water hammer. Alternate keepfill will ensure these requirements remain satisfied. Therefore, the margin of safety as defined in the bases for the Technical Specifications are not impacted.
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3 SE No.: 95-0059 Source Document: DCP 93-0098, Rev. O Description of Change This design change installs a flow orifice in the suction line of the Reactor Core Isolation Cooling (E51) system waterleg pump 1E51-C003. The modification also adds test taps in the pump suction and discharge lines 4
to 1E51-C003.
y Summary !
I. No. The modification does not affect the E51 system control logic and i does not affect the automatic system response to an accident. The i modification is designed per ASME Code requirements. The modification does not affect the performance requirements of the ;
waterleg pump to provide keepfill for the E51 system. The E51 i system performance remains unchanged. Therefore, tl.t probability of occurrence or the consequences associated with an accident or i malfunction of equipment has not changed. 1 II. No. The modification installs equipment associated with the testing of the 1E51-C003 waterleg pump. The testing will be performed during :
cold shutdown conditions. All components added are designed to meet ASME Code requirements which assure that the new design meets the requirements of the existing system. Therefore, the possibility of j an accident or malfunction of a different type than any previously '
evaluated in the USAR is not created.
III. No. The E51 system performance requirements remain unchanged. Technical Specification 6.8.4.a requires that engineered safety feature systems be monitored for leakage remains satisfied. The modification is designed per ASME Code requirements. Therefore, no margin of safety has been changed.
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1 SE No.: 95-0060 Source Document: DCP 93-0096, Rev. 0 Description of Change This design change installs a flow orifice in the suction line of the Low Pressure Core Spray (E21) and Residual Heat Removal A (E12A) waterleg pump lE21-C002. This modification also adds test taps in the pump suction and discharge lines.
Summary I. No. The modification does not affect E21 or E12A systems contr01 logic and does not affect the automatic system response to an accident.
The modification is designed per ASME Code requirements. The modification does not affect the performance requirements of the waterleg pump to provide keepfill for the E21 and E12A systems and to provide sealing water to Feedwater Leakage Control (FWLC) system.
E21 and E12A system performance remains unchanged. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The modification installs equipment associated with the testing of the lE21-C002 waterleg pump. The testing will be performed during cold shutdown conditions. All components added are designed to meet ASME Code requirements which assure that the new design meets the requirements of the existing system. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The E21 and E12A system performance requirements remain unchanged.
The Technical Specification requirements associated with FWLC remain I unchanged. Technical Specification 6.8.4.a requires that engineered safety feature systems be monitored for leakage remains satisfied.
The modification is designed per ASME Code requirements. Therefore, no margin of safety has been changed. l 1
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1 SE No.: 95-0061 Source Document: DCP 93-0182, Rev. 0 Description of Change i
This design change replaces the Post-Accident Monitoring Recorders 1D23-R0090A/B and 1D23-R0170A/B with Johnson /Yokogawa R1800 digital I recorders. The new recorders have math capabilities which are used to create new averaging channels for suppression pool, drywell, and containment average temperatures to support performance of the Plant Emergency Instolctions. The modification replaces only the recorder ,
portion of the Regulatory Guide 1.97 variable instrument loop and does not modify the sensors, cables, and power supplies currently feeding the existing recorders.
l Summary I I. No. The new recorders meet the original design specifications for I l
seismic, electrical and physical separation, and equipment qualification of the existing recorders. Failure of the recorder is not an initiator to any of the accidents described in the USAR. The modification does not degrade or prevent operator actions during an accident. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The new recorders meet the original design specifications for seismic, electrical and physical separation, and equipment qualification of the existing recorders. The potential for common ,
mode failure has been minimized by the design of the recorder !
hardware and software, the verification and validation of the software, the testing of the software, and the design and testing of {
the hardware to demonstrate its resistance to EMI/RFI. Therefore, !
the possibility of an accident or malfunction of a different type l than any previously evaluated in the USAR is not created. '
III. No. This modification does not impact Technical Specifications and associated bases with respect to the required number of post-accident channels operable, minimum channels operable, action statements, or the frequency of channel checks or calibrations. No margins of safety are impacted.
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1 SE No.: 95-0062 Source Document: Potential Issue Form (PIF) 95-0494 Description of Change This PIF analyzes the "use-as-is' disposition of maintaining valve IP22-F0645 and 4ts associated one-half inch diameter piping disconnected until the end of Refueling Outage 6.
Summary I. No. The Mixed ced Demineralizer and Distribution (P22) system is not related to any initiation factors associated with accidents or transients described in the USAR. Valve IP22-F0645 and its associated piping does not interface with any plant structures, systems, or components other than P22. The disconnected piping cannot damage any systems, structures, or components. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not been increased.
II. No. Valve IP22-F0645 and its associated piping provides an unused service connection to the P22 system, which does not interface with any other plant structures, systems, or components. The disconnected piping cannot damage any systems, structures, or components. Therefore, the possibility of an accident or a malfunction of equipment of a type different than previously evaluated is not created.
III. No. This PIF is related to an unused valve and piping of the P22 system.
The pressure boundary of the P22 system is maintained. Thus, the reliability, pressure integrity, and operability of the P22 system will not be adversely affected. Therefore, no margin of safety has been reduced.
SE No.: 95-0064 I
Source Document: DCN 4903 i Description of Change ;
This drawing change updates various plant drawings by revising incorrect valve symbology. ,
Summary i
I. No. There will be no physical changes to the plant associated with implementation of this drawing change. The changes involve system diagram drawing changes only. The changes were analyzed to be consistent with the original FSAR analysis assumptions. The design, material, and construction standards applicable to the Standby Liquid Control (SLC), High Pressure Core Spray (HPCS), and Low :
Pressure Core Spray (LPCS) systems are not modified and the inherent ;
reliability of the systems to perform required safety functions '
remains intact. Therefore, the probability of occurrence or the l consequences of an accident or malfunction of equipment important to )
safety previously evaluated in the USAR have not increased. i l
II. No. HPCS and LPCS systems are part of the injection network for '
emergency core cooling and are designed to provide protection against postulated loss-of-coolant accidents. The SLC system provides an alternate means of reactor shutdown, independent of the ,
Control Rod Drive (CRD) system. The drawing changes do not make any !
physical modification to the plant and are within the bounds of the original licensing basis. The reliability of the ,C, HPCS, and ,
LPCS systems to perform required safety-related .ctions will not be altered. The safety injection, core coo .and decay heat removal capability of these systems are ur .ed . Therefore, the possibility for an accident or malfunctior m equipment of a different type previously evaluated in the USAR will not be created.
III. No. There are no physical changes to the plant. Existing Technical i Specifications will be unchanged. Therefore, no margins of safety have been reduced.
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SE No.: 95-0065
) Source Document: PAP-0101, Rev. 8, TC-1 PAP-1910, Rev. 3, TC-3 PAP-1914, Rev. 5, TC-3 Description of Change These procedure changes evaluate changes to the Fire Protection Program with respect to organization and administration. Specifically, PAP-1910 is being listed as the main controlling document of the Fire Protection Program. All other implementing procedures and instructions are sub-tier to PAP-1910.
Summary l I. No. These changes update and modify the USAR and implementing procedures
! to reflect organizational changes to the administrative aspects of the Fire Protection Program. No functions or activities described
- were eliminated. This change did not introduce any new fire hazards
' to the plant or create any new or modify any existing safe shutdown circuits or equipment. The accidents listed in the USAR and their associated analyses were not affected. All fire protection and safe shutdown systems and equipment will still operate as previously
] analyzed. Therefore, this change does not increase the probability l or the consequences of an accident or of a malfunction of equipment
- important to safety previously evaluated in the USAR.
II. No. The required operating modes and functions of any system important
! to safety or radiological dose mitigation as they relate to safe l shutdown in event of a fire are not changed. Operation of fire l systems in response to a fire are not affected. Therefore, this
- change does not create the possibility of an accident or a i
malfunction of equipment important to safety of a different type j than any previously evaluated in the USAR.
! III. No. Only administrative and audit aspects of fire protection are contained within the Technical Specifications. These consist of review and audit responsibilities, the need for administrative !
procedures, and the associated reoorting requirements. This change !
is consistent with Section 6.0 of the Technical Specifications.
Changes to the Fire Protection Program will not affect any other j system contained in the Technical Specifications. Therefore, no ;
margin of safety has been reduced. '
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SE No.: 95-0066 Source'Documen: USAR Change Request 94-037 Description of Change This change request clarifies the current use and application of the Control Room Isolation Status Panel to ensure that it is consistent with the Containment Isolation system design bases. In addition, clarifying information associated with the Hydrogen Igniter control circuitry was also addressed.
Summary I. No. The information presented on the isolation status panel does not affect the operation of the Containment Isolation system, nor does it impact the design bases of the Containment Isolation system. The isolation status panel is only an operator aid for quick reference of isolation valve position status. The clarifice. tion of the use of the status panel is an administrative action and cannot impact the USAR accident analyses. The added clarification for the method of Hydrogen Igniter operation does not change the design bases of the Hydrogen Control system described in the USAR. The control logic will still support the required mitigation of a degraded core accident described in USAR Sections 6.2.8.1 and 6.2.8.3. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
l II. No. The changes made to USAR Section 6.2.4.2.1 and Table 6.2.3.2 were made to provide specific detail on the current use and design application of the Coitrol Room Isolation Status Panel. The changes
- were found to not affect current containment isolation design bases.
t In addition, the amended wording for the Hydrogen Igniter control logic is for clarification and does not change hydrogen control i
design bases requirements reflected in USAR Section 6.2.8.1.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not i
created.
I III. No. The changes were consistent with the design bases of the Containment i
Isolation sy tem and the Hydrogen Control system. As a result, no margin of safety has been reduced.
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SE No.: 95-0067 Source Document: SMRF 95-5029, Rev. O Description of Change This design change modifies the underground Fire Protection (P54) system yard main piping by installing a new sectional control valve in the 8' line west of the Diesel Generator Building. The design will utilize mechanical joint pipe fittings, which will change the piping Line Specification as presently identified on USAR Figure 9.1. The new fittings provide flexibility for pipe movement / alignment in place of the current run of Yoloy welded piping in the line. The capacity of the underground mains are unchanged. Material meets the requirements of the design parameters, i.e., pressure, temperature, etc., of this fire protection piping and NFPA Standard 24, In tallation of Private Fire Service mains and Their Appurtenances.
Summary I. No. This modification does not change the fire hazard in any plant area.
The Fire Protection system will maintain its current design requirements as defined wi,hin the USAR. In the event of a fire, the ability to perform safe shutdown functions will be maintained by the Fire Protection system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The design change only impacts the Fire Protection system and is not functionally related to any known accident mode for plant features important to safety. The Fire Protection system will maintain its current design as defined within the USAR. Therefore, the implementation of this design change cannot create any new accident or malfunction of equipment beyond those previously postulated within the USAR.
III. No. The consequences of a fire are unaffected by this change since the Fire Protection system will maintain its current design as defined within the USAR. Therefore, no margin of safety has been reduced.
i j SE No.: 95-0068 Source Document: DCN 4889 Description of Change l
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, This drawing change updates various plant one-line electrical drawings to l l better reflect the as-built configuration of the distribution )
transformers 1R25-S0025, -S0027, -S0033, and -S0035. The changes are editorial in nature and include typographical error correction, and additional information of the transformer input circuit breakers and input cable size. ,
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. No. Implementation of this change updates various electrical one-line drawings. The changes are editorial. No hardware, device, circuit, etc., of the above transformers will be physically altered. The system operability, capability and reliability will remain unchanged. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. Implementation of this change updates various electrical one-line drawings. These changes are editorial and do not affect the i operation of any equipment required for plant safe shutdown.
Therefore, this change will not create the possibility of an I accident or malfunction of equipment of a different type than any previously evaluated.
III. No. This drawing change is only an editorial change. Operability and j reliability of the transformers (1R25-S0025, -S0027, -S0033, -S0035) '
will remain unchanged. Technical Specifications are not affected. i Therefore, this change will not reduce any margins of safety. l l
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SE No.: 95-0069 S_ource Document: USAR Change Request 95-046 Description of Change l
This change request evaluates the practice of opening the drywell !
personnel airlock shield doors during plant startup and shutdown. The i doors are susper.ded from a monorail structure su;. orted from a platform I at 620' elevacion in containment. Analyses indicate that with the shield doors open during Operational Modes 1, 2, and 3, several members of the 620' elevation platform and the monorail structure could exceed design stress allowables. This evaluation resulted in a 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.: 95-0070 Source Document: Potential Issue Form (PIF) 95-1045 Description of Chance This PIF analyzes the 'use-as-is' disposition of improperly sized filter elements 1C11-D0010A/B at the suction of the Control Rod Drive (CRD) water pumps 1C11-C0001A/B.
Summary I. No. There will be no physical changes to the plant due to the i
'use-as-is' disposition of this PIF. The changes are analyzed to be consistent with the original plant construction. The design, material, and construction standards applicable to the CRD system are not modified and the inherent reliability of the system to
- perform required safety functions remains intact. Therefore, the .
probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR have not been changed.
II. No. The changes evaluated by this PIF do not make any phyrical modifications to the plant and are within the bounds of the original licensing basis. The reliability of the CRD system to perform required safety-related functions will not be altered. The safety injection, core cooling, and decay heat removal capability of other safety-related systems are unaffected. Therefore, the possibility i
for an accident or malfunction of equipment of a different type previously evaluated in the USAR will not be created.
i III. No. Therc are no physical changes to the plant which will result from i tbc implementation of the PIF. Existing Technical Specifications will be unchanged. Therefore, no margins of safety have been
- reduced.
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SE No.: 95-0071, 96-0004 l j Source Document: DCP 95-5044, Rev. 0 l 1
l Description of Change
! This design change permanently removes several cards in the Steam Bypass !
and Pressure Regulation (CBS) system and the Turbine control (N32) system ;
that deal with the load demand error signal which is not used at Perry.
1 Summary j i
! I No. The cards have no interface with equipment important to safety. The
! _ unused cards develop a signal for use in the Reactor 4 Recirculation (B33) system's automatic load following mode; however, i i this mode of operation is not, nor ever will be, used at Perry. '
l Card removal eliminates any possibility of an unnecessary pressure ;
i regulator trip due to an inadvertent fault detection from load !
I demand error. Therefore, the probability of occurrence or the l consequences associated with an accident or malfunction of equipment <
has not changed.
! II. No. The cards have no interface with the equipment important to safety.
' Removal does not affect the fault detection circuitry for the flow l
' demand signal or bypass valve demand signal since only the cards !
associated with load demand error are removed. Removal does not l j
cause or contribute to the occurrence of an accident or malfunction
- of equipment of a different type than previously evaluated in the USAR.
III. No. The cards have no interface with the equipment important to safety.
The safety portions, bypass valve control, and turbine overspeed protection, reside on electronic cards separate from the unused cards. There is no overlap of function. Thus, removal of the unused cards does not reduce any margin of safety.
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SE No.: 95-0072 Source Document: DCP 94-0111, Rev. O Description of Change This design change permanently installs 80 strap-on Resistance Temperature Devices (RTD) to Residual Heat Removal (RHR) and Emergency Service Water (ESW) system piping. The RTDs are only used for performance testing.
Summary I. No. The installation is nonsafety-related and the RTD outputs have no interface with equipment important to safety. The RTDs are only utilized to monitor RHR heat exchanger performance testing. There are no operational control functions, alarms, interlock functions or indications performed by this instrumentation. The RTDs are mounted (strapped) to existing RHR and ESW piping and the armor cable is routed directly to various termination points in the RHR rooms. No piping boundaries or structures are breached by this installation.
The cables are terminated on fabricated supports located inside the web of structural columns, one platform, and on non-removable handrails. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
I7. No. The installation is nonsafety-related and the RTD outputs have no interface with equipment important to safety. The RTDs are only utilized to monitor RHR heat exchanger performance testing. There are no operational control functions, alarms, interlock functions, or indications performed by this instrumentation. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The installation is nonsafety-related and the RTD outputs have no interface with equipment important to safety. The RTDs are only utilized to monitor RHR heat exchanger performance testing. There are no operational control functions, alarms, interlock functions, or indications performed by this instrumentation. This installation has no active function. Therefore, no margin of safety has been changed.
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SE No.: 95-0073 i Source Document: DCP 93-0095, Rev. 0 l
l Description of Change This design change installs a flow element in the suction line of the Residual Heat Removal (RHR-E12) system water leg pump 1E12-C003.
Additionally, test taps will be added in the pump's suction and discharge lines to improve performance testing. This water leg pump provides keepfill to RHR Loops B and C.
Summary I. No. This design change does not affect the E12 system control logic and does not affect the automatic s'/ stem response to an accident. The modification does not affect t).e performance requirements of the waterleg pump to provide keepfill for the E12B and E12C systems and to provide sealing water to Feedwater Leakage Control (FWLC). The E12B and E12C system performance remains unchanged, the modification is designed per code requirements and components added to the Engineered Safety Features (ESP) boundaries will be monitored for component and overall system leakage through the inservice test progr.2. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The modification installs equipment associated with the testing of the 1E12-C003 waterleg pump. If the components added by this modification did fail, the redundant Emergency Core Cooling Systems (ECCS) are available to provide accident mitigation as required.
All safety-related components added are designed to meet ASME Code requirements which assure that the new design meets the requirements of the existing system. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The E12B and E12C system performance requirements remain unchanged, therefore the Technical Specification safety margin is maintained.
Adherence to the ASME Code ensures that the margin of safety associated with Technical Specification 4.0.5 is maintained. The Technical Specification requirements associated with FWLC remain unchanged. Technical Specification 6.8.4.a requires that ESF systems be monitored for leakage will be performed for the components added in this modification. Therefore, no margin of safety has been changed.
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SE No.: 95-0074 Source Document: USAR Change Request 95-052 Description of Change This change request revises USAR Section 3.1.2.6.2.1.2 to state the Fuel Pool Cooling and Cleanup (FPCC) system will be "normally operated" to agree with other sections in the USAR.
Summary I. No. This change to the USAR will have no impact on the design of the Fuel Pool Cooling and Cleanup system. The operational requirements on the system will change to permit the system to be secured for maintenance. Administrative controls will ensure the pool temperatures and level are maintained to the Technical Specification criteria, Based on a review of the accidents analyzed in USAR Chapter 15 and the designed function of the system, the FPCC system does not act as an accident initiator or mitigator. The change to the USAR will not result in the FPCC system becoming or influencing an initiator, or mitigator of any accident or malfunction of safety equipment analyzed in Chapter 15 of the USAR. Therefore, the probability of an accident of malfunction of equipment as previously evaluated in the USAR is not increased.
II. No. The change to the USAR will permit the FPCC system to be secured for maintenance. Once secured, if the FPCC system cannot be made o
J available to satisfy functional requirements of maintaining pool water inventory and temperature, the Residual Heat Removal (RHR) system will provide augmented fuel pool cooling. The RHR system can be aligned for the Fuel Pool Cooling Assist mode within one shift.
Therefore, the change to the USAR does not create the possibility of an accident or a malfunction of equipment of a different type than those previously evaluated in the USAR.
III. No. The FPCC system does not assist in the safe shutdown of the plant and is not included in any Technical Specification basis. The shutdown of the FPCC system will not result in temperatures greater than Technical Specification or USAR limits because the RHR system can provide augmented fuel pool cooling. Therefore, this change will not reduce the margin of safety as defined in any bases for any of the Technical Specifications.
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SE No.: 95-0075 {
Source Document: Physical Security Plan, Rev. 20 i Description of Change l
This evaluation analyzes changes made to the Physical Security Plan '
(PSP). The changes have been evaluat.ed to ensure that the effectiveness of the Perry Nuclear Power Plant Security Plan has not been reduced and i to ensure that the requirements of 10CFR73, Physical Protection of Plants i and Materials, are met. Site Protection must be contacted for further !
details since this is considered " SAFEGUARDS' information.
1 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 of a different type.
III. No. The PSP changes do not reduce any margin of safety.
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SE No.: 95-0076 Source Document: DCN 4966
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Description of Change This drawing change revises plant drawings by indicating an increase in the valve stroke times for the Residual Heat Removal (RHR) minimum flow l bypass valves 1E12-F0064A,B,C from 14 seconds to 15 seconds.
Summary I. No. Analysis showed that the increase in the valve stroke time would not increase the potential or severity of an accident. The review determined that the design function of the valves was to provide containment isolation, pump protection, and system pressure boundary l protection. None of these are impacted by the change. The review I also considered impact on ECCS LOCA analysis and found that there was a negligible impact. The change has no impact on radiological-consequences of an evaluated accident as specific stroke times for these valves were not an input into any radiological consequences. !
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. Analysis showed that no new equipment types or new system interactions are created. The original plant design basis is maintained. There are no hardware changes to the plant.
Performance of the affected system is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type that any previously evaluated.
l III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. Although the stroke time for the subject valves is increased, the capability of the RHR system to provide rated flow as required by the ECCS/LOCA analysis is not compromised. This change results in no effect on the output /results of the LOCA analysis. The containment isolation function of these valves are not affected. Therefore, no reduction of safety margin is caused by this change.
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SE No.: 95-0077 Source Document: DCN 5004 Description of Change This drawing change revises plant drawings by indicating an increase in the allowable stroke times for the Reactor Core Isolation Cooling (RCIC) minimum flow valve lE51-F0019 from 5 to 8 seconds, and the first )
isolation valves for the Inboard MSIV Leakage Control (E32) system l lE32-F001A,E,J,N from 20 to 22 seconds. I Summary I. No. Analysis showed that the increase in valve stroke time would not increase the potential or severity of an accident. The RCIC valve was reviewed for impact on flow requirements, pump protection impact, and containment isolation impact. No adverse impacts were noted. The E32 valves were reviewed for impact on system operation, system interfaces, and containment isolation. Again, no impacts were noted. The change has no impact on radiological consequences of an evaluated accident as specific stroke times for these valves ,
were not an input into any radiological consequences. Therefore, I neither the probability of occurrence nor the consequences of an I accident or malfunction of equipment will be in aased by this change.
II. No. Analysis showed no new equipment types or new system interactions are created. The original plant design basis is maintained. There are no hardware changes to the plant. Performance of the affected systems is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke times for the affected valves'does not change or violate any system design basis. The change results in no effect on the output /results of the LOCA analysis. Closure circuitry and isolation functions of these valves are not affected
'by the change. Therefore, no reduction in any safety margin is ,
caused by this change. !
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SE No.: 95-0078 Source Document: DCN 4940 Description of Change This drawing change revises plant drawings by indicating an increase in the stroke times for the minimum flow valves for the Low Pressure Core Spray (LPCS) system 1E21-F00ll from 22 to 23.5 seconds and the High Pressure Core Spray (HPCS) system 1E22-F0012 from 7 to 8 seconds.
Summary I. No. Analysis showed that the increase in valve stroke time would not increase the potential severity of an accident. The evaluation reviewed the valves' function for containment isolation, pressure integrity, pump protection, and capability to mitigate a design basis LOCA. It was determined that the change did not impact containment isolation, pump protection, LOCA mitigation, or pressure integrity. Present USAR accident analysis does not explicitly model closure times of containment isolation valves, nor were they re<:uired to be modeled per the guidelines of Standard Review Plan 15.6.5. This change does not increase the probability of radioactive release since the probability of a HPCS/LPCS pipe break or component failure is not increased. 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. Analpis showed that no new equipment types or new system interactions are created. The original plant design basis is maintained. There are no hardware changes to the plant.
Performance of the affected systems is not altered as a result of this change. Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke time for the affected valves does not change or violate any system design basis. Although the min.
flow valve stroke times are increased, the capability of the LPCS and HPCS systems to provide rated flow and pump protection as required is not compromised. These valve will continue to automatically close to provide containment isolation when necessary.
Therefore, no reduction of safety margin is caused by this change.
SE No.: 95-0079 Source Document: NR 95WS-091, Rev. 0 Description of Change This nonconformance report analyzes the "use-as-is' disposition of Class A fill material fail'ing to meet the gradation requirements for the material set forth by Installation Standard Specification SP-2154.
Review of the Foundation Design Report (Structural Calculation 1:29.1.11) has shown that deviation of approximately 2% in the fines is considered a minor deviation, and will not affect the overall acceptance of the Class A fill material.
Summary I. No. The acceptance of the slightly lower value for the material gradation results for the Class A fill is justified since the material continues to meet the design, material and construction standards established in the Foundation Design Report. The resulting backfill will continue to fulfill all of its functions of filtering subsoil and Class B fines, providing a drainage medium and providing adequate load bearing capacities. Therefore, there is no increase in the probability of occurrence or the consequences of an ,
accident or malfunction of equipment important to safety previously ;
evaluated. '
II. No. The fill in question is designed to perform specific functions and :
the Foundation Design Report has confirmed that the properties which <
allow those functions to be performed are not changed. Based on l that fact, the fill material remains adequate to perform said functions. Therefore, the possibility for an accident or malfunction of a different type than previously evaluated is not created.
III. No. This analysis confirms the adequacy of the subject fill material and ensures that all original design functions continue to be met or exceeded. Since the Foundation Design Report values bound the range of values listed in the Specification, the acceptance of the slightly lower value does not affect any established safety margin.
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SE No.: 95-0080 Source Document: DCN 5010 e
Description of Change This drawing change revises various B-022 series drawings due to transcription errors made when transferring gamma radiation doses / dose rates to the tables on these drawings from the base calculations.
Summary I. No. This change had no effect upon previously qualified equipment since the previously used radiation qualification environments were conservatively calculated. No physical changes occurred to the plant, to previously defined accidents, or to equipment important to safety. These changes were determined to be editorial in nature and were concluded not to result in an increase to any probabilities or consequences of any previously analyzed accident or malfunction of equipment important to safety.
II. No. This change represents an editorial change to the B-022 drawings to establish conformance to the calculations which are used as the basis for the respective drawings. No physical changes to the plant have occurred. Therefore, this change will not result in any new accidents or malfunctions of equipment important to safety.
III. No. This drawing change is considered an editorial revision. There is no increase in any parameter for which components located in the ,
areas in question were designed to withstand. Therefore, no margin of safety has been reduced.
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SE No.: 95-0081 l Source Document: DCN 4984 1
l Description of Change This drawing change revises plant drawings by indicating an increase in the maximum allowable stroke time for the Reactor Water Cleanup (RWCU) containment isolation valves. The valves and their new stroke times are as follows: RWCU pump suction inboard and outboard containment isolation valves to 20 seconds (lG33-F0001 and 1G33-F0004), RWCU pump discharge inboard and outboard containment isolation valves to 15.5 seconds (lG33-F0053 and 1G33-F0054), RWCU return to feedwater inboard and outboard containment isolation valves to 20 seconds (lG33-F0040 and 1G33-F0039), and the RWCU blowdown to condenser inboard and outboard containment isolation valves to 20 seconds (lG33-F0028 and 1G33-F0034).
Summary I. No. Analysis showed that the increase in the valve stroke times would not increase the potential or severity of an accident. Present USAR accident analysis does not explicitly model closure times of containment isolation valves, nor were they required to be modeled per the guidelines of Standard Review Plan 15.6.5 for radiological consequences. The change was reviewed for impact in the Steam Tunnel, containment, RWCU/RCIC pump rooms, Auxiliary Building hallway, as well as impact on the RWCU system and the Standby Liquid Control system. The review concluded no impact on present design basis. 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. Analysis showed that no new equipment types or new system interactions are created. The original design basis is maintained.
There are no hardware changes to the plant. Performance of the affected system is not altered as a result of this change.
Therefore, this change will not create the possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The new maximum allowable stroke times for the affected valves does not change or violate any system design basis. All valves continue to automatically close to provide containment isolation and prevent radiological effects from exceeding the guidelines specified by 10CFR100. The change does not change the way the valve performs and safety function of the valves is not changed. Therefore, no reduction of safety margin is caused by this change.
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i SE No.: 95-0082 Source Document: DCN 4987 Description of Change This drawing change documents the permanent installation of two !
Mechanical Foreign Items (MFI) placed on the Standby Liquid Control (SLC) system that facilitate monitoring and controlling of SLC system pressure between the storage tank isolation valves and the pumps.
Summary I. No. This drawing change will not alter the control, operation or expected response of the SLC system. The changes cannot creste an i environment that will be detrimental to the SLC system or any system l it interfaces with. The changes will not alter, degrade or prevent I actions described or assumed to occur as discussed in the USAR. The l changes will not alter any assumptions previously made in evaluating the radiological consequences that are described in the USAR. '
Therefore, neither the probability of occurrence nor the a
consequences of a previously analyzed accident or malfunction of equipment important to safety will be increased.
II. No. The changes will have no adverse affects on any systems, structures <
or components, or the way they will react to normal and abnormal transients. The changes cannot result in any new equipment '
failures, no new initiators, or contributors for any new event.
Therefore, this drawing change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The changes to the SLC system will not degrade its capability to mitigate the effects of postulated transients. Thus, there will be no reduction in the margins of safety provided by Technical !
Specification 3.1.5 and its bases.
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SE No.: 95-0083 Source Document: DCN 4988 Description of Change This drawing change documents the permanent installation of the Betz Chemical Addition system, which had been previously installed under the Mechanical Foreign Items (MFI) Program. The Betz system was installed to provide a more reliable and effective means of controlling Circulating Water (101) system water chemistry. The Betz system is nonsafety-related l and located entirely in the Circulating Water Pump House. The building l is nonsafety-related, non-seismic, and as such all equipment located !
within it is nonsafety-related.
Summary I. No. There are no accidents evaluated in USAR Chapter 15 that can be ,
affected by the inclusion of the Betz system as permanent plant '
equipment. The chemical being utilized has been evaluated and found compatible v;th the Circulating Water system. Failure of the Betz chemical storage tank has been evaluated and found to have no adverse affect on Control Room habitability. The conclusions presented in USAR Sections 2.2.3 and 6.4 are not altered. The Betz system will not change, degrade, or prevent actions described or assumed in any accidents described in the USAR. The Betz system has no interface with structures, systems or components important to safety. The permanent installation does not increase the dose to the public nor will it increase onsite doses that would impede actions necessary to mitigate the consequences of a malfunction of equipment important to safety. Therefore, neither the probability of occurrence nor the consequences of previously analyzed accident or malfunction of equipment important to safety will be increased.
II. No. The introduction of the Betz system, including 1000 gallons of phosphonic acid, has been shown to not impact equipment that is important to safety. No system, safety or nonsafety-related, will be adversely impacted by the Betz system. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. Neither the original Acid Addition (P83) system nor the 101 system are addressed in the Technical Specifications. The Betz System has been shown to not impact equipment important to safety. Failure of the Betz chemical storage tank has been evaluated and will not adversely impact Control Room habitability. Therefore, no margins of safety have been reduced.
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- SE No.: 95-0084 Source Document: PTI-P72-P0002, Rev. 3 Description of Change
- This periodic test instruction evaluates the methodology used to quantify l t
the groundwater inflow into the Underdrain (P72) system. This instruction was revised to measure ground water level using the existing 1
. piezometers at the beginning and the end of the test to help quantify '
4 ground water inflow.
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Summary i
i I. No. The Underdrain system is used to mitigate the consequences of the i accidents described in USAR Sections 2.4.13.5.2 and 10.4.5.3.1 only l l and has no affects on the accident initiators. The gravity drain j
- portion of the Underdrain system is designed to handle the volume of i water generated by the events. The gravity drain portion of the
! underdrain is no way altered or effected by this instruction
- revision. USAR Chapter 15.7.3 addresses Postulated Radioactive i Releases Due To Liquid-Containing Tank Failures. In this accident, l the dedicated P72 radiation monitors trip off the P72 pumps or the .
! pumps are manually tripped, using dedicated disconnects. In either !
4 case, operation of the pumps is stopped. This instruction revision I does not impact this mitigation activity. Therefore, the !
4 probability of occurrence or the consequences associated with an ,
i accident or malfunction of equipment has not changed.
- II. No. This instruction revision is limited to minor methodology changes that make no changes to the design or operation of the P72 system.
Accident analysis is not impacted. Therefore, the possibility of an
! accident or malfunction of a different type than any previously l evaluated in the USAR is not created.
] III. No. The Underdrain system is not addressed in the Technical
- Specifications. This instruction revision doe not alter the design i
or operation of the P72 system. Therefore, no margin of safety has been changed.
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SE No.: 95-0085 Source Document: DCP 95-0031, Rev. O Det.cription of Change This design change revises control wiring to delete the leakage treatment function of the inboard leakage control system of the containment airlock door. This is being done in order to avoid creation of a flow path from containment whenever an upstream inflated seal is not in place, e.g.,
during door maintenance. The outer airlock door leakage control system valve line-up was revised to increase the availability of the system and avoid susceptibility to single active failure.
Summary I. No. The airlock leakage control lines are an accident mitigation system and are not associated with initiation of accidents. The changes to the controls and valve positions do not have any adverse effect on any interfacing components. T) account for all operating modes, airlock door connections are assigned as bypass flow paths as a part of this change. No change is made to the allowable containment total leakage and bypass leakage limits. The inflatable seals on the airlock doors are conservatively tested every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The history of this surveillance testing has shown that the inflatable seals are providing a much tighter leakage barrier than originally anticipated in the system design. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change does not have potential to impact the basic functions of l the airlock or interfacing systems. Thus, a different type of accident / malfunction is not created.
III. No. The margin of safety is not reduced since Technical Specification and bases address the allowable leakage limits and general containment isolation criteria. As stated above, this criteria has not been revised.
SE No.: 95-0086, 96-0019 Source Document: SMRF 95-6041, Rev. O and Rev. 1 Description of Change This design change modifies Residual Heat Removal (RHR-E12) system check valve lE12-F0046B by removing its internals. The internals are being removed because the check valve function is not required.
Summary I. No. The check valve lE12-F0046B is not required to check flow during normal operation. The purpose of IE12-F0046B is to prevent hydraulic interaction between RHR Loops B and C when the loops are piped with a common return line. At Perry, the loops are not piped with a common return line. Other lines which might cause backflow through this line with the check valve internals removed have been evaluated. The only source of backflow is during the steam condensing mode of operation of RHR, during which a small backflow through the RHR pump could be experienced. This backflow would have no detrimental 'ffect on the pump. This evaluation is supported by GE letter RTR h03, dated 1/22/91. Check valve lE12-F0046B is capable of failure in such a manner as to impede flow through the line. By removing the internals of lE12-F0046B this failure mode is removed, and overall system reliability is removed. No credit is taken in any of the accident analyses for this check valve. It does not perform a containment isolation function. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Removal of the internals of check valve lE12-F0046B does not involve an initiator or failure not considered in the USAR. This change has no effect on operating procedures. System functions are maintained.
The existing piping maintains the original system design margins and the original system functions. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. There are no changes to any operating procedures made as a result of the proposed changes. No procedures or instructions governed by the Technical Specifications are adversely impacted. The operation of the RHR system is unaffected. Therefore, no margin of safety has been changed.
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SE No.: 95-0087 Source Document: USAR Change Request 95-0058 j Description of Change This change request revises USAR Sections 1.8, 2.3, 7.1 and 7.3 to clarify the location and description of the onsite meteorological measurements plan. This change also corrects the description of the plant's seismic recorders.
Summary i
I. No. There are no equipment changes or modifications to the plant as'a i result of this USAR change. The Meteorological system and Seismic !
systems cannot be the initiator of any accident. Therefore, the '
probability of occurrence or the consequences associated with an I accident or malfunction of equipment has not changed.
l II. No. The Meteorological and Seismic systems have no interface to any system, structure, or component that is important to safety. These !
systems were not changed from their installed design. Therefore, j the possibility of an accident or malfunction of a different type i than any previously evaluated in the USAR is not created.
l III. No. The Technical Specifications and bases for the Meteorological and l Seismic systems is unchanged by this change request. There are no !
equipment changes or modifications to the plant as a result of this change. The ability to evaluate the need to initiate protective measures to protect the health and safety of the public is not changed. Therefore, no margin of safety has been changed.
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i SE No.: 95-0088 i 1 Source Document: USAR Change Request 95-016 I Description of Change a
This USAR change clarifies / corrects three USAR sections / figures. I Section 6.3.2.2.1 is being made consistent with other USAR sections with respect to the closure of containment isolation valve lE22-F023.
Figure 6.3-69 is being corrected to include a reference line that was i i inadvertently omitted during its development. Section 3.9.6 is being !
l clarified with respect to the actual code being used.
a l Summary I. No. These change are administrative. The design or operation of the plant has not been impacted. No design analysis has been altered. '
1 i NRC regulations remain satisfied. Accident analysis remain unchanged. Therefore, the probability of occurrence or the
- consequences associated with an accident or malfunction of equipment
} has not changed. t II. No. These change are administrative. The design or operation of the !
4 plant has not been impacted. No design analysis has been altered. >
j' NRC regulations remain satisfied. Accident analysis remain unchanged. Therefore, the possibility of an accident or malfunction !
- of a different type than any previously evaluated in the USAR is not
- created. i l
i III. No. These change are administrative. The design or operation of the !
plant has not been impacted. No design analysis has been altered.
NRC regulations remain satisfied. Accident analysis remain j unchanged. Therefore, no margin of safety has been changed.
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SE No.: 95-0089 '
Source Document: DCP 93-0096, Rev. 1 Description of Change This design change installs a flow orifice in the suction line of the Low i Pressure Core Spray (E21) and Residual Heat Removal A (E12A) waterleg !
pump 1E21-C002. This modification also adds test taps in the pump )
suction and discharge lines.
Summary I. No. The modification does not affect the E21 and E12A systems control logic and does not affect the automatic rystem response to an accident. The modification is designec er ASME Code requirements.
The modification does not affect the pt tormance requirements of the waterleg pump to provide keepfill for the E21 and E12A systems and l to provide sealing water to the Feedwater Leakage Control (FWLC) '
system. E21 and E12A system performance remains unchanged. The .
componentr being added to the Engineered Safety Features (ESP) l boundaries will be monitored for component and overall system i leakage through the inservice test program. Therefore, the t probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
4 II. No. The modification installs equipment associated with the testing of the 1E21-C002 waterleg pump. If the components added by this i modification did fail, the redundant Emergency Core Cooling Systems (ECCS) are available to provide accident mitigation as required. All safety-related components added are designed to meet ASME Code requirements which assure that the new design meets the requirements of the existing system. Therefore, the possibility of an accident-or malfunction of a different type than any previously i evaluated in the USAR is not created.
III. No. The E21 and E12A system performance requirements remain unchanged.
Adherence to the ASME Code ensures that the margin of safety 4 associated with Technical Specification 4.0.5 is maintained. The !
Technical Specification requirements associated with the FWLC system '
remain unchanged. Technical Specification 6.8.4.a requires that ESF systens be monitored for leakage will be performed for the components added in this modification. Therefore, no margin of safety has been changed.
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SE No.: 95-0090 Source Document: DCN 4464 Description of Change 1
This drawing change corrects errors on P&ID D302-107 and Instrument Loop t Drawing 803-146, Sheet 15, Condensate Demineralizer. The change will make the drawings comply with the actual field conditions.
l Summary l
! I. No. Two demineralizer drawings are being revised. The actual field ,
( installation is correct and operates correctly. This activity does !
! not affect the Chapter 15 analysis, physically modify the plant, or its operating program. Therefore, the probability of occurrence or !
I the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This activity corrects drawing errors for the condensate ,
demineralizer bypass valve. The drawing change will not create any new accident scenarios. Therefore, the possibility of an accident i or malfunction of a different type than any previously evaluated in the USAR is not created.
i III. No. This activity is an change to system documentation and does not physically alter the system or its operation. Therefore, no margin of safety has been changed.
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SE No.: 95-0091 Source Document: SOI-Fil/15, Rev. 5 Description of Change This system operating instruction revision' evaluates the installation of
! sa jumper to defeat the Refueling Platform Bridge Reverse Motion Interlock Number 2. This instruction places administrative controls on the Reactor Mode Switch to prevent the plant from entering an operational condition in which the Reverse Motion Interlock 2 would be required.
1 Summary I. No. The jumper being installed has no affect on the red zone interlocks, core alteration interlocks, or winch / hoist interlocks. No equipment important to satety required by any plant procedure / instruction is affected by this jumper. Therefore, this action does not alter the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. Since the overlapping administrative control prevents the plant from entering the operational mode in which the interlock is required by the USAR, there is no opportunity created in which the jumpering of the reverse motion interlock number 2 (as performed by this instruction) could create the possibility of an accident of a ,
different type than evaluated in the USAR. Therefore, this action !
does not create the possibility of an accident or a malfunction of {
equipment important to safety of a different type than any previously evaluated in the USAR.
III. No. The Reverse Motion Interlock Number 2 is not associated with the Core Alterations interlocks (required by Technical Specification 3.9.6) or the One-Rod-Out interlock (Technical -
Specification 3.9.1). Therefore, all equipment required by '
Technical Specifications will remain operable while the jumper is installed. Therefore, no margin of safety is reduced.
SE No.: 95-0092 Source Document: OCP 96-0609, Rev. 6 Description of Change This design change removes abandoned piping and a 3' manual isolation valve from the Instrument Air' (P52) syste.n to facilitate the installation of a new Emergency Service Water (ESW) keepfill water supply line to be installed via Design Change Package (DCP) 94-0139.
Summary I. No. The 1PS2-F1021 valve is currently not functional nor has it ever been. Removal of this valve will not change the original FSAR analysis assumptions. This modification will ensure that the PS2 system will continue to meet its applicable design, material, and constructions standards. The performance of the Instrument Air system is not impacted. The removal of valve 1PS2-F1021 does not have any effect on the radiological consequences of any accident described within the USAR. There is no possible impact to plant equipment important to safety. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The removal of the valve will not impact the PS2 system design basis as dt;scrli~d in the USAR. USAR accident analysis will not be impacted. Therefore, the possibility of an accident or malfunction of a different type than any previously. evaluated in the USAR is not created.
III. No. The design or operation of the PS2 system is not impacted by this modification. USAR accident analysis is not affected. Therefore, no margin of safety has been changed.
SE No.: 95-0093 Source Document: SCRs 0-95-1002 through 0-95-1005 ELI-R24, Rev. 4, TC-12 VLI-M29, Rev. 4, TC-3 Description of Change These setpoint changes and instruction changes will increase the relative humidity of the Unit 1 Computer Room.
Summary I. No. USAR Chapter 9.4.12.2.9, Control and Computer Rooms Humidification System, describes the equipment and setpoints for both Unit 1 and 2.
Unit 2 Computer Room is not in service and is currently acting as a storage room. Unit 2 Computer Room has no supporting functions with the operating Unit 1 Computer Room and has, therefore, been placed in a capacity considered abandoned with no relevance to plant operations. In an effort to eliminate nuisance alarms, the Unit 1 Computer Room will raise its high humidity alarm and its humidity controller setpoint. These actions have been evaluated to be well within the design and environmental parameters of the room.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These changes alter the humidification alarms of the Units 1 and 2 Computer Rooms and increase the humidification setpoint for the Unit 1 humidity. The new setpoints will cover the expected humidity swings found in the Computer Room Ventilation system. The computer room humidification process does not affect any other controls that could be related to safety. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Increasing the alarm span for the computer room humidification control and increasing the humidity controller for the Unit 1 Computer Room, adequately supports the humidification process required for computer operation. The effects upon the Plant Computer operating criteria of the Technical Specifications are not affected by these changes. Therefore, no margin of safety has been changed.
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l SE No.: 95-0094 Source Document: USAR Change Request 95-059 l Description of Change !
This change request corrects an inconsistency in USAR Sections 9.5.9.2.2, l 8.3.1.1.2.6, and Table 8.3-1 relating to the start time of the :
Division III Emergency Service Water (ESW) pump following a High Pressure i Core Spray (HPCS) Diesel Generator (DG) Loss Of Offsite (LOOP)/ Loss Of !
Coolant Accident (LOCA) response. USAR Table 8.3-1 correctly states the i time delay for the Division III ESW to be 33 seconds. However, the !
description in USAR Sections 8.3.1.1.2.6 and 9.5.9.2.2 incorrectly '
reflect a 20 second time' delay. '
Summary I. No. This change clarifies the response time for the initiation of ESW l through the HPCS DG jacket cooling water system. The HPCS system, !
which is supported by the jacket water cooling system, is an accident mitigating system which is normally not in operation and is installed to respond to an emergency or accident condition. The l HPCS diesel generator is provided with a sufficient heat sink to '
allow a hot HPCS dissel engine to start and operate for at least !
2 minutes without Emergency Service Water flow through the jacket water flow through the jacket water cooling system heat exchanger. -
Hence, the HPCS DG will not be affected by increasing the time in ,
which the DG operates without ESW from 20 seconds to approximately j 46 seconds. Therefore, the probability of occurrence or the I consequences associated with an accident or malfunction of equipment !
has not changed.
II. No. There is no change to the performance of any component previously required to function following an accident. The resulting delay time is enveloped by the qualification of the component. Hence, the operability of the system is not compromised. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The HPCS DG was previously demonstrated that the DG is provided with a sufficient heat sink to allow a hot HPCS diesel engine to start and operate for 2 minutes without Emergency Service Water flow through the jacket water cooling system heat exchanger. Hence, increasing the time in which the DG operates without ESW from 20 seconds to approximately 46 seconds does not decrease the margin of safety since the resulting delay is still within the acceptance criteria of he component. Therefore, no margin of safety has been changed, l
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, SE No.: 95-0095 l Source Document: SCR 1-95-1355 !
Description of Change i
This setpoint change will adjust the turbine load limit with -
- recirculation run back from the present setpoint of 5.1'HgA to 5.6"HgA. !
l This change will permit the operation of the Main Condenser slightly !
above its nominal vacuum value of 5"HgA up to the new value of 5.5"HgA. !
4 Implementation of this setpoint change will enhance plant operations to l operate up to 100% rated power during extreme summer conditions, i a
- Summary I I. No. Raising the existing setpoint of the 5.l'HgA turbine lead limit setback with recirculation run back to 5.6'HgA for limited summer j conditions has no effect on the overall steam system performance.
] The Main Condenser will be permitted to operate slightly above the i'
nominal value of 5.9'HgA only to support 100% power operation as necessary during abnormal conditions. The setpoint change and new ;
! operating criteria (vacuum) for the Main Condenser is considered !
I nonsafety and has no impact whatsoever on the original design '
specifications such as seismic, separation, or environmental o
criteria. No new missiles are created due to this modification.
l The setpoint change ar.d new operating criteria (vacuum) for the Main l Condenser has no impact on radiological dose with respect to malfunction of equipment. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of l equipment hs not changed. I II. No. This change considered turbine generated missiles. In accordance with the Main Turbine manufacturer, General Electric, approval has been granted to operate the Main Condenser above the pre-described nominal value of 5.0'HgA for limited durations at high load levels.
GE has concluded that no new threat to the reactor, fuel, condenser, bypass system, safety systems, recirculation system, and MSIVs exists with this change. There are no related malfunctions with respect to the setpoint change or new operating conditions. This nonsafety setpoint is only related to an attempt to prevent a Main Turbine trip on a slow rising degraded Main Condenser. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specification 3/4.3.3.2 remains unchanged. There is no impact upon the design of the Main Condenser. No new missiles are created. Therefore, no margin of safety has been changed.
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SE No.: 95-0098 '
Source Document: DCN 5003 i
- Description of Change l
- This drawing change adds a note to drawing D219-101, Electrical Grounding .
1 Schematic, to include GE Specification 22A2736, "Special Wire and Cable", l and GE Installation Instruction 22A4185, " Power Generation Control l
Complex Installation' as references.
- Summary i ;
i I. No. The instrument grounding system is designed and installed to provide
! personnel safety and to minimize the probability of electromagnetic i interference (EMI) causing a malfunction within any instrumentation j
! and control system. Installation and performance of the system were reviewed and evaluated early in the plant operating history to !
assure that the system conforms to GE Specification 22A2736 and GE l Instruction 22A4185. This change adds these documents to the i drawing as references. There are no physical or functional changes l in the system. The instrument grounding system contains no active i components and is not the initiator of any accidents or malfunctions !
of equipment important to safety. The drawing change has no impact ;
on the operation of the system. Therefore, the probability of !
occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
l II. No. This drawing change provides additional reference information only.
No physical or functional changes are made in the instrument grounding system. The system is installed to minimize EMI in low voltage instrumentation and control circuits. The system has no interfaces with any other system. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The overall instrument grounding system-to-earth resistance was measured at 0.2 Ohms, well within the GE specification requirement of 1 Ohm or less. This low resistance plus the physical configuration of the system ensure that any effects due to EMI do not impact the operability of any instrument channel. This change i makes no physical or functional modifications to the system. l Therefore, no margin of safety has been changed. !
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SE No.: 95-0099 Source Document: PTI-C11-P0009, Rev. O Description of Change This periodic test instruction evaluates the installation and removal of l chemistry sampling equipment on the Control Rod (C11) drive pump suction l and drive water filters. The sampling equipment is being used to assess the performance of the C11 filters and to characterize the particulates in the C11 water.
Summary I. No. This test instruction installs sampling equipment into existing sampling points in the C11 system. The portion of the system the equipment is connected to is nonsafety. Should failure of the rig occur, the C11 pumps have sufficient capacity to make-up the loss of flow. The basic design and operation of the C11 system will not be affected. The scram function will not be impacted. Hence, accident analysis remains unchanged. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This test instruction installs sampling equipment into existing sampling points in the C11 system. Should failure of the rig occur, the C11 pumps have sufficient capacity to make-up the loss of flow.
The basic design and operation of the C11 system will not be affected. The scram function will not be impacted. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This test instruction installs sampling equipment into existing sampling points in the C11 system. Should failure of the rig occur, the C11 pumps have sufficient capacity to make-up the loss of flow.
The basic design and operation of the C11 system will not be affected. The scram function will not be impacted. Therefore, no margin of safety has been changed.
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SE No.: 95-0100~
Source Document: PTI-P72-P0004, Rev. 3 Description of Chance This periodic test instruction verifies the physical condition of the Plant Underdrain (P72) system. The object of the test is to verify that the porous concrete pipe, the weepholes, and the gravity drain inverts are free of obstruction and capable of performing their design functions.
This revision to the instruction will extend the performance frequency to annual.
Summary I. No. There are two design basis accidents involving the Plant Underdrain system described in USAR. The accidents are a yard break in the Circulating Water pipe outside the plant near the Steam Tunnel and Auxiliary Buildings, and failed expansion joints occurring inside the Turbine Building via flow through a fracture in the building base mat. Additionally, USAR Section 15.7.3 describes a postulated accident where radioactive water enters the groundwater from the Radwaste Building. The Underdrain system is used in these accidents to mitigate the consequences of the accidents only, and it has no affect upon the accident initiators. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Experience has shown that sedimentation is in fact a slow process.
The reduction in the frequency of manhole inspections will not affect the sedimentation rate, it just decreases the frequency at which it is monitored. Due to the slow build-up rate of sedimentation, the inspection frequency reduction will not affect the operability of the static components in the Plant Underdrain system. This instruction revision is limited to extending the performance frequency. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Plant Underdrain system is not addressed in the Technical Specifications. The safety function of the system is to prevent the build-up of hydrostatic pressure beneath the safety class building foundations to assure a sufficient factor of safety against dynamic instability is maintained. Dynamic stability is maintained as long as groundwater level is maintained at or below 590' elevation. This makes the ground water elevation of 590' elevation specified in the USAR the implied margin of safety for the P72 Plant Underdrain system. Since the gravity drain pipe normally sees no flow, no credible process exists for blockage to occur. Extending the Plant Underdrairi inspection frequency to annual will not reduce any margin l Of safer s.
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SE No.: 95-0101 g Source Document: DCN 4782 i
i Description of Change This drawing change revises B-022-007 to include the environmental
- conditions in zone Auxiliary Building (AB) 9 due to an Reactor Core Isolation Cooling (RCIC) system line break in zone AB-3. The j
environmental conditions that result from an RCIC line break in AB-3 l result in environmental conditions that are more severe for zone AB-9 l than that currently shown. As a result of this change, an increase in
- the maximum temperature by 60F occurs in zone AB-9. '
i Summary i
I. No. This change has no effect upon previously qualified equipment since l the newly incorporated environmental conditions could be withstood '
by all of the critical safety components in the zone. No physical changes occurred to the plant as a result of this drawing change.
This change had no effects upon any previously defined accidents or on equipment important to safety. Therefore, this change was concluded not to result in an increase in the probabilities or the consequences of any previously analyzed accidents or malfunctions of i
equipment important to safety.
II. No. All critical safety-related components remain well within their environmental limits, thus no change to performance of any component previously required to function is affected. In addition, no physical changes to the plant occur as a result of this change.
This change will therefore, not result in any new accidents or malfunctions of equipment important to safety.
III. No. The margin of safety for components is considered the difference between the condition in which the components will fail and the condition for which the component was demonstrated to be capable of withstanding. Since there is no increase in any parameters for which the components were designed to withstand, the activity has no effect on operability / failure point of any safety-related component.
Based on this, no margin of safety has been reduced.
SE No.: 95-0102 Source Document: USAR Change Request 95-073 Description of Change This change request clarifies Sections 9.1.4.2, 9.1.4.2.10.4, 9.1.4.2.10.6, 9.1.4.2.10.7, 9.1.4.2.10.8, 9.1.4.2.10.9, and 9.1.4.2.10.11 for specific activities dealing with refueling.
Summary I. No. The changes made to the USAR do not alter the design, function, or operation of any plant system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The changes made to the USAR are essentially administrative and do not change or affect plant equipment / operation. Therefore, the creation of a new accident or equipment malfunction not previously evaluated is not possible.
III. No. The changes made to the USAR are consistent with plant design basis documents. The changes do not alter the design, function, or operation of any plant system. Therefore, no margins of safety will be reduced.
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l SE No.: 95-0103 l Source Document: SCRs 1-95-1344, -1345, -1348, -1349 and -1350 l Description of Change l
l These Setpoint Changes (SCRs) will decrease the conductivity value for I
the Reactor Water Cleanup (G33) Reactor Water and Control Rod Drive (CRD) failed low alarms and add the associated leave-as-is tolerance and reset values. With the current setpoints for the RWCU and CRD conductivity low l alarms, there is the possibility that the alarm could activate falsely.
When RWCU is functioning properly, a conductivity value of 0.054 -
0.058 mhos/cm is common. This is only 0.004 - 0.008 mhos/cm difference and does not allow for the normal loop and calibration tolerances in the instrument loop. This condition allows the annunciator to alarm even if no problem exists. By lowering the alarm value in a conservative direction, the nuisance alarm will be eliminated, but retain the ability to detect a loop failure.
Summary I. No. These SCRs will alleviate a nuisance alarm condition, but yet allow the detection of a malfunctioning conductivity monitoring circuit.
These SCRs affect only a low alarm that alerts operation of the possibility that the associated conductivity cell may have failed.
The SCRs or alarm functions have no relevance to the safe operation of the plant, nor do they affect any plant instrumentation or accidents as described in the USAR. These SCRs have no impact upon plant chemistry. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These SCRs lower the low limit for the detection of a failed conductivity monitoring circuit for RWCU and CRD water. This alarm currently exists using a narrower operating span. Increasing the span for the alarm, covers the expected conductivity ranges found in RWCU and CRD water, yet maintain the ability to detect a loop failure. These setpoints affect only current alarms and does not alter or modify the detection of dissolved ionic species that could be found in the reactor coolant. These SCRs do not affect any controls that could be related to safety. This activity will not create the possibility of an accident or a malfunction of equipment important to safety of a type different than previously evaluated in the USAR.
III. No. Decreasing the low alarm limit for the RWCU and ORD water conductivity monitoring circuits, adequately supports the plant l chemistry program. The plant chemistry program and the Technical Specifications are not affected by these setpoint changes. This i activity does not reduce any margin of safety.
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SE No.: 95-0104 Source Document: DCN 5106 {
Description of Change This drawing change revises P&ID D352-241, Service and Instrument Air Supply - Unit 2, by removing computer points P51-EA001, PS2-EA001, P51-BC001, PS2-BC001, PS2-BC002, and P52-BC003. These computer points are not utilized.
Summary I. No. Removing the computer points identified above will not affect the l automatic operation of the Fervice (P51) and Instrument Air (PS2) i systems. Operations personnel never relied on these computer points to operate the P51 or PS2 systems since other instrumentation is !
available to directly indicate the Unit 2 P51 and PS2 parameters. '
Therefore, the probability of occerrence or the consequences associated with an accident or malfunction of equipment has not changed. ;
II. No. Removing the computers point identified above will not affect the automatic operation of the PS1 and PS2 systems. Operations ,
personnel never relied on these computer points to operate the P51 !
or PS2 systems since other instrumentation is available to directly indicate the Unit 2 P51 and P52 parameters. Therefore, the possibility of an accident or malfunction of a different type than ;
any previously evaluated in the USAR is not created. I l
III. No. Removing the computer points identified above will not affect the '
automatic operation of the P51 and PS2 systems. Operations personnel never relied on these computer points to operate the P51 or P52 systems since other instrumentation is available to directly indicate the Unit 2 P51 and P52 parameters. Therefore, no margin of safety has been changed.
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SE No.: 95-0105 l
Source Document: USAR Change Request 95-078 <
Description of Change This change request documents the containment spray flow rates of >
5800 GPM per train (11600 GPM total) in the containment negative pressure analysis contained within the USAR. A qualitative analysis indicating ;
that the containment negative pressure analysis remains valid when considering a flow rate of 11,600 GPM combined with a 950F initial containment temperature and assuming a 90% spray efficiency.
Summary I. No. The existing plant configuration, operation, and system flow rates are adequately encompassed by the results of the existing containment negative pressure analysis considering the qualitative analysis performed. The item did not change any equipment or require the revision of any existing analyses. Therefore, the probability of occurrence or the consequences associated wi',h an accident or malfunction of equipment has not changed.
II. No. The analysis showed that no new equipment types or new system interactions are created and that the nominal flow rates are unchanged. No new equipment failure modes were identified as a result of the change. No changes were made to any plant system.
The change in the flow rate was not the result of an equipment change. Therefore, the possibility of an accident or nelfunction of I a different type than any previously evaluated in the USAR is not created.
III. No. The existing plant configuration, operation, and system flow rates are adequately encompassed by the results of the existing containment negative pressure analysis considering the qualitative analysis performed. The item did not change any equipment or require the revision of any existing analyses. Therefore, no margin of safety has been changed.
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SE No.: 95-0106 Source Document: CHI-0046, Rev. O Description of Change This instruction provides the guidance for the temporary installation of corrosion coupon racks and corrosion test coupons for six nonsafety-related systems. The instruction will provide a means for assessing corrosion rates by evaluating representative coupons after being exposed to system water for various lengths of time. The typical duration of a test is 90 days. The coupon racks will be removed from the systems when not in use. The racks will be installed in small diameter branch lines ranging in size from 3/4" to 2'. The racks will be designed in a manner consistent with the plant systems being monitored.
1 Summary I. No. The rack design is compatible with the operating parameters of the affected systems. The flow requirements to operate the racks are l insignificant relative to the system flows. Should the racks become disengaged, the resultant leaks are bounded by more severe leaks already analyzed in the USAR. Each system has sufficient make-up capability to overcome the leak volume. The design function of each of the affected systems will remain unchanged. The portions of the affected systems involved with these changes are nonsafety-related and not required to safety shutdown the plant. The installation is structurally acceptable. Therefore, the probability of occurrence or the consequences of any accident or malfunction previously evaluated in the USAR has not changed.
)
II. No. The coupon racks will be designed and installed in accordance with the applicable design and operating parameters. Any potential flooding concerns are bounded by existing evaluations. The portions of the systems involved do not support the safe shutdown of the plant. The installation will not impact the operation of any equipment important to safety. The coupon rack installation will not affect the operation of the affected systems or any other plant systems. Therefore, the possibility of a different type of accident or malfunction of equipment important to safety is not created.
III. No. The coupon racks will be designed and installed in accordance with the applicable design and operating parameters. Any potential flooding concerns are bounded by existing evaluations. The portions of the systems involved do not support the safe shutdown of the plant. The installation will not impact the operation of any equipment important to safety. The coupon rack installation will not affect the operation of the affected systems or any other plant systems. Therefore, no margin of safety has been reduced.
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SE No.: 95-0107 i Source Document: DCN 4512 l
Description of Change 1
] This drawing change revises P&ID D302-201, Circulating Water System, to eliminate abandoned instrumentation. Flow indicator IN71-R0206 was
- removed from service and abandoned in place during Refueling Outage (RFO) 4.
Summary s
j I. No. Flow indicator 1N71-R0206, associated supports, tubing and mounting l
hardware have been abandoned in place since RF04. Removing this )
installation from service was evaluated by Safety Evaluation '
94-0079. This change only addresses the removal of the instrument from a design drawing. As previously evaluated, there is no impact upon the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Flow indicator 1N71-R0206, associated supports, tubing and mounting hardware have been abandoned in place since RF04. Removing this l installation from service was evaluated by Safety Evaluation i 94-0079. This change only addresses the removal of the instrument from a design drawing. As previously evaluated, there is no impact upon the plant. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Flow indicator 1N71-R0206, associated supports, tubing and mounting hardware have been abandoned in place since RF04. Removing this installation from service was evaluated by Safety Evaluation 94-0079. This change only addresses the removal of the instrument from a design drawing. As previously evaluated, there is no impact upon the plant. Therefore, no margin of safety has been changed.
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SE No.: 95-0108 Source Document: FTI-E0013, Rev. 3 Description of Change This fuel technical instruction incorporates the use of the fuel bundle lift hook as an available option for the transfer of new fuel between the crating area and new fuel inspection stand.
Summary -
I. No. The fuel lift hook is used in the Fuel Handling Building and is not intended to carry loads over irradiated fuel or to carry loads in the spent fuel pool. Therefore, the probability of a fuel handling accident outside or inside containment is unaffected. The fuel lift >
hook is appropriately rated for the load imposed by a new fuel bundle. The latching mechanism is manual engagement vice pneumatic actuation; the fuel assembly is securely locked in place until -
manual disengagement. The probability of dropping a new fuel bundle is not increased when using the fuel lift hook to transfer ,
unirradiated fuel assemblies from their shipping containers to the i new fuel inspection stand. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment is not increased.
II. No. The only credible accident considered is dropping a new fuel bundle which has minimal radiological consequences and is completely bounded by the fuel handling accident outside or inside containment.
The only credible malfunction is the mechanical malfunction of the fuel lift hook. The fuel handling procedures described in USAR 9.1.4 are unchanged and the same malfunction can occur with the general purpose grapple. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in '
the USAR is not created.
III. No. The handling of new fuel is not controlled by and does not form the bases for any Technical Specification. Failure of the fuel lift hook is bounded by existing analyses. Therefore, no margin of l safety has been changed. >
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SE No.: 95-0109 Source Document: PAP-0802, Rev. 5 Description of Change This procedure change analyzes increasing fuel enrichment to 4.5% U-235.
The higher enrichment enables fuel to operate in the reactor longer and results in higher burnup fuel. In support of the increased enrichment, the criticality analysis for the spent fuel pool racks was updated.
E.gnary I. No. Using fuel with higher U-235 enrichment does not change any design, material and construction standards. The criticality analysis re-performed on the spent fuel pool racks confirms the safe storage of higher enrichment fuel. There is no increase in the dose to the public above the licensing limits in 10CFR20 and 10CFR100 or onsite doses that would impede actions necessary to mitigate the consequences of accidents which consider fuel with a specific burnup, e.g., Loss of Coolant Accident, Control Rod Drop Accident, and Fuel Handling Accident, are insignificantly affected.
Additionally, system designs which are based on source terms previously established remain bounded. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No changes to the physical plant are required to support operation with higher enriched fuel. There is no increase in the dose to the public above the licensing limits in 10CFR20 and 10CFR100 or onsite doses. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. There is no increase in the dose to the public above the licensing limits in 10CFR20 and 10CFR100 or onsite doses. All original acceptance limits used in the original licensing basis are maintained. The margins to safety remain acceptable with the use of fuel with increased enrichment.
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l SE No.: 95-0110 Source Document: DCP 95-0055, Rev. O Description of Change This design change replaces the Reactor Recirculation Pumps (lB33-C001A/B) shaft mechanical seals with a new improved seal design, and adds a new manual vent valve on the second stage instrument sensing line of each seal assembly with the objective of increasing pump reliability.
Summary I. No. This design change maintains the fundamental two-stage seal design.
The pressure retaining components adhere to applicable ASME Code requirements. The new design does not adversely affect the pump slow coastdown capability or pump performance as a fission product barrier. In addition, the vent valve arrangement adheres to existing design criteria for unisolable lines between drywell and containment. Potential failure modes and effects as a result of this change have been evaluated as acceptable. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. Since these changes do not create any new systems or alter any system interconnections, the redundancy and independence of the affected systems is maintained and susceptibility to common mode or common cause failures is not possible. No new failure modes are introduced by the design change and the probability of existing failures occurring is not significantly altered by this modification. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. By adherence to the ASME Code for pressure boundary components, Technical Specifications 4.0.5 and 3/4.4.8 are not affected.
Drywell and reactor coolant leakage limits associated with Technical Specifications 3/4.4.3 and 3/4.6.2.2 are maintained by this change.
Further, since Reactor Recirculation system functions and pump slow coastdown capability are maintained, Technical Specification 3/4.4.1 is not affected either explicitly or implicitly. Therefore, the activity does not reduce any margin of safety.
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SE No.: 95-0112 Source Document: PAP-0510, Rev. 5, TC-3 PAP-0514, Rev. 4, TC-5 Description of Change These procedure changes incorporate the use of a bioassay program composed of an internal screening component and a diagnostic bioassay component to determine individual doses due to the inhalation and ingestion of radioactive material. The internal screening analysis (passive monitoring) will be used to screen individuals and determine the need for performing the diagnostic bioassay analysis (whole body count).
Summary I. No. The changes are administrative and programmatic. The use of a bioassay program composed of an internal screening component and a diagnostic component to determine individual doses due to the inhalatico and/or ingestion of radioactive material does not alter the design or the operation of plant 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 has not increased.
II. No. The changes are administrative and programmatic. The use of a bioassay program composed of an internal screening component and a diagnostic component to determine individual doses due to the inhalation and/or ingestion of radioactive material does not impact the design or operation of the plant. Therefore, the possibility of l an accident or malfunction of a different type than previously !
evaluated in the USAR has not been created.
III. No. The administrative section of the Technical Specifications require that procedures be written and approved to implement the requirements of 10CFR20. The use of a bioassay program composed of an internal screening component and a diagnostic component to determine individual doses due to the inhalation and/or ingestion of radioactive material are consistent with the requirements of 10CFR20. Therefore, no margin of safety has been reduced.
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SE No.: 95-0113, 96-0089 >
Source Document: Emergency Plan, Rev. 13 '
Description of Change This revision to the Emergency Plan incorporates various changes to the '
plan affecting event classification, facility staffing, conduct of drills, and plant equipment testing.
Summary I. No. This revision does not direct or impact the operation or design of any plant structure, system or component. 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. :
1 II. No. This revision does not alter the design of the plant; the type, !
frequency or consequences of an accident; or direct plant mitigating l actions. Theref,re, it will not create the possibility for an '
accident or malfunction of a different type than previously evaluated.
III. No. This revision 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.: 95-0114 Source Document: TM 1-95-020 Description of Change Temporary Modification (TM) 1-95-020 installs temporary jumpers in the plant Instrument Air (PS2) system to facilitate repairs to an air distribution system isolation valve while maintaining the PS2 distribution system in service.
Summary I. No. The temporary modification described re-routes portions of the PS2 distribution system to allow normal continued operation of plant instrument air loads during repair of PS2 sy.= tem components.
Failure of the TM to supply PS2 system lor.ds is anveloped within the USAR Loss of Instrument Air accident analysis. Therefore, the probability of occurrence or the conseo';ences associated with an accident or malfunction of equipment haa not changed.
II. No. The temporary modification described re-routes portions of the P52 distribution system to allow normal continued operation of plant instrument air loads during repair of PS2 system components.
Failure of the TM to supply PS2 system loads is enveloped within the USAR Loss of Instrument Air accident analy:,is. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The PS2 distribution system is not addressed by Technical Specifications. Components supplied by the instrument air distribution system fail in their safe shutdown condition upon a loss of instrument air. This 'IN is bounded by existing USAR analyses. Therefore, no margin of safety has been changed.
s SE No.: 95-0115 i Source Document: DCP 95-5053, Rev. 0 l
Description of Change !
This design change converts 2500 square feet of the Fuel Handling '
Building from the Protected Area to the Vital Area. -
i Summary C s
I. No. This change converts 2500 square feet of the Fuel Handling Building (
620' elevation from the Protected Area to the Vital Area. The i original intent was to utilize the area to support Unit 2 ;
construction. Consequently, the area was partitioned from Unit 1 !
with a temporary barrier identified in the Plant Security Plan. l This change relocates the plant security boundary to the building i perimeter to permit the use of the entire 620' elevation floor '
space. The design change maintains the integrity of the security i system. This change does not increase the severity or nature of any j fire related challenge to any equipment important to safety or other l plant feature. Personnel access and intrusion detection will continue to be controlled as required by the Security Plan. l Therefore, the probability of occurrence or the consequences !
associated with an accident or malfunction of equipment has not !
changed. !
II. No. The security monitoring system will continue to support plant operations and operate as intend 2d by the original design. This l change maintains the integrity of the security system. Therefore, i the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. There is no relationship between the design change and the basis for any Technical Specification. The security system is not discussed in the Technical Specifications and will continue to meet the requirements identified in the Plant Security Plan. The Plant Operations Section (POS) will maintain this room as a combustible free zone. This design change has no impact on the administrative aspect of the Fire Protection Program described in Technical Specifications 6.5.1.6, 6.5.2, 6.8, or 6.9.4. This change will not affect the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, no margin of safety has been changed.
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l SE No.: 95-0116 l
Source Document: PAP-Oll4, Rev. 1, TC-1 Description of Change This procedure change evaluates the reassignment of the Radiation Protection Manager responsibilities defined in Regulatory Guide 1.8 from Plant Health Physicist to the Manager, Radiation Protection; and the title change of the Plant Health Physicist to Superintendent, Health Physics.
Summary I. No. The reassignment of the Radiation Protection Manager i responsibilities defined in Regulatory Guide 1.8 to the Manager, l' Radiation Protection and the title change of the Plant Health Physicist to the Superintendent, Health Physics are administrative l and programmatic. The education and experience requirements of the I Radiation Protection Manager defined in Regulatory Guide 1.8 remain.
The design and operation of the plant equipment important to safety, i and all supporting analyses have not changed. USAR accident analysis has nct changed. Therefore, the probability of occurrence or the radiological consequences of an accident or a malfunction of equipment important to safety previously evaluated in the USAR has not changed.
II. No. The reassignment of the Radiation Protection Manager responsibilities defined in Regulatory Guide 1.8 to the Manager, Radiation Protection and the title change of the Plant Health Physicist to the Superintendent, "aalth Physics are administrative and programmatic. The education and experience requirements of the Radiation Protection Manager defined in Regulatory Guide 1.8 remain.
These administrative and programmatic changes do not create an accident or malfunction of equipment of a different type than previously evaluated in the USAR.
III. No. The reassignment of the Radiation Protection Manager responsibilities defined in Regulatory Guide 1.8 to the Manager, Radiation Protection and the title change of the Plant Health Physicist to the Superintendent, Health Physics are administrative and programmatic. The education and experience requirements of the Radiation Protection Manager defined in Regulatory Guide 1.8 remain.
These administrative and programmatic are consistent with the requirements of 10CFR20. Therefore, no margin of safety has been reduced.
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SE No.: 95-0117 Source Document: TM l-95-023 Description of Change This Temporary Modification (TM) installs a freeze seal to support valve testing per SVI-G33-T9131. The freeze seal location is associated with the Reactor Water Cleanup (RWCU-G33) bottom head drain line from the Reactor Recirculation (B33) system and reactor vessel, downstream of closed valves 1G33-F0101 and -F0103.
Summary I. No. The accident of interest is a LOCA within the Reactor Coolant 1 Pressure Boundary (RCPB), more specifically, an un-isolatable bottom head drain (B33/G33) line break or loss of freeze seal. The freeze seal will be placed on the 3" RWCU piping while the reactor vessel is being maintain between 70 F and 1400F with only static head pressure (approximately 50 PSID) during refueling (Mode 4/5).
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 exams 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 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 40 GPM will leak into the drywell with the valves closed as !
previously stated. Any 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 associated with an accident or malfuncticn of equipment has not changed.
II. No. With the plant in Mode 4/5 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 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 duration and a design bases LOCA.
Although the prerequisites for a LOCA inside drywell appears to be different (recctor at power verses the reactor depressurized and fuel isolated by valves) the design bases LOCA bounds the freeze seal failure. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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i SE No.:
95-0117 (Cont.) ;
Summary (Cont.)
j III. No. The treeze seal on the 3" pipe is an industry proven method of
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isolating water systems that have limited isolation. Industry' tests have proved that deviation below the transition temperature and back-to-normal temperature do not change the crystal structure or i '
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 exams for indications, i 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. l Therefore, no margin of safety has been changed. l i
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SE No.: 95-0118 Source Document: DCP 95-0066, Rev. O Description of Change This design change revises the means of establishing the maintenance cross-tie between Unit 2, Division III Battery 2E22-S005 and the Unit 1 Division III DC Bus ED-1-C (1R42-S0037). In addition, the 480 VAC power source to the Unit 2 Division III Battery Charger 2E22-S006 will be changed from 2R24-S0029 (a non-diesel backed motor control center) to a non-1E source (distribution panel F2-D12/2R25-S004). The unique function of the battery charger (2E22-S006) is to maintain battery 2E22-S005.
During a cross-tie, only battery 2E22-S005 will be connected to bus ED-1-C, the rest of the Unit 2, Division III DC system will be isolated.
The nonsafety and safety-related equipment separation requirements per IEEE 384-1974 and NRC Regulatory Guide 1.75 are not compromised.
Summary I. No. Implementation of this change restores the maintenance cross-tie between the Unit 1, Division III 125 VDC Bus ED-1-C and the Unit 2, Division III Battery 2E22-S005. There is no impact on the system operability, capability and reliability. No potential effects exist. Therefore, neither the probability of occurrence nor the consequence of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. This modification replaces the Unit 2 Division III DG Control Panel 2E22-P002 with a smaller panel to restore the Division III 125 VDC system cross-tie capability. The Unit 1 Division III DC system original function and operation remain unchanged. Therefore, this modification does not create the possibility of an accident or malfunction of different type than any previously evaluated.
III. No. Technical Specifications 3.8.2.1c.1 and 3.8.2.2.c.1, either the Unit 1, Division III Battery (1E22-S005) or the Unit 2, Division III Battery (2E22-S005) is required to support operability of the Unit 1 Division III High Pressure Core Spray system. Since all of the nonsafety-related circuits and equipment must be separated from the circuits and equipment used for Unit 1, Division III HPCS system before any cross-tie can be performed, separation requirements per IEEE 384-1974 are maintained. There are no potential effects on the basis any Technical Specification. Hence, this modification will not reduce any margin of safety.
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SE No.: 95-0119 Source Document: PAP-0101, Rev. 8, TC-3 PAP-0502, Rev. 11, TC-1 PAP-0507, Rev. 10, TC-5 PAP-0522, Rev. 7, TC-3 NQI-0507, Rev. 12, TC-3 NOI-0510, Rev. 9, TC-3 Description of Change These procedure and instruction changes incorporate the elimination of formal periodic reviews for various Perry Operations Manual documents and the reduction in the number of approval signatures for the Perry Operations Manual documents.
Summary I. No. The changes involve revisions to administrative processes. The technical content of Perry Operations Manual documents have not been affected. Commitments to Regulatory Guide 1.33 have been maintained. There is no impact upon the design or operation of the plant. Therefore, the probability of occurrence or the consequcaces associated with an accident or malfunction of equipment has not changed.
l II. No. The changes involve revisions to administrative processes. The technical content of Perry Operations Manual documents have not been affected. Commitments to Regulatory Guide 1.33 have been maintained. There is no impact upon the design or operation of the plant. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The changes involve revisions to administrative processes. The technical content of Perry Operations Manual documents have not been affected. Commitments to Regulatory Guide 1.33 have been ,
maintained. There is no impact upon the design or operation of the l plant. Therefore, no margin of safety has been changed.
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- SE No.: 95-0120
! Source Document: Potential Issue Form (PIF) 95-0111 i Description of Change j
This potential issue provides clarification as to the intended compliance with IEEE 338-1477, " Criteria for the Periodic Testing of Nuclear Power
- Generating Station Safety Systems", and the Power Systems Branch
! Technical Dosition (PSB-1) for the degraded voltage protection scheme.
! Each of two groups of three degraded voltage relays per division has been
- identified as a channel in Table 3.3.3-1 Section D of the Technical Specifications. The two relay groups individually meet the generic
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I definition of a channel and the design basis described in the Perry Supplement to Safety Evaluation Report 2 (SSER) as 'two out of two I
coincident logic". This is also consistent with Branch Technical l Position (PSB-1), ' Adequacy of Station Electric Distribution System i Voltages", requirement that the Class 1E undervoltage protection scheme i " include coincidence logic on a per bus basis to preclude spurious trips
- of the offsite power source". The design basis, therefore does not
! require channel independence as specified in IEEE 279-1971, Section 4.6, j for the channels as described in the degraded voltage protection scheme.
Summary i
I. No. The applicalle USAR Chapter 15 accidents bre LOOP and LOCA.
Clarifying compliance with IEEE 338-1977 maintains the current probability of occurrence of these accidents. No new components have been added. Therefore, the failure modes for the degraded voltage protection scheme have not changed and periodic testing remains consistent with the applicable standards. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No new accidents can be created as the degraded voltage protection scheme malfunctions are enveloped by previously analyzed USAR analysis for LOOP and LOOP /LOCA. As the configuration of the degraded voltage protection scheme has not been modified, there are no new failure modes associated with providing clarification of conformance to IEEE 338-1977 and PSB-1. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The technical basis applicable to this activity is conformance with IEEE 279-1971 and 338-1977, and Regulatory Guide 1.22 and Technical Specification Table 3.3.3-1. As discussed in the description, the design basis of the degraded voltage protection scheme is satisfied by the intended compliance with IEEE 338-1977 as modified by PSB-1.
Therefore, clarification with regard to its applicability to the degraded voltage protection scheme as governed by PSB-1 for testing does not reduce any margin of safety.
SE No.: 95-0121 Source Document: USAR Change Request 95-084 )
Description of Chance This change request revises various USAR figures which depict containment isolation valve information. Information from the plant P& ids was incorporated to better illustrate the isolation valve configurations at )
the containment /drywell penetrations.
Summary i
I. No. There will be no physical changes to the plant with implementation of this change request. The changes involve the incorporation of
'as-built" information to more clearly illustrate the plant configuration. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR cannot be attributed to these changes. l II. No. The change request is limited to editorial drawing / figure revisions that have no affect on system operability / function or physical configuration of the plant. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. These revisions are limited to editorial changes to various USAR ;
figures. None of the changes alter any of the related system functions or responses. Therefore, no margin of safety has been changed.
SE No.: 95-0122 Source Document: Physical Security Plan, Rev. 21 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. i Summary I. No. The PSP describes the comprehensive Physical Security Program and j 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. No. The PSP does not direct the operation of plant systems or equipment .
and, therefore, does not create the possibility for an accident or i malfunction. !
i III. No. The PSP changes do not reduce any margin of safety. j r
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SE No.: 95-0123 Source Document: Potential Issue Form (PIF) 95-1877 Description of Change This PIF evaluates the "use-as-is' disposition of the continued use and addition of aluminum tags to the drywell and containment. The significance of non-evaluated aluminum tags within the containment and drywell is the corrosion rate and the resultant hydrogen generation rate in a post-accident steam environment.
Summary I. No. Aluminum tags do not alter the function of any plant equipment and do not interact with the containment or drywell atmcsphere except after an accident has occurred. This disposition provides the allowance for the continued use and the addition of aluminum tags placed within the containment and drywell. Plant operation with these identification tags in-place has not, and will not affect the ability of the hydrogen control equipment to perform its intended function since the increase in hydrogen production is negligible.
Since the function of the hydrogen control equipment is not affected, the containment barrier will still function as designed and will continue to maintain the radiological consequences as previously analyzed. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The aluminum corrosion and resu' .it hydrogen release are presently evaluated in the USAR and its s.gporting analyses. All equipment designed to control the generation of hydrogen remains unaffected by this disposition. No new accident initiators are created.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. ,
l III. No. The addition of aluminum tags produces a very small differential j increase in the overall hydrogen production for events under 10CFR50.44 and 50.46. Technical Specification 3/4.6.7, Containment Hydrogen Recombiner Systems, which encompasses the Containment l
Hydrogen Recombiner system, the Combustible Gas Mixing system, and '
the Containment and Drywell Hydrogen Ignition system, remains unaffected and unchanged. The USAR states that, 'The production of l hydrogen from the corrosion or aluminum is negligible compared to other sources", continues to remain valid and unchanged. Therefore, ,
the 'use-as-is" disposition would not reduce any margin of safety.
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l SE No.: 95-0124 '
Source Document: TXI-0233, Rev. O Description of Change l l
This temporary test instruction (TXI) will be perfcrmed to govern !
shutting down the Turbine Lube Oil system Emergency Bearing Oil !
l Pump (EBOP) and Turning Gear Oil Pump (TGOP) when the Main Turbine is in l operation. Within the TXI, the low bearing oil pressure trip will be l defeated while securing the EBOP.
Summary l I. No. The low bearing oil pressure trip specifically is not relied upon in I any accident analysis. The turbine is designed to withstand normal .
conditions and anticipated transients including those resulting in !
! turbine trip.without a loss of structural integrity. The USAR i I
analysis of plant response to the turbine trip is bounding. The C Main Turbine low bearing oil trip is not required for the safety of the plant as discussed in USAR Chapter 6 and l'SAR Sections 7.4 and 7.6, and therefore cannot effect analyses performed on those systems required for safety. Therefore, the probability of !
occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. An analysis of plant response to the Main Turbine trip is performed i in USAR Section 15.2.3 for a variety of initiating causes, and an increased potential for low bearing oil pressure is bounded by this analysis. Defeating the turbine trip on low bearing oil pressure does not result in an accident not already evaluated in the USAR.
Failure of the turbine to trip is not credible. Direction within i the TXI instructs the operator to manually trip the turbine if bearing oil pressure is lost. The Main Turbine is not required for safety and defeating the low bearing oil pressure trip has no effect upon the operation of those systems required for safety. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Main Turbine trip on low bearing oil pressure is not a Technical Specification required function. Defeating the low bearing oil pressure trip will not effect any Technical Specification related i setpoint or function. Defeating the low bearing oil pressure trip affects no other turbine trips, and so does not reduce the margin of safety for turbine overspeed protection. Therefore, no margin of safety has been changed.
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SE No.: 95-0125 Source Document: SMRF 95-5080, Rev. 0 l
Description of Change l This design change removes Control Room valve position indicating meters 1E12-R608A(B) and 1E12-R609A(B) from Control Room panel 1H13-P601. These l valve position indicators are associated with one-inch Motor Operated :
Valves (MOVs), 1E12-F073A(B) and 1E12-F074A(B). These meters require frequent repair and calibration to provide meaningful indication which requires time to fix and exposes maintenance personnel to unnecessary dose.
Summary l
I. No. The following accidents analyzed in the USAR apply to the Residual '
Heat Removal (RHR-E12) system: Inadvertent Shutdown Cooling Operation and Failure of Shutdown Cooling. This modification does l not impact either of these events. The change will not affect RHR system performance. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This modification removes the valve position indicating meters for the RHR heat exchanger vent valves from Control Room panel H13-P601.
Disconnecting these meters has no impact on the operation of the RHR valve control circuitry or any other RHR control circuitry. ,
This design change will not affect the functional performance of the RHR or any other system. Therefore, the possibility of an accidant or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Modifications made to the RHR wiring does not impact the safety-related function of the RHR valves or any other RHR control logic. The RHR system design criteria and safety functions are maintained. Therefore, no margin of safety has been changed.
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SE No.: 95-0126 Source Document: DCP 92-0056, Rev. 1 Description of Change This design change removes the internals from the Liquid Radwaste system OG50-F0641 and -F0642 check valves to prevent the internals of the valves from plugging with filter aid material.
Summary I. No. Removal of the check valve internals does not adversely ahect the design or operation of the radwaste processing system. Pressure integrity of the valve / system is unaffected by this change. The flatbed filter system and sub-system components are nonsafety and are not relied upon by equipment important to safety. Failure of this equipment could not initiate and is not used to mitigate any of the accidents described in the USAR Chapter 15. The filter aid tank contains condensate water which is radioactive; however, it's volume and concentration are conservatively bounded by other liquid radwaste system tanks. Therefore, it's failure does not increase the radiological consequences of a tank rupture as described in USAR Chapter 15, nor would it impede the operator actions necessary to mitigate the consequences of any accident. Therefore, the probability of occurrence or the consequences of an accider .
malfunction of equipment important to safety previously evalv ,ed in l the USAR is not increased. l l
The check valves are the only equipment modified by this activity.
II. No.
The flatbed filter system is nonsafety and is not required for safe shutdown of the reactor. Failure of this equipment does not have the potential to threaten the fuel, reactor coolant pressure boundary, nor does it have an effect on the function nor can it degrade the performance of any systems, structures, or components important to safety. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR is not created. l III. No. The equipment affected by this change is not governed by Technical Specifications. USAR accident analysis is not affected. Therefore, no margin of safety has been reduced.
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SE No.: 95-0127 Source Document: DCP 86-0338, Rev. 4 Description of Change This design change included installation of a motor-operated flow control !
l valve with flow straightening vanes in the backwash inlet line to the l Condensate Demineralizer (N24) Anion Regeneration Tank to facilitate variable backwash flow rates. This design change was partially implemented. The resin regeneration system is no longer being used. I Revision 4 of the design change deletes straightening vane installation.
Summary !
I. No. The chemical regeneration portion of the Cobirnsate Demineralizer system does not contribute to the occurrence of any accidents evaluated in the USAR, and has no mitigating functions. The installed components which will be abandoned cannot cause or influence other system components to operate outside their design limits. 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 will not create any new systems, add any new active equipment, or compromise the function of any active systems, structures or components. The affected components do not interface with any equipment important to safety. Condensate, condensate demineralizer, and interfacing system operations will not be adversely impacted by these changes. Therefore, the possibility of an accident or malfunction of equipment of a different type than any previously evaluated is not created.
III. No. The Condensate Demineralizer system and regeneration process are not covered in the Technical Specifications. The changes will not degrade the capability of any system to mitigate the effects of postulated accidents. Thus, no margin of safety is reduced.
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SE No.: 95-0128 Source Document: DCP 95-5090, Rev. O Description of Chance This design change installs a Potable Water (P71) line and a Sanitary Sewer (P66) line to support an office trailer being installed on the east side of the Maintenance Building.
Summary I. No. The P71 and P66 systems are not relied upon directly in any of the accidents evaluated in USAR Chapter 15. Since these systems are not accident initiators, the addition of these new lines to an office trailer cannot increase the probability of an accident described in Chapter 15. Components being added as part of the new P71 and P66 lines are consistent with those already installed. Neither the P71 nor the P66 systems are involved in coping with any radiological consequences. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This modification does not alter the relationship of the P71 and P66 systems to plant safety features. The modification will not i degrade the performance of the P71 or P66 systems since the new lines will be designed and constructed more conservatively than the balance of the systems. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The P71 and P66 systems are not directly associated with any plant systems important to safety. . This modification will not alter that status. The Technical Specifications do not address the P71 or P66 systems. Therefore, no margins of safety are affected. i l
SE No.: 95-0129 Source Document: DCP 95-5076, Rev. O Description of Chance This design change installs a new Potable Water (P71) line to a toilet trailer complex being placed at the east end of the Unit 2 Turbine Building. The trailers will be used to support craft activities during Refueling Outage 5.
Summary l I. No. The P71 system does not have any direct connections to other plant i systems important to safety. It is not relied upon in any accident l
or malfunctions previously evaluated. The P71 system is not an accident initiator. The new line will be constructed similarly to that of the original design. Therefore, the probability of occurrence or the consequences associated with an accident or t malfunction of equipment has not changed.
l t II. No. The P71 system has no direct connections to systems important to !
safety. The new line will be designed and installed similarly to the existing P71 components. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The P71 system has no direct connections with other plant systems important to safety. The P71 system is not addressed in the l Technical Specifications. Therefore, no margin of safety has been changed.
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- SE No.
- 95-0130 Source Document: Potential Issue Form (PIF) 95-1103 l 1 Description of Change l This potential issue evaluates deletion of Instrument Air (PS2) lines routed to sluice gates P45-D0004A and P45-D0004B as depicted on P&ID D-302-791, Emergency Service Water. The-subject sluice gates were l formerly equipped with inflatable seals. The seal removal was evaluated l as acceptable at the time of the work, and therefore this change is !
basically editorial in nature to provide a more accurate reflection of !
the installed condition.
Summary i
I. No. The purpose of the sluice gates are to open automatically upon I receipt of a low water level signal from the Emergency Service !
Water (ESW) Pumphouse forebay should normal ESW supply from the '
intake tunnel be interrupted. The removal of the PS2 air line in the subject P&ID will not change the operation of the gates, which ;
had the seals previously removed. This drawing change does not )
effect the design, material, and construction standards applicable I to the gates and the ESW System. The revision to the P&ID does not !
affect automatic initiation of the sluice gate nor the ability to i perform its function. Verification of the ability to verify automatic operation of the sluice gates will be maintained.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No equipment or physical work is being conducted or added to the plant as part of this potential issue disposition. The drawing change is editorial in nature. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The ESW system remains unaffected by this disposition. The sluice gate operation will continue to function from the ESW forebay water level actuation as designed. The extent of leakage that flows i across the closed sluice gates with the seals removed and the ;
resulting increase in ESW supply temperature has been previously l evaluated to be acceptable. Therefore, no margin of safety has been j changed, i
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SE No.: 95-0131 Source Document: Potential Issue Form (PIF) 95-2267 Description of Change J
This potential issue evaluates the "use-as-is" disposition of the ;
unavailability of the lower drywell cooling unit 1M13-B0001 and associated fans 1M13-C001A/B. The subject unit supplies cooling air to i the Reactor Pressure Vessel (RPV) skirt area, RPV pedestal area (under l vessel), drywell bioshield inner wall, and general lower drywell areas.
As a result of the unavailability of this cooling unit, ambient i temperatures in the subject areas have increased.
Summary I. No. Failure of the drywell lower air coolers is not an initiator to any accidents or transients evaluated in the USAR. The increased temperature as a result of operation without lower drywell cooling does not adversely affect the structural integrity of the components in the areas cooled. The integrity of the control rod drive housing lower supports has beca demonstrated to be not impacted. Higher localized temperatures as a result of operating without the lower drywell coolers does nat adversely affect any EQ-qualified equipment. The structural integrity of the reactor coolant system is not adversely affected. USAR accident analysis remains unchanged. Therefora, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
1 II. No. Structural integrity of the reactor vess and its support structure has been demonstrated. The drywell cooling system is nonsafety-related and not required to mitigate the consequences of an accident. Higher localized temperatures as a result of operating without the lower drywell coolers does not adversely affect any EQ-qualified equipment. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluatei in the USAR is not created.
III. No. While the drywell cooling system performance is reduced by this loss of available capacity, the average air temperature of the drywell is maintained. The impact on equipment qualification of affected drywell components has been determined to be unaffected. The structural integrity of the RPV is maintained and the integrity of the ASME Section III components which attach to the vessel are not adversely impacted. Thus, Technical Specification 3/4.4.8 is not affected. Therefore, no margin of safety has been changed.
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i SE No.: 95-0132 i Source Document: NR 92-N-056, Rev. 2 }
4 Description of Change -
i This Nonconformance Report (NR) identifies a broken test line inside the ;
high pressure condenser. The broken test line has been repaired by i securing the broken end to an adjacent cross brace with a oipe clamp. ;
2 This change secures the line against movement during condenser operation.
j The accessible broken end leading to the extraction steam piping was i plugged and seal welded to eliminate the steam jet that would result
! under normal plant operation. This will prevent steam flow from the line i and its impingement on any nearby internal condenser components. ;
i k Summary I
l l I. No. The connection to the steam piping has been plugged and seal welded i in accordance with ISS-2000 and ASME/ ANSI B31.1. A normally closed j valve, IN36-F0560A, on the outside of the turbine skirt prevents air i from leaking into the condenser. Potential leakage through this j normally closed valve would result in a small amount of in-leakage into the condenser through the 1/4" tube - thus, potentially j '
j decreasing the condenser vacuum slightly. However, this added i leakage is bounded by other more severe failure modes as discussed 1
in USAR Section 15.2.5. This change does not involve equipment required for mitigating accidents. There are no increased radiological consequences of any accident previously evaluated in
- the USAR. As stated in USAR Section 10.4.1.4, the Main Condenser is l not required to support the safe shutdown of the reactor, or to j support the operation of reactor safety features. Therefore, the j probability of occurrence or the consequences of an accident or
- malfunction of equipment important to safety previously evaluated in j the USAR is not increased.
~i II. No. As stated previously, the only adverse effect expected is a slightly e
increased condenser stem flow rate. In the highly unlikely event j that the test piping should break off during normal operation,
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damaged (leaking) tube (s) can be identified and plugged on-line as described in USAR Section 10.4.1.2.e. None of the affected components are required for the safe operation or shutdown of the plant. Therefore, the possibility of a different type of accident or malfunction of equipment previously evaluated in the USAR is not created.
III. No. The NR disposition does not affect any system or plant operation.
Should the current configuration result in tube leaks (via impact from a dislodged or broken test line or clamp component) plant operation will be maintained within the limits defined in Technical Specification Section 3.4.4. In addition, the seal-welded plug and the root valve are designed in accordance with the requirements of ISS-2000 and ASME/ ANSI B31.1 to design conditions more severe than those imposed by the condenser. Based on the above, no margin of safety is affected.
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4 SE No.: 95-0133 i Source Document: NR 92-N-057, Rev. 2 i
Description of Change i This Nonconformance Report (NR) identifies a broken test line inside the j
high pressure condenser. This broken test line has been repaired by l removal of the line, and plugging and seal welding the coupling i
! connection to the 42' Main Steam (Nil) piping from the inside of the 42' l piping to eliminate the steam jet that would result under normal plant operation. This change prevents steam flow from the line and its
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- impingement on any nearby internal condenser components.
1 Summary 4
I. No. The connection to the steam piping has been plugged and seal welded
- in accordance with ISS-2000 and ASME/ ANSI B31.1. A normally closed
] valve, IN36-F0558A, on the outside of the turbine skirt prevents air 1 from leaking into the condenser. Potential leakage through this l normally closed valve would result in a small amount of in-leakage into the condenser through the 1/4" tube - thus, potentially decreasing the condenser vacuum slightly. However, this added j leakage is bounded by other more severe failure modes as discussed i in USAR Section 15.2.5. This change does not involve equipment i
required for mitigating accidents. There are no increased ,
radiological consequences for any accident previously evaluated in '
the USAR. As stated in USAR Section 10.4.1.4, the Main Condenser is >
] not required to support the safe shutdown of the reactor, or to support the operation of reactor safety features. Therefore, the '
probability of occurrence or the consequences of an accident or a I
malfunction of equipment important to safety previously evaluated in the USAR has not increased.
II. No. The condenser function is not altered by this change and no new equipment is being added. Equipment affected by the disposition of this NR is adequately designed to meet the requirements imposed by the condenser vacuum and Main Steam system. None of the affected components are required for the safe operation or shutdown of the plant. Therefore, the possibility of an accident or malfunction of equipment important to safety different than any previously evaluated in the USAR is not created.
III. No. The NR disposition is in accordance with the requirements of ISS-2000 and ASME/ ANSI B31.1. The condenser is not discussed in any of the Technical Specifications. Therefore, no margin of safety has been reduced.
SE 3.: 95-0134 l Source Document: USAR Change Request 95-092 i Description of Change This USAR change incorporates Regulatory Guide 1.101, Revision 3 with {
respect to the use of an acceptable alternative methodology for the '
development of Emergency Action Levels (EAL).
]
l Summary I. No. This Regulatory Guide deals with the development of EALs. It does not address or implement any changes to the plant or its operation. :
Accident analysis remains unchanged. This change does not impact i the ability of the emergency response organization to properly l classify an event or to assess its offsite consequences. The ;
effectiveness of the Emergency Plan, per 10CFR50.94(q), has not been reduced. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
l II. No. This Regulatory Guide deals with the development of EALs. It does not address or implement any changes to the plant or its operation. ;
Accident analysis remains unchanged. This change does not impact '
the ability of the emergency response organization to properly classify an event or to assess its offsite consequences. The effectiveness of the Emergency Plan, per 10CFR50.94(q), has not been reduced. Therefore, the possibility of an accident or malfunction i of a different type than any previously evaluated in the USAR is not '
created.
III. No. This Regulatory Guide deals with the development of EALs. It does not address or implement any changes to the plant or its operation.
Accident analysis remains unchanged. This change does not impact the ability of the emergency response organization to properly classify an event or to assess its offsite consequences. The effectiveness of the Emergency Plan, per 10CFR50.94(q), has not been reduced. Therefore, no margin of safety has been changed.
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SE No.: 95-0135 Source Document: DCN 4645 Description of Change This drawing change clarifies design pressures for the Service Water (P41) system return side piping and the Cooling Tower blowdown line
[ Circulating Water (N71) system] which are shown on several P& ids. The basis for the revised / reduced design pressures are calculations P41-27 and N71-13 which determined the maximum possible operating pressures for the various lines in question. The "new' design pressures simply correspond to the maximum operational pressures that can be seen in any of the respective lines. The revised design pressure information to be incorporated on the P& ids has no effect on the operation of the P41/N71 systems.
Summary I. No. These changes, which primarily clarify system operating / design data, have no affect on the integrity, function, operability or reliability of either system. The new defined design pressures are bounded by the original design pressures. These systems are nonsafety, non-radiological and are not required for mitigation of an accident or transient. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These changes have no effect on present system or plant operation.
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. Plant operation and effluents remain unaltered by these changes. As such, the Technical Specifications remain unaltered. Therefore, no margin of safety has been changed.
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- SE No.: 95-0136 Source Document: DCN 4824
] Description of Change i
This drawing change revises P&ID D912-608, Controlled Access and
- Miscellaneous Equipment Area HVAC System, by adding Master Parts I
! List (MPL) numbers to the fume hoods.
4 Summary i
i I. No. The nonsafety fume hoods located in Control Complex 599' elevation l
- are not associated with the initiation or mitigation of any of the t
- USAR accidents. The fume hoods are not classified as, or support ,
- any, equipment important to safety. This editorial change does not l t
affect containment integrity. Therefore, the probability of j occurrence or the consequences associated with an accident or i malfunction of equipment has not changed. ,
II. No. This activity does not result in any physical modifications to the j
' subject laboratory fume hoods, nor does the change affect any plant )
equipment. These hoods do not initiate, or support any equipment i that would initiate an accident. Therefore, the possibility of an
! accident or malfunction of a different type than any previously
- evaluated in the USAR is not created.
III. No. The fume hoods are neither safety-related nor do they support any equipment important to safety. No physical modifications are
$ performed in associated with this editorial change. Therefore, no margin of safety has been changed.
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3 SE No.: 95-0137
! Source Document: DCN 4830 Description of Change
- This drawing change will correct the direction of the exhaust damper ;
control signal arrows as identified on P&ID D912-619, Diesel Generator !
l Building Ventilation (DGBV) System, in order to show the as-built '
] condition.
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S_ummary
- 1. No. The DGBV system operates whenever the diesel generators operate.
i Specifically, the exhaust louvers require control signals indicating
- a diesel start in order to operate. The control signal to one of l
- the exhaust louver motor operators was not properly depicted on i
- drawing D912-619. Correcting the direction of the exhaust damper i j control signal arrows does not affect any equipment associated with l l the initiation or mitigation of any of the USAR accidents. This 3ctivity does not result in any physical modifications to the Diesel 32nerator Room Ventilation system. This activity does not degrade j the diesel generators ability to perform their safety-related function. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment i has not changed.
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II. No. This activity does not result in any physical modification to the i system. Correcting the direction of the exhaust damper control signal arrows on a P&ID does not degrade the design, safety function, or introduce new interfaces. Therefore, the possibility
- of an accident or malfunction of a different type than any
} previously evaluated in the USAR is not created.
l III. No. This activity does not result in any physical modifications to the i system. Consequently, correcting the direction of the exhaust damper control signal arrows does not reduce any margin of safety.
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! SE No.: 95-0138 i Source Document: DCN 4899 Description of Change i !
i This drawing change corrects a note on P&ID D302-671, Reactor Water i Cleanup (RWCU) System, concerning the containment isolation signal to s valves 1G33-F001, -F004, -F028, -F034, -F039, -F040, -F053, and -F054. '
i This is an editorial drawing change only with no field work being j performed. :
Summary i I. No. RWCU containment penetration inboard and outboard valves as noted i above are designed to isolate on an automatic isolation signal to j
! prevent the release of radiation during a design basis accident from ,
the reactor containment. There is no impact on the RWCU containment i isolation valves' ability to perform their design intent of .
} containment isolation during an accident. There are no physical ;
}
j changes to the plant configuration. Therefore, the probability of occurrence or the consequences associated with an accident or !'
malfunction of equipment has not changed.
l II. No. This drawing change does not affect the physical installation of the i
RWCU containment isolation valves or their ability to perform their
! design function. Therefore, the possibility of an accident or i
! malfunction of a different type than any previously evaluated in the i USAR is not created.
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i III. No. The drawing change corrects a drafting error only. As such, there are no changes to the limiting conditions for operations, l surveillance requirements, or design basis of the Technical Specifications. Therefore, no margin of safety has been changed.
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.; SE No.: 95-0139 q Source Document: DCN 3709 Description of Change i
This drawing change revises P&ID D302-963, Leak Detection System. The 4~
revision added missing valves to a table that identifies valves that I
close upon receipt of a steam leak detection high temperature isolation
! signal.
i Sumary
- I. No. This drawing change simply reflects the as installed design and t i plant conditions. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment
{
has not changed.
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i II. No. This drawing change makes the drawing consistent with other design base documents and the system design basis. Therefore, the i possibility of an accident or malfunction of a different type than 4 any previously evaluated in the USAR is not created. "
- III. No. Technical Specifications 3/4.3.2 and 3/4.6.4 are not impacted. The j drawing simply is being made consistent with other design base
- documents. Therefore, no margin of safety has been changed.
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SE No.: 95-0140 Source Document: DCN 4272 Description of Change '
This drawing change will add the two pitot traverse test locations to the P&ID D912-617, Fuel Handling Building Ventilation System.
Summary I. No. There is no field work associated with this change. The two pitot I traverse test locations are an integral feature of the existing Fuel Handling Building Ventilation system, which are furnished to facilitate flow testing of the system. There is no change to the functionality, operation, or system logic of the existing system as a result of this change. Per Section 9.4.2 of the USAR, this system does not support the safe shutdown of the plant. However, the ventilation system is required to mitigate a fuel handling accident.
There is no affect on any equipment associated with the initiation or mitigation of this event. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change does not affect the design, operation, or alter the performance of any existing plant equipment or system. This drawing change does not impact the design safety function of any equipment important to safety or introduce any new system interfaces. t Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. There is no physical modification to the Fuel Handling Ventilation system. The design change only reflects the as-built configuration of the system. Since there is no alteration to the design, operation, or system logic no margin of safety is reduced.
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SE No.: 95-0141 Source Document: PAP-0911, Rev. O SOI-M25/26, Rev. 4, TC-3 Description of Change This procedure and system operating instruction change permits the utilization of Generic Letter 91-018 and the increased in-leakage allowance of USAR Table 6.4-5 for the Control Room. The changes permits inspection, repair, and restoration of the Control Room Boundary (CRB) with the plant at power by permitting controlled breach of the boundary.
Summary I. No. The in-leakage during an accident condition is maintained by ensuring that closure is capable within the limitations provided by USAR Table 6.4-5. The shielding provided to the Control Room operators will not be affected. The breach will be controlled by fire impairments when applicable. Habitability of the Control Room will be maintained within the guidelines of GDC 19 of 10CFR50.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Loss of the Control Room is already analyzed in the USAR. The provisions of the referenced documents will minimize the potential for such an event by permitting inspections and repairs using approved materials. The shielding provided to the Control Room operators will not be affected. The breach will be controlled by fire impairments when applicable. Habitability of the Control Room will be maintained within the guidelines of GDC 19 of 10CFR50.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. The in-leakage during an accident condition is maintained by ensuring that closure is capable within the limitations provided by USAR Table 6.4-5. The shielding provided to the Control Room operators will not be affected. The breach will be controlled by fire impairments when applicable. Habitability of the Control Room will be maintained within the guidelines of GDC 19 of 10CFR50.
Therefore, no margin of safety has been changed.
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SE No.: 95-0142 Source Document: DCN 4285 Description of Change This drawing change corrects the valve symbol for 1E51-F0510 from a gate valve to a globe valve on P&ID D302-632, Reactor Core Isolation Cooling (RCIC) system. .
Summary I. No. Valve 1E51-F0510 is a globe valve installed in the steam supply line to the RCIC turbine. During normal plant operation, this valve is maintained in the open position and is used to isolate and throttle steam flow to the RCIC turbine. The valve's safety isolation function and design capability to throttle steam flow is not affected as a globe valve. The editorial correction of the valve symbol from a gate valve to a globe valve on the P&ID does not affect the installed configuration or function of 1E51-F0510. There is no affect on the RCIC system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The change is an editorial drafting change only. It does not affect the function, operation, or accident response capabilities of the RCIC system. Therefore, this change will not create the possibility of an accident or malfunction of a different type than any .
previously evaluated in the USAR.
III. No. The editorial correction of the valve symbol does not affect the configuration of the RCIC system. Therefore, no margin of safety has been changed.
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SE No.: 95-0143 Source Document: DCN 4398 Description of Change This drawing change revises various ventilation system drawings to l incorporate design information.
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I. No. The design of the system in question as a result of this change are l l unaffected. The fume hoods and associated dampers addressed by this l change are nonsafety-related, serve no safety functions, and do not l affect any equipment important to safety. The changes do not degrade or prevent actions described or assumed in any accident discussed in the USAR. The fume hoods and associated dampers will ;
continue to meet the original design for materials and i constructions. Therefore, the probability of occurrence or the ,
consequences associated with an accident or malfunction of equipment !
has not changed.
II. No. The overall design of the systems affected as a result of this drawing change are not impacted. Since the fume hoods and associated dampers addressed by this change will continue to meet i the original design intent for these systems, no new failure mechanisms have been created. This change is not associated with ]'
any physical modifications to any plant equipment which could initiate an accident. Therefore, the possibility of an accident or a malfunction of a dif'lerent type than any previously evaluated in the USAR is not created.
i III. No. The fume hoods and associated dampers addressed by this-change are nonsafety-related, serve no safety-related functions and do not :
affect any equipment important to safety. The design of the systems affected are maintained. The fume hoods and associated dampers will continue to meet their original design intent. Additionally, the bases for Technical Specifications 3/4.11.2.4 and 3/4.11.2.5 are not impacted. Consequently, this change does not reduce any margin of safety.
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SE No.: 95-0144 Source Document: DCN 4407 Description of Change This drawing change revises P&ID D302-172, Two-Bed Demineralizer (P21)
System, to depict the proper port configuration for solenoid valves OP21-F376 and OP21-F311.
Summary I. No. Valves OP21-F375 and OP21-F310 are drain valves which allow flushing / draining of P21 system piping following maintenance activities. This is accomplished by energizing its associated solenoid valve (0P21-F376 or OP21-F311) to isolate the supply air and vent air off thereby causing the drain valve to open. The P21 system is nonsafety-related. There are no physical modifications made to any plant system as a result of activities associated with this drawing change. Depicting the proper as-built configuration of valves OP21-F376 and OP21-F311 does not affect any equipment associated with the initiation or mitigation of any of the USAR acc Nnts. The original design intent of the P21 system is mair.. led. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The solenoid valves addressed by this drawing change will continue to meet the original design intent for the P21 system. No new failure mechanisms have been created. There are no physical modifications made to any plant system. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Two-Bed Demineralizer system is nonsafety-related. The solenoid valves serve no safety-related function and do not affect any equipment important to safety. The overall design of the P21 system is unchanged. Therefore, no margin of safety has been changed.
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i SE No.: 95-0145 Source Document:
s DCN 5200 '
Description of Change This drawing change revises P&ID D302-131, Condenser Air Removal System.
The revision incorporates the changes made to the plant per Mechanical Foreign Item (MFI) 1-89-244, which provided manual control capabilities for the hot water heat exchanger drain valve IN26-F0200. A three-way isolation valve has been added to the system to allow isolation of the level control signal from the level controller to the drain valve, and allows reduced plant instrument air to provide a manual signal to the ,
drain valve. This capability is to be used during startup conditions and ;
low power operation, when the water level in the hot water heat exchanger '
is unstable.
Summary I. No. This drawing change makes permanent MFI 1-89-244 which permits '
manual control of 1N26-F0200. No new failure modes or effects have i been created that affect equipment important to safety or safe shutdown capability. Although operation of the system has been slightly altered (addition of manual control for valve IN26-F0200),
the automatic and normal level control functions, and other ,
instrumentation associated with alarms and controls for the hot :
water heat exchanger have not been impacted. Therefore, the !
probability of occurrence or the consequences associated with an !
accident or malfunction of equipment has not changed. l II. No. This drawing change makes permanent MFI 1-89-244 which permits manual control of 1N26-F0200. No new failure modes or effects have been created that affect equipment important to safety or safe '
shutdown capability. Although operation of the system has been slightly altered (addition of manual control for valve IN26-F0200),
the automatic and normal level control functions, and other instrumentation associated with alarms and controls for the hot water heat exchanger have not been impacted. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The drawing revision does not make any changes to equipment that is safety-related or required for safe plant operation or safe shutdown of the plant. The equipment related to the manual operation of the hot water heat exchanger, and the hot water heat exchanger itself are not mentioned in the Technical Specifications. Therefore, no margin of safety has been changed.
SE No.: 95-0146 Source Document: DCN 3547 Description of Change This drawing change revises P&ID D806-022, Plant Radiation Monitoring.
The change involves the correction of a continuation flag that incorrectly depicted a drawing reference.
Summary I. No. The drawing change has had no impact upon system operability and was only correcting a drawing error. The drawing revision did not introduce any new failure modes or accident initiators. The operation of the radiation sampling system and monitoring capabilities have not changed. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The revision to the continuation flag on P&ID D806-022 does not make any physical changes to the plant. The equipment involved with the drawing revision is nonsafety-related and is not associated with safety operation or safe shutdown of the plant. The system operation was not affected. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The drawing change does not make any physical changes to the plant, nor was the operation of the system changed or impacted. The effluent radiation limits or sampling methods given in Technical Specification 3/4.11.2 were not changed. Therefore, the drawing change does not reduce any margin of safety.
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% Source Document: I DCN 3813 3 o 4C t Description'of: Chance:
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,,i$ y Q. 5. 7f, - - Mis' drawing change revises various Condensate System Filtration (N32)
- l psyng.p,. ge change updates the valve limit switch and solenoid valve if E. tinterlock logic associated with various system control valves in order 2put;them intoLeompliance with the current plant conditions and the p# ~ - '
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s Gas-built: elementary wiring diagrams which correctly depict the system
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4 . .I <U No. Misidrawing change was essentially editorial and involved no I
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% g imarks for the valve interlock logic. The N23 system is xx kly nonsafety-related and is not relied upon for safe shutdown, or safe g~& 4 (operation of the plant. Therefore, the probability of occurrence or ithe: consequences associated with an accident or malfunction of
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Onhw . equipment has not changed.-
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drawingwiring the; elementary change corrected diagrams the that were valve interlock determined logic to agree to agree with m gV" ' ' 'thalplant: configuration. The change made no physical alteration of Rb*W
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in
- the" plant.c: Merefore, the possibility of an accident or malfunction
!.of a different type than any previously evaluated in the USAR is not
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compliance with.the plant configuration and the elementary wiring a --diagrams...The operation of'the system was not altered. There was h',- .a 7.no impact"upon the Technical Specifications. Therefore, no margin l l4 ,m ?, :ofJaafety has been changed.
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1 SE No.: 95-0148 Source Document: DCP 92-0097A, Rev. O Description of Change This design change modifies the control circuitry of the Residual Heat Removal (RHR) heat exchanger outlet valves 1E12-F003A/B to automatically stroke open on receipt of a Low Pressure Coolant Injection (LPCI) 1 initiation signal. I Summary I. No. These modifications cannot effect Chapter 15 analyzed accidents, and will not adversely affect any other mode of RHR. The design change l maintains the requirements of the original equipment design and construction codes, system design bases and equipment qualification requirements. No common mode / common cause failures are probable due to the separation criteria applied the design. The design continues to meet single failure criteria. Compliance with Regulatory Guide 1.75 and IEEE Standard 384 for physical independence is not compromised. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to i safety is not increased. '
II. No. The design change creates no now systems, introduces no new equipment types, and continues to maintain the requirements of RHR !
accident mitigation. The design meets the single failure criteria !
(i.e. redundant RHR loops), and does not impact channel and electrical separation requirements. Thus, common mode / common cause l
failures are not credible. Therefore, the possibility for an '
accident or malfunction of a different type is not created.
III. No. These modifications cannot effect Chapter 15 analyzed accidents, and will not adversely affect any other mode of RHR. The design change ,
maintains the requirements of the original equipment design and '
construction codes, system design bases and equipment qualification requirements. No comnon mode / common cause failures are probable due i
to the separation criteria applied the design. The design continues to meet single failure criteria. Compliance with Regulatory Guide 1.75 and IEEE Standard 384 for physical independence is not compromised. Therefore, no margin of safety has been changed.
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l SE No.: 95-0149 Source Document: DCP 92-0060, Rev. 1 3
Description of Change a
I This design change converts Motor Operated Valves (MOV) 0242-F0315A, OP42-F0315B, and OP42-F0315C to manually operated valves. Valves, actuators, and motors will remain installed. Circuits are lifted to j~ disable both the power and control function of the MOVs. The cables and raceways are spared in place, and the valves will be operated manually by their handwheels. This design change removes the automatic control l features for the MOVs. l Summary-1
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i I. No. The accident relevant to this design change is a loss of cooling to the Control Complex chillers. A failure analysis was performed and shows adequate cooling will be maintained. The electrical power has i been removed from these valves. Since the desired position for safe shutdown is normally open, no closure is required during loss of power or LOCA. Therefore, the probability of occurrence or the i
consequences associated with an accident or malfunction of equipment
, has not changed.
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- II. No. This design change eliminates the auto mode of operation. This l 4
design change is such that the failure of OP42-F315A, B, or C will not prevent the system from performing its safety function.
i Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not
, created.
i III. No. This change does not affect the previously evaluated manual mode of ;
I operation. Flow through the Control Complex chiller will be I
! controlled by existing plant procedures. The Control Complex !
j chillers will still function as designed. Therefore, no margin of '
- safety has been changed.
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I SE No.: 95-0148 i Source Document: DCP 92-0097A, Rev. O Description of Change This design change modifies the control circuitry of the Residual Heat !
Removal (RHR) heat exchar.ger outlet valves lE12-F003A/B to automatically stroke open on re:itpt of a Low Pressure Coolant Injection (LPCI) initiation signal.
j Summary I. No. These modifications cannot effect Chapter 15 analyzed accidents, and will not adversely affect any other mode of RHR. The design change maintains the requirements of the original equipment design and construction codes, system design bases and equipment qualification requirements. No common mode / common cause failures are probable due to the separation criteria applied the design. The design continues to meet single failure criteria. Compliance with Regulatory Guide 1.75 and IEEE Standard 384 for physical independence is not compromised. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety is not increased.
II. No. The design change creates no new systems, introduces no new ;
equipment types, and continues to maintain the requirements of RHR 1 accident mitigation. The design meets the single failure criteria (i.e. redundant RHR loops), and does not impact channel and !
electrical separation requirements. Thus, common mode / common cause ;
failures are not credible. Therefore, the possibility for an j accident or malfunction of a different type is not created. ;
III. No. These modifications cannot effect Chapter 15 analyzed accidents, and will not adversely affect any other mode of RHR. The design change maintains the requirements of the original equipment design and construction codes, system design bases and equipment qualification requirements. No common mode / common cause failures are probable due to the separation criteria applied the design. The design continues to meet single failure criteria. Compliance with Regulatory Guide 1.75 and IEEE Standard 384 for physical independence is not compromised. Therefore, no margin of safety has been changed.
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SE No.: 95-0150 Source Document: DCN 4461 Description of Change This drawing change corrects drafting errors on P&ID D302-212, Service Water System. The errors involve incorrect Master Parts List (MPL) numbers.
Summary:
I. No. These drafting changes do not alter the design, materials or construction standards of the plant. They do not alter the function of any plant components. Therefore, these changes cannot affect the probability or consequences of any accident or malfunction previously evaluated in the USAR.
II. No. These drafting changes cannot alter the design or function of any plant components. Therefore, this drawing change cannot create any new type of accident or malfunction.
III. No. This drawing change corrects drafting errors. It does not alter the design or function of any plant equipment. It will have no effect 1
on the Technical Specifications. Therefore, no margin of safety has ;
been reduced.
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SE No.: 95-0151 Source Document: DCN 5210 Description of Change i i
This drawing change revises P&ID D302-335, Zinc Injection Passivation System, to show that the reactor water zinc sample station is abandoned '
in place. This station is isolated and is not required, i Summary I. No. The position of the valves associated with the abandoned station are controlled under approved plant procedures such that the station is i isolated from the Reactor Coolant System ;RCS). RCS water chemistry 1 is adequately monitored by grab samples and thus, the integrity of the RCS is not affected. The station is not connected to any system j
required for safe shutdown of the plant. Therefore, neither the '
probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased.
II. No. The drawing change does not require any field work or testing. The zine analyzer and associated equipment will be isolated from plant systems by manual valves. The connections do not affect the reactor coolant pressure boundary. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. The Zinc Injection system is not addressed in Technical Specifications. Grab samples provide an acceptable alternative to on-line analysis. This drawing change does not alter any water chemistry requirements or associated limits. Thus, Technical Specifications 3/4.4.4 and 3/4.4.5 are not affected. Therefore, this change does not reduce any margin of safety. l I
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SE No.: 95-0152 Source Document: DCN 4729 Description of Change This drawing change revises P&ID D302-791, amergency Service Water (ESW)
System, by removing remove the cap shown on normally open valve 1P45-F0605. Piping isometrics SS-304-796-103.2 and SS-304-797-102.2 are also being revised to not show caps. These valves are used as vent valves for the ESW standpipes ar.d are used to prevent waterhanner.
Summary i I. No. This drawing change does not alter the design of the ESW system or the function of any of it's components. There is no impact upon the design or function of any other plant system which may rely upon the ESW system. Therefore, this change cannot affect the probability or consequences of any accident or malfunction of equipment previously evaluated in the USAR.
II. No. This drawing change does not alter the design or function of the ESW system or any other plant systems relying upon the ESW system.
Therefore, this change cannot create any new type of accident or malfunction not already evaluated.
III. No. This drawing revision makes no changes to the design or function of any plant system. The basis for Technical Specification 3/4.7.1 is unaffected. Therefore, no margin of safety has been changed.
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I SE No.: 95-0153 Source Document: DCN 5129 Description of Change i
l This drawing change corrects a drafting error on P&ID D302-359, (
Division III Diesel Lube Oil System, which incorrectly showed the l Division III diesel lube oil strainer sump drain line connected to valve IE22-F558.
l Summary I. No. This drawing change does not alter the physical plant. Neither the
, method through which the Division III diesel generator system ,
i performs its safety functions, nor any USAR prescribed methodology associated with system operations is affected by this change. The I change corrects a drafting error such that the existing design configuration is correctly depicted on associated drawings.
Therefore, neither the probability of occurrence nor the j consequences of a previously analyzed accident or malfunction of :
i equipment will be increased.
l II. No. The drawing change does not require any physical changes to the plant and does not represent an operational or procedural change. l Further, correction of the drafting error will not affect plant i procedures or operating methods. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluatM III. No. This change does not impact the operation of the Division III diesel generator or its supporting systems. Technical Specifications 3/4.5, ECCS Systems, and 3/4.8, Electrical Power Systems, are not affected. Therefore, there is no reduction in any margin of safety.
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SE No.: 95-0154 Source Document: DCN 5112 Description of Change 1
This drawing change corrects a valve numbering error on P& ids D302-009 and B208-141 Sheet 123, Offgas (N64) System, which incorrectly showed two valves with the same Master Parts List (MPL) number.
Summary I. No. This change does not alter the physical plant. Neither the method through which the N64 system performs its functions, nor any USAR prescribed methodology associated with system operatiens is affected by this change. No plant procedures are required t< De revised since the valve to be re-numbered is not specifica12y addressed in any existing procedures. In addition, no other systems are affected by this change. Therefore, neither the probability of occurrence-nor the consequences of a previously analyzed or malfunction of equipment accident will be increased.
II. No. The proposed drawing change does not require any physical changes to the plant and does not represent an operational or procedural change. Further, correction of the numbering error will not affect plant procedures or operating methods. Therefore, this change does not create the possibility of an accident or malfunction of a different type than any previously evaluated. l III. No. The N64 system is nonsafety-related and is not a Technical Specification system. Correcting of the valve numbering has no impact on the operation of the N64 system or any other plant system.
Therefore, there is no reduction in any margin of safety.
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l SE No.: 95-0155 1 Source Document: DCN 4491 )
Description of Change This drawing change updates various design drawings to show the .tework of two steel beam connections. This rework consisted of replacing the untorqued A325 bolts with A325-F torqued friction type bolts. This rework was necessary to eliminate overstress in some of the Auxiliary 2
Building 574'-10" elevation platform steel members. The overstresses were found acceptable for past operability using the actual material strength of the steel members and increase in allowable stresses given in the American Society of Mechanical Engineers Code Section III, Appendix F, Section F-1334.
Summary I. No. Past operability and long term continued operation of the 574'-10" platform steel members has been evaluated and qualified in calculation file code 4:14.1 Rev. 11 through 13 for all loading
. conditions resulting from accident conditions previously evaluated in the USAR. There is no affect on any fission oroduct barriers, and no increase in dose to the public or onsite doses that would impede actions necessary to mitigate the consequences of accidents.
Therefore, there is no increase in the probability or consequences of an accident or malfunction of equipment previously evaluated in the USAR.
II. No. As stated above, the platform design remained adequate for past operability and long term continued operation. There is no change to the function of any existing equipment, and there are no new loading conditions as a result of this change. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Based on calculation file code 4:14.1, Rev. 11 through 13, the safety function performed by the platform remains unaffected since the steel members were found to be adequate based on the actual material strength properties. There is no reduction in any margin of safety.
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l l SE No.: 95-0156
! Source Document: DCN 3193 l
Description of Change l This drawing change revises P&ID D302-613, Nuclear Closed Cooling System, i to show that the three drywell coolers, M13-B001, M13-B002 and M13-B003, i each have two independent coils.
l Summary I. No. All of the changes involved in this drawing change are editorial in I nature. It revises the Master Parts List (MPL) designations of the l
' coolers to correctly show their configuration. These changes have no effect on the design or operation of any plant equipment. l l Therefore, this change cannot affect the probability or consequences of any accident or malfunctions of equipment previously evaluated in l the USAR.
II. No. This change constitutes documentation revisions only. It clarifies the MPLs of three coolers. As such, it will not have any effect on I
' the operation of any plant components. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The components involved in this change are not specifically addressed in the Technical Specifications. The changes are editorial in nature and will have no effect on the operation of any plant components. Therefore, no margin of safety has been changed.
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i SE No.: 95-0157 Sjurce Document: DCN 5122 Description of Change This drawing change revises P&ID D302-161, Makeup Water Pretreatment !
(P20) System, to show the coagulator effluent turbidity, influent flow I recorder, and conductivity instrumentation abandoned in place. This instrumentation is obsolete. Other methods, such as plant rounds and )
j l periodic sampling, can be used to monitor the P20 system performance. i Summary I I. No. The P20 system will continue to meet its original functional intent with the instrumentation abandoned in place. This instrumentation is not needed to obtain the water chemistry monitoring results since the Chemistry Unit can obtain the results via plant rounds that are i controlled by existing plant procedures. Abandoning the ,
instrumentation in place does not create any new failure modes or
- effects. The abandoned instrumentation is nonsafety-related and is not related to systems, structures, or components that are important f to safety. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment '
has not changed.
II. No. The Chemistry Unit has the capability to monitor the water chemistry parameters required to operate the P20 system. This monitoring is achieved through existing plant procedures and instructions. This !
drawing change does not involve any hardware changes. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The instrumentation that was abandoned by the drawing change is not mentioned in the plant Technical Specifications. P20 system ;
operability is maintained. Therefore, no margin of safety has been l changed.
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l SE No.: 95-0158 Source Document: DCN 4426 Description of Change This drawing change revises P& ids D302-603 and -604, Reactor Recirculation (B33) System, to show the jet pump flow indicators 1B33-R0611A, B, C, and D the same as on other controlled drawings such as elementary drawing B208-016-B02. The purpose of these indicators is to provide indication to the Control Room operator of actual flow through the diffuser and to check the operation of the other jet pumps. There is no control or alarm function involved with this loop.
Summary I. No. This is a drawing only change. No physical changes were made to the plant. The design basis and function of the B33 jet pumps is not affected by this drawing change. The drawing change merely adds design basis information that is bounded by USAR accident analysic and shown on other design basis documents. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changej.
II. No. The information being added to the P&ID drawings iJ consistent with the B33 system design basis. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The basis of the function of the jet pump indicators is determined by engineering considerations for protection and calibration of equipment that is not safety-related. No physical changes were made to the plant. This is a document only change. Therefore, no margin of safety has been changed.
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SE No.: 95-0159 Source Document: DCN 3689 ,
Description of Change This drawing change corrects a discrepancy on P&ID D302-651, Fuel Pool !
Cooling and Cleanup System. Specifically, the P&ID shows the siphon l break lines terminating at the fuel transfer /dryar storage pool, but the lines actually terminate in the separator storage well.
l Summary l
I. No. This drawing change shows the as-built termination point of the two la siphon breaker lines. These siphon breaker lines are installed ,
l to prevent inadvertent pool drainage in the event of a pipe rupture.
This change does not prevent the siphon breaker lines from i
I performing their design function. There are no physical plant changes resulting from this drawing change. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR has not changed.
II. No. This change eliminates a discrepancy between several engineering documents. It does not alter the function, operation, or design basis of the Fuel Pool Cooling and Cleanup (G41) or the Residual Heat Removal (E12) systems. There are no physical plant changes associated with this drawing change. The change is limited to i providing correct and consistent design information. Therefore, the '
possibility of an accident or malfunction of a different type than i any previously evaluated in the USAR is not created. I III. No. Technical Specifications 3/4.4.9, 3/4.9.9, and 3/4.9.11 are not impacted by this change. This drawing change is limited to providing engineering documentation that is accurate and consistent, and represents the current acceptable design and licensing basis for the Fuel Pool Cooling and Cleanup and the E12 systems. The performance, operational requirements, design bases, and safety limits of these systems have not changed. Accordingly, no margin of safety has been reduced.
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SE No.: 95-0160 Source Document: DCN 4336 Description of Change This drawing change revises P&ID D302-793, Emergency Service Water ,
Operating Data, to depict the correct normal operating flow rate of the l Emergency Service Water (ESW) flow to the Fuel Pool Cooling Heat l Exchanger. The flow rate shown is 2000 GPM, with a note stating that the minimum design basis flow is 500 GPM. Since the flows to the Fuel Pool Cooling Heat Exchanger are not throttled, they are typically higher than the 500 GPM. To avoid potential confusion indicating that the 2000 GPM is minimum, the drawing is being revised to show the design minimum value of 500 GPM.
I Summary i I. No. The safety-related function of the ESW system is a source of cooling and makeup water to cope with various plant accident conditions.
ESW is also a safety-related source of cooling water to the Fuel Pool Heat Exchangers. Calculations indicate that the minimum ESW flow required to maintain fuel pool temperature below 1500 is 500 GPM. This drawing change does not change this value. ESW system design or operation are not affected. Therefore, the probability of occurrence or the consequences associated with an i accident or malfunction of equipment has not changed.
II. No. As described above, this change does not alter the design or operation of the ESW system. The system will niaintain its safety functions as intended and described in the USAR. 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. Operation of the ESW system is unchanged. Technical Specification 3/4.7.1 is not affected. The expected performance of the ESW system will be more clearly described as a result of this change, thereby avoiding confusion as to the ESW operability. Calculations indicate that the 500 GPM flow is adequate to maintain fuel cooling as described in Technical Specification 5.6.3. Therefore, no margin of safety has been reduced.
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i SE No.: 95-0161 Source Document: SCRs 1-94-1953 and 1-94-1954 !
Description of Change These setpoint changes will decrease the setpoint from 68 GPM to 59 GPM .
and increase the leave-as-is-zone from 15.4 GPM to 18.8 GPM for the Reactor Water Cleanup (RWCU) Isolation Differential High 1E31-N0609A and 1E31-N0609B trip function, i Summary j I. No. These setpoint changes do not affect the design basis of the RWCU '
system, or any other safety-related or nonsafety-related plant ;
system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These setpoint changes lower the setpoint in the conservative direction for the RWCU system high 6ifferential flow signal in order to maintain the Technical Specification allowable limit of 68 GPM.
The leave-as-is value is being increased for calibration purposes and requires that the setpoint be lowered by 9 GPM in order to maintain the margin between the existing allowable value and the upper leave-as-is value. Therefore, the possibility of an accident or malfunction of a different type than.any previously evaluated in the USAR is not created.
III. No. The purpose of the instrumentation is to detect and isolate the RWCU system based on sensing a gross failure such as a line break or other loss of flow. Even through the setpoint change is in the conservative direction, it is well above the normal flow rates of 5-14 GPM. RWCU design is not affected by this change. Therefore, no margin of safety has been changed. !
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SE No.: 95-0162 Source Document: DCN 3426 j Description of Change t
This drawing change revises various drawings associated with the diesel generators. The change includes revising various Master Parts List (MPL) t designations and deletion of several pressure gauges.
Summary I. No. The operation and design of the diesel generators are maintained.
The changes do not degrade or prevent actions described or assumed in USAR Chapter 15. The injectors, instrument snubbers, and previously deleted pressure gauges addressed by this change will continue to meet the original design for materials and construction.
Therefore, the probability of occurrence or the consequences !
associated with an accident or malfunction of equipment has not changed.
II. No. The design or operation of the diesel generators is not impacted by this change. No new failure mechanisms have been created. This change has not made any physical modification to plant equipment.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specifications 3/4.5.1 (ECCS Operation) and 3/4.8.1 (AC Sources) have not been affected. The injectors, instrument snubbers and previously deleted pressure gauges addressed by this drawing change will continue to meet their original design intent.
Therefore, no margin of safety has been changed. ,
SE No.: 95-0163 Source Document: DCN 4748 Descrir; tion of Change This drawing change revises P&ID D302-961, Leak Detection (E31) System, by restoring an inadvertently removed piping cross reference flag. The piping is cross referenced from P&ID D912-604. The change is an editorial drawing change only with no field work being performed.
Summary I. No. The change restores a piping cross reference flag on P&ID D302-961.
There is no physical change to the plant configuration or hardware.
There is no impact to the Leak Detection or the Containment Vessel and Drywell Purge systems' abilities to perform their design function. Therefore, the probability of an occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR has not changed.
II. No. The change is limited to a drafting correction to reinstate an inadvertently removed piping cross reference flag on P&ID D302-961.
This change does not change the physical installation of the Leak Detection or the Containment Vessel and Drywell Purge systems, nor does it change their ability to perform their design functions.
Therefore, this change will not create the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR.
III. No. The change ccrrects an error on P&ID D302-961. This change does not change the physical installation of the Leak Detection or the Containment Vessel and Drywell Purge systems, nor does it change their ability to perform their design functions. Therefore, no margin of safety has been reduced.
SE No.: 95-0164 Source Document: Im 92-SO931, Rev. 3 NR 92-S-232, Rev. 2 i Description of Change These Nonconformance Reports (NR) evaluate the temporary "use-as-is' disposition of a six-inch relief discharge header that does not conform to the ASME Code. The NRs require that the Steam Condensing Mode (SCM) 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 safety 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 different type than any previously evaluated.
III. No. The margin of safety as defined in the bases to Technical Specifications 3/4.6.3 and 3/4.6.1.5 refers to design parameters for the suppression pool and to the structural integrity of containment, respectively. 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.
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SE No.: 95-0165 Source Document: DCP 95-5096, Rev. O i
- Description of Change ,
i This design change reconfigures the one-out-of-two-taken-twice control [
logic associated with the automatic startup of the Reactor Core Isolation
- Cooling (RCIC-E51) system, upon receipt of reactor low water level i signals. The one-out-of-two-taken-twice features are retained. The .
1 relay contacts making up this logic are rewired so that the two contacts '
! associated with Division I low level trip units are wired in parallel with each other and the two contacts associated with Division II low
- level trip units are wired in parallel with each other. The new contact j configuration eliminates an inadvertent startup of the RCIC system if an
! instrument power supply failure occurs.
l Summary
- I. No. RCIC operation is associated with the safe shutdown function to i maintain vessel inventory. It is not a potential initiator of any j accident. Existing analyses do not relay upon nor take credit for RCIC in the mitigation of accidents or transients. Per a failure mode and' effects analysis, the design change has no potential for 1 causing increased failures of associated reactor level .
i' l instrumentation, power supplies, relays and wiring. This design and
- installation ensures that equipment qualification is maintained,
] combustible loading is not increased and there is no other effect on i structures, systems or components used to mitigate the consequences i of accidents evaluated in the USAR. Therefore, the probability of
- occurrence or the consequences of an accident or malfunction of '
- equipment important<to safety previously evaluated in the USAR is j not increased.
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II. No. The RCIC initiation logic supports the plant response to an i accident. It is not capable of causing an accident. Per analysis, 4
the logic changes have no potential for causing increased failures
- of associated reactor level instrumentation, power supplies, relays
! and wiring. The method of design and installation ensures that j equipment qualification is maintained, combustible loading is not 4
increased and no other safe shutdown circuits are affected. There i is no impact on the ability of the system to provide makeup water to the vessel. Therefore, the possibility of an accident or
?! malfunction of a different type than any evaluated previously in the
) USAR is not created.
i III. No. Engineering evaluation concludes that the system will acceptably support the response to all plant accider.ts, transients and conditions without diminished capability. The minimum number of RCIC operable channels per trip system, and associated operability and action statements have not been changed. The operability requirements for Emergency Core Cooling Systems which includes RCIC in operational Modes 1, 2 and 3 are unchanged. The margin of safety has not changed.
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-SE No.: 95-0166 Source Document: DCN 4861 Description of Change This drawing change revises P&ID D302-811, Containment Integrated Leak Rate Testing (E61), to correct a discrepancy by showing system pressure ;
taps as being sealed with threaded caps instead of. flanges. -
Summary 1
I. No. The overall design of the E61 system is unchanged as a result of l this drawing change. The threaded caps are passive components that t are not part of an active system. The pressure taps will continue to meet the original design for materials and construction. The use ;
of threaded caps in this application will not affect containment '
integrity or any of the plant equipment credited with mitigation of any of the USAR accidents. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR has not changed.
II. No. The overall design of the E61 system is unchanged as a result of this this drawing change. The threaded caps are passive components and not part of an active system. Since the pressure taps will continue to meet the original design intent for the E61 system, no new failure mechanisms have been created. This drawing change is i not associated with any physical modification to any plant equipment '
which could initiate an accident. The change does not affect any equipment important to safety. Therefore, the activity will not create the possibility of an accident or malfunction of a different type than previously evaluated in the USAR.
III. No. Technical Specifications 3/4.6.1 and 3/4.6.4 are concerned with testing and maintaining the containment leak tight. The threaded ,
j caps are acceptable containment pressure boundary components. They will not impact the regular containment leakrate testing (either Type A or Type C). Hence, the leaktightness of the containment and ,
the ability to test it are not impacted by using threaded caps in this application. Therefore, no margin of safety has been reduced.
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1 SE No.: 95-0167 l Source Document: USAR Change Request 95-094 i
- Description of Change This change request corrects textual errors which describe the independently powered control loop for the Hot Surge Tank (HST) level control valves 1N21-F220 and 1N21-F230. No hardware changes in the field
] will be made by this change request.
Summary i I. No. There is no accident in USAR Chapter 15 which discusses the loss of
{ HST level control. The addition of an independent control loop and
! power source will increase the reliability of the Condensate system i and will in no way impact overall Condensate system pe:-formance.
Normal flow paths of the Condensate and Feedwater systems will not be altered or degraded. The provision of an ladependently powered control loop to the Condensate system will not adversely affect i
' interfacing systems. No new radiological flow paths are introduced.
By adding a similar design independent control, the failure of one l j
power supply does not adversely impact Condensate System control, i j The equipment specified has the same failure modes as that which is '
4 presently installed. Therefore, the probability of occurrence or the consequences associated with an accident or .nalfunction of l equipment has not changed.
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II. No. There is no adverse impact in the control of heated and dearated condensate from the Direct Contact Heater to the HST. No credible accident may be expected as a result of the implementation of this activity. The design integrity of the Condensate system remains unaltered. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not j created.
III. No. HST level control and the IN21-F220/F230 control valves are not covered by Technical Specifications. The portion of the Condensate i system covered by the change does not affect any Technical i
Specification. The design integrity of the Condensate system
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remains unaltered. Therefore, no margin of safety has been changed.
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i SE No.: 95-0168 Source Document: DCN 4029 j
Description of Change This drawing change corrects a drafting error which shows duplicate i valves on P& ids D302-643, Residual Heat Removal (RHR) System, and I D302-655, Fuel Pool Cooling and Cleanup (FPCC) System.
Summary I. No. This drawing change corrects two P& IDS to eliminate duplicate valves. There is no physical change to plant configuration or hardware. As such, there is no impact on the design of the RHR or FPCC systems. Therefore, the probability of occurrence or the t consequences associated with an accident or malfunction of equipment has not changed.
II. No. Only one set of valves were accounted for in the original design and are installed. The elimination of the incorrectly depicted duplicate set of valves from the P& ids cannot impact the design, operation, or function of the RHR or FPCC systems. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change eliminates duplicate valves erroneously shown on one of two P& ids. As such, there is no change to the limiting conditions for operation, surveillance requirements or design basis of the Technical Specifications. The design, operation, or function of the RHR or FPCC systems have not been impacted. Therefore,-no margin of safety has been changed.
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SE No.: 95-0169 i Source Document: DCN 4850 i Description of Change ,
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This drawing changes revises P& ids D302-641, Residual Heat Removal (RHR)
System, and D302-705, Low Pressure Core Spray (LPCS) System, to reflect i the proper configuration of the packing gland for valves 1E12-F0021 and !
1E21-F0012. ,
Summary I. No. The appropriate notes on the P& ids associated with valves 1E12-F0021 l and 1E21-F0012 are being revised to allow the packing gland (s) to be '
on the downstream side of the disk provided the packing is tested for leakage. Since.the design of the valve used is such that the design flow is in a direction that places the packing in the containment leakage paths, this alternative is acceptable provided 3 the packing is monitored for leakage. These changes do not affect '
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any equipment associated with the initiation or mitigation of any of I the USAR accidents. The valves and systems associated with the
! l affected drawings will continue to meet the original design for i materials and construction. Therefore, the probability of I occurrence or the consequences associated with an accident or 1 malfunction of equipment has not changed. 1 II. No. The valves and systems associated with the affected drawings will continue to meet the original design. No new failure mechanisms have been created. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The overall design of the RHR and LPCS systems remain unchanged.
The changes addressed by this drawing change do not affect any equipment important to safety. Technical Specifications 3/4.4,9 (RHR), 3/4.9.11 (RHR), and 3/4.5 (LPCS) have not been impacted.
Therefore, no margin of safety has been changed.
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SE No.: 95-0170 Source Document: USAR Change Request 95-095 Description of Change This change request eliminates USAR Table 13.5-1 which is a listing of plant sections responsible for procedure / instruction preparation.
Summary I. No. This change request revises the representation of an administrative process. The process still complies with Regulatory Guide 1.33 and NUREG-0800. Qualified personnel will still prepare the procedures / instructions. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change request revises the representation of an administrative process. The process still complies with Regulatory Guide 1.33 and NUREG-0800. Qualified personnel will still prepare the procedures / instructions. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change request revises the representation of an adainistrative process. The process still complies with Regulatory Guide 1.33 and NUREG-0800. Qualified personnel will still prepare the procedures / instructions. Therefore, no margin of safety has been changed.
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1 SE No.: 95-0171 Source Document: SMRF 95-5087, Rev. 0 '
Description of Change l This design change modifies the Two-Bed (P21) and Mixed Bed (P22) Water Treatment systems' conductivity instrument loops by replacing six conductivity cells and conductivity recorder OP21-R0025 with new temperature compensated cells, three conductivity analyzers, and a digital recorder.
Summary I. No. The replacement conductivity instrumentation will continue to '
perform to the original design parameters. The modification will not introduce any new failure modes or effects that would increase ,
the likelihood of a malfunction of equipment that is important to safety. The water treatment instrumentation is nonsafety-related ;
and is not required for safe shutdown or safe operation of the '
plant, and is not used for accident mitigation. The replacement of the existing conductivity recorder with a new digital recorder was evaluated for EMI/RFI emissions and found to be well below the levels defined in EPRI TR-102323 guidelines for acceptable equipment emissions. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment i has not changed.
II. No. The new conductivity equipment will not adversely impact the function and operation of the P21 and P22 systems. Any failures related to this new instrumentation do not impact or change any i
systems, structures, or components that are important to safety. No new common mode failures of redundant safety-related systens, components, or structures are introduced as a result of the new design. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specification 3/4.4.4 and Table 3.4.4-1, which govern the ,
plant chemistry, are not affected. The design and operation of the '
water treatment systems were not altered. Therefore, no margin of safety has been changed.
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I I SE No.: 95-0172 Source Document: DCN 4326 4 Description of Change This drawing change updates two environmental zone drawings to reflect i revised temperature data. The revised temperatures will have no effect l upon equipment located within the areas in question.
l Summary I a
I. No. This change revises two drawings. As indicated above, the changes
] will have no effect on the operation of any plant components.
l Therefore, the probability or consequences of any accident or i malfunction of equipment previously evaluated in the USAR has not j changed.
i II. No. As indicated above, the changes being made will not affect the i
operation of any plant equipment. Therefore, the possibility of an accident or malfunction not already evaluated in the USAR is not
- created.
i III. No. The designs of the systems in the areas affected by this drawing change are unaffected. The bases for Technical Specification
) 3/4.6.2.6 is not impacted. Therefore, this change does not reduce any margin of safety.
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SE No.: 95-0173 Source Document: DCN 4403 Description of Change This drawing change updates various Generator Hydrogen Purity (N35) drawings. The drawing change is made to reflect actual field wiring for the hydrogen purity low / upscale local annunciator.
Summary I. No. This instrumentation performs nonsafety-related alarming and indication functions for the Hydrogen Purity system, and does not provide any functions that could impact any USAR accident. Failure of the instrumentation is not an initiator or mitigator to any of the accidents described in the USAR. USAR Section 10.2.5.3 states that the N35 system serves no safety-related function and that failure of the system will not compromise any safety-related equipment or prevent safe shutdown. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The Hydrogen Purity system instrumentation design, function, and operation is not impacted by this drawing change. Failure of the alarm circuit does not impact or change any safety-related or important to safety equipment design, function or operation. No new common mode failures of redundant, safety-related equipment are introduced by the drawing change. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Hydrogen Purity instrumentation alarm and associated setpoint are not in the Technical Specifications. N35 design, function or operation are not affected. Therefore, no margin of safety has been changed.
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SE No.: 95-0174 Source Document: DCN 4410 Description of Change This drawing change revises P&ID D302-632, Reactor Core Isolation Cooling (RCIC) System, by removing the trip of the gland seal air compressor 1E51-C004 from the compressor logic. This change only reflects actual field conditions, the compressor trip was never installed. This is consistent with the BWR6 design provided by General Electric.
Summary I. No. The P&ID logic change being made is associated with the nonsafety-related control logic of the RCIC system gland seal air compressor. This logic change does not impact the safe shutdown functions of the RCIC system. Further and specific to RCIC, it was identified that there are no USAR Chapter 6 or Chapter 15 accident / transients that rely on or take credit for RCIC operation to mitigate the consequences of an accident / transient. Since this logic change is associated with the nonsafety function of the gland seal air compressor, it is not a potential initiator of an accident.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. A failure effect analysis was performed and conmluded that the failure of the gland seal air compressor to automatically trip due to high air temperature will not create a new malfunction of equipment important to safety. The failure of 1E51-C004 does not impact the safe shutdown operation of the RCIC system. 1E51-C004 is nonsafety-related and this logic change does not affect any safety-related system or associated circuitry. This drawing change reflects the as-built plant conditions, and no physical changes are required. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The RCIC gland seal air compressor high temperature monitoring function for discharge air is not required to be operable for RCIC turbine operation and is not included in Plant Technical Specifications. In addition, the drawing change does not affect any RCIC initiation logic. Therefore, no margin of safety has been reduced.
SE No.: 95-0176, 96-0010 Source Document: Project 95-017-091, Chemical Decontamination l 1
Description of Change This project evaluates the performance of a CITROX chemical decontamination of the Fuel Pool Cooling and Cleanup (FPCC) system.
Summary I. No. The chemical decontamination of the FPCC system is expected to remove portions of the corrosion layer and contained activity from '
the internal surfaces of the system. The chemical process for this application has been determined acceptable for the materials it will contact. The decontamination activity will not impact the original design of the FPCC system, or degrade any structure, system, or component of the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The materials within the FPCC system have the exact or similar specifications of the materials tested with the Citrox process, and '
were found to have acceptable corrosion rates even at the elevated process concentrations. Once the decontamination activities are completed, the FPCC system will be restored to an operational status '
per plant procedures. Therefort, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The FPCC system is utilized to maintain the fuel pool temperatures.
The decontamination evolution will be conducted in a timely manner to ensure the system is restored within adequate time to maintain the pool temperatures. The activity will not impact the design of the FPCC system. Therefore, no margin of safety has been changed.
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i SE No.: 95-0177, 96-0018 Source Document: Project 95-017-091, Chemical Decontamination a
Description qf Change:
j This project evaluates the performance of a CITROX chemical ,
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1 decontamination of the Reactor Recirculation (B33) system. l:
Summary I. No. The chemical decontamination of the B33 system is expected to remove portions of the corrosion layer and contained activity from the {
internal surfaces of the system. The chemical process for this application has been determined acceptable for the materials it will {
l contact. The decontamination activity will not impact the original design of the B33 system, or degrade any structure, system, or component of the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The materials within the B33 system, reactor vessel and fuel pools have the exact or similar specifications of the materials tested with the Citrox process, and were found to have acceptable corrosion rates even at the elevated process concentrations. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
1 III. No. The decontamination will take place during Operational Condition 5 l and no applicable Technical Specification will be affected. The >
decontamination activity will not impact the design of the B33 I system. Therefore, no margin of safety has been changed. -
SE No.: 95-0178, 96-0024 Source Document: Project 95-017-091, Chemical Decontamination Description of Change This project evaluates the performance of a CITROX chemical decontamination of the Reactor Water Cleanup (RWCU) system.
Summary I. No. The chemical decontamination of the RWCU system is expected remove portions of the corrosion layer and contained activity from the internal surfaces of the system. The chemical process for this application has been determined acceptable for the materials it will contact. The decontamination activity will not impact the original design of the RWCU system, or degrade any structure, system, or component of the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The materials within the RWCU system, reactor vessel and fuel pools have the exact or similar specifications of the materials tested with the Citrox process, and were found to have acceptable corrosion l i
rates even at the elevated process concentrations. Therefore, the I possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. ;
III. No. The decontamination will take place during Operational Condition 5 !
and no applicable Technical Specification will be affected. The decontamination activity will not impact the design of the RWCU system. Therefore, no margin of safety has been changed.
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SE No.: 95-0179 Source Document: USAR Change Request 95-096 Description of Change The change request makes the following changes to the USAR:
- 1. Reassigned the Eadiation Protection Manager responsibilities defined in Regulatory Guide 1.8 from the Plant Health Physicist to the Manager, Radiation Protection.
- 2. Changed the title of the Plant Health Physicist to the Superintendent, Health Physics Operations.
- 3. Updated the the Radiation Protection Section organization.
- 4. Deleted the use of Albedo dosimeters for determining neutron dose.
- 5. Deleted the reference to recording direct reading dosimeter (DRD) readings on radiation work permits.
- 6. Deleted the reference to the Emergency Plan for defining the Radiation Protection organization.
4 Summary !
I. No. This change request makes the changes that are administrative and programmatic. The design and operation of the plant equipment I important to safety, and all supporting analyses remain unchanged. '
USAR accident analysis has not changed. Therefore, the probability of occurrence or the radiological consequences of an accident or malfunction of equipment important to safety previously evaluated in I the USAR have not changed.
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II. No. The changes are administrative in nature. This change request does not affect the design and operation of the plant. No plant design analyses have been altered. USAR accident analysis has not been affected. Therefore, the possibility of an accident or malfunction I
of a different type than any previously evaluated in the USAR is not created.
III. No. The changes are administrative and programmatic in nature. Plant design analyses have not been altered. Accident analysis remains unaffected. Therefore, no margin of safety has been reduced.
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SE No.: 95-0180 Source Document: USAR Change Request 95-097 r Description of Change f
This USAR change describes a management reorganization involving the '
onsite Regulatory Affairs Section. ;
Summary $
I. No. This change does not alter the plant. in any way. No functions or activities have been eliminated. Tne onsite personnel involved '
continue to meet the applicable A!4SI N18.1-1971 qualification '
requirements for their positions. The change complies with Technical Specifications. Therefore, neither the probability of i 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 '
sccident or malfunction of a different type than any previously i evaluated will not be created. ;
III. No. There are no changes being made to the physical plant. The i personnel qualifications continue to meet the requirements of ANSI N18.1-1971 and the Technical Specifications. Therefore, this i change will not reduce any margin of safety. i I
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SE No.: 96-0001 Source Document: DCP 92-0135, Rev. O Description of Change This design change is Phase II of a three-phase program to upgrade the Perry Plant Computers. The Phase II upgrade will replace the existing Honeywell 4400 Process Computer and its data acquisition system and integrate existing 3D-MONICORE and balance of plant functions onto a common computer platform. The Process Computer (C91) system is nonsafety-related as well as its software applications.
Summary I. No. The Process Computer has no control interface with any safety or nonsafety plant system. The Process Computer is a non-radioactive system and will not increase present radiation levels nor create new radiation sources beyond acceptable limits. The computer does not play a role in mitigating the consequences of an accident described in the USAR. The Process Computer will not affect any equipment important to safety nor will it impact any current (nonsafety, safe shutdown, or safety) system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The new Process Computer does not change plant operations, failure modes or reliability. The plant computer system is not a variable factored into accident analysis. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The computer system will be used to verify compliance with Technical Specifications. The 3D-MONICORE software will supply data related to Sections 2.1.1 (Core Thermal Power), 2.1.2 (MCPR), 3.2.1 (APLHGRs), 3.2.3 (LHGR) and Table 4.3.1.1.1 (Heat Balance). The l replacement of the Process Computer does not impact the above sections in any manner. Therefore, no margin of safety has been changed.
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SE No.: 96-0003
[ Source Document: PSTG, Rev. 4 Description of Change l The Perry Specific Technical Guidelines (PSTG) is the document which l
incorporates the BWROG Emergency Procedure Guidelines (EPG). This revision to the PSTG incorporates various changes due to the new fuel type in use and modifications made to the plant.
Summary I. No. The PSTG provides the basis for the Plant Emergency Instructions (PEI). The PEIs provide actions to shutdown the reactor and to restore and maintain core cooling and containment integrity for a wide range of accidents and transients. Operation of the systems and components as directed will not affect the various accident initiators previously evaluated in the USAR.
Therefore, this revision does not alter the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The PSTG, and the PEIs which are created from it, are not used prior to the beginning of an accident or transient; therefore, they cannot create the possibility of the accider.t or transient. Operation of the systems and components as directed will not affect the various accident initiators previously evaluated in the USAR. 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. This PSTG revision does not create an adverse impact upon the plant design or any analysis used in the design. The PEIs which are i created from the PSTG are symptom oriented, and administered to l respond to a wide range of accidents and transients. The PSTG as modified by this change does not adversely affect systems or components required for safety. Therefore, no margin of safety is j impacted. ;
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SE No.: 96-0005, 96-0103 Source Document: DCP 93-0125, Rev. 2 Description of Change This design change modifies the Primary Containment Personnel Airlocks by: 1) removing the existing hydraulic system components formerly supporting automatic airlock operation, 2) removing the s:isting handwheel locking device, 3) converting the existing door linkage-type mechanisms to gear-type mechanisms to control the activation of the ball valves' (this requires relocating the ball valves slightly), 4) installing modified latch pins containing springs and a reaction collar, and
- 5) installing a cew handwheel locking system that includes a new Programmable M9 6 ~ontroller (PLC).
Summary I. No. These modifications are being performed to improve the reliability of the operation of the airlocks. This activity does not impact the pressure boundary function of the airlocks. After modifications are complete and before the airlock doors will be placed back into operation, testing in accordance with 10CFR50, Appendix J, must be performed to ensure that the airlock doors and door seals will function as containment pressure boundaries. Containment integrity will continue to be ensured by the existing airlock mechanical interlock system. Inadvertent overriding of the mechanical interlock system is prevented by the new handwheel locking system. i The weight and center of gravity of new and existing components are I essentially the same and, thus havo negligible effect on the. !
existing dynamic (seismic) qualification of the doors. No new loads are added to the doors. Additional electrical and instrument air loads imposed by this activity are insignificant and have been ,
judged to have no adverse impact on existing equipment. Therefore, I the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Failures of the components of the modified airlock door (including the new PLC) produce the same results as failures of the current door components. Failure of the nonsafety-related circuit or the nonsafety-related instrument air associated with the locking cylinder will cause the spring loaded cylinder locking pin to retract and the handwheel will be free to operate (as is currently allowed). The personnel airlocks serve to mitigate the consequences l of accidents by providing containment integrity and pressure boundary. Loss of a nonsafety-related circuit or nonsafety-related instrument air have been considered in the original design and can not create a new accident. Thus, this activity does not create the i possibility of a new accident or malfunction of a different type than previously evaluated in the USAR.
SE No.: 96-0005 (Cont.)
Summary (Cont.)
III. No. Technical Specifications 3/4.6.1.1 and 3/4.6.1.2 provide l restrictions on primary containment integrity and containment leakage rate. Only one closed door is required to maintain the !
integrity of the Containment, and the mechanical interlock system supports this requirement. As previously discussed, the containment i pressure boundary is not affected. Therefore, no margin of safety I has been changed.
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SE No.: 96-0006 Source Document: l DCP 94-0172, Rev. 2 ;
Description of Change This design change establishes a Vehicle Barrier System (VBS) in response to the 1994 amendment to 10CRF73, " Protection Against Malevolent Use of Vehicles at Nuclear Power Plants." The purpose of the VBS is to ensure that a four wheel drive vehicle cannot be malevolently used to transport personnel and explosives into direct proximity of plant vital areas.
Summary I. No. This modification adds passive and active arriers at the plant protected area boundary in response to aIn..ded 10CFR73. In addition, alarms are added to the Centra' Alarm Station and the
! Secondary Alarm Station. No modificati' to any systems, l
structures, or components identified t cident initiators in the USAR are affected by this activity. Therefore, the probability of j
occurrence or the consequences of an accident or malfunction of
' equipment important to safety previously evaluated in the USAR has not increased.
II. No. No interactions with safety-related equipment are created by this activity. The barriers are anchored to resist the impact loading of the land based vehicle. Based on this anchorage, the threat of the !
barriers becoming a missile hazard due to high wind loading is !
removed. Sufficient distance exists between the passive barriers '
and the inner protected area fence to prevent any drifting of snow caused by the barriers from impacting the protected area fence.
Installation activities will be coordinated with plant Security to ensure that appropriate measures are in placc at all times.
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 modification is intended to enhance the plant's security by modifying the site Security Plan in response to amended 10CFR73.
The activity will not detract from any existing portions of the site Security Plan. Additionally, with the VLS fully implemented, all atiuctures will continde to perform their intended functions within all design allowables and requirements, as defined in USAR Sections 3.2 and 3.8. Therefore, this change will not reduce the margin of safety as defined in the bases of any Technical Specification or the USAR or other licensing documents.
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l SE No.: 96-0007 Source Document: TM 1-96-001 Description of Change A butt-welded spool piece of Reactor Water Cleanup (RWCU) piping (1-G33-3-RWCU-105-AB) will be temporarily removed and replaced with a flanged spool piece to facilitate chemical decontamination. There will be no increase in the number of field welds. The piping where the spool piece is to be installed is in the RWCU blowdown line to the Liquid Radwaste system, which is utilized for level control during plant heatup, shutdown, or refueling operations. Following completion of the chemical decontamination, the flanged spool piece will be replaced with a butt-welded spool piece, thereby restoring the piping run to its original configuration.
Summary I. No. The purpose of the RWCU system is to provide continuous purification of reactor water and maintain water clarity during refueling. The RWCU system also: i) discharges excess reactor water during startup, shutdown, and hot standby; ii) minimizes thermal stratification in the main recirculation piping and the reactor pressure vessel during periods when tha recirculation pumps are unavailable; and iii) provides sample connections for continuous water quality monitoring. The evaluat'.vu ut the chemical decontamination of the RWCU system is addresred in Project 95-017-01. This evaluation addresses the spool piece temporary installation and restoration for the purposeo of connection to the decontamination skid. The temporary spool siece will meet the SP-2000 L1-4 piping specification requirements, and therefore the integrity of the system will be maintained. During the course of the temporary installation, there is a deviation from SP-2000 in that it is preferred that all joints be welded except at flanged equipment connections. The section where the spool piece is to be installed will be isolated, and associated isolation valves tagged-out during the work effort. This temporary modification does not effect the design, material, or construction standards applicable to the RWCU system. Therefore, this change does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated in the USAR.
II. No. No new radiation source is expected as a result of installing the temporary modification. Additionally, the piping integrity will be naintained in the temporary condition and following restoration after chemical decontamination is completed. The originally evaluated function of this system is being maintained. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not creatW.
III. No. The temporary piping section will will meet the intent of SP-2000 L1-4 requirements. Therefore, the activity will not reduce the margin of safety.
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SE No.: 96-0008 Source Document: USAR Change Request 96-009 Description of Change This change request incorporates the performance of a high pressure water leak test for the stem and bonnet of feedwater outboard gate valves as an
- alternative to a Type C air test.
{ Summary 1 !
I. No. This change is a change in test methodology only and acceptance
- criteria of zero leakage is not affected. Therefore, the i probability of occurrence or the consequences of an accident or i malfunction of the equipment important to safety previously )
evaluated in the USAR has not increased. i II. No. The alternative test method will ensure that leakage of the components is within acceptable limits. Test parameters are i
controlled through approved plant instructions which maintain i
affected components within design equipment qualifications, Therefore, this change does not create the possibility of an l accident or malfunction previously evaluated in the USAR.
- III. No. Performance of a Type C air test or the alternative method for these components will ensure that srecondary containment bypass leakage
- totals are maintained within Technical Specification limits. The i acceptance criteria of zero leakage imposed by this change does not reduce the margin of safety as defined in the bases for Technical l Specifications.
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SE No.: 96-0011 l Source Document: TM 1-96-002 1
Description of Change This temporary modification evaluates the use of freeze seal to support installation of piping associated with DCP 90-0123 (Turbine Cuilcing Sample Panel). The freeze seal will be performed on a 2" t nonsafety-related stainless steel pipe. I Summary I. No. The installation of the freeze seal will not damage the piping system. If a leak were to occur due to freeze seal failure, the l
flooding concern would be bounded by a Circulating Water joint i failure break inside the Turbine Building. Nondestructive Examinations will be performed both before and after the freeze seal to ensure that the pressure integrity of the piping system is maintained. Piping integrity will not be compromised. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Freeze seal failure is bounded by the flooding analysis of the 1 Turbine Building / Turbine Power Complex. Valves associated with the l Condensate Storage Tank (CST) will be closed, as necessary, to limit I the flooding concern if a pipe break were to occur during the freeze seal process. This freeze seal does not introduce a different type of malfunction on any nonsafety or safety-related pipe within the Turbine Building. Permanent sump pumps will lessen any flooding I concerns in the Turbine Building. Therefore, the possibility of an !
accident or malfunction of a different type than any previously evaluated in the USAR is not created. i III. No. The pressure and temperature limits (class 1 piping) referenced in Technical Specification Bases B 3/4.4.6 does not apply to this freeze seal activity, since this activity is being performed on a nonsafety system. The freeze seal which effects the CST header (2' pipe) is an industry proven method of isolation of water systems that have limited isolation. Industry tests have proven and is acceptable to freeze stainless steel pipe since the line is seamless austenitic stainless steel with acceptable low-temperature toughness. Nondestructive Examinations (PT exams for indications, variations, changes and outside diameter differences) performed both before and after the freeze seal will ensure that the pressure integrity of the piping system is maintained. Therefore, no margin of safety has been changed.
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l-2 SE No.: 96-0012 j Source Document: TM l-96-004 i i
- Description Of Change l l
} This Temporary Modification (TM) will lift a lead to defeat the Low Electro-Hydraulic (EHC) Pressure Main Turbine Trip during data collection l
l for Work Order 96-064. The lead will be lifted as a precaution to i prevent a spurious turbine trip while EHC pumps are shifted for data l collection and troubleshooting. The TM and work order will be used to augment troubleshooting of an unexpected Low EHC Pressure First Hit i
indicating light which Control Room personnel received when shifting EHC 1
- pumps.
1 Summary t
]
I. No. A turbine trip signal is defeated to prevent a potential spurious, j unwanted, Main Turbine trip during data collection. No i safety-related equipment is affected by this item, and it will not
- compromise the safe shutdown of the reactor. Therefore, the j probability of occurrence or the consequences associated with an q accident or malfunction of equipment has not changed.
- II. No. Disabling the Low EHC Pressure Main Turbine Trip will only prevent I
unwanted spurious trip signals while shifting EHC pumps.
j Administrative controls will be in place to trip the Main Turbine <
1 manually if needed. Main Turbine trip event as evaluated in USAR l
} Chapter 15 remains bounding. No safety-related equipment function will be compromised by this temporary modification. Therefore, the
! possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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- III. No. The Main Turbine EHC system is not safety-related, and the Low EHC
] Pressure Trip is likewise not safety-related. The trip is provided 1
to protect the plant equipment, but it does not effect any i safety-related components nor does it have a direct function in the 4
i safe shutdown of the reactor. Administrative controls are in place to trip the turbine manually if necessary. 'Therefore, no margin of 4
safety has been changed.
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SE No.: 96-0013 Source Document; DCN 5204 '
Description of Change This drawing change revises drawing D511-072, Reactor Building-Steel Framing, to show the grating on top of the drywell bulkhead plate as optional.
Summary I. No. Calculation 3:71, Rev. O has qualified the grating for all the inertial loads resulting from loss of coolant accidents inside containment without the benefit of the grating saddle clips and studs. There is no plant equipment supported from or in close proximity of this grating that could be damaged such that a loss of coolant accident could be initiated. Removal or partial removal of this grating can in no way degrade any of the fission product barriers, or cause an increase in dose to the public, or an increase in onsite doses that would impede actions necessary to mitigate the consequences of an accident. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR is not increased.
II. No. There is no safe shutdown equipment that could be damaged as a result of this grating not having any hold down sa3dle clips, and there is no accident initiator or failure created an a result of this drywell bulkhead grating being removed or partially installed.
There are no new equipment failure modes created as a result of this i grating being removed or partially removed. Therefore, the l possibility of an accident or malfunction of equipment different !
than any previously esaluated in the USAR is not created.
III. No. The safety function performed by the bulkhead plate is unaffected by the removal or partial removal of the grating. No functional change was made to any plant equipment or systems. There is no change in initial conditions or system response time for any equipment or systems required for safety. Therefore, the change does not reduce any margin of safety.
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SE No.: 96-0014 Source Document: DCN 5277 Description of Change This drawing change revises P&ID D912-603, Drywell Cooling (M13) System, to eliminate the operating airflow rates for Flags 3, 4, and 5. The change is being in order to clarify airflow rates associated with different M13 operating configurations.
Summary I. No. The design and operation of the M13 system has not been altered.
Drywell temperature limits have not changed. USAR accident analysis is unaffected. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The design and operation of the M13 system has not been altered.
Drywell temperature limits have not changed. USAR accident analysis is unaffected. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The design and operation of the M13 system has not been altered.
Drywell temperature limits have not changed. USAR accident analysis is unaffected. Therefore, no margin of safety has been changed.
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1 SE No.: 96-0015 Source Document: DCP 95-6105, Rev. O i Description of Change !
This design change removes the pitot tube flow e1.ements used to provide local indication of Circulating Water flow to the Main Condenser. The ;
i measurement is not required by any procedule and is not monitored by 1 operator rounds. The modification will remove the elements f rom the pipe, cover the element hole with a blind flange, discard the root valves, cut back the tubing to the nearest support, and remove the indicators from the instrument rack. r Summary I. No.
The only accident in the USAR evaluated related to the circulating Water system is the Turbine Building flooding event. Since the pipe modifications will be nede in accordance with the original design code, ANSI /ASME B31.1, the pipe integrity is maintained. The instrumentation provides no signals to any equipment, hence, its i removal has no effect on plant operation or reliability. Therefore, I the probability of occurrence or the consequences associated'with an accident or malfunction of equipment has not changed.
II. No. This configuration change creates no new equipment types and maintains the original plant design basis. These changes meet the requirements of ANSI /ASME B31.1, which is the original construction code. Therefore, the modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the USAR. !
III. No. This modification complies with the original construction code, ANSI /ASME B31.1, and does not affect the operation of the system.
Therefore, no margin of safety has been reduced. I
l SE No.: 96-0016 Source Document: SVI-PS2-T9306, Rev. 4, TC-1 Description of Change This surveillance instruction provides the necessary direction to install temporary air jumpers around the instrument air containment penetration during testing without reliance on the Temporary Modification Program for the activity.
Summary I. No. The instruction change described routes instrument air into the containment during shut down plant conditions with the instrument air containment penetration isolated to permit normal operation of components dependent on instrument air. Failure of the air jumpers is enveloped by the USAR's Loss of Instrument Air Accident Analysis.
Therefore, the probability of occurrence or the consequences associated with an accident or nalfunction of equipment has not changed.
II. No. The instruction change described routes instrument air into the containment during shut down plant conditions with the instrument air containment penetration isolated to permit normal operation of components dependent on instrument air. Failure of the air jumpers is env .eped by the USAR's Loss of Instrument Air Accident Analysis.
Therefore, the possibility of an accident or malfunction of a !
different type than any previously evaluated in the USAR is not I created.
III. The Instrument Air Distribution system is not addressed by Technical l Specifications. Components supplied by the Instrument Air Distribution system fail in their safe shutdown condition upon a loss of air. The USAR analysis remains bounding. Therefore, no margin of safety has been changed.
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SE No.: 96-0017
- Source Document
- PTI-G41-P0004, Rev. O Description of Change This Periodic Test Instruction (PTI) provides a controlled method to lower the spent fuel storage pool and cask pit level to support the
, replacement / testing of relief valve 0G41-F746A. The testing is required by Technical Specification 4.0.5 and the requirements are delineated in
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the Pump and Valve Inservice Test Program. Reducing the water level in the spent fuel pool is required to provide an adequate drain of the piping for valve replacement.
Summary I. No. The USAR accident analysis applicable to the spent fuel storage pool is described in USAR Section 15.4.7. The analysis addresses the potential accident of dropping a fuel assembly onto the fuel racks and the associated radiological consequences. This PTI prohibits fuel handling during performance of the instruction. Therefore, the initiator associated with this accident scenario has not been affected. The administrative controls established in the PTI ensure that water level and temperature are maintained to ensure adequate cooling and radiation shielding. Therefore, the conditions established by the PTI are bounded by USAR analysis. Therefore, the ;
probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The PTI utilizes the nonsafety cask pit drain pump to reduce level in the spent fuel storage pool. A malfunction of equipment in the cask pit drain piping can be addressed by operator action to close the safety boundary valve 0G41-F581. Operation of the valve is within its design basis and would re-establish the pool boundary.
Additionally, nonsafety pump suction and discharge isolation valves are also located at the pump which could be used to re-establish the pool boundary. Therefore, the activity does not create the possibility of an accident or a equipment malfunction important to safety different than any previously evaluated in the USAR.
III. No. The instruction limits the reduction of the spent fuel storage pool water level to a maximum of l'-5' which ensures the Technical Specification 3/4.9.9 limit of 23 ft. above the top of the spent fuel is maintained. The bulk water temperature is maintained within the normal operating band (<l270F). Supplemental cooling is also available to maintain the bulk water temperature <l50 F which is within the bounds of the plant design. Therefore, no margin of safety has been changed.
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SE No.: 96-0020 Source Document: DCP 96-5001, Rev. O Description of Change This design change installs a 4' diameter spectacle blind downstream of valve 1N27-F0150 and upstream of the High Pressure (HP) condenser (vacuum .
side during normal plant operations). This change will eliminate seat '
leakage associated with valve IN27-F0150 and positively control feedwater flow bypassing the reactor, downstream of the feedwater flow venturies. i The line being isolated was originally installed to allow wet lay-up of the Feedwater system with hydrazine. Since hydrazine wet lay-up has never been used and no procedures exist to allow or require its use, there is no need to maintain this line in service. The installation of this nonsafety, non-seismic modification is in accordance with ASME/ ANSI B31.1 and meets the pressure rating of the Feedwater (N27) system.
Summary t
I. No. The accidents of interest are USAR 15.2.5, Loss of Condenser Vacuum, and 15.6.6, Feedwater Line Break - Outside Containment. The installation of this nonsafety, non-seismic modification is in accordance with ASME/ ANSI B31.1. It meets the pressure and temperature requirements of the Feedwater system, therefore, system pressure integrity is maintained and the piping system is not expected to fail. The modification will not cause system vibration, initiate a water hammer event, or degrade in any manner such that a pipe break in the system could occur resulting in an accident as mentioned above. The modification does not impact or interfaca any equipment important to safety. There is no safety-related equipment nor equipment important to safety in the proximity of where this modification is to be installed. No malfunctions of equipment are being introduced as a result of this modification. For these reasons, doses to the public and onsite personnel are not increased.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not ;
changed.
II. No. The modification to the Feedwater system does not involve an initiator or failure of a different type not considered in the USAR.
The modification was reviewed with respect to malfunctions not described in the USAR or past single failure analysis. During operation with the spectacle blind flange installed, the new joints ;
are not expected to fail to the point of catastrophic line break since the flanges are designed to the pressure and temperature requirements of the feedwater system. In service leak testing will be performed on the welds of the new flanges. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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1 SE No.: 96-0020 (Cont.)
Summary'(Cont.)
i III. No. The Technical Specifications and other related safety bases (USAR,
! SER, etc.) were reviewed with respect to this modification. The installation has no impact upon these bases. The installation is in 1
accordance with ANSI B31.1 and will maintain the pressure boundary l
integrity of the Feedwater system. Stresses within the piping system remain within acceptable limits. No over pressurization
- condition will exist as a result of this modification. Tnerefore, i no margin of safety has been changed.
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l SE No.: 96-0021 Source Document: DCN 5281 l i
1 Description of Change !
This drawing change revises P&ID D302-672, Reactor Water Cleanup System, l
- to include a temporary spool piece installed for the fifth refueling ;
outage decontamination project as a permanent plant component within the Reactor Water Cleanup (RWCU) system.
Summary l ,
! I. No. The spool piece conforms to requirements of the existing line i i' specification with the exception of the flanged connections. These connections however were evaluated through a Specification j Deviation (SDA) and were accepted. This change did not impact the ,
function or design of any safety-related equipment or equipment i important to safety. This change had no effect upon any previously !
defined accidents or on equipment important to safety. Therefore, this change will not increase the probabilities or consequences of ,
any previously analyzed accidents of malfunction of equipment !
j important to safety.
II. No. The permanent installation of the spool piece does not change the ;
system's operation or function, or any other interfacing system's i operation or function. The spool piece conforms to the existing i line specification with the exception of the flanged connections. '
l These, however, were evaluated and found to be acceptable under a i i SDA. Therefore, the possibility of an accident or malfunction of equipment important to safety of a type different than previously i i evaluated is not created.
III. No. The permanent installation of the spool piece conforms to the i applicable specifications and design requirements. It does not -
.I result in a change to the operation or function of the plant. ;
Therefore, the margin of safety has not been reduced. '
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SE No.: 96-0022 Source Document: Physical Security Plan, Rev. 22 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.
SE No.: 96-0026 Source Document: DCN 5305 Description of Chance This drawing change revises drawing D201-146, Electrical Fire Barrier Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated barrier on conduits IP57F2B and 1R33F1051B located in the Auxiliary Building Residual Heat Removal (RHR) A Room (Fire Zone 1AB-lb) and the 599' elevation corridor (Fire Zone 1AB-2). Calculation SSC-004 was prepared to demonstrate that in the event a fire in Fire Zones 1AB-lb or 1AB-2 which initiates a spurious closure of Division II safety-related instrument air isolation valve IP57-F0152. the components and circuits for Division I isolation valves IP57-F0151.and 1P57-F020A will remain free of fire damage and the Division I iralation valve will remain open.
The safe shutdown Method A Automatic P. pressurization System (ADS) air supply will remain available prov % for long term reactor vessel depressurization.
Summary I. No. Removing the requirement for a fire barrier can only affect the protection provided the conduits during a fire, it will not affect the probability of the occurrence of an inadvertent safety / relief valve opening event. Since accidents are not assumed to occur coincidentally with a fire, removing the requirement for a one-hour rated fire barrier does not affect the ability of the ADS valves to function under all other design conditions (including accident conditions). Hence, in the event of a fire, the capability of the Method A ADS to function as designed to perform safe shutdown functions and to minimize radioactive releases to the environment will be maintained. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with valve 1P57-F015B are made by this activity, and thus no new system interaF. ions are created. Removing the requirements to have a rated fire barrier on these conduits does not affect their ability to function under required operational conditions. Other means of long-term vessel depressurization are available in the case of a fire in Fire Zone 1AB-lb or Fire Zone 1AB-2. The capability of the Method A ADS to function as designed is not affected by this activity. The change will not expose any systems important to safety to any possible failure modes beyond those previously analyzed in the USAR. Therefore, the implementation of this change ;
cannot create any new accident or malfunction beyond those '
previously postulated within the USAR.
4 SE No.: 96-0026 (Cont.)
Summarg (Cont.)
III. No. Since this activity makes no changes to the ADS valves or air supply, it does not affect the requirements for operability of the ADS valves. Since analysis has shown that removing the requirement for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barriers will not affect the ability of the plant to achieve and maintain safe shutdown, Appendix R requirements are not adversely affected. Therefore, no margin of safety has been re/ aced.
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SE No.: 96-0027 Source Document: DCN 5306 Description of Change This drawing change revises drawing D201-146, Electrical Fire Barrier i Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated i barrier on conduits 1R33R2069B, 1R33R1072D and 1R33R2073D located in the Intermediate Building 620' elevation corridor (Fire Zone IB-3). These conduits contain circuits for main steam line pressure instruments ,
1B21-N076A, 1B21-N076C, 1B21-N076B and IB21-N076D. Calculation SSC-006 l demonstrates that the loss of the circuits from the main steam line l pressure instrumentation will not adversely impact the ability to close '
the Main Steam Isolation Valves (MSIVs) as required to achieve and maintain safe shutdown in the event of a fire in Fire Zone IB-3.
Summary I. No. No changes to the conduits other than the requirement for a fire barrier have been made. Removing the requirement for a fire barrier I can only affect the protection provided to the conduit during a !
fire, it does not affect the probability that a fire will occur. l This activity will not increase the probability of the occurrence of a fire. Because the MSIVs will close as designed, this activity '
does not affect the consequences of accidents previously evaluated i in the USAR. The MSIVs will function as intended, and a fire '
induced fault on the low main steam line pressure circuits will not cause a malfunction of the MSIVs. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with the MSIV trip on main steam line pressure are made by this activity, and thus no new system interactions are created. The design of the circuits to initiate MSIV closure are such that a fault on the circuit can only cause the solenoid on the MSIV to de-energize and the MSIV to close.
Since in the case of a fire, the plant will initiated shutdown which includes closure of the MSIVs, closure of the MSIVs is not considered to be a malfunction. The MSIVs are designed to close upon de-energization of the solenoid (for any reason), so there is not an occurrence of a different type than already analyzed for the plant. Therefore, the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR has not been created by this activity.
III. No. This activity makes no changes to the primary containment and drywell isolation instrumentation, or operability and stroke times i for the MSIVs. Since evaluation has shown that removing the requirement for one-hour rated fire barriers will not affect the ;
ability of the plant to achieve and maintain safe shutdown, Appendix R requirements have not been adversely affected.
Therefore, no margin of safety has been reduced.
SE No.: 96-0028 Source Document: DCN 5249 Description of Change This drawing change revises drawing 808-302, Functional Control Drawing -
Reactor Protection System, to correct IC71-N005A-D pressure switch channel designators for the Turbine Control Valve (TCV) fast closure signals.
S.,jLmmary I. No. The arrangement shown on the 808 drawing is clearly in error.
Correcting the lettering error for the channel designator does not affect the function / operation of the Turbine Control Valves. No physical changes are being made to the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Correcting the 808 drawing will have no impact on either the turbine trip or generator load rejection accident analysis contained in USAR Chapter 15. No physical changes are being made to the plant.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The drawing change will not change any plant component. No new components are introduced into any plant system. No components are being removed from any plant system. The operation of all systems will remain unchanged. No additional stresses thermal, mechanical, or electrical will be added to any plant system or component.
Therefore, no margin of safety has been changed.
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SE No.: 96-0029 Source Document: DCP 96-5005, Rev. O Descriptior: of Change This design change installs vibration monitoring equipment on the Reactor Recirculation (B33) system pumps and motors. This design installation will facilitate monitoring pump and motor performance, and will aid in troubleshooting potential problems which could affect the pump shaft and their mechanical seals.
Summary I. No. Accidents related to the B33 system previously evaluated in the USAR are Recirculation Pump Trip (Section 15.3.1), Recirculation Flow Control Failure (Section 15.3.2), Recirculation Pump Seizure (Section 15.3.3) and Recirculation Pump Shaft Break ,
(Section 15.3.4). The vibration monitoring equipment is installed l so that it cannot physically affect any B33 system component.
Hence, the above accidents will not be impacted. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Radiological consequences of accidents associated with the B33 system are listed in USAR Sections 15.3.1.5, 15.3.2.5, 15.3.3.5 and 15.3.4.5. The vibration monitoring equipment has no impact on B33 system operation. The installation of the vibration monitoring
- equipment does not change the function of the B33 system in any way.
Hence, the radiological consequences of the previously listed events i are not impacted. Therefore, the possibility of an accident or l malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The sensors are attached to the recirculation pumps and motors such that no damage will occur, should the vibration monitoring equipment fail. The additional weight of the sensors was found to be negligible. No piping boundaries or structures are breached by the installation. Cables and conduit are installed to allow appropriate clearance between B33 system equipment and vibration monitoring equipment. Therefore, no margin of safety has been changed.
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d SE No.: 96-0030 j Source Document: USAR Change Request 96-036 1
l Description of Change i
This USAR change replaces the reference to the the specific number of shift crews that is required during plant shutdown conditions, with the a following statement: "During plant shutdown conditions, the number of shift crews may be less than five in order to better accommodate outage i workload and scheduling requirements."
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! I. No. This USAR change deletes the reference to having four shift crews during shutdown conditions. The specific number of shift crews is l 1
in no way related to the anticipated operational transients and
- design basis accidents that have been analyzed in the USAR.
t Required minimum shift manning levels remain unchanged. Operator. ;
l actions during an accident are not altered or degraded. Assumptions '
I made in evaluating radiological consequences during an accident are i not altered. Fission product barriers are not impacted. Shift working hours and overtime limitations will still adhere to the i l requirements of Generic Letter 92-12. Therefore, the probability of i
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occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
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. II. No. This USAR change deletes the reference to having four shift crews ;
j during shutdown conditions. Individual crew manning levels will ;
l- remain in compliance with Technical Specifications and other i
i regulatory requirements. Shift working hours and overtime ;
limitations will still adhere to the requirements of Generic Letter 82-12. Therefore, the possibility of an accident or i malfunction'of a different type than any previously evaluated in the j USAR is not created.
- III. No. Technical Specification 6.2.2 describes the minimum shift crew
- composition but does not provide a limitation on the minimum number of shift crews that must be maintained. This Technical j Specification also references Generic Letter 82-12 for limitations i on overtime worked be unit staff members. Perry will continue to i comply with these requirements. Therefore, this USAR change does
, not reduce any margin of safety.
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SE No.: 96-0031 Source Document: FCR 22611 Description of Change This Field Clarification (FCR) evaluates the provision of the technical !
requirements for the safe rigging and transporting gates 1G41-E001A and 4 1G41-E002A between the Inclined Fuel Transfer System (IFTS) pool and the dryer storage pool during Refueling Outage 5 (RF05) with fuel stored in the upper containment storage rack.
Summary I. No. The plant's safe load handling requirements imposed on the lifting and transporting of the IFTS pool gates exceeds the requirements of NUREG-0612. Additionally, the redundant load handling requiremente by the FCR exceeds the requirements of NUREG-0612 for critical lifts. The safe load handling requirements provided in the FCR preclude the possibility of a single failure path resulting in a drop of the gates. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The requirements of the FCR exceed those of NUREG-0612. The FCR provides additional load handling requirements to ensure no single failure path exists that could result in a drop of the gates.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
l III. No. By following requirements of a Category B lift as defined in l approved procedures and by providing a dual redundant load path, I required by the FCR, the potential for a load drop is precluded.
The requirements of NUREG-0612 remain satisfied. Therefore, no margin of safety has been changed.
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SE No.: 96-0032 Source Document: Potential Issue Form (PIF) 96-0762 Description of Change This potential issue evaluates the temporary "use-as-is' disposition -
involving a Division I Diesel Generator (DG) jacket water system leak. ;
Water was found to be slowly seeping from the top of the exhaust shroud from a pin hole sized crack at the toe of a weld. The weld attaches a small support block for the jacket water discharge manifold to the 4
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exhaust shroud.
Summary I. No. This disposition allows for the continued operability of the Division I DG until the end of Refueling Outage Six. At that time, the defect will be reworked to conform with the original design requirements. The cause of the seepage discovered is attributed to a subsurface defect at the toe of the weld that propagated to a through wall defect. The defect propagation is judged to have been driven by a combination of residual stresses in the weld and thermal stresses. Further propagation is not likely as the residual stresses in the weld reduce as the defect propagates. The DG jacket vater system is designed to accommodate minor inventory loss through pump seals, mechanical joints and evaporation. The worst case assumption assumed the leakage rate exceeded the observed leakage rate by a factor of 100. With this the DG could still operate for seven days and not compromise the DG jacket water system.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This dispositior. does not alter the fundamental design or function of Division I DG or its ancillary systems. Operation of the Division I DG in its present material condition cannot adversely impact any interfacing safety system. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. As stated above, assuming worst case failure, the operability of the Division I DG is not compromised. Therefore, no margin of safety !
has been changed.
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l 1 SE No.: 96-0033 Source Document: FCR 22637 f i
Description of Change 1
f This field clarification evaluates the performance of a freeze seal to l support local leak rate testing. The freeze seal will be placed on an l ASME III, Class NB, 6' Reactor Coolant Pressure Boundary (RCPB) pipe. l Summary !
I. No. This freeze seal will be placed on the RCPB pipe when the plant is !
shutdown and in Modes 4 or 5. Reactor coolant temperature and l pressure will be below high energy line break limits. NDE testing
- will be performed prior to and after the freeze seal to ensure no ;
degradation of the piping. Since the valves upstream of the freeze t seal, on the reactor side will be maintained in the closed position, :
l the potential for vessel drain down is prevented if a catastrophic '
failure of the piping occurs at the freeze seal. Therefore, the !
probability of occurrence or the consequences associated with an l accident or malfunction of equipment has not changed.
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II. No. Since the RCPB will be maintained in an isolated condition, failure ;
l of the freeze seal would not cause a loss of reactor coolant ;
I inventory. Small amounts of valve seat leakage can be made up via j the Emergency Core cooling Systems. The impact of the freeze seal on the piping is negligible. Therefore, the possibility of an accident or malfunction of a different type than any previously l evaluated in the USAR is not created.
III. No. The pressure and temperature limits referenced in Technical Specification Bases 3/4.4.6 is not affected. Industry tests has shown that deviation below the transition temperature and return to normal temperature does not change the crystal structure or the
' characteristics of ferretic material. Therefore, performance of l this freeze seal does not reduce any margin of safety. !
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SE No.: 96-0034 Source Document: DCP 95-0022, Rev. 1 j DCP 95-0022B, Rev. 1 l
Description of Change The modifications replace a section of Service Water (SWS-P41) system supply side Fiberglass Reinforced Plastic (FRP) piping, with carbon steel piping and resolved a critical portion of the overall SWS 3
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reliability issue by removing the ' weak link' of the system. Abandoned ;
Unit 2 Emergency Service Water (ESW) piping is being utilized by these l modifications. The modifications include the addition of cathodic protection and provisions to prevent flooding in the Unit 2 Auxiliary Building (AB) and pipe chase, which could potentially affect safety-related components in Unit 1.
Summary !
I I. No. The replacement carbon steel piping provides the same design function as the original piping and does not adversely impact the operation of the SWS or its design functions. The current performance of the SWS is maintained. Precautions and modifications -
during outage and non-outage construction activities preclude damage j to safety-related Structures, Systems, and Components (SSC) from construction activities, load drops, flooding, and tornado missiles to ensure the operability of safety-related systems. This ensures that safety-related systems are available to perform their safety-related functions at all times. The final tie-in of the new SWS piping will be completed during an outage when operation of the SWS is not required. The Unit 2 ESW piping being utilized for this modification does not provide any safety-related function. 1 Therefore, this activity will 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 r ....acement piping is designed to the original performance ;
requirements for the system. The new piping is designed to meet ANSI B31.1 and does not affect the design function or operation of the system. This activity, including installation activities, does not affect the function or degrade the performance of any SSC.
Provisions are included in the design of the modification to prevent the effects of flooding, adverse weather, and connecting the new piping to existing system piping from initiating an accident. Thus, the activity does not cause any SSC to operate outside its design basis or fail to perform its safety function. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the USAR is not created.
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l SE No.: 26-0034 (Cont.)
Summary (Cont.)
III. No. Technical Specification 3.7.1.1, ESW system (Loops A, B, C), ensures l that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. During the SWS modifications, the alternate keepfill water supply from the Fire Protection system will be used to maintain the operability of the ESW system. Temporary cooling for !
required heat loads will be provided while the SWS is out-of-service during the outage. Replacing SWS piping assures the availability of the SWS to perform the ESW system keepfill function. The design and operation of the SWS is not adversely affected by these modifications. Therefore, no margin of safety has been reduced.
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l SE No.: 96-0035 I l Source Document: USAR Change Request 96-043 '
Description of Change '
i This change request adds the following Regulatory Guides to USAR {
Table 1.8-2: !
i 1.8, " Personnel Selection and Training',
l 1.26, " Quality Group Classifications and Standards for Water, Steam, and '
Radio-Waste-Containing Components of Nuclear Power Plants", and 1.29, ' Seismic Design Classification'. ,
USAR Table 1.8-2 provides a listing of Perry's commitments to QA-related regulatory guides. Regulatory Guides 1.8, 1.26, and 1.29 were inadvertently omitted when USAR Table 1.8-2 was created. These regulatory guides and Perry's degree of conformance to them are currently i
listed in USAR Table 1.8-1.
l l Summary I l I. No. This USAR change adds three regulatory guides to USAR Table 1.8-2. !
i These regulatory guides are currently listed in Table 1.8-1. There i l is no change to the degree of commitment to these regulatory guides.
i i There are no changes to the design-or operation of the plant or upon '
i any accident analysis. Therefore, this change does not increase the
, probability of occurrence or the consequences of an accident or l malfunction of equipment previously evaluated in the USAR.
II. No. This USAR change adds three regulatory guides to USAR Table 1.8-2.
These regulatory guides are currently listed in Table 1.8-1. There 1
! is no change to the degree of commitment to these regulatory guides. J There are no changes to the design or operation of the plant or upon any accident analysis. Therefore, this administrative change does not create the possibility of an accident or malfunction of equipment of a different type than any previously evaluated in the USAR.
l l III. No. This USAR change adds three regulatory guides to USAR Table 1.8-2.
l These regulatory guides are currently listed in Table 1.8-1. There is no change to the commitment to these regulatory guides.
Therefore, no margin of safety has been reduced.
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l SE No.: 96-0036 Source Document: PDB-F0001, Rev. 3 Description of Chance 1
This plant data book revision updates the Core Operating Limits Report to reflect the thermal limits for Cycle 6 operation.
Summary I. No. The only change to the plant is in the introduction of a new fuel ,
configuration and core design. The potentially limiting plant transients and accidents hau been evaluated using the same limits on consequences as previously evaluated in the USAR, and as approved in the NRC Safety Evaluation for GESTAR II. The fundamental sequences of accidents and transients have not been altered. The General Electric document Jll-02581SRLR, Rev. O, " Supplemental I Reload Licensing Report for PNPP Unit 1 Reload 5 Cycle 6", documents the results of the GESTAR II analysis for Cycle 6. Therefore, the probability of occurrence or the consequences a;..ociated with an accident or malfunction of equipment has not chat';ed.
II. No. Plant operation will still conform to the analyzed envelope of USAR Chapters 4, 5, and 15, with the Core Operating Limits Report revision. Conformance with the GESTAR II analysis has been maintained. The GESTAR II analysis has been accepted by the NRC as comprehensive for ensuring that fuel designs will perform within acceptable bounds. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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 MCPR Safety Limits, and the bases of the other fuel-related Technical Specifications. Therefore, no margin of safety has been changed.
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SE No.: 96-0037 Source Document: DCP 96-6004, Rev. O Description of Change i This design change terminates the Fire Protection (P54) and Potable !
! Water (P71) service to the Unit 2 Circulating Water Pumphouse. This ;
! change is being performed due to the pumphouse being an abandoned i building with little or no heat. l l
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I. No. The Unit 2 Circulating Water Pumphouse (CWPH) is an abandoned !
l structure which no longer has any plant related function, and does i not contain any equipment important to safety. The building no )
longer requires the two support systems. She capping off of the two lines will not alter the operation of the remaining portions of the >
l systems. The change will not alter the relationship between the two i i affected systems and the rest of the plant. Therefore, this J l modification cannot affect the probability or consequences of any l accident or malfunction of equipment previously evaluated in the l l
USAR.
- II. No. The Unit 2 CWPH has no remaining plant functions, a'4d contains no ;
equipment important to safety. The capping of the cwo lines will not alter the operation of the remainder of the sy;tems. The j relationship between the two systems and the bala4ce of the plant i
will be unchanged. Therefore, no new accident o, malfunction types -
will be created by this modification.
l III. No. The Technical Specifications do not address tra P71 system. The P54 I system as discussed in Section 6.0, Administr:4tive Controls, l involves review, audit and reporting requirenants for the Fire Protection Program. The P54 system will cont inue to meet all such requirements. Therefore, the applicable marnins of safety regarding the P54 and P71 systems will remain unchangeil by this modification.
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l SE No.: 96-0038 Source Document: USAR Change Request 96-050 Description of Change This USAR change revises USAR Chapters 10 and 15. The changes are an output from the Steam Bypass and Pressure Control (N32/C85) System Operation and Test Review Program (SOTR). These USAR changes are only intended to improve the USAR text.
Summary I. No. The control valve opening time, USAR Section 10.2.1, is limited to providing stable pressure control. The intent of changing the time from 7 to 10 seconds reflects the original design value. The Section 10.2.2.4.1 change corrects an error in the description of the logic for the loss of 125 VDC to the Turbine Electro-Hydraulic Control (EHC) system. The supplied method of providing redundant pressure regulator channels (error detection) is equivalent to the method used for USAR analyses. These USAR changes have no connection to any equipment important to safety. None of the changes to the USAR text have an affect on the radiological consequences of accidents evaluated in the USAR. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed. !
II. No. This USAR change is an administrative revision. No physical changes !
were made in the plant. All postulated failures are bounded by the USAR analyses in Chapter 15. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
l III. No. The control valve full open stroke time and the loss of 125 VDC turbine trip are not Technical Specification related. The Technical Specification basis for the Steam Bypass system is mitigating the Feedwater Controller Failure accident analyzed in USAR Chapter 15.
The " margin of safety" as related to the Feedwater Controller accident is the ability of the steam bypass valves to open. This USAR change does not have any affect on the ability of the bypass valves to operate, thus does not reduce the margin of safety.
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l SE No.: 96-0039 Source Document: SVI-E51-T9106, Rev. 4 SVI-E22-T94( , Rev. 5 SVI-E51-T9115, Rev. 1 SVI-E12-T9402, Rev. 6 SVI-E51-T8106, Rev. 2 SVI-E12-T9403, Rev. 7 SVI-T23-T1201, Rev. 4 SVI-E12-T9107,.Rev. 6 SVI-T23-T1202, Rev. 1 SVI-E12-T8408, Rev. O SVI-T23-T1203, Rev. 1 TAI-1120-1, Rev. 2 SVI-T23-T9002, Rev. 0 TAI-1120-2, Rev. 2 SVI-T23-T9003, Rev. O TAI-1120-3, Rev. 2 SVI-E51-T9101, Rev. 5 TAI-1120-4, Rev. 2 SVI-E12-T9102, Rev. 5 TAI-1120-5, Rev. 1 SVI-E21-T9103, Rev. 5 TAI-1120-7, Rev. 1 SVI-E12-T9105, Rev. 7 PDB-G0001, Rev. 1 SVI-N27-T9121, Rev. 5 PTI-T23-P0001, Rev. 1 SVI-N27-T9414, Rev. 4 PTI-T23-P0002, Rev. 1 Description of Change These instruction revisions incorporate a modification which altered the containment isolation valve arrangement for penetration P106.
Summary I. No. Design Change Package (DCP) 95-0017 added 1E51-F040 as a second containment isolation valve to Reactor Core Isolation Cooling (RCIC) penetration P106. This change was implemented in accordance with the requirements of 10CFR50, Appendix A, General Design Criterion 56. Replacement valve 1E51-F040 is designed to meet or exceed the original system performance requirements. Therefore, the probability of occurrence or the consequences of an accident or malfunction of the equipment important to safety previously evaluated in the USAR has not increased as a result of this change.
II. No. RCIC system operation was not affected by installation of replacement valve 1E51-F040. These instruction changes simply reflect DCP 95-0017. The changes do not affect the function or degrade the performance of any structure, system, or component. !
Therefore, these changes do not create the possibility of an accident or malfunction of a different type than previously evaluated in the USAR.
III. No. Replacement valve 1E51-F040 does not affect the ability of the i Reactor Core Isolation Cooling system to perform its' design !
function. The allowable leakage values specified for this valve ensure that secondary containment bypass leakage rates will not exceed values assumed in the accident analysis. Implementation of these changes do not impact RCIC operation or the allowable leakage values. Therefore, no margin of safety has been reduced.
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I SE No.: 96-0040 '
Source Document: Potential Issue Form (PIF) 96-1040 Description of Change This potential issue disposition allows the continued use of feedwater check valve, IN27-F559A,' until the end of Refueling Outage Six (RF06) with circumferential indications in the weld transition zone between the valve body casting wall and the stellite seat area.
Summary I. No. The existence of circumferential indications along the interior wall of the valve body at the transition to the valve seat does not l affect the pressure retaining capability of the valve or compromise l the integrity of the reactor coolant pressure boundary. The disposition does not affect valve closing or penetration leak tightness. ASME Code requirements are maintained. Therefore, i neither the probability of occurrence nor the consequences of an j accident or malfunction of equipment will be increased.
II. No. This disposition relates to the internals of a single check valve l
and as such does not reduce the redundancy or independence of existing structures, systems and components.- The single failure criteria is maintained. Thus, there is no increased potential for common mode or common cause failures. ASME Code requirements and l the safety functions of the affected component are maintained.
Therefore, this disposition does not create the possibility of an accident or malfunction of a different type than any previously j evaluated. '
l III. No. By adherence to the ASME Code for pressure boundary components, l
Technical Specifications 4.0.5 and 3/4.4.8 are not affected. The i l disposition does not adversely affect the leak tightness of the '
valve and thus, Technical Specifications 3/4.6.1.2.e and 3/4.6.4 are i not affected. Since the valve design functions are maintained and '
ASME Code requirements are met, this does not reduce any margin of safety.
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i SE No.: 96-0041 Source Document: Potential Issue Form (PIF) 96-1031 Description of Change This potential issue evaluates the 'use-as-is' disposition of operation with three 18' diameter stainless steel expansion bellows and two 10' diameter stainless steel expansion bellows associated with the Extraction Steam (N36) system having approximately la long through-wall cracks.
Summary l
I. No. Expansion bellows failures, if they were to occur, would have the extraction steam contained within the Main Condenser. The Main Condenser will function to limit any radiological consequences.
Actual loss of condenser vacuum to the point of plant shutdown is not likely. Catastrophic failures of the expansion bellows is bounded by the Loss of Condenser Vacuum and Loss of Feedwater Heating accidents. The Offgas (N64) system will continue to operate as designed. No other systems, structures, or components important to safety are affected by the failure of the expansion bellows.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The catastrophic failure of the expansion bellows within the Main Condenser is bounded by the Loss of Condenser Vacuum and Loss of Feedwater Heating accidents. No new failures or new initiator are expected with this 'use-as-is' disposition. The Main Condenser is not required to affect or support the safe shutdown of the reactor, or to support the operation of reactor safety features. Therefore, l the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The approximately l' long cracks as found on the 10" and 18' diameter expansion bellows within the condensers.have been judged by Engineering to have a marginal effect on the overall capabilities of the expansion bellows. The catastrophic failure of the expansion bellows within the Main Condenser is bounded by the Loss of Condenser Vacuum and Loss of Feedwater Heating accidents. No new failures or new initiator are expected with this "use-as-is' disposition. Therefore, no margin of safety has been changed.
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l l SE No.: 96-0043
! Source Document: EI-0200, Rev. 0 l
Description of Change This instruction provides the mechanism for the onsite placement and retention of Emergency Service Water (ESW) sediments in the chemical cleaning lagoon.
Summary l
I. No. The onsite placement and retention of ESW sediments does not modify plant systems, components, or procedures. It does not rely on, utilize, or indirectly interact with any equipment important to safety. It is bounded by the liquid radwaste tank rupture USAR accident scenario. The criterion specified for the placement and retention of ESW sediments ensures that the concentrations of radioactivity in any release of ESW sediment radioactivity to unrestricted areas is below the concentrations evaluated in the USAR scenario. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR has not changed.
II. No. The onsite placement and retention of ESW sediments does not rely on, utilize, or indirectly interact with any equipment important to safety. Therefore, the possibility of a different type of accident or malfunction is not created.
1 III. No. The onsite placement and retention of ESW sediment is subject to controls that ensure that Technical Specification requirements remain satisfied. Therefore, no margin of safety has been reduced.
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4 SE No.: 96-0044 Source Document: SMRF 96-5008 Description of Change
" This design change installs a vacuum breaker assembly, consisting of two 3/4" check valves attached to a tee in a parallel arrangement, on a horizontal portion of the 12" High Pressure Core Spray (HPCS) suppression
- pool test return line at approximately elevation 625', This modification will resolve water hammer on the 4' HPCS minimum flow by-pass line when i
the E22-F0012 bypass valve opens during conditions of testing, normal operations and accident conditions.
l Summary I. No. The modification will eliminate the potential for water hammer in
B&PV Code,Section III, Subsection NC, 1974 Edition through Winter 1975 addenda. Piping evaluations conclude that piping stresses will remain within the ASME Code allowables. The
- modification will have no effect on existing pipe supports nor on j their loadings. This modification will not affect the integrity of the containment boundary. The check valves on the vacuum breaker
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assembly will be closed when the HPCS line is pressurized. The vacuum breaker assembly is oriented to minimize wetting and spraying of equipment, should a check valve fail to close. This modification will not change, degrade, or prevent any actions described or assumed in the accidents described in the USAR. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The HPCS line is part of the containment pressure boundary, and thus serves an accident mitigation function. The piping modifications have been evaluated per the requirements of ASME B&PV Code,Section III, Subsection NC, 1974 Edition through Winter 1975 i Addenda. The HPCS line is air-tested, and no credit is taken for '
submersion of the pipe end in the suppression pool. Therefore, the integrity of the containment boundary remains the same. The vacuum breaker assembly is oriented to minimize wetting and spraying of equipment should a check valve fail to close. Check valve failure to close will have no measurable effect on system flow performance and any water which does leak by will be directed to the suppression pool. Therefore, the possibility of an accident or malfunction of a ;
different type than any previously evaluated in the USAR is not '
created.
III. No. Technical Specifications 3/4.5.1, ECCS - Standby, and 3/4.5.2, ECCS - Shutdown, are not impacted by this modification. This modification maintains the operability of pipe supports associated with the HPCS system. This modification will have no effect on the design basis of the HPCS containment penetration, containment isolation systems or the containment. Therefore, no margin of safety has been changed.
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SE No.: 96-0045 Source Document: DCN 5355 Description of Change '
This drawing change updates P&ID D302-360, Division III Diesel Generator Jacket Cooling Water System, by incorporating a pressure setting change I and a configuration revision to the Division III Diesel Generator (DG) '
Jacket Cooling Water Expansion Tank.
Summary I. No. The change of the pressure setting for the Division III DG Cooling Water Expansion Tank filler cap and correction of the cap / overflow piping configuration meets the original design of the system. The overall DG performance and the cooling water system performance is unaffected. The change to the filler cap cannot impede actions necessary to mitigate the consequences of accidents, change l
assumptions previously made in evaluating radiological consequences of an accident, or affect any fission product barriers. The changes do not introduce any new failure mechanisms to the Division III DG Cooling Water system. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The design of the Division III Diesel Generator Cooling Water system ,
as described in the USAR is unaffected. All safety-related aspects I of the diesel generator and support subsystems are unchanged. No l new accident initiator or failure mode is introduced. Therefore, i the possibility of an accident or malfunction of a different type ;
than any previously evaluated in the USAR is not created.
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III. No. Technical Specifications 3/4.8 and associated bases are in no way impacted. The Division III DG performance is unaffected. Accident analysis remains unchanged. Therefore, no margin of safety has been changed.
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f SE No.: 96-0046 Source Document: USAR Change Request 96-067 Description of Change This change request analyzes the operation of the plant at power using the Unit 2 startup transformer as the immediately available (primary) source of offsite AC power. The main and auxiliary transformers will be i used as a delayed source of offsite AC power. I Summary I. No. The offsite source configuration described above is in compliance with GDC 17 and Regulatory Guides 1.32 and 1.93. Electrical system ;
operation is not adversely impacted. USAR accident analysis is not :
affected. Therefore, the probability of occurrence or the i consequences associated with in accident or malfunction of equipment '
has not changed. !
II. No. The offsite electrical configuration remains in compliance with GDC 17, and Regulatory Guide 1.32 and 1.93. The electrical system operation has not been affected. Worst case component failure has been evaluated resulting in no adverse impact upon the plant.
Accident analysis remains unchanged. Therefore, the possibility of l an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specification 3/4.8.1.1 is not affected. Worst case component failure was evaluated with no adverse impact upon the i design or operation of the plant. The electrical configuration I remains in compliance with the applicable regulations. Therefore, I no margin of safety has been changed.
SE No.: 96-0047 l Source Document: DCN 5357 Description of Change This drawing change revises drawing D201-146, Electrical Fire Barrier Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated barrier on Tray 1305, located in the Emergency Service Water (ESW) Pump House (Fire Zone ESW-la). Tray 1305 contains Method B circuits for the Division II screen wash system and traveling screens, and a circuit for the Division II ESW pump discharge isolation valve (1P45-F130B) and limit switch (1P45-N131B). In the event of a file in this area, safe shutdown is achieved utilizing either the Method A Method B equipment, depending on the location of the fire. Calculation SSC-005 documents the technical basis to conclude that the screen wash system and traveling screen circuits and components are not required to achieve and maintain safe shutdown. This calculation also demonstrates that, in the event of a postulated fire in the Emergency Service Water Pump House, adequate separation exists between those components required to achieve and maintain safe shutdown such that one train will remain available.
Because of the low fire load, their spatial separation and the ceiling height in excess of 60 feet, a fire in the ESW-la Fire Zone would not impact both ESW safety shutdown trains.
Summary I. No. No changes to the raceways other than the requirement for a fire barrier and no physical changes to the circuits are made. As removing the requirement for a fire barrier can only affect the protection provided to the c.rcuits during a fire, it does not d
affect the probability of the occurrence of any event. Removing the requirement for a one-hour rated fire barrier for these conduits does not affect the ability of the conduits or their associated circuits to function under all other design conditions (including accident conditions), so the ESW system will function as required.
In the event of a fire, the capability of the ESW system to function designed to perform safe shutdown functions and to minimize radioactive releases to the environment will be maintained.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with the ESW system are made by this activity. No new system interactions are created by this activity. Removing the requirement to have a rated fire barrier on these raceways can only affect whether or not circuits could be damaged in a fire and does not affect their ability to function under other conditions. Since these circuits will function under current design conditions and because other events are not postulated to occur concurrently with a fire, at least one train of ESW will remain available in the case of a fire in Fire Zone ESW-la.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
SE No.: 96-0047 (Cont.)
Summary (Cont.)
l III. No. Since this activity makes no changes to ESW system components, this l activity does not affect the requirements for operability of the ESW l system. Since this evaluation has shown that removing the l requirement for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barriers will not affect the I ability of the plant to achieve and maintain safety shutdown, l Appendix R compliance is ensured Therefore, no margin of safety I has been changed. '
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I SE No.: 96-0048 Source Document: DCH 5358 Description of Change This drawing change revises drawing D201-146, Electrical Fire Barrier Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated barrier on conduits 1P45R13A, 1P45R39A and 1R332200A, located in the Control Complex 574' elevation (Fire Zone CC-la) . These circuits only provide remote indication of the flow on the outlet side of the Emergency Closed Cooling Water (ECCW) heat exchanger. Any fire induced faults will not impact the operation of the Emergency Service Water (ESW) or ECCW systems as analyzed for safe shutdown. Calculation SSC-015 demonstrates than an alternative means of verifying Emergency Service Water flow will be available in the event that a fire in Fire Zone CC-la disables the Method A flow transmitters for remote indication. Since there are no accidents requiring Control Room evacuation postulated during the fire in Fire Zone CC-la, the remote shutdown panel would not be used. In the event that a fire occurred, the operation of the Division I ESW system could be verified by flow indication through the other heat exchangers.
Therefore, it is not necessary to provide protection for conduits IP45R13A, 1PeJ39A and 1R332200A in Fire Zone CC-la.
Summary I. No. No changes to the conduits other than the requirement for a fire barrier and no physical changes to the circuits are made. As removing the requirement for a fire barrier can only affect the protection provided to the circuits during a fire, it does not affect the probability of the occurrence of any other event.
Removing the requirement for a one-hour rated fire barrier does not l affect the ability of the associated circuits to function under all '
other design conditions (including accident conditions), so the ESW system will function as designed. In the event of a fire, the capability of the ESW to function designed to perform safe shutdown functions and to minimize radioactive releases to the environment will be maintained. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with the flow instruments are made by this activity. Thus, no new system interactions are created by this activity. Removing the requirement to have a rated fire barrier on these conduits can only affect whether or not these conduits could be damaged in a fire and does not affect their ability to function under other conditions. Since these conduits and circuits will function under current design conditions and because other events are not postulated to occur t
concurrently with a fire, and other means of verifying ESW flow are available in the case of a fire in Fire Zone CC-la, the capability of the ESW system to function as designed is not affected by this activity. Therefore, the possibility of an accident or malfunction l of a different type than any previously evaluated in the USAR is not created.
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SE No.: 96-0048 (Cont.)
Summary (Cont.)
III. No. Since this activity makes no changes to ESW flow instruments or the ESW system components, this activity does not affect the requirements for operability of the ESW system. Since this evaluation has shown that removing the requirement for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barriers will not affect the ability of the plant to achieve and maintain safety shutdown, Appendix R compliance is ensured.
Therefore, no margin of safety has been changed.
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SE No.: 96-0049 Source Document: Potential Issue Form (PIF) 95-2191 Description of Change l
This potential issue evaluates the "use-as-is" disposition of having ;
several flow restricting orifice plates installed in various plant ,
systems which could be subjected to stresses that could cause plate !
deformation. The orifice plates are non-ASME Code components, will not cause the degradation of any system components, and will not prevent the systems from performing their intended functions. ]
1 Summary l
I. No. This disposition will not alter the control, operation or expected l response of the systems in question. The disposition cannot create i an environment that will be detrimental to the involved systems or '
any system they interface with. The disposition will not alter, i degrade or prevent actions described or assumed to occur as l described in the USAR. Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or l malfunction of equipment important to safety will be increased.
II. No. This PIF disposition will have no adverse affect on any system, structure, or componer' or the way they will react to normal and ,
abnormal transients. .. .sposition cannot result in any new l equipment failures, no new initiators or contributors for any new event. Therefore, this disposition does not create the possibility !
of an accident or malfunction of a different type than any l previously evaluated. j III. No. This disposition will not degrade the systems' capabilities to mitigate the effects of postulated transients. Thus, there will be no reduction in the margins of safety.
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t SE No.: 96-0050 '
Source Document: Potential Issue Form (PIF) 96-0883 Description of Change ,
This potential issue evaluates the 'use-as-is" disposition for the 24" Circulating Water (N71) system Cooling Tower blowdown / overflow line.
Physical inspection of the piping confirmed that leakage / seepage exists at several locations. Continued use of the line in its present condition will have no effect on system or plant operation. ;
t Summary I I. No. Use of the line in its present condition has no effect on N71 system !
or plant operation. Any leakage out of the line during normal plant :
operation will have no effect on N71 system flow to the Main or Auxiliary Condensers. As a result, the probability of a Loss of ;
Condenser Vacuum accident is not increased. Potential flooding is considered bounded by the design basis flooding accidents. Plant equipment that is important to safety would remain unaffected. The ,
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N71 system is nonsafety-related and is not required for safe shutdown of the reactor. The N71 system is also non-radiological.
Leakage or failure of the line has no radiological consequences.
Therefore, use of the line does not increase the probability of occurrence or consequences of any accident or malfunction of l equipment important to safety previously evaluated in the USAR.
II. No. Use of this line in its present condition has no impact on system or !
plant operation. Associated leakage / seepage does not create an unreviewed environmental question. The N71 system is nonsafety, non-radiological and not required for safe shutdown of the reactor.
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. Plant operation and effluents remain unaffected by the use of the !
line. Failure of the line would result in the commencement of an {
orderly plant shutdown in accordance with plant procedures.
Flooding remains bounded by existing USAR analyses. Therefore, no margin of safety has been reduced.
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SE No.: 96-0051 l Source Document: PAP-0101, Rev. 8, TC-6 Description of Change This proce<**s: change describes a management reorganization involving the onsite computer personnel.
Summary I. No. This change does not alter the plant in any way. No functions or activities have been eliminated. The onsite personnel involved continue to meet the ANSI N18.1-1971 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 ANSI N18.1-1971 and the Technical Specifications. Therefore, this change will not reduce any margin of safety.
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i SE No.:.
96-0052 ,
Source Document: DCN 5301 Description of Change This drawing change revises P& ids D302-791 and -792, Residual Heat Removal System, by eliminating the demineralized wet lay-up requirements '
of the Emergency Service Water (ESW) portion of the Residual Heat Removal (RHR) system heat exchangers.
Summary
! I. No. Based on the frequency that the ESW e, em is chemically treated, I
stagnant water will not cause degradation beyond design corrosion
! allowable of the RHR heat exchanger. Testing (performance testing and eddy current examination) and water box / tube inspections show that the RHR heat exchanger still meets its design requirement to remove heat for all modes of operation. The elimination of wet lay-up cannot cause a loss of shutdown cooling capability and cannot j cause a LOCA event to occur. Overall system performance of the ESW l
and RHR systems is maintained. The RHR and ESW system will not be ,
operated any differently than the way the systems have been operated in the past. Water is still present in the ESW side of the RHR heat exchanger thus, a water hammer event will not occur. Therefore, the probability of occurrence or the consequences associated with an l accident or malfunction of equipment has not changed.
l l II. No. This change does not involve an initiator or failure that could l create the possibility of a new accident. No hardware is being added or modified by this change. Demineralized wet lay-up is no l longer necessary to maintain the RHR/ESW systems functional.
Overall system performance of the ESW and RHR systems is maintained.
The RHR and ESW system will not be operated any differently than the l way the systems have been operated in the past. Therefore, the possibility of an accident or malfunction of a different type than l any previously evaluated in the USAR is not created.
l III. No. The RHR heat exchangers remain capable of performing their design functions. Cooling capacity (ESW) for safety-related equipment is not reduced during normal and accidents conditions. Heat removal capability for removing decay heat remains unchanged. Therefore, no margin of safety has been changed.
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l SE No.: 96-0053 Source Document: DCN 5412 Description of Change This drawing change revises drawing D201-146, Electrical Fire Barrier 4
Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated
, barrier on conduits 1M25C51X, IM25C19X, 1R33C5459X, 1R33C5460X and 1R33C5461X, routed through Fire Areas ICC-4a, ICC-4e, ICC-4f, CC-6 and j ICC-6 in the control Complex. These conduits contain circuits associated with interlocks between differential pressure instruments M25-K070A/B and M26-K050A/B and the trip function for the Control Room Ventilation (M25) 1 supply fans (M25-C001A/B), circuits for the supply fan flow control 1 dampers (M25-F260A/B), and circuits for instrumentation annunciation. '
Calculation SSC-007 demonstrates that while fire induced faults of the subject circuits would disable certain annunciation and control functions
- in the M25 system, operator action can be taken, consistent with l
i established procedures, to place the M25 system in the Recirculation '
Mode. Once in the Recirculation Mode, the affects of fire induced faults on the subject circuits will not prevent M25 operation, to maintain Control Room habitability. Therefore, the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barrier protection provided on the subject conduits is not required to achieve
- and maintain safe shutdown following a fire in the above listed fire creas. ,
Summary I. No. No changes to the conduits other than the requirement for a fire barrier and no physical changes to the circuits are made. As ;
removing the requirement for a fire barrier can only affect the !
protection provided to the conduits during a fire, it does not affect the probability of the occurrence accident of any or event requiring Control Room isolation. Since accidents are not assumed to occur coincidentally with a fire and single failure need not be postulated, removing the requirement for a one-hour rated fire barrier for these conduits does not affect the ability of the equipment or its associated circuits to function under all other i design conditions (including accident conditions). Hence, the Control Room isolation mode of M25 will function as designed. In the event of a fire, the capability of the M25 nystem to function as designed to support safe shutdown functions will be maintained.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with this equipment are made by this activity. Removing the requirement to have a rated fire barrier on these conduits can only affect whether or not these i conduits could be damaged in a fire and does not affect their i ability of the equipment to function under other conditions. Since ;
these conduits and circuits will function under current design i conditions and because other events are not postulated to occur j
l SE No.: 96-0053 (Cont.)
l Summary (Cont.)
concurrently with a fire, Control Room isolation capability is not adversely affected and will function as designed. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Since this activity makes no changes to M25 system or operation in the Recirculation Mode, this activity does not affect the requirements for operability of the M25 system. Since this evaluation has shown that removing the requirement for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barriers will not affect the ability of the plant to achieve and maintain safe shutdown, Appendix R compliance is assured.
Therefore, no margin of safety has been changed.
l SE No.: 96-0054 i Source Document: DCN 5413 Description of Change 1
This drawing change revises drawing D201-146, Electrical Fire Barrier Details, to reflect the deletion of the requirement for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated j barrier on conduits 1M25C37A, 1M25C140A, 1R33C2463A, 1R33C3276A, i 1M25C56B, 1M2EC144B, 1R33C2461B and 1R33C3282B, routed through Fire l Areas 1CC-6 and 2CC-6 in the Control Complex, 679' elevation. These !
conduits carry air control solenoid and limit switch circuits for Control l Room Ventilation (M25) system outside air isolation dampers OM25-F020A '
and OM25-F020B. Calculation SSC-014 demonstrates that fire induced faults of the subject circuits could disable outside air isolation dampers OM25-F020A or 0M25-F020B. However, this calculation confirms that redundant outside air isolation dampers OM25-F010A or 0M25-F010B would remain available following a fire in Fire Area 1CC-6 or 2CC-6. j Therefore, the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated fire barrier protection provided on the i subject conduits is not required to achieve and maintain safe shutdown following a fire in Fire Areas 1CC-6 or 2CC-6.
Sammary I. No. No changes to the conduits other than the requirement for a fire barrier and no physical changes to the circuits are made. As removing the requirement for a fire barrier can only affect the protection provided to the conduits during a fire, it does not affect the probability of the occurrence of any accident or event requiring Control Room isolation. Since accidents are not assumed to occur coincidentally with a fire and single failure need not be postulated, removing the requirement for a one-hour rated fire barrier for these conduits does not affect the ability of the equipment or its associated circuits to function under all other design conditions (including accident conditions). Hence, the Control Room isolation dampers will function as designed. In the event of a fire, the capability of the M25 system to function as designed to support safe shutdown functions will be maintained.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. No physical changes to the circuitry associated with these dampers l are made by this activity. Removing the requirement to have a rated fire barrier on these conduits can only affect whether or not these l conduits could be damaged in a fire and does not affect their I ability of the dampers to function under other conditions. Since i these conduits and circuits will function under current design conditions and because other events are not postulated to occur concurrently with a fire, Control Room isolation capability is not adversely affected and will function as designed. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
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1 SE No.: 96-0054 (Cont.)
Summary (Cont.)
III. No. Since this activity makes no changes to Control Room isolation I dampers or their operation, this activity does not affect the requirements for operability of the M25 system. Since this evaluation has shown that removing the requirement for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated i
fire barriers will not affect tne ability of the plant to achieve ;
and maintain safe shutdown, Appendix R compliance is assured.
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i SE No.: 96-0055 Source Documegt,; PSTG, Rev. 4, TC-1 i Description of Change This Perry Specific Technical Guidelines (PTSG) change updates entry conditions due to various setpoint. changes and reinstates step PC/H-5 (containment spray using only those pumps not required for adequate core !
cooling).
Summary i
The containment spray initiation pressure limit and termination I. No.
pressure limit used in step PC/H-5 are the same as those used in the containment pressure control guideline. The containment spray system is operated in the same manner for primary containment hydrogen control as it is for primary containment pressure control.
! For accidents previously evaluated in the USAR, step PC/H-5 will not l be performed since in all analyzed accidents, primary containment ;
hydrogen concentration stays below the Hydrogen Deflagration '
l Overpressure Limit (HDOL). Therefore, this action does not alter ,
l the probability or consequences of an accident or malfunction of '
l equipment important to safety previously evaluated in the USAR.
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l II. No. The containment spray initiation pressure limit and termination pressure limit used in step PC/H-5 are the enme as those used in the containment pressure control guideline. The containment spray system is operated in the same manner for primary containment l
hydrogen control as it is for primary containment pressure control.
Therefore, this action does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR.
l III. No. The PSTG as modified by this change does not adversely affect I systems or components required for safety. Therefore, no margin of i safety has been reduced.
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SE No.: 96-0056 Source Document: NR 94-S-521, Rev. 1 Description of Change This Nonconformance Report (NR) documents a condition involving the Offgas (N64) system in which pieces of a stainless steel Steam Jet Air Ejector (SJAE) bellows are missing and are presumed lost in the B N64 loop. The disposition of Revision 0 required debris associated with the i
missing bellows to be located and removed from the piping system. Work I was performed to locate and retrievn the missing bellows pieces. !
However, only a few small pieces were found. Approximately 20% of the l original bellows is still missing. Since the bulk of the missing mass of i the bellows remains to be found, the conparative significance of the j identified pieces is judged negligible. Therefore, disposition is being '
revised to permanent 'use-as-is'.
Summary I. No. The Offgas system will continue to meet the required design and construction standards with the exception of cleanliness 'D' requirement per PAP-0204. The operation of the system with debris is not expected to have any effect on the overall system performance or reliability. Piping pressure boundary wall thinning caused by this condition to a point where a detonation could breach the pressure boundary of the N64 system is not feasible. This condition clearly has ao effect on the probability of a seismic event, and the consequences will remain unchanged. The N64 system remains i available as designed to control the release of gaseous effluent I from the Main and Auxiliary Condensers. This disposition does not I impair the capability to detect a failure within the Offgas system, to isolate the Offgas system in the event of a failure, or affect the function of the N64-related indicators or annunciators it the main Control Room. Therefore, the probability.of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Allowing the use of the N64 system with debris is in no way related to an activity initiator or failure not considered in the USAR.
Tube leakage within the N64 preheater is considered normal. The existing debris is not expected to increase the potential for tube failures within the preheater. The operation of the system with debris is not expected to have any effect on the overall system performance or reliability. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Technical Specification B3/4.ll.2.6 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 flammability limits of 4% hydrogen. The Offgas system will continue to meet the required design and construction standards, maintaining the original designed factors of safety.
Therefore, no margin of safety has been changed.
1 SE No.: 96-0057 Source Document: NR 94-S-482, Rev. 3 Description of Change This Nonconformance Report (NR) documents the permanent use of a replacement valve for 1G41-F0522. This valve is in the fuel pool cooling return line to the upper pools. The original valve was stainless steel while the replacement is carbon steel.
Summary I. No. This change maintains the ASME Code requirements. The replacement does not adversely affect the valve's function as a process flow path and containment isolation boundary. The new valve is of a similar design configuration such that no new failure modes are introduced. In addition, component redundancy and independence are not reduced by this single check valve replacement. The system's ability to provide cooling for spent fuel is not adversely affected.
Therefore, neither the probability of occurrence nor the consequences of a previously analyzed accident or malfunction of equipiuent will be increased.
II. No. The replacement valve does not create any new systems, add any new equipment types or compromise any existing systems, structures or components. In addition, the change affects a single active check valve and does not add any new interconnection with other components or systems. Thus, there is no increase possibility of common mode or common cause failures. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than any previously evaluated.
III. No. By adherence to the ASME Code for pressure boundary components, Technical Specification 4.0.5 is not affected. This disposition maintains containment integrity and associated valve leakage requirements and thus, Technical Specifications 3/4.6.1.2.b and 3/4.6.4 are not affected. The new valve will provide the necessary system flow capability such that requirements of Technical Specification 3/4.9.9 are maintained. Further, since ASME Code requirements and system / valve design requirements are maintained, this change will not reduce any ma*]in of safety.
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SE No.: 96-0058 Source Document: DCN 5383 Description of Change This drawing change revises P&ID D302-001, System Diagram Symbols, to clarify the guidance associated with the use of pipe caps and plugs.
Summary I. No. This drawing change is editorial only. The change will be to more clearly define the use of pipe cap and plug symbols on P& ids. This editorial change does not impact the operability of functionality of any system. All systems will continue to function as designed. No physical changes to plant equipment are made by the change.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The change is editorial only. The change does not involve a failure not considered in the USAR. No change to plant hardware is involved. The change creates no system interactions. The capability of any system, structure or component is not reduced.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The piping system diagrams are not governed by the Technical Specifications. The change does not cause physical changes to the plant or to any operating procedures. Therefore, no margin of safety has been changed.
l SE No.: 96-0059 Source Document: DCN 5380 !
Description of Change This drawing change serves to document the permanent installation of the sealant injection adapters and to identify valve IN36-F0511 as disabled due to sealant injection repair or ' killed". The permanent configuration :
of the subject valve will be documented on the applicable plant drawings.
Summary I. No. The leak sealant repair implemented for 1N36-F0511 was performed in accordance with established plant and engineering approved procedures, and has been shown not to have an adverse impact on the structural integrity of the valve or the affected piping. The i portion of piping in which the disabled valve is installed has been '
permanently abandoned under an approved design change. The system configuration resulting from this change does not increase the probability of occurrence consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. The change serves to document the permanent installation of the sealant injection adapters and to identify the valve as disabled due to sealant injection repair. The portion of piping in which tha disabled valve is installed has been permanently abandoned under an approved design change. The change does not alter the function er operating characteristics of the affected systems and creates no new ;
system interactions. Therefore, the change does not create the i possibility for an accident or malfunction of a different type than My previously evaluated in the USAR. I III. No. Operation of the Auxiliary Steam system, Extraction Steam system and the Direct Contact Heater is not governed by the Technical Specifications. The leak sealant repair was performed in accordance with established plant and engineering approved vendor procedures which ensure that the leak sealant repair and associated hardware that will remain installed permanently will not have an adverse impact on the affected component or interfacing systems. The activity represents no threat to the Feedwater system, reactor water chemistry or reactor coolant pressure boundary. Therefore, the change does not reduce any margin of safety.
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SE No.: 96-0060 :
Source Document: Emergency Plan, Rev. 13, TC-1 l Description of Change I I
This change to the Emergency Plan revises various plan-related k activities. The changes include the use of the County Decontamination :
Centers, clarification of notifications to company management, and I deletion of certain training requirements.
Summary .
t I. No. This revisite does not direct or impact the operation or design of ;
any plant structure, system or component. Accident initiators are j 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.
II. No. This revision 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.: 96-0062 Source Document: DCN 5018 Description of Change <
This drawing change revises P&ID D302-301, Hydrogen Supply System, by -
removing two 1/2' to 3/4" expanders and a section of 3/4" piping. The expanders and associated piping are not installed. A bushing, integrated into the design of valves 1N35-F503A and 1N35-F503B accomplishes this function.
I. No. The correction of this drawing does not affect the operation of the ,
nonsafety-related Hydrogen Supply system. The function of the expander shown currently on the drawing is accomplished by a :
subcomponent (bushing) incorporated in the valve design used for 1N35-F503A and B. No field work is required to accomplish this change. This activity dova not affect plant systems, structures, and components assumed to L' accident initiators. Therefore, this activity does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated in !
the USAR.
II. No. This activity is a drafting change to a plant drawing to delete the ;
expander shown on the P&ID. It does not affect function or operation of the Hydrogen Supply system as evaluated in the USAR, because the function of the expander is accomplished by a subcomponent (i.e., a bushing) in the valve design utilized for 1N35F503A and B. No new system interactions are created by this activity. No field work is required to implement this change.
Since the function and operation of the Hydrogen Supply system are not affected by this change, this change will not create the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR.
III. No. The Technical Specifications do not address the Hydrogen Supply system. The function of the expanders is accomplished through a bushing in the valves downstream of the expanders. Thus, system operation and function are not affected by this activity.
Therefore, no margin of safety has been reduced.
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i SE No.: 96-0063 i
- Source Document: DCN 5059 1
j Description of Change
' This drawing change revises P&ID D302-653, Fuel Pool Filter Demineralizer System, by removing the Fail Open (FO) and Fail Close (FC) designations !
! from various Fuel Pool Cooling and Cleanup (G41) system valves. This is an editorial drawing change only with no field work being performed. The '
design, function, and operation of these valves is not affected by the
- editorial changes.
j _ Summary 4
I. No. No physical work is being performed by this change. The change corrects an error by removing the "F0" or "FC' designation of the l nonsafety-related air operated valves. There is no impact to the ,
Fuel Pool Cooling and Cleanup system operation. Since there is no ;
physical change to the plant configuration or methodology by this '
l change, the radiological consequences as previously analyzed will ,
, not increase. Therefore, the probability of occurrence or the
! consequences of an accident or malfunction previously evaluated in ;
j the USAR has not changed. I II. No. The change is limited to drawing revisions that reflect the proper
- failed position of Fuel Pool Cooling and Cleanup system l nonsafety-related valves. The changes do not alter the valves or
- the ability of the system to perform its design function.
! Therefore, this change will not create the possibility of an accident or malfunction of a different type than any previously
- evaluated in the USAR.
III. No. This change corrects a drawing error concerning the failure position of nonsafety-related valves. As such there is no change to the j limiting conditions for operations, surveillance requirements, or
, design basis of the Technical Specification. Therefore, no margin j of safety has been reduced.
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1 SE No.: 96-0064 i Source Document: DCN 5068 1
Description of Change j j This drawing change resolves a drawing discrepancy between drawings i
D921-614, Reactor Building, and D302-651, Fuel Pool Cooling.and Cleanup System, concerning the elevation of the spent fuel pool overflow line.
) The results of a survey indicated that the overflow elevation was i- 688'-10". The design, function, and operation of the Fuel Pool Cooling l.
and Cleanup (G41) system is not affected by this editorial change.
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- Summary 5
i I. No. This change corrects an elevation error on design drawings. There j is no physical change to plant configuration or hardware. There is
- no impact to function, operation, or accident response capabilities
{ of the Fuel Pool Cooling and Cleanup system. The correction on the drawings cannot cause safety equipment to malfunction. Therefore, the drawing change does not increase the probability of an
- occurrence or the consequences of an accident or equipment i
d malfunction beyond what was previously evaluated in the USAR.
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! II. No. The change is limited to a drafting correction to reflect the l correct elevation of the overflow line. This drawing change does 1 j not affect the physical installation of the spent fuel pool or its l 1 ability to perform its design function. Therefore, this change will j not create the possibility of an accident or a malfunction of
- equipment of a different type than any previously evaluated in the i USAR.
I III. No. The change corrects a drawing error concerning the elevation of the l spent fuel pool overflow line. Technical Specifications relating to i the spent fuel pool relate to maintaining a minimum water level and i a limiting maximum temparature. The maximum level is not a Technical Specificatier, concern. As such, there is no change to the j limiting condiaions fo: uperations, surveillance requirements or j
design bases of the Technical Specification. Therefore, no nargin of safety has been redu:ed.
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SE No.: 96-0066 Source Document: DCN 5142 Description of Change This drawing change revises P& ids D302-791, Emergency Service Water System, and D302-671, Reactor Water Cleanup System, by adding / correcting Emergency Response Information System (ERIS) data points. The revision accomplishes an editorial drawing change only with no field work being performed. The design, function, and operation of tt; Emergency Service Water (ESW) and Reactor Water Cleanup (RWCU) systems cre not affected by these editorial changes.
Summary I. No. This drawing change corrects an error and an omission on design drawings in the ESW and RWCU systems. These computer points are already connected. The drawing updates do not change any safety-related functions of the systems or the plant. There is no physical change to plant configuration or hardware. As such there is no impact to the ESW or RWCU syste~1s' ability to perform their design functions. Therefore, the probability of an occurrence or the consequences of an accident or malfunction beyond what was previously evaluated in the USAR has not changed.
II. No. No physical work is being performed. The change corrects ERIS computer input points in the RWCU and ESW systems. This change does not physically affect either system's ability to perform their design functions. Therefore, the possibility of a different type of accident or malfunction of equipment important to safety previously evaluated in the USAR is not created.
III. No. This change corrects drawings errors concerning the ERIS computer input signal locations. There are no signal changes or setpoint changes. As such there is no change to the limiting conditions for operations, surveillance requirements, or design bases of the Technical Specifications. Therefore, no margin of safety has been reduced.
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- SE No.
- 96-0067 Source Document: DCN 5087 Description of Change This drawing change revises vendor drawings 4549-40-343 and 4549-40-359
- to incorporate alternate disc material for valves 1E22-F0026, 1E22-F0031, i 1E51-F0502, 1G33-F0039, 1G33-F0040, 1G33-F0051A, and 1G33-F0051B. The i valve vendor is now providing a replacement valve disc made of SA105
- (forging material) in lieu of the originally supplied SA216 WCB (casting i material). These two materials have similar chemistry and mechanical l
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properties with identical minimum yield and tensile strength requirements. The hardfacing on the replacement disc will be of the same material as of the original disc. This disc replacement does not alter
, the' function of the component.
i l Summary 1
l I. No. This drawing change updates the appropriate vendor drawings to indicate the replacement valve disc material. As stated in CERF 1167, the form, fit, and function of the new valve disc have not changed. The new material is a vendor recommended replacement for these valves. The new material meets all the same code
] requirements as the original disc. As such, this activity does not increase the probability of occurrence or the consequences of an 1
- accident or malfunction previously evaluated in the USAR.
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j II. No. The replacement forged disc will function the same as the old cast disc. Material properties meet ASME Code requirements and have 4
equivalent attributes. There will be no perceived operational or
, response difference of the valves with the new discs. Therefore, this change will not create the possibility of an accident or malfunction of a different type than any previously evalua!cd in the i j USAR.
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, III. No. This drawing change updates vendor drawings to reflect the material change that was authorized by CERF 1167. The replacement material has been evaluated as an equivalent material to ensure proper function of the valve. As the valve with the replacement disc will function as before, there will be no change to the limiting conditions for operations, surveillance requirements, or design bases of the Technical Specification. Therefore, no margin of safety has been reduced.
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SE No.: 96-0068 Source Document: DCN 5331 Description of Change This drawing change corrects Master Parts List (MPL) numbers and wire i marks on various plant drawings to reflect the as-built condition of temperature switches 1N21-N0185A/C, and computar points 1N21-C007/C009. ,
I Summary l
l I. No. The Main Turbine is not required for safety-related shutdown of the )
reactor and is not affected by this drawing change. There are no physical changes being made to the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The Main Turbine emergency trip system design basis, function, and operation is not impacted by this drawing change. No new common mode failures of redundant safety-related systems are introduced by the drawing change. No new accident initiators are introduced by this drawing change. No new equipment malfunctions are introduced by this drawing change. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The Technical Specifications and associated bases are not affected.
Therefore, this change does not dacrease any margin of safety.
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SE No.: 96-0069, 96-0081 Source Document: DCN 5375 Description of Change
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This drawing change revises P& ids D302-671 and -672, Reactor Water Cleanup System, by incorporating four high point vent valves on various impulse lines associated with panel 1H22-P0002.
Summary l i
I. No. The valves are manual, non-active, normally closed valves utilized !
as high point vents located in instrumentation lines. Their only '
function is to be opened during instrument line filling and venting operations. Otherwiec, during normal plant operations, they are closed and their function is to maintain pressure boundary. The only purpose of this drawing change is to incorporate these valves on drawings D302-671 and D302-672 in order to reflect the as-built condition of the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The valves are manual, non-active, normally closed valves utilized i as high point vents located in instrumentation lines. Their only {
function is to be opened during instrument line filling and venting l operations. Otherwise, during normal plant operations, they are i closed and their function is to maintain pressure boundary. The only purpose of this drawing change is.to incorporate these valves on drawings D302-671 and D302-672 in order to reflect the as-built condition of the plant. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The valves are manual, non-active, normally closed valves utilized as high point vents located in instrumentation lines. There only function is to be opened during instrument line filling and venting operations. Otherwise, during normal plant operations, they are closed and their function is to maintain pressure boundary.
Therefore, no margin of safety has been changed.
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- 96-0070
- Source Document
- Potential Issue Form (PIF) 95-2231
! Description of Change
- This PIF evaluates an error in the Station Blackout (SBO) analysis for i Perry. The Safety Relief Valve (SRV) lo-lo set lift setpoints had been i set incorrectly in the analysis input. Also, a change in the Heat I
Capacity Limit (HCL) curve had been made since the original SB0 analysis was performed. _ Evaluation of the SRV setpoint error indicates virtually no impact upon the peak suppression pool temperature in the SB0 analysis.
l The effect of updating the HCL curve indicates that the time to onset of manual depressurization should be reduced by about 30 minutes and would i occur at a suppression pool temperature of ~1260F vice 1320F. This time j reduction has no adverse impact upon the SB0 analysis.
f Summary i
i I. No. The input error to the SB0 analysis relates to reactor vessel j pressure relief after the event has occurred. The new HCL is an accident coping mechanism after the onset of the event. Neither one
- of these changes would affect the final peak suppression pool temperatcre. The reason is that the decay heat is the same in all i
cases, thus the total energy deposition to the suppression pool is j the same. There is no breach of containment, hence no increased i does to plant personnel and to the public. Therefore, the i
probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
i II. No. The changes have predictable effects upon the SB0 accident response, 1
and the suppression pool temperature is bounded by the existing :
, licensing analysis. A change in the SRV setpoints nor the timing of I manual depressurization has little influence upon the total energy i transfer from the vessel to the suppression pool. Hence, the peak pool temperature at the end of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is virtually unchanged.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created, i
l III. No. The correction the SRV setpoint error and use of the new HCL curve j j
has no impact upon the peak suppression pool temperature.
i Sensitivity calculations have been performed with further shows
! these changes do not adversely impact the SB0 analysis. Therefore,
- no margin of safety has been reduced. ;
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SE No.: 96-0071 I Source Document: Emergency Plan, Rev. 13, TC-2 Description of Change This change to the Emergency Plan revises various plan-related activities. The changes include the revised Emergency Action Level entry criteria, revised the fission product barrier figure, and incorporated various grammatical corrections.
Summary I. No. This change does not direct or impact the operation or design of any pl: 1. structure, system or component. 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.
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.: 96-0074 Source Document: Improved Technical Specifications Change Request 96-020 Description of Change The proposed Technical Specification changes are divided in three categories; technical, administrative, and editorial. With the exception of increasing the minimum diesel generator day tank volume in SR 3.8.1.4, the proposed changes are administrative or editorial.
Summary I I. No. The diesel generators are not an initiator of any accident, but are, rather, mitigators of accidents. The proposed change to the minimum i day tank volumes does not change this. The increase in the minimum l
! day tank volume does not modify the design of the diesel generators or their ability to perform their design function. Since the surveillance frequencies for the diesel generators and associated equipment are unchanged, equipment reliability and availability is unchanged. The other proposed changes are either administrative or l
editorial in nature. Neither of these types of changes impact the plant design or operation. Thus, the proposed activity does not increase the probability of occurrence or the consequences of an i
accident or a malfunction of equipment important to safety previously evaluated in the USAR.
II. No. No new equipment is introduced into the plant. No change in operation of existing equipment is proposed. The size of the diesel generator day tanks is not increased, hence there is no increase in the fire loading in the rooms. No change in operation of existing equipment is proposed. The other proposed changes are either administrative or editorial in nature. Neither of these types of changes impact the plant design or operation. Thus, the proposed activity does not create the possibility of an accident or malfunction of equipment to safety of a different type t'. tan any previously evaluated in the USAR.
III. No. The bases for the Improved Technical Specifications, as well as the SER, SSER, Standard Review Plan, were reviewed. No margin of safety for the Improved Technical Specifications was found to be reduced.
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SE No.: 96-0075 Source Document: TM 1-96-015 Description of Change This temporary modification installs a plug in both the upstream and downstream instrument taps of the Motor Feed Pump balance line flow element, 1N27-N600. The temporary modification also cuts and removes sections of the instrument piping including valves 1N27-F788 and 1N27-F789. The remaining line segments will then be capped off. The installation of this temporary modification results in disabling of the local indication of the balance disc position indicator located at 1N27-N601 and process computer point BA099.
Summary I. No. The temporary modification meets or exceeds the applicable piping specification (G1-4) requirements, and as such conforms to the original piping design specifications for this system. This change does not impact the function of Feedwater system, or function or design of any safety-related equipment. This change had no effect upon any previously defined accidents or on equipment important to safety. Therefore, the probability or consequences of any previously analyzed accidents of malfunctions of equipment important to safety has not changed.
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1 II. No. The temporary modification does not change the Feedwater system's '
operation or function, or any other interfacing system's operation or function. The temporary modification conforms to the existir.g line specification. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The temporary modification conforms to the applicable specifications and design requirements. It does not result in a change to the operation or function of the plant. Therefore, no margin of safety has been reduced.
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SE No.: 96-0076 Source Document: TM 1-96-016 Description of Change This Temporary Modification (TM) bypasses faulty level switches on Main, Reheat, and Extraction Steam (N22) system drain valves 1N22-F0170 and 1N22-F0300B. The TM also bypasses an interlock with extraction block valves 1 Nil-F0110B/1 Nil-F0115B due to misalignment of limit switches on valve IN22-F0010 which permits these valves to close.
Summary I. No. This Temporary Modification (TM) meets the existing design, material, and construction standards for the N22 system.
Additionally, the overall N22 system performance will not be impacted. There are no accidents described in the USAR which are impacted by this change. There is no equipment that is important to safety (ITS) that is impacted by this change. However, the malfunction of the involved equipment can pose challenges to safety systems should the nonsafety equipment malfunction.
Valve IN22-F0300B removes condensation from the supply of the Reactor Feed Pump Turbine (RFPT) steam line. Credit is being taken for cycling this valve once per day per plant rounds to mitigate water induction into the RFPT; thereby, minimizing a loss of I feedwater flow event. The TM on this nonsafety system will not degrade any structure, system, or component such that any safety function will be altered from the current description in the USAR.
Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The TM bypasses certain control functions inherent in the existing design of the N22 system. However, no different accidents or malfunctions are created. The likelihood of water backing up into the evaluated components has been addressed and has been determined to be no more consequential than any of the existing accidents evaluated in the USAR. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Portions of the N22 system are covered by the Technical Specifications. Namely the main steam stop valves, the turbine stop i and control valves, and the main steam outboard isolation valve before seat drain valves. No portion of the involved Technical Specifications are being impacted by this TM. Therefore, no margin of safety has been changed.
SE No.: 96-0077 Source Document: USAR Change Request 96-095 Description of Change This change request removes valve P42-F0551 from the " Intended Use of Light (s)" column for Lighting Unit 1R71-S0205 and adds it to the
" Intended Use of Light (s)" column for Lighting Unit 1R71-L0206A. The change corrects a discrepancy in USAR Table 9A.3-2 which will describe the actual plant condition.
Summary I. No. Appendix R Emergency Lighting Unit 1R71-LO206A is a two lamp fixture that is located near valve P42-F0551. Lighting Unit 1R71-LO206A is located above valve P42-F0300A with one of its two lamps aimed at valve P42-F0300A and the other lamp illuminating an access aisle for both valves P42-F0551 and P42-F0300A. With this configuration and based on a walkdown of the area, valve P42-F0551 receives sufficient light from the Emergency Lighting Unit 1R71-LO206A. This identification of applicable emergency lighting units cannot increase the probability of a fire. This change does not alter the design or operation of the plant. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change is limited to emergency lighting components required for access or illumination of safe shutdown / safety-related equipment in various plant areas. None of these emergency lighting components can initiate a design basis accident or transient. There is no effect on the physical or schematic configuration of the required emergency lighting. Therefore, the possibility of an accident or malfunction of a different type than any previously evalaated in the USAR is not created.
III. No. This change ensures a complete and correct listing of emergency lighting components / systems is properly documented in the USAR. The Appendix R emergency lighting units are not addressed in the Technical Specifications. This change does not affect the actual function of the lighting system, any operational limit, or the bases of any Technical Specification. Therefore, no margin of safety has been changed.
SE No.: 96-0079 Source Document: DCN 5424 SCRs 0-96-1000 through 1014 SCRs 1-96-1042 through 1055 Description of Change These changes revise various Liquid Radwaste Sump (G61) system P& ids to correct contact position data of several relays. Additionally, changes to reset information and instrument zero elevations will be made to the Master Setpoint List to improve the accuracy of the G61 instrument loop calibrations.
Summary I. No. These changes do not impact the basic det gn and operation of the G61 system. No new failure mechanisms have been introduced. USAR accident analysis has not been affected. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. These changes do not impact the basic design and operation of the G61 system. No new failure mechanisms have been introduced. USAR accident analysis has not been affected. Therefore, the possibility of an accident or malfunction of a different type than any i previously evaluated in the USAR is not created. '
III. No. These changes do not impact the basic design and operation of the G61 system. No new failure mechanisms have been introduced. USAR accident analysis has not been affected. Therefore, no margin of safety has been changed.
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i SE No.: 96-0082 Source Document: DCN 4968 Description of Chance This drawing change revises P&ID D308-861, Plant Foundation Underdrain (P72) System, to reflect the correct orientation of an instrument tap and its isolation valve.
Summary I. No.
This drawing change corrects the orientation of an instrument tap and its isolation valve. The depiction of this configuration on a '
drawing cannot impact the function or operation of the P72 system. l The flooding analyses contained in the USAR are unaffected. !
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment has not changed.
II. No.
This drawing change corrects the orier. .Jit.) of an instrument tap and its isolation valve. The depictici ' this configuration on a l
drawing cannot impact the function or opiration of the P72 system.
The flooding analyses contained in the USAR are unaffected. i Therefore, the possibility of an accident or malfunction of I equipment of a type different than previously evaluated has not been created.
III. No. This drawing change corrects the orientation of an instrument tap and its isolation valve. The depiction of this configuration on a drawing cannot impact the function or operation of the P72 system. i The flooding analyses contained in the USAR are unaffected.
Therefore, no margin of safety has been reduced. l
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i j SE No.: 96-0083 l Source Document: DCN 5414
] Description of Change j
This drawing change revises P&ID D302-651, Fuel Pool Cooling and Cleanup I
j System, to reflect fuel storage, fuel transfer, and dryer storage pool separation, and physical routing of piping. This is a drawing
] presentation change only and involves no physical changes to the plant.
Summary h I. No. The revision of the P&ID provides an enhanced depiction of the existing plant. The change does not alter the configuration of the 4
plant, nor does it change the previously perceived arrangement,
- size, and functioning of the containment upper pools. The change has no affect on any accident analysis. Therefore, the probability l of occurrence or the consequences associated with an accident or 4
malfunction of equipment has not changed.
! II. No. The revision of the P&ID is being made to clarify the physical 3 arrangement of the upper pool and the dryer storage pool. There are j no changes to the physical plant. Accident analysis is not 1 affected. Therefore, the possibility of an accident or malfunction j of a different type than any previously evaluated in the USAR is not s i
i created.
4 III. No. The revision of the P&ID represents a figure enhancement which more
- clearly depicts the existing plant configuration. There is no j affect on the actual plant, any setpoints, or any existing
, procedures, including Technical Specifications. Therefore, no j margin of safety has been changed.
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1 SE No.: 96-0086 Source Document: USAR Change Request 96-104 Description of Change This change request replaces references to USAR Figure 10.2.9, which is an obsolete figure, with references to USAR Figures 10.1-2 and 10.1-10, which are current with the as-built plant design.
Summary I. No. This change request eliminates an obsolete USAR drawing. The current as-built design drawings reflecting the system in question are maintained within the USAR. This is considered an administrative change. There is no change to the design or I operation of any plant system or component. Accident analysis is I not affected. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change request eliminates an obsolete USAR drawing. The current as-built design drawings reflecting the system in question i are maintained within the USAR. This is considered an j administrative change. There is no change to the design or '
operation of any plant system or component. Accident analysis is not affected. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change request eliminates an obsolete USAR drawing. The current as-built design drawings reflecting the system in question are maintained within the USAR. This is considered an administrative change. There is no change to the design or operation of any plant system or component. Accident analysis is not affected. Therefore, no margin of safety has been changed.
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SE No.: 96-0087 Source Document: DCN 5102 Description of Change i
This drawing change revises various plant drawings to indicate the {
removal of temporary startup strainers from several plant systems. !
l Summary I. No. The drawing change corrects the respective system drawings for the as-built plant configuration. By design, these startup strainers are intended to be removed from their respective systems prior to commercial operation. No credit has been taken for temporary startup strainers in any USAR accident analysis. The operation of I the respective systems is not affected. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. The startup strainers are intended to be removed from their respective systems prior to plant commercial operation. This drawing change does not alter the plant design or operation.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
l III. No. The startup strainers are not addressed in any of the Technical Specifications. System operation is not impacted. Accident analysis is not affected. Therefore, no margin of safety has been changed.
SE No.: 96-0090 Source Document: DCN 5030 Description of Change This drawing change updates plant drawings to revise references to safety-related instrumentation that actuates flood alarms upon detection of high water level inside plant areas. Flood level switches are nonsafety, yet seismicly qualified.
Summary I. No. Moderate energy line breaks which cause flooding have been analyzed to be detected and isolated in 30 minutes. This flooding will not adversely impact any safety-related equipment. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. Moderate energy line breaks which cause flooding have been analyzed to be detected and isolated in 30 minutes. This flooding will not adversely impact any safety-related equipment. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. Flood level indication is not contained in the Technical Specifications. With isolation of the leak within 30 minutes, the flood will have no impact on safety-related equipment. Therefore, no margin of safety has been changed.
SE No.: 96-0092 Source Document: SOI-P22, Rev. 5, TC-13 Description of Change This system operating instruction change incorporates the option of operating the Mixed Bed Demineralizer (P22) system in manual with various shutdown features bypassed.
Summary I. No. Operation of the P22 system in this configuration is acceptable since the conductivity alarms will still be available and administrative controls require system shutdown if a high conductivity alarm is received. The design of the P22 system is not affected. Failure of the P22 system will not adversely impact plant safety. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
, II. No. Operation of the P22 system in this configuration is acceptable since the conductivity alarms will still be available and i administrative controls require system shutdown if a high I conductivity alarm is received. The design of the P22 system is not affected. Failure of the P22 system will not adversely impact plant safety. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. .
III. No. Operation of the P22 system in this configuration is acceptable since the conductivity alarms will still be available and administrative controls require system shutdown if a high conductivity alarm is received. The design of the P22 system is not affected. Failure of the P22 system will not adversely impact plant safety. Therefore, no margin of safety has been changed.
SE No.: 96-0104 S,ource Document: USAR Change Request 96-126 Description of Change This change request revises the title of the offsite review organization from the Nuclear Safety Review Committee to the Company Nuclear Review Board (CNRB).
Summary I. No. This change is administrative only. The functions, responsibilities, and authorities of the offsite review organization have not changed. The design or operation of the plant are unchanged. Accident analysis is unaffected. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This change is administrative only. The functions, responsibilities, and authorities of the offsite review organization have not changed. The design or operation of the plant are unchanged. Accident analysis is unaffected. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This change is administrative only. The functions, responsibilities, and authorities of the offsite review organization have not changed. The design or operation of the plant are unchanged. Accident analysis is unaffected. Therefore, no margin of safety has been changed.
SE No.: 96-0105 Source Document: USAR Change Request 96-127 Description of Change This change reflects the change in name and reporting responsibilities of the supply organization. The Perry Supply Section (PSS) will now adrainistratively report to the Director, Supply Chain Organization, who does not report to the Vice President, Nuclear. To maintain a reporting line to the Perry Plant organization, PSS will also report to the Director, Perry Nuclear Services Department regarding the performance of activities related to the procurement, receipt, storage and issue of items used at the Perry Nuclear Power Plant. The Perry Supply Section will work to Perry Nuclear Power Plant procedures and instructions and be subject to the Perry Quality Assurance Plan.
Summary I. No. This is an administrative change. There is no change to any plant system, structure or component. The functions and activities of the PSS have not changed. Therefore, the probability of occurrence or the consequences associated with an accident or malfunction of equipment has not changed.
II. No. This is an administrative change. The design or operation of plant system,s structures and components will not be affected. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. This is an administrative change that renames the Perry Supply Section. The purchasing, receipt, storage, issue, desior or operation of items used at the Perry Nuclear Power Plar ; te not affected. Therefore, no margin of safety has been changea.
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SE No.: 96-0108 Source Document: USAR Change Request 96-130 !
i Description of Change This change request revises the Main Turbine valve testing frequency.
The actual frequency will not be stated in the USAR. The frequency will be stated in the Operational Requirements Manual (PDB-R0001). ,
Summary I I. No. Increasing the Main Turbine valve testing frequency will not alter the probability of generating a turbine missile. There are no hardware or physical changes to the plant as a result of this i change. No additional failure mechanisms for this equipment are )
being introduced by this change. The turbine overspeed system will not ba affected. Therefore, the activity does not create the i possii,ility of an accident or malfunction of equipment to safety of a differ 2nt type than any previously evaluated in the USAR. 3 j
II. No. There are no hardware or physical changes to the plant as a result of this change. No additional failure mechanisms for this equipment I are being introduced by this change. Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created.
III. No. The monthly and quarterly valve test frequency results in a turbine ;
missile probability of 9.2E-6 per year. This is less than the probability specified by the USAR. Therefore, the action will not reduce any margin of safety.
l SE No.: 96-0115 Source Document: DCN 4199 Description of Change !
This drawing change revises various Piping and Instrument Diagrams (P&ID) to correct the construction coordinates.
Summary I. No. This change is strictly editorial. It does not alter any equipment !
or operating practices. Therefore, neither the probability of l' occurrence nor the consequences of a previously analyzed accident or malfunction of equipment will be increased by this change.
II. No. The editorial change does not alter or adversely affect any ,
equipment in the plant. Therefore, this change will not create the l possibility for an accident or malfunction of a different type than ;
any previously evaluated. j III. No. This change is editorial and has no affect on Technical :
Specifications. Therefore, this change will not reduce the margin !
of safety as defined in the bases for any Technical Specification, j l
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SE No.: 96-0124 Source Document: USAR Change Request 96-142 Description of Change This change requesc is an editorial change request. The changes to the USAR include but are not limited to revisions to page numbering, correction of grammar, updating lists of figures. None of the changes alter the design or operation of any plant system or component.
Summary I. No. This USAR change request is an editorial change. None of the :
changes described within the change request alter the design, function, or operation of any plant systems or components. USAR analyses are not impacted. USAR accident analysis remains unchanged. Therefore, the probability of occurrence or the ,
consequences associated with an accident or malfunction of equipment has not changed.
II. No. This USAR change rcquest is an editorial change. None of the .
changes described within the change request alter the design, !
function, or operation of any plant systems or components. USAR '
analyses are not impacted. USAR accident analysis remains -
i unchanged. Therefore, the possibility of an accident or malfunction I of a different type than any previously evaluated in the USAR is not .
created. I III. No. This USAR change request is an editorial change. None of the changes described within the change raquest alter the design, function, or operation of any plant ' systems or components. USAR analyses are not impacted. USAR act;ident analysis remains unchanged. Therefore, no margin of safety has been changed.
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