ML20116L684
| ML20116L684 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 02/13/1996 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20116L682 | List: |
| References | |
| NUDOCS 9608190156 | |
| Download: ML20116L684 (189) | |
Text
{{#Wiki_filter:_ I l RIVER BEND STATION ! DOCKET NO. 50-458 l LICENSE NO. NFP-47 l t 10CFR50.59
SUMMARY
REPORT , t i i t This report contains a brief description of design changes and procedure changes !' made at River Bend Station during the reporting period through February 13,1996, , The summaries depict changes made to the facility as described in the River Bend Updated Final Safety Analysis Report (USAR) for which an evaluation was : determined to be required. It also contains evaluations for tests conducted which are ! not described in the USAR. The safety evaluations included in this report were l performed in accordance with 10CFR50.59 and determined that none of the changes j resulted in an unreviewed safety question. ! l l q d i I l 1
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I j 9608190156 960812 ; PDR ADOCK 05000458 R PDR Revision 8 August 1996 m -
TABLE OF CONTENTS SECTION TITLE PAGE I UPDATED SAFETY ANALYSIS 1 REPORT CIIANGES ; 11 LICENSE AMENDMENT REQUESTS 122 III PROCEDURE CilANGES 134 IV MODIFICATIONS TO TIIE PLANT 150 V CONDITION REPORTS 159 i VI TEMPORARY ALTERATIONS 171 i VII MISCELLANEOUS EVALUATIONS 185 ' b r l i l l \
SECTION I UPDATED SAFETY ANALYSIS REPORT CHANGES Section 1 Page 1 of 187
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i I I Channe Nussber/USAR Section: Licensing Change Notice (LCN) 1.2-009 ; Pages: 1.2-39 Figures: 1.2-1 2.3-42 9.2-6 ' 9.2 20 1.2-2 2.3-43 9.2-10 9.2-22 2.1 2 2.3-44 9.2-25 j 9.2-23 2.1 2.4-4 10.4-3b ' 9.3-21 2.3-36 2.5-24 13.3 27 9.3-22 2.3-37 2.5-72 2.4-6 i 9.3-23 2.3-38 2.5-82 l 9.3-28 2.3-39 2.5-85 i 9.3-28A 2.3-40 2.5-91 - 10.4-19 2.3-41 3.5-16.4-1 l Description and Basis for Channe: { This modification demolishes the sewage treatment plant (STP) and builds a wastewater treatment ; plant (WWTP) southwest of the clarifiers within the boundaries of River Bend Station property l lines. The existing STP was appraised to be in poor condition due to its using outmoded j technology and consuming excessive amounts of power. He new WWTP will optimize the , sanitary waste treatment process. ; Summary of Safety Evaluation: l i ne new WWTP, as well as the existing STP, are not safety related facilities. They are neither l used for safe operation or safe shutdown of the plant, nor do they house systems for safe operation l or safe shutdown of the plant. There is not a safety related system, structure or component (SSC) l sufficiently close to the existing STP and new WWTP to be affected by any seismic response of the { STP or the WWTP. The physical separation also precludes any soil / structure interactions. Load l changes due to shielding of gradient winds would be insignificant compared to design basis tornado : wind speeds. The design loads on safety related SSC associated with tornado wind pressure and depressurization effects will remain unchanged. Should part of the WWTP or STP become ! detached and airborne in a tornado, the resulting missiles would be enveloped by those postulated , for construction debris. The STP demolition and the new WWTP installation will not adversely ! impact the runoff characteristics, flooding characteristics, or the flooding and ponding thresholds 1 of the site in safeguarding equipment and plant stmetures impoitant to safety. Therefore, this ! modification does not increase the probability of an accident or a malfunctian of any safety related , SSC previously evaluated in the SAR. The incidental fire hazards associated with the STP and l WWTP will not pose a serious fire hazard to the surrounding plant structures due to physical ; separation. De new WWTP remains designed as a " clean" sanitary waste treatment system and is ? not intended to interact with contaminated or radiological effluents. He existing and new features , of the sanitary waste system are not designed nor anticipated to be a radioactive effluent pathway, i Train Number 1, the part of the sanitary waste system sening the protected area, is not directly l connected to any radioactive systems in the designated radiologically controlled areas. He ( possibility of the system becoming contaminated from senice water leakage or other contaminated ! liquid system leakage is extremely remote The demolition of the STP and the installation of the l WWTP does not change the radiological consequences of an accident at the site boundary. j herefore, this modification does not increase the consequences of an accident or a malfunction of l any safety related SSC previously evaluated in the SAR. The STP and WWTP do not inwalidate l
' Section 1 - Page 2 ofI87 l l
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t I existing seismic and tornado missile design considerations or criteria, probable maximum l' l precipitation flooding analysis conclusions, or pose a serious fire hazard to the surroundmg plant ; structures. The existing STP and new "VWTP are located outside of the protected area and will ; not adversely affect any safety related SSC. Herefbre, this modification does not create the i possibility of an accident or a malfunction of any safety related SSC different from any previously l evaluated in the SAR. He STP and WWTP are not governed by any technical specification. l l Therefore, this change does not reduce the margin of safety as defined in the basis of any technical f
- specification. For these reasons, this modification does not constitute an unreviewed safety
i question. j i ; t Pages: 42 ! Channe Number /Initiatine Document: LCN 1.8-027 (QAD-14) 42a i Table: 1.8-1 i Description of Channe: i This safety evaluation was performed when a discrepancy was revealed that operability evaluations
- did not meet the requirements of Quality Assurance Directive 14 (QAD-14). QAD-14 was revised to bring it in line with ANSI 18.7-1976. This rnision also provided guidance on suitable j
. documentary evidence per NRC Generic Letter 91-18. !
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- Summary of Safety Evaluation
l ! I This revision to the procedure does not alter the design, function or method of performing the ~ i functions of structures, systems or components at River Bend Station (RBS). This revision { strengthens and in no way deletes or reduces any part of the existing Quality Assurance Program. This change is administrative only, and as such, does not constitute an unreviewed safety question. , i The change will align the RBS QA Program with NRC accepted practice as described in NRC l i Generic Letter 91-18 which allows for continued operation while an operability determination is ! j- being made, provided the licensee has reasonable expectation that the structure, system or j component (SSC) is operable and that the determination process will support that expectation. The i probability or consequences of an accident previously evaluated in the SAR will not be increased i because no required QA Program activities have been deleted as a result of this change. His ! change does not degrade any SSC reliability or its original design specifications, therefore, this l change does not create the possibility of an accident which is different than any previously l evaluated in the SAR. Since no QA Program activities have been deleted, the change will not i increase the probability of radiological consequences due to the malfunction of equipment : 2 important to safety as previously evaluated in the SAR. For the same reasons, the change will not . j increase the consequences of a malfunction of safety related SSCs presiously evaluated in the SAR 3 and will not increase the possibility of a malfunction of safety-related SSCs presiously evaluated in j the SAR. This change does not degrade any SSC's reliability or it's original design specifications; l' therefore, the change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewcd safety question. i
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l Channe Number /USAR Section: LCN 1.8-030 Table 1.8-1 Description and Basis for Channe: l l During construction of River Bend Station, a requirement was added to the SAR directing the Quality Assurance (QA) program to transmit their records to the Permanent Plant Files (PPF) within 30 days of completion of associated activities. This requirement was intended to ensure protection of QA records while in interim storage before transmittal to PPF It compensated for inadequate fire protection in areas designated as interim records storage during construction. This condition no longer exists. The SAR requires interim storage areas to have one-hour fire rated ! cabinets and to have either an autcoatic sprinkler or a combination of two or more of the following: 1) automatic fire alarms, 2) hose stations, 3) portable extinguishers. All interim storage areas now meet or exceed these fire protection requirements as specified in the SAR. This change allows QA records to be stored in interim storage facilities for periods longer than 30 days. Summary of Safety Evaluation: This change does not address modifications to equipment or systems. It does not impact any activity associated with design, material, and construction standards associated with the modification of equipment or systems. The activities associated with this change do not include the use of calibrated instrumentation or impact system operation or design. Since this change does not affect equipment or modify systems, this change will not increase the probability of an accident previously evaluated in the SAR. His change is associated with records and administrative functions. It does not address, impact or create any activities that could potentially result in an i accident evaluated in the SAR and as a result cannot alter radiological consequences or plan a 2 direct or indirect role in mitigation of the radiological consequences of an accident. Consequently, this change does not increase the consequences of an accident previously evaluated in the SAR. This change revises an administrative procedure and does not add new equipment or change the physical configuration of existing equipment. Therefore, this change does not increase the possibility of an accident that is different from any previously evaluated in the SAR. His change does not impact or address any equipment determined to be important to safety. Accordingly, this change neither increases the probability of a malfunction of equipment important to safety nor
- increases the consequence of a malfunction of equipment important to safety, nor creates the possibility of a malfunction of equipment important to safety. The requirements prosided for quality assurance in the technical specifications have been relocated to other documents. Since this subject has been removed from the Technical Specifications, it cannot reduce the margin of safety as defined by the basis to any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
Channe Number /USAR Section: LCN 1.8-031 Table 1.8-1 Description and Basis for Change: The reason for this change is to allow storage of non-permanent records in facilities which do not meet the requirements of a single storage facility as committed to in the SAR on ANSI N45.2.9-1974 and 1979. Presently the SAR describes and allows records to be stored in interim storage until transmittal to the Permanent Plant File (PPF). The change will allow non-permanent records, those records which do not require lifetime retention, to be stored in what is now described as Section i Page 4 of 187
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- interim storage until final disposition. All interim records storage areas meet or exceed fire l protection requirements specified in SAR Sections 1.8, Table 1.8-1. l Sa==ary of Safety Evaluation:
- The changes to Section 1.8 of the SAR do not address modifications to any equipment or systems and do not impact any activity associated with design, material and constmetion standards associated with che modification of equipment or systems ne activities associated with this ;
change do not include the use of calibrated instrumentation or impact system operation or design. l The changes are administrative in nature and do not impact any equipment which has been ! determined to be important to safety, and as a result, cannot increase the probability of a ; malfunction of any equipment or systems important to safety. Therefore, these program changes i do not increase the probability of an accident or a malfunction of equipment important to safety - previously evaluated in the SAR. These changes to the SAR do not address, impact or create any l activities which could potentially result in an accident evaluated in the SAR and as a result cannot j alter radiological consequences or play a direct or indirect role in mitigating the radiological i consequences of an accident. The changes are associated with the storage of non-permanent j records and, as such, are administrative in nature. The changes cannot alter any assumptions ! previously made in the evaluation of radiological consequences of an accident or affect any fission : product barriers described in the SAR. These changes will have no affect or impact on any equipment important to safety. Since these changes do not impact systems, structures or l components / equipment, the changes cannot increase the consequence of a malfunction ofITS ; equipment. Herefore, these changes do not increase the consequences of an accident or a : malfunction of equipment important to safety previously evaluated in the SAR. The changes do [ not create a condition that could initiate an accident, so, the changes do not impact or create any ! activities which could cause an accident different than any accident previously evaluated in the l SAR. The changes do not affect any equipment, any activities associated with plant equipment or , affect the way plant equipment is operated. Here will be no new equipment, no equipment . configuration changes, nor are there system operation changes being introduced that could create the possibility of a malfunction of a different type previously evaluated in the SAR. Therefore, j these changes do not create the possibility of an accident or a malfunction of equipment important ; to safety different from any previously evaluated in the SAR. These changes do not affect or alter l Technical Specification requirements, so, the changes do not reduce the margin of safety as defined l in the basis of any technical specification. For these reasons, this modification does not constitute ! an unreviewed safety question. i Channe Number /USAR Section: LCN 2.3-10 Page 2.3-23 Description and Basis for Channe: This modification replaced Esterline-Angus Multi-Point recorders in the meteorological monitoring i system with Westronics 3200 series recorders. He primary system recorders were replaced with a single recorder and the secondaiy recorders were replaced with a single recorder. He new primary , recorder will trend all primary channel outputs and the new secondary recorder will trend all the secondary outputs. i Section I Page 5 of187 l 1 i N -:' ? m _ w- w _ _ _ _ . _ __ _ _ . _ _ _ - ____ _ _ _ _ _ _ . _ _<_ -- _____- ___ .__ _ ___
Summary of Safety Evaluation: The instruments afTected by this modification were not postulated to cause any accident described [ in the SAR. Therefore, this modification did not increase the probability of an accident presiously evaluated in the S AR. He onsite meteorological data provides an acceptable basis for making conservative estimates of atmosphere dispersion conditions used for estimating accident , consequences and routine gas releases to offsite locations. The recorders are not used for mitigating the consequences of an accident. No pressure boundaries were affected by this modification. Therefore, this modification did not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The replacement recorders fit in the existing non-seismic, non-safety-related panels using vendor supplied mounting adapters. ne new configuration does not adversely affect the structural integrity of the panels. Electrical independence and physical separation from equipment important to safety were maintained. Since the number of recorders was reduced from six to two, the overall loading on the instrument bus was not adversely impacted. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety. There were no new failure mechanisms related to the installation of the replacement recorders. Therefore, this modification did not create the possibility of an accident or a malfunction of equipment important to safety difTerent from any previously evaluated in the SAR. This modification did not afTect any equipment which is addressed in any technical specification or the basis of any technical specification. Therefore, this modification did not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification did not constitute an : unreviewed safety question. j i Channe Number /USAR Section: LCN 3.2-010 Pages: 5.2-42 5.2-43 i 5.2-49 7.3-14 ) 7.3-15 7.3-17 7.6-5 l 7.6-6 7.6-7 l i 7.6-21 Tables: 3.2-1 Sheet 18 5.2-8 Figures: 7.6-1 Sheet 1 7.6-1 Sheet 2 7.6-9 Sheet 1 7.6-9 Sheet 2 9.4-7a 9.4-7e Description and Basis for Change: This modification will delete existing Riley differential temperature switches (DTS) in the leak detection system providing system isolation signals for mains line, RCIC, RWCU and RHR Systems. For each DTS, there are corresponding redundant, safety related ambient temperature 4 Section i Page 6 of 187
switches which also provide the system isolation signal. His modification also installs four new ambient temperature switches (Class IE). The four new switches will not perform any isolation function or meter indication and are being installed so as to monitor their performance and prove their reliability for possible future usage in the plant. Deletion of the DTSs will not affect the redundancy of the ambient temperature loops. Summary of Safety Evaluation: The differential temperature instruments are redundant to Technical Specification requirements to have area temperature monitors, level instruments and high flow instruments, and do not contribute to the primary success paths to prevent or mitigate design basis accidents and transients. He changes will not reduce the capabihty of the leak detection system to perform its intended design function. Deletion of the differential temperature switches will not increase the probability of occurrence of a line break accident because the corresponding redundant ambient temperature l
. switches will continue to isolate the affected system in case of a leakage. For the ambient and the differential temperature switches, credit is not taken in any transient or accident analysis in the SAR since bounding analyses are performed for large breaks such as main steam lio : breaks.
- Derefore, this modification does not increase the probability of an accident previousiy aa!uated in the SAR.
His modi 6 cation will not affect the reliability of any safety related valves or systems (including the leak detection system) required for safe shutdown. Deletion of these switches: 1) will not cause the malfunction of any equipment, leak detection system or isolation actuation function because there are redundant ambient temperature switches to detect the leakage, and 2) does not create any new failure modes for the existing equipment important to safety nor does it change any existing failure modes. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This modification does not create new interfaces with other equipment important to safety. This modification will not affect the reliability of any safety related valves or system (including leak detection system) required for safe shutdown of the plant. The changes will not affect the redundancy of the ambient temperature loops and the ability to isolate the affected lines. This modification does not increase the radiation sources or affect the release paths or accident mitigation equipment. The system will still function as designed to isolate the applicable lines. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Deletion of differential temperature switches in the leak detection system will not reduce the capability of the system to perform its intended function. There is no change to the operation of
- any other safety related system. The equipment deleted and added by this modification does not create any new interfaces with any equipment important to safety, nor does it change any cxisting interfaces. Performance of the intended function for RCIC, RWCU. RHR and main steam systems is unaffected by this modification because the function performed by the operation of these systems is not changed. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the S AR.
Equipment affected by this modification is not included in the basis for any technical specifications. This modification will not have any effect on the margin of safety of any structures, systems, and components which are addressed in the basis for the Technical Specifications. This modification does not change any existing setpoints or allowable values. The differential Section I Page 7 of187
temperature instruments are redundant to technical specifications required area temperature
- monitors, level instruments and high flow instruments, and do not contribute to the primary success paths to prevent or mitigate design basis accidents and transients. Therefore, this modification does not reduce the margin of safety as defmed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
Channe Number /USAR Section: LCN 9.5-094 and LCN 3.6-002 Pages: 9.5-24a 9.5-26 Figures: 9.5-2c 9.3-8a Description and Basis for Channe: This modification adds a level indicator to the Division I and 11 standby diesel generator jacket water standpipes at a location where it can be read by an operator standmg on the floor adjacent to the standpipe. This new indicator will provide water level measurement over the full range of the standpipe. De existing indicator is deleted by this modification. Drain connections are added to the standpipes to provide a means of obtaining periodic water samples for analysis. The new level ' indicators 1EGT-LI24A and 1EGT-LI24B will be tapped offof the existing process valves. The process valves will become normally open and standpipe water sampling capability will be provided from the instrument tubing drain valves at the instrument stand. Summary of Safety Evaluation: De standby diesel generators and thejacket water standpipe are not postulated to cause any accident. The sole purpose of the new level indicators is to provide operations and maintenance personnel with a means for easily determining water levels in the diesel generatorjacket water standpipe. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. The diesel generators' ability to minimize the consequences of an accident is not changed or challmged. Therefore, this modification does not increase the consequences of any accident previously evaluated in the SAR. This modification meets or exceeds the original system design requirements. There are no new failure modes created. No new system interfaces or initiating events are introduced. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The new indicators are seismically qualified and mounted for mechanical pressure boundary integrity. This modification does not cause interaction with other safety related component or equipment in the area. The structural-integrity of the jacket water system is maintained. Therefore, this modification does not increase the probability of a malfunction of any safety related system, stmeture, or component (SSC) previously evaluated in the SAR. This modification will not affect the ability of any safety related system in performing its function to mitigate the consequences of postulated accidents and transient events. This modification does not impact the perfonnance of any barriers used to mitigate the consequences of a malfunction of any safety related SSC. Derefore, this modification does not increase the consequences of a malfunction of any safety related SSC presiously evaluated in the SAR. This modification does not create any new failure modes. Derefore, this modification does
. not create the possibility of a malfunction of any safety related SSC different from any presiously . evaluated in the SAR. There are no technical specifications that address thejacket water standpipes. Therefore, this modification does not reduce the margin of safety as defined in the ; basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question.
Section I Page 8 of187
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Channe Number /USAR Section: - LCN 3.11-002 Pages: 3-xliii(a) , 3.9A-27 -i 3.11-10 ! 3.11 l 3.11-13 ; 3.11-14 7.1-14 l 7.2-14 l Tables: 3.9B-3a l 3.11-9 ! 6.2-52 j Description and Basis for Channe: [ This change eliminates the mechanical equipment qualification (MEQ) program. The MEQ [
, ' program duplicates existing programs. In most cases, other programs provide a better means of l ensuring continued equipment operability either through more frequent testing of the equipment or j through periodic testing which demonstrates the operability cf the equipment and identifies any -I degradation of nonmetallic materials which are the subject of the MEQ program.
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- Summary of Safety Evaluation:
l I Elimination of the MEQ program will not alter the equipment or its function to the plant's physical l configuration. Other programs currently in place at RBS verify continued operability of the MEQ j components, or ensure that aging mechanisms will be detected and corrective maintenance < performed. Existing design, procurement, maintenance, and testing programs provide reasonable l assurance that the criteria applied to mechanical equipment through the MEQ program are l evaluated through other plant programs and the MEQ program performs a redundant activity. In
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many instances, it was found that the alternate programs provided a greater assurance of continued j operability of the component than that provided through the MEQ program. Therefore, this change j does not increase the probability or con equences of an accident previously evaluated in the SAR.
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Elimination of the MEQ programs will not result in any changes to plant structures, systems, or components. Other programs provide tee same assurances as the MEQ of the suitability of the ! I nnnnwallic portion of ccitain active safety related equipment located in a harsh envirorment for its intended function. Therefore, this change de:s not create the possibility of an accident different from any previously evaluated in the SAR. It was determined that most of the equipment in the MEQ has at least one, if not several, surveillance or periodic testing activities that addressed the maintenance identified in the MEQ program. For those components not already covered by
, alternate surveillance or testing programs, special preventative maintenance items were created.
Items originally covered by the MEQ but that had no active safety post accident function were deleted from the MEQ program. New or replacement evw,gsgr.ts that would have been evaluated by the MEQ program are now addressed by procurement and storage procedures. 'Iherefore, this change does not increase the probability of a malfunction of equipment important to safety. h equipment will not be changed in any manner by the deletion of the MEQ program. hrefore, this change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. This change does not create the possibility of a malfunction of
' equipment important to safety different from any previously evaluated in the SAR. All evaluations ' performed to satisfy General Design Criteria 4 of 10CFR50 Appendix A will be retamed A Section I Page 9 of187 4
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i i i preventive maintenance program, surveillance test program and inservice test program are in place I which provide assurance of continued equipment operability. Procurement controls are established which ensure design requirements for new and replacement equipment are satisfied. Derefore, this change does not reduce the margin of safety as dermed in Technical Specifications or the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l t i Channe Number / USA R Section: LCN 3.11-003 Sections: 3.11.1.2.4 l 3.11.2.1 j 3.11.2.2.1.1 l 3.11.3.1 j 3.11.3.3 , Tables: 3.11-10 ! 3.11-11 ! 3.11-12 j 3.11-13 : 3.11-14 Figure 3.11-6 Description and Basis for Channe: i This change evaluates two proposed changes to the SAR. The first change consists of describing the revisions to the equipment qualification master equipment list (EQMEL) that will occur as a !
. result ofits incorporation into the site wide component data base. Several fields that were part of the EQMEL only are now contained in other areas of the CDB and the tables included in the SAR q describing codes for these fields are no longer required. The second revision to the SAR consists !
i of a change in the method of documenting the qualification basis for RBS equipment. He RBS EQ program is migrating to a document called an equipment qualification assessment report ] (EQAR). The EQAR provides a text based discussion of the basis for considering each EQ piece , of equipment at RBS to be qualified. His migration will elimmate the need for a system
- component evaluation worksheet (SCEW) which is described in Section 3.11.3.2 of the SAR. A :
description of the EQAR is being added to the SAR as a result of the documentation format j change. There are no technical changes associated with the SAR revision. The only change is in ! j- the description of the documentation format. i Summary of Safety Evaluation: 1
- The proposed change to the SAR consists of a change to document formats. Here is no technical i l or physical change to the plant as a result and no changes are made in commitments or quality
- levels of existing components. Equipment will continue to function as designed. Therefore, this )
change does not increase the probability or consequences of an accident or a malfunction of _ equipment important to safety previously evaluated in the SAR. He change alters the internal
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- ' documentation format for EQ equipment and does not alter the design requirements for it. It does not change in any manner the equipment itself. Therefore, this change does not create the _
possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR or reduce the margin of safety as defined in the basis of any technical specincation. For these reasons, this modification does not constitute an unresiewed safety question. .- Secten 1 - Page 10 of187
I i Channe Number /USAR Section: LCN 3.8-009 Pages: 3.8-14 l' 3.8-30 3.8-36 3.8-106 3A-iii ! 3A.29-1 ' 3A.30-1 6.2-116 Description and Basis for Change: l This modification replaces the bellows components of containment piping penetration IKJB*Zl9 and 1KJB'Z20. He replaecment components are similar in form, fit and function to the original equipment but differ by the manner in which they are installed. He original expansion joints were , installed as whole, cylindrical components together with the piping. He replacement bellows are installed in are segments, or clamshells, around the existing piping. The replacement bellows are ' fabricated and cut into segments in the shop and then welded back together around the existing piping in the field. Design of the replacement bellows utilized the computer codes ME101 and i NISA 11 to reconcile the effects of the new components with the existing pipe stress analyses and to help design the testing mechanism for the installed components. l l Summary of Safety Evaluation: The computer code ME101 is endorsed for use at all the other Entergy nuclear plants for the application used in this modification. It was verified for use at RBS by comparison to results of ; other computer codes currently listed in the RBS SAR. NISA II was verified and is maintained by Entergy. No postulated accident will occur more frequently due to a change in the primary containment boundary. None of the mechanisms whose postulated failure forms the bases for the occurrence of accidents evaluated in the SAR is affected by a change of the primary containment bellows components. The change in bellows material is an upgrade to a material more resistant to - the mechanism presumed responsible for cracking in the equipment being replaced. The .. replacement bellows are functionally equivalent to the original equipment in performing as the primary containment boundary. Herefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. Although the primary containment boundary is involved in the control ofleakage during and after all postulated accidents in the SAR, there is no functional change to the barrier. The replacement bellows are similar in size and shape to the original equipment and are designed to act in the same way as the original components. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. , here are no new failure modes introduced as a result of this modification due to the similarity between original and replacement components. The functioning components are substantially ; 3 unchanged. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. There is no change in the performance of the primary containment boundary or in availability to test the boundary for integrity as defmed in Technical Specifications as a result of this i modification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l Section I Page 11 ofI87 ; I
h l ? l.- i c . Channe Number /U$AR Section: LCN 3.9A-011 Triole 3.9A-11 Sheet 1 , i ~ Description and Basis for Ch_ga.ge: n i This modificatim .eplaced the 10 ft-lb,1700 rpm motor and associated overload heaters on valve j
- IB21*MOVF/19 with a 25 ft-lb,1700 rpm motor and the required new size overload heaters and 'l
! trip coils. Le existing valve stem was replaced with a valve stem made from a higher strength - i
- material. Before this modification, this motor operated valve (MOV) had met the established valve l i operability criteria for thrust but not the desired margin criteria. His modification was made to l allow setting the MOV torque switch higher to meet the margin criteria to assure long term proper l function under design basis conditions. 3
!, i Summary of Safety Evaluation: ! !- . This valve was not associated with the initiation of any accident evaluated in the SAR. This i ' modification did not adversely affect an equipment considered important to safety. Therefore, this j! modification did not increase the probability of an accident or a malfunction of equipment : i j important to safety previously evaluated in the SAR. His valve performs functions associated i with the operation of the main steam drains, contamment isolation and main steam positive leakage l control systems. All of the required functions of this valve remained unchanged by this l ! modification. This modification maintained the integrity of the pressure boundary. The balance of : the valve components and the actuator rating was evaluated and found acceptable to withstand the l l higher thrust. All environmental qualifications requirements were maintained. The electrical ! supply was evaluated with respect to the increased loads and found to be acceptable. The i replacement thermal overload heaters were sized according to original design requirements. The l 9 stroke time, control logic and leakage requi ements of the valve remamed unchanged. System design requirements, redundancy and separation requirements were maintained. Herefore, this :
- modification did not increase the consequences of an accident or a malfunction of equipment !
C . important to safety previously evaluated in the SAR. No new failure modes were created by this l , modification. Herefore, this modification did not create the possibility of an accident or a j 4 malfunction of equipment important to safety different from any previously evaluated in the SAR. a j nis modification did not affect any technical specification or the basis of any technical ; specification. Therefore, this modification did not reduce the margin of safety as dermed in the
- basis of any technical specification. For these reasons, this modification did not constitute an unreviewed safety question.
i ! Channe Number /USAR Section: LCN 3.9A-012 Table 3.9A-11 Sheet 1 Description and Basis for Channe: { His modification replaced the 10 ft-lb,1700 rpm motor and associated overload heaters on valve l 1B21*MOVF085 with a 25 ft-lb,17C'0 rpm motor and the required new size overload heaters and i ~
, - trip coils. De existing valve stem wac replaced with a valve stem made from a higher strength 1
- material. - Before this modification, this motor operated valve (MOV) had met the established valve operability' criteria for thrust but not the desired margin criteria. This modification was made to l l allow setting the MOV torque switch higher to meet the margin criteria to assure long term proper
- functon under design basis conditions. j
- i Secten I . Page 12 of187 l 4
I, . l l I !
! Summary of Safety Evaluation: j
; His valve was not associated with the initiation of any accident evaluated in the SAR. This !
. modification did not adversely affect an equipment considered important to safety. Herefore, this j modification did not increase the probability of an accident or a malfunction of equipment l important to safety previously evaluated in the SAR. His valve performs functions associated ,
4
- with the operation of the main steam drains, containment isolation and main steam positive leakage ;
- control systems All of the required functions of this valve remamed unchanged by this i modification. This modification maintamed the integrity of the prwsure boundary The balance of the valve components and the actuator rating was evaluated and found acceptable to withstand the i j higher thrust. All environmental qualifications requirements were maintamed The electrical supply was evaluated with respect to the increased loads and found to be acceptable. The replacement thermal overload heaters were sized according to original design requirements. The i j stroke time, control logic and leakage requirements of the valve remamed unchanged. System i 3
design requirements, redundancy and separation requirements were maintamed Therefore, this e j modification did not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. No new failure modes were created by this : I modification. Therefore, this modification did not create the possibility of an accident or a ! malfunction of equipment important to safety different from any previously evaluated in the SAR. 3 This modification did not affect any technical specification or the basis of any technical j 3 specification Herefore, this modification did not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification did not constitute an } } unreviewed safety question. l i
- i 4
i Channe Number /USAR Section: LCN 3.9A-013 Table 3.9A-11 , r Description and Basis for Channe: I { This modification replaces the 10 ft-lb,1700 rpm motor and associated overload heaters on valve ! , IB21*MOVFOl6 with a 25 ft-lb,1700 rpm motor and the required new size overload heaters and l trip coils. %c existing valve stem is replaced with a stem made from a higher strength material. : Before this modification, this motor operated valve (MOV) had met the established valve operability criteria for thrust but not the desired margin enuria. His modification is made to I allow setting the MOV torque switch setting higher to meet tce margin criteria to assure long term l proper function under design basis conditions. ! Summary of Safety Evaluation: i i i . This modification does not affect the operation or function of the valve or systems in which the valve performs a function. He Limitorque actuator has been evaluated for use with the 3 rephe-t motor and has been determined to be acceptable. He additional weight of the motor has been considered for seismic loading of the valve / actuator assembly and piping. All
. replacement valve parts meet or exceed the original design and applicable code requirements. He
- new motor and stem will allow the actuator torque switch to be set to seat the valve with a higher thrust. %e replacement thermal overload heaters are sized according to original design requirements. This valve is provided to assist in mitigating the consequences of a line break ,
accident and is not associated 'with any accident initiators. Therefore, the change will not increase the probability of an accident previously evaluated in the SAR. He modification does not increase 1 the amount of radioactive material or adversely affect the pressure boundary of any system that Section I . Page 13 of I87
contains, stores, processes or transports radioactive material. The valve's stoke time, control logic, and leakage requirements remain unchanged. Therefore, the consequences of a line break accident, for which these valves provide a mitigating function, will remain unchanged. This modification does not add nor delete any equipment or constitute a new test of equipment. Therefore, this modification does not create the probability of an accident which is different than any previously evaluated in the SAR. The change provides the valve with tne capability of closing with increased thrust, prosiding additional assurance that the valve will successfully perform its safety function ; under all design basis conditions. Here is no increase in the probability of a malfunction of a i safety related SSC previously evaluated in the SAR. No new failure modes are created since l . standard components which meet the original design and Code requirements are used. System , design requirements, redundancy, and separation requirements are maintained. Therefore, the :
- consequences of any malfunction of this valve will not be increased. This modification does not create the probability of a malfunction of a safety related SSC different than any presiously !
evaluated in the SAR. Neither the Technical Specifications nor their bases are affected by this j modification. Therefore, the modification does not reduce the margin of safety as defined in the i basis to any technical specification. For these reasons, this alteration does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 3.09B-019 Table 3.9B-2g i Descrintion and Basis for Change: , I MR 95-0003 modified the River Bend Station main steam safety relief valves (SRVs) to improve l the seat tightness of Crosby Valve Company's ASME Section I and Section 111 Main Steam and ) Pressurizer Valves. Additional information was required in item #6 of Table 3.9B-g to reflect the new stress analysis for the valves' discs. l Summary of Safety Evaluation:
- Although several analyzed events in the SAR result in SRV actuation, only the inadvertent opening of an SRV accident is initiated by an SRV. The cause of this accident is not specifically identified, but it could result from operator error or a single failure within the SRV relief mode control circuitry. The replacement of the disc insert and necessary minor matching of parts would not ;
affect an inadvertent opening of an SRV. Herefore, this modification does not increase the l probability of an accident previously evaluated in the SAR. The modified SRVs retain the same design and operational characteristics as the original SRVs with the exception ofimproved seat leakage. Therefore, this modification does not increase the consequences of an accident presiously evaluated in the SAR. Two potential accidents were reviewed. Inadvertent opening of multiple SRVs and multiple SRVs failing to open could not occur. These accident could be caused by the ' I pneumatic operators or their control circuits, neither of which are affected by this modification. They could also be caused by the setpoints drifting lower and higher, respectively. Spring preload, which would cause setpoint drifting, is set and verified following implementation of this modification. %erefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. He modified SRVs retain the same design and operational characteristics as the original SRVs with the exception ofimproved seat leakage. Improved seat leakage could have a positive impact in decreasing the probability of malfunction of equipment important to safety which currently must be used to deal with the valve leakage. Herefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the S AR. This modification does not affect the relief mode of SRV operation. Also, each SRV is essentially an independently operating component and. Section i Page 14 of 187 I
1 4 as such, does not affect other equipment important to safety. Therefore, this modification does not I increase the consequences of a malfun tion of equipment important to safety previously evaluated l in the SAR. No new failure modes are -dded and no new or different system interfaces are added i Therefore, this modification does not create a dfunction of equipment important to safety l
- different from any previously evaluated in the SAR. His rnodification had been evaluated to not a decrease the basis for any technical specification and as such no margin of safety has been reduced. j For these reasons, this modification does not constitute an unreviewed safety question. j 4 5
- j. Channe Number /USAR M f ding: LCN 4.3-002 Pages: 4.1-3 l
. 4.1-4 i j 4.1-6 i
, 4.2 2 i 1 4.2-3 I 4.2-4 l 4.2-5 ) i Figures: 4.2-4 ! 4.2-5 4.2-5a i ! Description and Basis for Channe: ! t . I
- During the last refueling outage, 29 General Electric original equipment control rod blades (CRBs) {
were replaced with advanced long life CRBs manufactured by ABB. The new ABB CRBs have > l the same general shape. Certain design changes were necessary to account for the total control rod . i- weight, neutron absorption cross-section, and inclusion of a metal absorber material. i l l Summary of Safety Evaluation:
)
' I The exterior of the new CRBs is compatible with the original CRBs. Seismic qualification of the I !- replacement CRBs conforms to RBS licensing requirements regarding scram times and structural l integrity. De new CRBs have a velocity limiter identical to the original CRBs. Herefore, this i change will not increase the probability of an accident previously evaluated in the SAR. Drop j. characteristics of the new CRB have been shown by testing to be identical to the original CRBs. j This change does not affect scram times, control blade envelope, surface finish, and rod worth. ! Therefore, scram reactivity is not affected by this change. Accident consequences that are !
- dependent on scram reactivity are also not affected by this change. Therefore, this change does not .
increase the consequences of any accident previously evaluated in the SAR. The mechanical failure modes for the new CRBs are the same as the original CRBs. The chance of through wall j cracking in the new CRBs is less that the original CRBs. De mechanical end-of-life and nuclear . end-of-life will continue to be the same. The new CRB dimensional is compatible with RBS CRB handling equipment. He material used in the new CRBs is less likely to be damaged or deformed under the design conditions. Therefore, this change does not create the possibility of an accident different than any presiously evaluated in the SAR. De fuel system is the only affected system.
- The change itself does not increase the probability of a failure of the new CRB. De design changes make the CRB more resistant to cracking. The only possible way a crack can lead to -
failure of a control rod is if the crack becomes large enough to wrap around the wing, branch out, and causes a piece of the blade to fall off andjam. The change will decrease the probability of this occurring. If a few ligaments between high exposure holes crack, structural integrity and performance of the CRB are maintamed. Seismic qualification verifies, in part, that the new CRBs are compatible with the existing plant configuration. Derefore, this change does not increase the Section i Page 15 of187 i j i i _ _ _ _ _ _ - . ._. --- . _ _ _ . _ _ ._. , _ , , _ , - . . , . __ - . . _ i
i 1 J l v .
;~ ;
probability of a malfunction of a safety related structure, system or component (SSC) previously I evaluated in the SAR. An increase of the consequences of a malfunction of any safety related SSC ! i would be related to scram reactivity and CRB drop velocity. The design of the velocity limiter and l
; total weight of the new CRB allows the drop velocity to be the same. Therefore, this change does l J
not increase the consequences of a malfunction of a safety related SSC previously evalua:cd in the i
! SAR. The CRB socket is compatible with the existing control rod drive (CRD) coupling spud. !
l ne maximum weight of the new CRB is within the design weight for the CRD and hydraulics . i system. The new CRB is compatible with the existing fuel channels and core internals. Form, fit, l j- function and behavior of the replacement CRBs are at least the same as the original CRBs. .! herefore, this change does not create the possibility of a malfunction of a safety related SSC j
- different from any previously evaluated in the SAR. His change neither directly nor indirectly i impacts any technical specification. Derefore, this change does not reduce the margin of safety as l defined in the basis of any technical specification. For these reasons, this modification does not ;
i constitute an unreviewed safety question. 4 i Channe Number /USAR Section: LCN 4.4-003 Pages: 4.4-5 15.0-11a i 15.0-12 ]' Figure 4.4-5 r i Description and Basis for Channe: i 1 l The changes consisted of setpoint and other control circuit card changes as well as SAR, COLR, j TRM and procedure changes as necessary to allow reactor operations at increased core flow (ICF)
- j. conditions (core flow up to 107% of rated). ,
l : ! Summary of Safety Evaluation: i Setpoint changes are consistent with the analysis performed and consistent margins-to-trip setpoints are maintained. Derefore, this modification does not increase the probability of an
- accident previously evaluated in the SAR. All setpoint and thermal limit changes maintain the i I required margin-to-safety limits. Herefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. Setpoint and thermal limit changes are made to insure plant parameters stay within the design basis. Derefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. No hardware changes are required to make the changes specified in this modification, all components have sufficient design margin to allow ICF. Therefore, this modification does not increase the
- probability of a malfunction of equipment important to safety previously evaluated in the SAR.
Setpoint and thermal limit changes are made and analysis performed to insure plant parameters
- stay within the design basis and no hardware changes are made. Derefore, this modification does
~
not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. Setpoint and thermal limit changes are made and analysis performed to insure plant parameters stay within the design basis and no hardware changes are made. The setpoint changes do not result in new failure modes. Derefore, this modification does not create a malfunction of equipment important to safety different from any presiously evaluated in the SAR. All setpoint and thermal limit changes maintain the required margin-to-safety limits. Herefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Section i Page 16 ofI87
I Channe Number /USAR Section: LCN 4.6-26b Figure 4.6-Sa
< Description and Basis for Chance: l This modification provided the reactor water cleanup (RWCU) pumps with a continuous source of ~ !
cooling water from the control rod drive (CRD) hydraulic system in order to increase seal j reliability and availability. A tap was made in the non-safety-related portion of the CRD pump i discharge piping. ! Summary of Safety Evaluation: ! The new piping materials and components are designed and specified in accordance with the same ! piping class as the existing CRD hydraulic system piping and components Piping stresses and l support loads have been evaluated and found acceptable. The pipe is seismically supported to Category II/I requircinents. The additional pressure exerted on the RWCU pump from the higher ! design and operating pressure of the CRD hydraulic system is relieved by installed relief valves. 3 The water draw from the CRD hydraulic system will not affect it's hg-i operation. The , continuous source of seal cooling water to the RWCU pumps will help cool the seals and extend - the life of the seals during times of pump idle. Therefore, this proposed change does not affect the , RDS system performance in a manner that leads to an increased probability of an accident or l malfunction of safety-related system, structures, or components previously evaluated in the SAR. The new lines are in a non-safety-related portion of the RDS that is not needed for a scram. He RWCU system provides no accident mitigation function; therefore, no change to the RWCU pump ; seal water could potentially increase dose to the public. Therefore, this change does not increase the consequences of an accident previously evaluated in the SAR. Neither is the possibility of an accident or malfunction of safety-related structure, system, or component which is different than ; any previously evaluated in the SAR created. The functions of the safety-related structure, systems, or components of the RDS system are not affected by the proposed nxxiification. Neither l does the RWCU system provide an accident mitigation function. Therefore, this modification does not increase the consequences of a malfunction of safety-related structure, system, or component , previously evaluated in the SAR. Technical Specifications require all control rod scram accumulators to be operable. This requirement is verified by ensuring that the indicated pressure is greater than or equal to 1520 psig unless the control rod is inserted and disarmed or scrammed. ! The flow of the additional line decreases the RDS pressure by approximately 10 psig, but the RDS lines nonnally run between 1750 and 2000 psig. Therefore the margin of safety as defined by ;
~
Technical Specifications is not reduced. Thus, this change does not constitute an unresiewui safety question. I Channe Number /USAR Section: LCN 4.6-29 Page 4.6-10 l
' Description and Basis for Channe:
This modification allowed the use of finer mesh filter elements in the control rod drive (CRD) i pump discharge filter housings. He allowable filtration rating was changed to a range of 50 microns absolute to 15 microns absolute. His modification improved water quality from CRD >
- pump discharge to improve component reliability and lower costs. :
i Section I Page 17 of187 i t
l 1 Susunary of Safety Evaluation: ;
. W affected filter elements were not postulated to cause any accident evaluated in the SAR. This [
modification did not adversely affect any component downstream from the filters since the existing l design and operating limitations on filter element differential pressure remained unchanged. The i failure probability of the affected system and surrounding equipment was thus uncharged by this j modification. Wrefore, this modification did not increase the probability of an accident or a j
- malfunction of equipment important to safety previously evaluated in the SAR. The CRD pump i discharge filters are not required to mitigate the consequences of any previously evaluated accident j and they are not required to support any equipment important to safety that mitigates accident i consequences. Although there is equipment downstream that is required to mitigate accident i consequences, that equipment was unaffected by this modification. Wrefore, this modification ;
did not increase the consequences of an accident or a malfunction of equipment important to safety ; previously evaluated in the SAR. This modification did not introduce any new failure modes. Therefore, this modification did not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This j modification did not affect any technical specification or the basis of any technical specification. Therefore, this modification did not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification did not constitute an unresiewed safety ; question. i
' Channe Number /ljSAR Section: LCN 4.6-030 SEN 95 0084 Figure 4.6-5c ;
l Description and Basis for Channe: ; his modification adds larger tubing and continuous vent lines from the upper (low pressure) instrument lines back to the scram discharge volume (SDV) header to prevent the formation of a vapor lock and to allow any water trapped in the diaphragm assemblics to drain. The increased i size of the instrument tubing and vents will climir. ate any capillary action that may be contributing to the water being trapped. Summary of Safety Evaluation: A review of the SAR indicates that failure of the SDV instrumentation is not an initiating event for ! ary accident evaluated in the SAR. Nor are these instrument modifications postulated to initiate l any previously evaluated accident. His change does not affect the instrument setpoint, setpoint bases or instrument calibration. The piping and tubing installed by this modification has been analyzed, qualified and supported to the same criteria as the existing tubing and piping. This modification does not affect any of the control rod drive, scram discharge volume or scram discharge instrument volume systems components' safety function. Herefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. None of the changes associated with this modification would reduce the ability of the instruments to perform their design functions ofinitiatmg a control rod
- block or a reactor scrum. Because all equipment will function as assumed in the accident analysis there is no change in the radiological consequences at the site boundary. None of the instmments am :equired to perform an action to mitigate an accident. hrefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. His modification does not add any new function to thi . cram Section I Page 18 ofI87
discharge instrument volume level transmitters. No new equipment failure modes are postulated to occur as a result of tlGs MR. This modification will not affect the operation of or cause any , malfunction to any adjacent systems, structures or components which are important to safety. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This modification has no effect on the safety function of the control rod drive, scram discharge volume or scram discharge instrument volume systems. Therefore, this modification does not reduce the margin of safety as dermed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 4B.00-006 Chapter 4 Chapter 5 Chapter 15 Section 4.2 Section SA Section 15.0 Page Page Pages 4.2-11 SA-1 15.0-1 Table 5A.1-1 15. 0-12 ; Section 4.3 Figure 5A.3-1 Table 15.0-3 Pages Figure 15.0-1 4.3-2 Chapter 6 4.3-4 Section 6.2.5 Section 15.4 4.3-10 Pages Pages Figure 4.4-5 6.2-98 15.4-15 6.2-99 Table 15.4-6 Section 4.4 6.2-101 Table 15.4-7 Pages 6.2-103a Table 15.4-11 4.4-1 6.2-106a 4.4-2 6.2-108 Section 15.7.4 6.2-109 Pages Appendix 4B 6.2-110 15.7-15 Pages 6.2-111 15.7-17 4B-ii Table 6.2-45 15.7-18 4B-1 Figure 6.2-68 Table 15.7-11 4 B-2 Figure 6.2-69 Table 15.7-13 4 4 B-3 4 B-3a Chapter 8 Section ISB 4B-3b Section 8.3 Pages 4B-4 Figure 8.3-14a Title Page Figure 4B.2-1 Figure 8.3-14b 15B-1 Figure 8.3-15 15B-2 Table 8.3-2a 15B-2a Table 8.3-2b 15B-3 Table 15B.1-1 Table 15B.3-1 Table 15B.3-2 Figure 15B.3-1 Figure 15B.3-2 Figure 15B.33 i Section 1 Page 19 of I87 i
Description and Basis for Chance: This change involved the insertion of the new fuel bundles into the reactor core. The core design for Cycle 7 consists of 392 partially bumed GE8 and 232 new gel 1 fuel bundles. This modification was made and the 50.95 safety evaluatN was performed to address the operation of the Cycle 7 reload core, core shuffle and the storage of spent fuel during RF-6 to ensure that the Cycle 7 reference core design meets all applicable safety and regulatory requirements. The safety evaluation specifically addressed design compatibility, thermal-hydraulic compatibility, over-pressure protection, spent fuel storage and handling ECCS requirements, and transient and accid:nt requirements. Also addressed were the required changes to the core operating limits report (COLR). Summary of Safe y Evaluation: The introduction of GElI fuel bundles does not impose any changes to mechanical fuel handling characteristics. GE l 1 is lighter (no additional scismic impact and no additional loads placed on the fuel storage racks) and requiro no new handling equipment. The mixing of GE8 and GElI have been neutronically accounted for based on NRC approved methods. The appropriate critical power correlation and loss coefficients are mcorporated appropriately into the thermal-hydraulics analysis methods which remain unchanged from Cycle 6, a fuel type dependent OLMCPR will be incorporated into the Cycle 7 COLR, and general stability analysis performed by the fuel vendor demonstrates compliance with stability criteria and determined it to be applicable to Cycle 7. Analyses have been performed to ensure sufficient shutdown margin will be maintained and that measures will be taken to preclude inadvertent core critically during RF-6 core shuffle. Spent fuel geometry and other storage conditions have not changed. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. Based on physical characteristics, failure of an individual fuel rod will not release more actisity for the gel 1 fuel; the Cycle 7 fuel and core design performance exceeds the applicable reactisity control requirements for shutdown margin, standby liquid control system and the spent fuel storage rack; gel 1 fuel lift analysis was compared to the GE8 fuel lift analyses and was shown that the potential for fuel lift ofThas not increased because of gel 1; the radiological consequence for gel 1 is bounded by the original SAR analysis which assumes a conservative noble gases and , halogen release; the radiological release for the mixed core of gel 1 and GE8 are well within the l 10CFR100 limits during a FHA; the maximum fission gas inventory for the GElI bundle will be i less relative to the GE8 due to the reduced fuel mass and decreased fission gas release fraction resulting from the lower temperature operation over most ofits life; and operation within the i COLR limits will assure acceptable consequences; the storage of fresh and irradiated fuel in the fuel building and the upper pool storage rack will not affect the precursors to any accident previously evaluated. Therefore, this modification does not increase the consequences of an accident previously evaluated in the S AR. Existing fuel handling procedures and monitoring will continue to provide protection against any fuel handling errors; the mixed core design will not create an accident of a different kind presiously i evaluated in the SAR; the GE I1 design part-length rods have been used successfully in other BWRs and industry experiences demonstrate that part-length rods cause no new interactions and thus no new types of accidents. There are no new system interactions or interconnections I Section 1 Page 20 of 187 l
I associated with reload of the Cycle 7 core design. Hus, the Cytle 7 core design cannot create the ) possibility of an accident of a different type than any evaluated previcusly in the SAR. All equipment important to safety will function in the same manner with the Cycle 7 reload core as with the presious core. There is no characteristic of the Cycle 7 core different from presious designs. The gel I will place no more loads on the fuel handling equipment and the assembly continues to be fully compatible with this equipment. Since the design criteria are met, the use of gel I will not increase the probability of a malfunction of the reactivity control system. Since the i same design method and same acceptance criteria are used, the probability of equipment ) l malfunction (in this case fuel failure) remains bounded by the SAR. No modification of fuel l l shuffle equipment / tools is required and no new equipment is needed for the shuffle or loading of GE8 and GElI fuels. Based on these discussions, the proposed Cycle 7 reload design will not J increase the probability of occurrence of malfunction of equipment important to safety evaluated j previously in the SAR. All equipment important to safety will function in the same manner with the Cycle 7 reload core as with the previous core design. He adoption of GElI fuel in the Cycle 7 core design will be transparent to plant equipment used to mitigate or prevent any potential accidents. Cycle 7 reload does not require new hardware or modification of existing hardware. Any individual fuel rod ( failure caused by malfunction of equipment will result in radiological doses bounded by the SAR. The SLC will shutdown the reactor during the unlikely event of an ATWS, and as such, the l consequences due to equipment malfunction remain the same. Based on these discussions, the proposed Cycle 7 reload core design will not create a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The Cycle 7 reload core design cannot create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated. Equipment important to safety will still be required to function in the same manner with the Cycle 7 core as with previous core designs. The change in core characteristics does not change any parameter that would affect the function of equipment important to safety. There are no new failure methods for equipment important to safety which are introduced by the Cycle 7 reload core design. Therefore, this modification does not create the possibility of a malfunction of equipment important to safety which is difTerent from any previously evaluated in the SAR. I The Technical Specifications safety limits and refueling operations remain unchanged from Cycle l
- 6. Analyses have been performed to ensure at least 1% shutdown margin will be maintained for any in-core fuel movement. The available shutdown margin of 1% meets the requirements of Technical Spccific.ations. Therefore, this modification does not reduce the margin of safety as defined in the basis to any technical specification.
J For the above reasons, this modification does not constitute an unreviewed safety question. Chance Number /llSAR Section: LCN 5.1-13 Figure 5.1-3b I Descrintion and Basis for Chance: Trip units B21-N693A and B21-N693B, which actuate on reactor pressurc vessel (RPV) water i level high (Level 8) and energize an alarm and trip the RCIC turbine, are shown receising signals fium Level Transmitters, B21-LTN091 A and B21-LTN091B. He trip units actually receive their Section i Page 21 of 187
l l l l level signals from the Transmitters, B21-LTN095A and B21-LTN095B, which are used for RPV l water levels 3 & 8. Summary of Safety Evaluation: The trip units shown on the nuclear boiler instrumentation do not cause an accident identified in the SAR, specifically a decrease in reactor coolant temperature, caused by feedwater controller failure. , The trip units and the associated level transmitters are used to mitigate the consequences of this accident. Therefore, the probability of an accident previously evaluated in the SAR is not increased. He RPV water level instrumentation provides signals to either initiate equipment or I trip equipment based on the specific RPV water level. The existing trip units initiate a closure of the reactor core isolation cooling system (RCIC) turbine steam supply valve upon a RPV high water level signal. This stops the source of water that is filling the RPV. There is no change to the i operation of the transmitters or trip units associated with RPV high water level and low water level. Therefore, the consequences of an accident previously evaluated in the SAR is not increased. No new failure mechanisms are introduced by connecting the trip units to the correct transmitters, which allows the existing safety equipment to continue to operate properly. Therefore, the change does not create the possibility of an accident which is different than any previously evaluated in the SAR. The trip units and transmitters are not being replaced nor is any field work being performed on these components. Therefore, no new failure mechanisms are introduced and the probability of a malfunction of a safety related structure, system, or component (SSC) previously evaluated in the SAR is not increased. There is no change to the operation of the transmitters or trip units associated with RPV high level and low level setpoints. Therefore, the consequences of a malfunction of a SSC previously evaluated in the SAR is not increased. Instead of receiving a signal from the wide range RPV level transmitters the trip units will receive a signal from the narrow range transmitters. These trip units will continue to function properly since they will continue to receive a signal from transmitters whose range includes the high level setpoint. There are no new operating scenarios of the trip units and RCIC turbine. Therefore, the possibility of a malfunction of a safety related SSC different than any previously evaluated in the SAR is not created. Technical Specifications states that the RPV high water level initiates a reactor scram approximately two feet above the normal operating level. This is intended to offset the addition of reactivity effect associated with the introduction of a significant amount of relatively cold water. The ability to initiate a scram is not affected by this SAR change since the RPV level transmitters and trip units used to ;nitiate a reactor scram is different. Therefore, the margin of safety as defined in the basis to any technical specification is not reduced. For these reasons, this change does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 05.01-016 Figure 5.1-3C Descrintion and Basis for Chance: This modification relocates Restricting Orifices 1RCS-RO109A,B,C, and D from their current location, upstream of Check Valves 1RCS-V3050A,B,C,D, to downstream of these check valves. The current location of the orifices can allow a check valve induced pressure perturbation to directly affect the reference legs and transmitters. After relocation downstream of the check valves cach orifice will function as a dampener to lessen the impact of any check valve induced pressure oscillations. The orifice relocation j has no adverse impact on system function or operation. The purpose of this modification is to reduce the possibility of a false water level indication and subsequent scram. Section i Page 22 ofI87 ) 1 I
Summary of Safety Evaluation: j By movb:g the flow orifices downstream of the Category I check valves, they become part of the reactor , coalant pressure boundary. However, they continue to have no safety function. The continuous backfill systs m does not represent a process variable, and it does not impact the initial conditions assumed in the safrty analysis. The instrument tubing is being installed with equal or better standards as compared to the existing tubing. Therefore, the probability of an accident previously evaluated in the SAR is not increased. The affected tubing is safety related, which means that it is required to mitigate the consequences of an accident for safe shutdown of the plant. The insertion of passive components such as a flow orifice does not increase the failure rate of the tubing in the vicinity of the new location of the components. The components are small enough that they do not afTect the dynamic loads on the tubing. The tubing and components are on the containment side and the containment pressurization characteristics of the instrument line break scenario are unaffected. Therefore, the consequences of an accident presiously evaluated in the SAR is not increased. All tubing associated with the continuous backfill system has a design pressure of 2000 psig. The reference leg root valves will be disabled in the open position, thereby minimizing the possibility of pressure boundary failure of the reference leg and its associated panel components. All tubing and supports are seismically mounted. Therefore, the possibility of an accident which is different than any previously evaluated in the SAR is not created. The relocation of the orifices enhances the backfill system and has no negative effect on any other system. The tubing is a passive component and failure of these components are not credible. Therefore, the probability of a malfunction of a safety related structure, system, or component (SSC) previously evaluated in the SAR is not increased. The consequences of a malfunction of a safety related SSC previously evaluated in the SAR is r.ot increased due to the fact that relocation of the orifices does not affect the backfill system. This modification affects the tubing and fittings only. Therefore, the possibility of a malfunction of a safety related SSC different than any previously evaluated in the SAR is not created. The relocation of the orifices does not change the design bases, functions, or operations of any safety related equipment and does not adversely affect any other safety related SSCs. Therefore, the margin of safety as defined in the basis to any technical specification is not reduced. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 5.218 Page 5.2-42 Descrintion and flasis for Chance: The SAR states there is a five GPM leak rate alarm associated with the undefined leakage into the containment floor drain sump. The source code for the Leakage Detection Computer system calculates unidentified leakage at a rate of 5 GPM based on the combines leakage into the drywell floor and the pedestal drain sumps. The source code also calculates total leakage at a rate of 25 GPM for the combined leakage into all containment, drywell equipment, and floor drain sumps. Summary of Safety Evaluation: The leakage detection computer and the components which supply inputs to this computer do not cause any accident identified in the SAR. This revision is a document change only and makes no physical change to the plant. Therefore, the probability of an accident previously evaluated in the SAR is not increased. 'Ihc containment floor drain sump monitors unidentified leakage, but not from the RCPB. Leakage into the containment floor drain sump is combined with leakage into the containment equipment drain, drywell floor, and drywell pedestal drain sumps to proside an alarm (2413) when total leakage exceeds 25 GPM. This alarm is generated from the Leakage Section i Page 23 of I87
Detection Computer which also provides a readout of the actual leakage from the containment floor drain sump as well as the other four sumps. Alarm 2413 is not affected by this revision. The function of the leakage detection capabilities that presently exist inside the drywell, and extemal to the drywell, will continue to provide the control room operators with the alarms needed to alert them to a potential degradation of the RCPB. Therefore, this revision will not increase the consequences of an accident previously evaluated in the SAR. The alarm does not exist in the plant and does not need to be added to the plant for the reasons stated above. Since the alann does not exist, the possibility of an accident widch is different than any previously evaluated in the S AR will not be created. Since the revision does not make any physical change to the plant nor does it affect the operation of any safety-related equipment the probability of a malfunction of a safety-related structure, system, of component previously evaluated in the SAR will not be increased. The consequences of a malfunction of a safety-related SSC previously evaluated in the SAR will not be increased due to the nonexistence of the alarm. The alarm functions as a leakage warning only and does not interface with the operation of any safety-related equipment. Therefore, this will not create the possibility for a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. There is no design or operational change associated with this activity. This design change will not reduce the margin of safety as defined in the basis for any technical specification. For these reasons, this alteration does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 5.4-111 Figures: 9.3-7j 5.4-8 Descrintion and Basis for Chance: his modification adds a new equipment drain line from reactor core isolation cooling system (RCIC) turbine bedplate to the existing floor drain hub. The drain line will be terminated in a raised-rim hub to prevent contamination of equipment drainage with floor surface drainage. The equipment drainage systems are not safety related and are designated nonnuclear safety (NNS). The pipe is seismically supported to protect auxiliary building safety related equipment. Summary of Safety Evaluation: The drain piping is routed directly to the floor drain which leads to the RCIC auxiliary building sump. The sump and its pump are sized to handle the calculated discharge from the equipment served by the sump. Therefore, this change does not effect any structure, system, component) l utilized in the accident analysis. Therefore, the consequences of an accident previously evaluated in the SAR is not increased, and the possibility of an accident which is different than any ; previously evaluated in the SAR is not created. Installation of the bedplate drain lines has no effect on any safety-related SSC and there will not be an increase in the probability of a malfunction other than that previously stated in the SAR. There will also be no increase in the consequences of a malfunction of a safety-related SSC previously evaluated in the SAR. In addition, the possibility of a malfunction of a safety-related SSC different than any previously evaluated in the SAR is not created. The margin of safety as defined in the basis to any Technical Specification will not be reduced because the change has no affect. For these reasons, this change does not constitute an unreviewed safety question. Section I Page 24 of 187
l Channe Number /USAR Section: LCN 05.04-113 Figure 5.4-11Sa i i Descriraion and Basis for Channe: ! This change replaced Temporary Strainer IWCS-STRT-1 A,lB, line IWCS-001-187-3, and Valves V175 and V173 with three inch pipe spools between the existing flanges. These strainers ! were required during initial start-up to protect the reactor water cleanup pumps from construction l debris. De strainer internals were removed after start-up testmg and line flushing. ne dead leg i portion of the strainers traps hot particles thus creatmg hot spots. Replacing the strainers with straight pipe spool of the same material and properties as the original pipe conforms to the original ,
' design configuration. !
Summary of Safety Evaluation: ! The portion of the system which contained the RWCU strainers was located in a portion of the f system isolated from the reactor. The piping and supports affected by this modification were l qualified per ASME Section Ill Class 3 requirements. The pipe spool used to replace the stramers ! is made of the same material and has the same properties thus maintaining the system design and ! safety margin. Therefore, the probability or consequences of an accident previously evaluated in { the SAR are not increased. This change does not introduce any new elements to the system , different from the original design configuration. Hus the possibility of an accident which is ! different than any previously evaluated in the SAR is not created. Since the RWCU pump is not j part of a safety system, removing the strainer which serves no filtration function did not increase l the probability of a malfunction of a safety related structure, system, or component (SSC). The l affected portion of the RWCU system does not fulfill any active safety function. The consequences of a malfunction of a safety related SSC are not increased. There are no adverse impacts to the qualifications required for these components. Hence, the possibility of a malfunction of a safety related SSC is not created. %c margin of safety as defined in the basis to any RBS Technical l' Specification is not reduced because no technical specification nor any basis for a technical specification is impacted by this modification. For these reasons, this alteration does not constitute an unreviewed safety question. l r Channe Number /USAR Section: LCN 5.4-120 MR 95-0016 Figures: 5.4-12a 5.4-12b Description and Basis for Channe: ! l Residual heat removal (RHR) system contains restricting orifices in test return line loops A and B , to provide system resistance to prevent pump runout during system operation. These orifices l contain multiple holes and provide a single stage pressure drop, a design which experiences cavitation at certain flow rates when associated with RHR test / suppression pool cooling and LPCS system test modes of operations. This modification replaces existing orifices with in-line drag resistors which produces multi-stage pressure drops which, in turn, will reduce or climinate the i cavitation. Summary of Safety Evaluation: A review of the SAR was performed and the only accidents identified for tle RHR were the loss of RHR shutdown cooling (SDC) due to a single failure or isolation of the suction litw, and RHR I Section I- Page 25 ofI87
. = .
l shutdown cooling increased due to misoperation of the R11R heat exchangers. This modification has no affect on the RHR heat exchangers' performance. Regarding the loss of shutdown cooling, this modification has no impact on the RiiR, SDC suction line. This modification is designed to meet all the material, quality and safety class, seismic, environmental, separation, system redundancy and reliability requirements of the lines in which the drag resistors are installed. This modification was also analyzed to ensure that the installation can withstand the effects of pipe movement, thermal expansion, LOCA, SRV discharge, safe shutdown carthquake, and suppression pool hydrodynamic loads which were considered in the original design. The operation, perfonnance characteristics and function of the RiiR and LPCS systems are not related to the causes of any accidents previously evaluated in the SAR, but rather, they function to mitigate the consequences of an accident. The modification installation meets all of the same requirements of the original design concerning materials, quality and safety class, seismic and suppression pool hydrodynamic loads, environmental conditions; has no effect on any control or automatic functions in the system or any electrical supply for the system; and physical separation and system redundancy is maintained and provisions for and intervals of system reliability testing are not affected. The malfunctioning of the in-line drag resistor due to fibrous material clogging the passages is not a credible event. It can be concluded that the probability of a malfunction or single component failure in the RliR or LPCS system is not increased nor the probability of a malfunction of other systems and structures which are either connected to or located in thnicinity of the modification. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR The RiiR system in the suppression pool cooling and LPCI modes of operation and the LPCS system are required to function to mitigate the consequences of many of the accidents / transients previously evaluated in the SAR to provide containment cooling and flooding of the RPV This modification does not affect the operability of these systems to function as required in these modes of operation. This modification does not affect the capability of the RiiR or LPCS systems in performing their required function or achieving the required system performance characteristics. Since this modification does not affect the reliability or availability of the minimum required system / flow paths to perform their design functions, it is concluded that the consequences of a malfunction are not increased. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This modification meets the originally provided design and functional requirements for the RHR and LPCS systems. The new restricting flow devices are installed in the existing piping and are designed to meet the same material, quality and safety class, seismic and suppression pool hydrodynamic loading requirements and emironmental conditions as the piping in which they are installed. No new control or alarm functions or hardware are added and no equipment or hardware of a different type than originally existing in the system is added. The assembly will not come apart during postulated seismic or dynamic events. Clogging of the in-line drag resistor holes with fibrous material is not a credible event. Therefore, this modification does not create the possibility of an accident or a malftmetion of equipment important to safety different from any previously evaluated in the SAR. System performance and availability are not affected. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question Section I Page 26 of I87
- - . - - . . ~. -.- - - -- .- -- - - . , _ . -
m ,. ; P i fhanne Number /USAR Section: . LCN 05.04-122a Figures: 5.4-12a 5.4-12c : 4 9.2-Id j 9.3.1b - r 9.3-le - f Description and Basis for Channe: } i Modification Requests 95-0009 and 95-0010 involve the design and installation of the suppression , i pool cleanup and decay heat removal system (SPC). He system is designcd (da;-adWg on the reactor operating mode and valve alignment) for demineralization and filtering of the suppression > pool water, removal of heat from the suppression pool or reactor vessel and control of suppression , pool or reactor vessel water level. l Summary of Safety Evaluation: ! The new piping and components installed under the mechanical portion of MR 95-0009 is not ; physically or functionally linked to any of the causes postulated to initiate or to contribute to the ; occurrence of anticipated or abnormal operational transients analyzed in the SAR. No raceway, ! i cables, or field mounted electrical components are postulated to cause an accident presiously described in the SAR. Penetration in "E" tunnel will not increase the probability of an accident ; previously reposted in the SAR and the secondary containment boundary is maintained. De RHR l "C" system should operate properly and perform its required ECCS function. Herefore, this i
- modification does not increase the probability of an accident or a malfunction of equipment !
important to safety previously evaluated in the SAR. The nature of these changes is such that i ., consequences of any failure of the safety-related equipment are not affected. Affected systems j t- response to accident signals, pressure bouixiary integrity of the RHR "C" suction and discharge l piping is preserved, and the -ad y containment boundary is maintained. De radiological ! consequences of t.n accident are not increased. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. There are no new system interfaces, no new types of components not
- previously used in the plant or acknowledged by the SAR. Herefore, no new failure modes are created for the modified system. Herefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The proposed modification will not affect the results of any accidem analysis that ;
have been previously performed. Important safety parameters are unaffected and unchanged for all ! analyzed events. Therefore, this modification does not reduce the margin of safety as defined in the !
- basis of any technical specification. For these reasons, this modification does not constitute an j unreviewed safety question. l i
Channe Number /USAR Section: LCN 5.4-124 Figures: 5.44 6.3-1 Description and Basis for Channe: , Two motor operated valves, one reactor core isolation cooling suction valve and one high pressure core spray suppression pool suction header isolation valve, were identified as susceptible to I
- pressure locking conditions due to the boiler effect. These valves are modified to install a small .
bypass line from an existing midsent drain for each of the valves into the system piping inboard to 'l Section i . Page 27 ofI87 i i i 5
I l the valve. For each of the two valves, a hot tap is required to the bypass line into the piping I system. As part of this hot tap, a normally open isolation valve is installed as part of the bypass line. The bypass line methodology is consistent with recommended solutions of NUREG/CP 0146, j
" Proceedings of the Workshop on Gate Valve Pressure locking and Thermal Binding." ,
1 Summary of Safety Evaluation: He new bypass lines are in accordance with the applicable codes and standards and current piping specifications. The installation of these bypass lines is in accordance with industry practice and guidelines. Installation and post modification testing will be performed in accordance with existing plant procedures. He addition of these bypass lines will not affect the function of the subject valves to provide containment isolation or to open for accident mitigation. These valves are leak tested in the reverse direction and may see a small increase in leakage due to this modification. However, this test in the reverse direction is very conservative as the valves have their greater scaling ability in the normal containment isolation direction. Any change in seat leakage due to this modification will be measured, recorded, and tracked according to plant procedure. This modification does not directly or indirectly adversely affect any other safety related equipment. This modification has no effect on any system pressure boundary. There is no increase in the calculated radiological dose for a design basis accident resulting with the malfunction of the subject gate valves. Therefore, this modification does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously analyzed in the S AR. Postulation of a new line beak is not required since the new bypass piping and the new tie-ins are 1" or under. This modification does not introduce any new high energy line breaks, sources of , flooding or combustible materials. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The modified valves will continue to provide their containment isolation function and I will continue to open for accident mitigation. His modification did not affect the level of detail contained in the Technical Requirements Manual. Therefore, this modification does not reduce the margin of safety as defined in any technical specification or the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. j i Chanee Number /USAR Section: LCN 5.4-126 Table 6.2-40 Sheet 7 i Figures: 5.4-12A 5.4-12B Description and Basis for Chance: j l This change removes Division I and Il residual heat removal (RHR) system Relief Valves RHS- l RV3A and RilS-RV3B that had been used for the RHR steam condensing mode of operation. These valves are no longer used because the steam condensing mode of the RHR system was abandoned. Each valve will be replaced with a welded-in piping spool piece which contains a welded blind plate for in-line isolation. These replacements will maintain piping system continuity . and relief valve branch line isolation. l Summary of Safety Evaluation: The replacement pipe spool pieces are designed and installed to meet the same material pressure , I boundary and pipe line isolation requirements as the Division I and 11 RHR Relief Valves RHS-RV3A and RHS-RV3B. RHR system modes of operation are not changed or affected by this modification. The pipe spool pieces and blind plates provide the same in-line pressure te adary l Section I Page 28 of 187 i
)
and pipe line isolation as the relief valves, however, since the relief valve function is no longer required, the blind plates are designed to the RHR system piping design pressure. No instrument setpoints or operational procedures are changed as a result of this change. Therefore, this change l does not increase the probability of an accident or a malfunction of equipment important to safety l previously evaluated in the SAR. The replacement pipe spool pieces mxt the RHR piping design and licensing requirements for pressure boundary integrity and containment isolation prosisions that are currently met by RIIR Relief Valves RHS RV3A and RHS-RV3B. He moderate energy through wall pipe crack postulation with regard to plant flooding and spray analysis as well as the , high energy pipe break pipe whips andjet impingement evaluation is not affected. The replacement : pipe spool pieces do use carbon steel pipe spool pieces to replace the RHR relief valves, but the added ccmbustible load to the auxiliary building is negligible. This change does not affect any RHR system operating modes and does not affect any RHR system parameters when it is required to perform its intended function herefore, this change does not increase the consequences of an accident or a malfunction of equipment to safety previously evaluated in the SAR. RHR piping is still protected from ovcrpressure by Relief Valves E12-RVF055A and E12-RVF055B, which have not been removed. The replacement pipe spool pieces do not affect RHR system operation including containment isolation and do not invalidate existing seismic and tornado design criteria. - Herefore, this change does not create the possibility of an accident or a malfunction of equipment i important to safety different from any previously evaluated in the SAR. The replacement spool ! pieces include a welded-in blind plate which provides the containment isolation function that is currently provided by the relief valve seats. The welded-in blind plate also provides a more i efficient leak tight barrier that the relief valve seats. This change has no effect on plant equipment or the design basis that is addressed in the Operating License. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 5.4-128 CR %0473 Figure 5.4-2a Description and Basis for Change: A thermocouple situated on the Phase A stator winding for reactor recirculation Pump A was found to be defective. This change swapped the failed thermocouple with a designated spare by switching the thermocouple leads inside the junction box mounted on the motor case. Summary of Safety Evaluation: ne deletion of the failed thermocouple and the use of the spare thermocouple is in accordance with i the design of the recirculation pump motor and has not been previously evaluated by the SAR. The l affected thermocouples are non-safety-related equipment and the changes made to the l thermocouples do not impact the original system design. No new credible failure modes which I initiate accident scenarios evaluated in the SAR are created by this modification. The swap of l Phase A stator winding thermocouples on reactor recirculation Pump Motor A does not adversely i affect any safety related structure, system or component (SSC). Therefore, this change does not increase the probability of an accident or a malfunction of any safety related SSC presiously evaluated in the SAR. He affected thermocouples are intended to monitor the recirculation pump motor stator winding temperature by providing recorder input to the control room and initiating an 1 alarm upon high stator winding temperature. None of the reactor recirculation pump trips described in the SAR are associated with the loss of the stator winding temperature thermocouples.
' The swap of these two thermocouples has no effect on the reactor coolant pressure boundary. No Section I Page 29 ofI87
new credible failure modes can be identified in which the use of spare thermocouple in recirculation pump motor can contribute to the consequences of an SSC malfunction. Herefore, this change ! does not increase the consequences of an accident or a malfunction of safety related SSCs previously evaluated in the SAR. The subject thermocouples are not required for safe shutdown of the plant. This change is to custing plant equipment within the original design as presiously i evaluated by the SAR. Herefore, this change does not create the possibility of an accident or a malfunction of safety related SSCs different from any previously evaluated in the SAR. These two thermocouples are not listed or referenced in the Technical Specifications, the Technical l Requirement Manual, or the Technical Specification Bases. Hence, these thermocouples are not required to be operable. This change has no adverse impact on the operability of any component or equipment which is required to be operable by Technical Specifications. Therefore, this change does not reduce the margin of safety as defined in any technical specification or the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed , safety question. Channe Number /USAR Section: LCN 6.1-005 Page 6.1-6 Table 6.1 3 Description and Basis for Channe: This change allows the post construction addition of unqualified coatings (2000 square feet total) inside containment. Unqualified coatings are those protective coatings installed inside containment j that do no meet Regulatory Guide 1.54 requirements. The 2000 square feet will be listed in the S AR and will provide an upper bound of additional unqualified enatings allowed inside containment. Unqualified coating additions will be tracked as they are approved by design change documents. Summary of Safety Evaluation: Unqualified coatings inside containment are not postulated to initiate any accident. Because the coatings are passive, a change in environmental condition has to occur to involve unqualified protective coatings. Therefore, the unqualified coatings alone do not increase the probability of any accident presiously evaluated in the SAR. A main steam line break (MSLB) is the most severe postulated accident due to the large amount of fibrous material and unqualified coatings (drywell ceiling paint) which could become debris. The amount of unqualified coatings removed from the drywell ceiling during a MSLB was calculated st 616 square feet. If a MSLB were to occur, it is ! , calculated that the ECCS suction strainers would become 11% blocked due to all loose debris in 3 containment. ECCS strainer blockage below 50% is allowable under any accident condition. The unqualified coatings will not block the ECCS suction strainers beyond acceptable limits. Herefore, this change will not increase the consequences of an accident previously evaluated in the SAR. Unqualified coatings are static and serve to protect structures, systems, and components (SSC) from corrosion and facilitate decontamination. They serve no active function during any accident event. Protective coatings in themselves cannot initiate any accident scenario. Therefore, the unqualified coatings do not create the possibility of an accident which is different than any previously evaluated in the SAR. The ECCS suction strainers in the suppression pool have been proven to remain capable of meeting design performance requirements with the addition of unqualified coatings inside containment. Small particles, less that 3/32", may filter through the ; ECCS strainer. However, these small particles will have no physical effect on the ECCS l components or systems. The ECCS components and systems will not be affected chemically by the ! Section i Page 30 ofI87
y 4 - i , i i j paint debris either. hrefore, this change does not increase the probability of a malfunction of any ! i safety related SSC previously evaluated in the SAR.' Protective coatings in themselves do not i mitigate the outcome of any accident event. De only safety related function of the protective : 1 coating is to remain adhered to its substrate during accident conditions. Herefore, this change will not increase the consequences of a malfunction of any SSC previously evaluated in the SAR.
- Protective coatings cannot create a malfunction of equipment due to their passive nature. h j j unqualified coating will remain intact on those surfaces to which they were applied until an !
- accident dislodges them. Review found dislodged coatings do not affect any equipment in the plant i
- other than the ECCS suction strainers.' hrefore, unqualified coatings do no create the possibility i of a malfunction of any safety related SSC different than any previously evaluated in the SAR. !
Unqualified coatings inside the containment have no adverse cEct on any technical specification. hrefore, this change does not reduce the margin of safety as defined in the basis of any technical ,
- specification. For these reasons, this modification does not constitute an unreviewed safety [
- question. ~[
- i. ;
Channe Number /USAR Section: LCN 6.2-043 Pages: 6.2-83 i 6.5-2 j 6.5-3
- 6.5-8 9.4-62 3
4 9.4-63 I 9.4-67 9.4-72 9.4-73 1 ! Figures: 7.3-7 Sheet I ! 7.3-9 Sheet 6 7.3-9 Sheet 7
- 7.3-9 Sheet 8
i 7.3-9 Sheet 9 7.3-9 Sheet 10 1 ] 7.3-9 Sheet 11 7.3-20 Sheet 3 I l 1 7.3-20 Sheet 7 I 9.4-7a j 9.4-7c ! Descrit, tion and Basis for Channe: j This change replaced switches in the annulus pressure control system. W Exhaust Low Flow : Switches lifVR*F556A and lilVR*F556B used to start the standby gas treatment system (SGTS) during low flow conditions were replaced with new differential pressure switches. These new , switches monitor the exact process parameters that must be monitored to start the SGTS and are >
- . indirectly able to determine losses of the annulus vacuum and normal heating, ventilation, and air conditioning (HVAC) system operations.
. Summary of Safety Evaluation: ;
This modification does not inen:as,c the probability of or the or consequences of an accident i previously evaluated in the SAR. Replacing the Q-Class I low flow trip switches, which start the - l Section I- Page 31 ofI87 .i
SGTS during a loss of annulus pressure, with Q-Class I differential pressure switches does not affect the safety of the system.' Starting the SGTS during a low annulus vacuum is not considered an engineering safety feature. Starting the SGTS during a loss of coolant accident or high annulus - efiluent radiation is considered an engineering safety feature, but this modification does not affect the startup. This modification will not increase the probability or consequences of a malfunction of any equipment important to safety previously evaluated in the SAR. The new differential pressure switches are not affected by inleakage variables and will be able to detennine when the annulus vacuum is reduced, whereas the flow trip switches were severely affected by inleakage variables. Redundancy and diversity are maintained so that a single failure would not prevent the actuation of the SGTS from starting. Therefore, this modification decs not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This change does not affect the parameters of any technical specification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 6.2-068 Pages: 1.2-27 3.1-20 6.2-82 7.2-9 7.2-11 7.3-14 7.6-12 10.4-9 11.5-13 Tables: 6.2-40, Sheet 1 6.2-40, Sheet 2 6.2-40 Sheet 18 11.5-1 Figures: 7.3-2, Sheet 4 15A.6-36 Description and Basis for Channe: This modification removes the four existing main steam line radiation monitors (MSLRM) and installs two new radiation monitors (lines A & C). The MSL high-high radiation and inoperable inputs to the reactor scram, the MSIV closure and MSL drain valves logic will be deleted. The B & D channels of MSLRM will be deleted. Alllogic inputs with the exception of the reactor water sample valves will be removed and the alarm circuits will be revised to remove the reactor scram alarm and alarms associated with channels B & D. A revision is to be made to the offgas pre-treatment radiation monitor high alarm nominal setpoint in order to comply with the NRC requirement that the Licensee standardize the radiation monitor high radiation alarm setpoint at 1.5 times the average process background radiation level. Summary of Safety Evaluation: The MSLRMs wn:t not cause an accident identified in the SAR, nor will the offgas pretreatment radiation monitor. The removal of the B & D channels of MSLRM will not create an accident of a different type than any previously evaluated in the SAR since existing failure modes and effects for the remaining A & C channels of MSLRM will not be changed by this modification. These channels will still trip the mechanical vacuum pumps, isolate the reactor water sample valves and Section I Page 32 of187
initiate a MSL high radiation alarm in the control room. Actuation of the MSL high radiation alarm from the A & C channels of MSLRM or the offgas high radiation alarm will provide adequate redundancy to ensure that the plant operators are aware of a high MSL radiation condition. This will ensure actuation of the alarms and subsequent prompt operator action. Additionally, the MSL and ofTgas radiation high-high and MSL radiation high setpoints are unchanged by this modification. The revision to the offgas pretreatment radiation monitor high radiation alarm setpoint does not add any new interfaces with equipment important to safety. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The power requirements of the NUMAC MSLRMs do not cause any adverse impact on the existing distribution system. Spurious trips will be eliminated with the new improved NUMAC MSLRM design. Spurious closure of these valves during normal plant operation will not create any abnormal operating conditions that could affect plant safety. The reactor water sample valves at the sample station will still close on a MSL high-high radiation signal. He offgas pretreatment radiation monitor high radiation signal provides an alarm output only. The setpoint change will stdl ensure a timely operator response to an ofTgas pretreatment radiation monitor high radiation alarm. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. The existing wiring and cables from the MSLRMs (A & C) and locations of the MSLRMs (A & C) themselves will not be changed by this modification. Therefore the existing redundancy and physical independence of the A & C channels of MSLRM will not be affected. The NUMAC LRM will still initiate a control room alarm when the MSLRM is inoperative. The existing failure modes of the reactor water sample valves and mechanical vacuum pumps will not be changed by this modification since there are no changes to the actual equipment; the changes are to the control circuits for the reactor water sample valves only. The revision to the offgas pretreatment radiation , monitor high radiation alarm setpoint will not affect the ability to monitor the consequences of i equipment malfunction; it will actually enhance this ability. The changes described above will not ! result in any changes to the radiological consequences at the site boundary. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The NUMAC LRMs are like-for-like replacements for the existing MSLRM. The safety-related function of closure of the reactor water sample valves and tripping of the mechanical vacuum pumps have not been altered. The mechanical vacuum pump will still be tripped on a high MSL radiation signal from channels A & C. The MSL radiation high-high setpoint has not been altered and no equipment will be required to operate in a different or abnormal manner due to this design change. He revision to the ofTgas pretreatment radiation monitor high radiation alarm setpoint , does not require change to any existing interfaces or the creation of any new interfaces with l equipment important to safety. Therefore, this modification does not create a malfunction of j equipment important to safety different from any previously evaluated in the SAR. The MSLRM high-high radiation inputs to the reactor scram and MSIV & MSL drain valves closure logics were previously deleted from the Technical Specifications by the Improved Technical Specifications but the margin of safety has not been reduced. The change to the offgas pretreatment radiation monitor high radiation alarm setpoint will not decrease any margin of safety. Herefore, this modification does not reduce the margin of safety as defined in the basis of any j technical specification. For these reasons, this modification does not constitute an unresiewed ' safety question. Section 1 Page 33 ofI87 l
F l l r i .i
-l ; Channe Number /USAR Section: LCN 6.3-27 ; Figures: 5.4-12c l 5.4-8 l
- 6.3-1 .j 6.3-4 ~
Description and Basis for Channe: } As a result of Generic Letter 95-07, several motor operated valves were identified as susceptible to !
; pressure locking and/or thermal binding. These valves were modified by installing a small bypass j i' line from an existing stem leakoffline or midseat drain for each of the su< ceptible motor operated j valves to an existing drain / test connection on the system piping inboard to the valve. He bypass l line methodology is consistent with the recc wr.cr.ded solution of NUREG/CP 0146, " Proceedings of the Workshop on Gate Valve Pressure Locking and Thermal Bindmg "
j Summary of Safety Evaluation: l
.. i
! The new bypass lines are in accordance with applicable codes and standards and current piping l
- specifications. De installation of these bypass lines is in accordance with industry practice and guidelines. Installation and post modification testing will be performed in accordance with existing !
l plant procedures. De addition of these bypass lines will not affect the function of the subject f valves to provide contamment isolation or to open for accident mitigation. A small leakage could [ 4 exist during HPCS and RCIC pump testing and will have a non-measurable e&ct upon the reactor coolant temperature and reactor vessel level. This modification does not directly or indirectly
- adversely affect any other safety related equipment. Derefore, this modification does not increase
,. the probability of an accident or the probability of a malfunction of equipment important to safety
- previously evaluated in the SAR. The containment isolation function of these valves will still be ;
- maintained by the outboard disc of their flex wedge design. Any change in seat leakage due to this -{ modification will be measured, recorded, and tracked according to plant procedure. This l modification has no effect on any system pressure boundary. Dere is no increase in the calculated : radiological dose for a design basis accident resulting in the malfunction of the subject gate valves. l Derefore, this modification does not increase the consequences of an accident or the consequences I of a malfunction of equipment important to safety previously analyzed in the SAR. Postulation of l a new line break is not required since the new bypass piping and the new tie-ins are 1" or under. l Therefore, this modification does not create the possibility of an accident or the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. The modified valves will continue to provide their containment isolation function. His
]
i modification does not reduce the margin of safety as defined in any technical specification or the i basis of any technical specification. For these reasons, this modification does not constitute an I i- unreviewed safety question. f l , Channe Number /USAR Section: LCN 06.07-013 Pages: 6,7-3 9.3-41 Descriotion and Basis for Channe: l The penetration valve leakage control compressors A and B are prosided to supply compressed air to the main steam leakage control system (MS-PLCS) and penetration valve leakage control system
. (PVLCS) after a design basis LOCA to minimize the release of fission products from containment.
Section I - Page 34 of I87 4
1 I l I t i
~ Also this system is a 1 ackup to the main steam safety relief system for safety relief valves These compressors are located in the auxiliary building. At pressure 105/110 psig, the capacity of l Compressor 'A' and 'B' was found to be 51.0 and 43.4 scfm respectively as compared to 60 scfm i noted in the SAR. Although the actual capacity of both compressors is more than the design basis l i
requirement, it is less than the SAR figure. An analysis was performed and established tha: the discrepancy will result in no adverse conditions. Both compressors are scheduled to be overhauled l during RF-6 (January-February,1996). j 4 Summary of Safety Evaluation: ) r LSV compressors are not postulated to cause an accident already described in the SAR. Deir j main function is to minimize the release of fission products to the en5ironment is case of a design , basis LOCA. To achieve this, about 36 scfm of compressed air is required per the referenced ! calculations. Therefore, the probability of an accident previously evaluated in the SAR is not ! increased. De LSV system is a safety related system designed to mitigate the consequences of a , LOCA. The leakage control system is intended to be manually initiated by the operator twenty l minutes after a derign basis LOCA to pressurize containment isolation valve bodies and pipe space i between the outboard and inboard isolation valves to minimize the release of fission gases to the ! environment to less than the 10CFR100 limits. These compressors also provide long and short i term demand of compressed air to SVV system should the non-safety-related SVV system fail or a loss of offset power occur coincident with the LOCA. Revising the SAR to list the design basis i requirement of compressed air from the LSV System rather than the compressor capacity does not ! result in a negative impact on the system's ability to perform its design function and thus does not l increase the consequences of an accident previously evaluated in the SAR. The compressors still i meet the design basis requirements for post LOCA demands and therefore, the possibility of an accident which is different than any previously evaluated in the SAR is not created. The proposed l activity does not change or modify any equipment important to safety. Since all capacity l requirements are met by the compressors, then the probability of a malfunction of a safety related [ structure, system, or component (SSC) previously evaluated in the SAR ;is not increased. As long as the compressor output is more than the design basis requirement, the LSV system is capable of performing its safety related function. Therefore, the consequences of a malfunction of a safety : related SSC previously evaluated in the SAR is not increased. Since the design basis output is met l by each compressor, the possibility of a malfunction of a safety related SSC different than any i previously evaluated in the SAR is not created. There is no margin of safety associated with the ! LSV system described in the Technical Specification Bases. Therefore, the margin of safety as ! defined in the basis to any Technical Specifications is not reduced. For these reasons, the ! modification does not constitute an unreviewed safety question. ! f Channe Number /USAR Section: LCN 7.1-7 Pages: 7.1 i 7.1-16 ; 7.1-17 ; 7.1-19 i Description and Basis for Channe: This modification adds pre-isolation temperature alarms for the main steam tunnel, main steam line f (MSL) shield wall, and MSL moisture separator and reheater. De auxiliary analog output from ; the trip units for isolations is used as an input to Opto 22 multiplexers using optically coupled, ! transformer isolated analog input modules. The connection from the auxiliary analog output of the I L ~ l Section I- Page 35 of187 i I
_ __ _ _ - ~ _ _ - . _ . - _ _ - _ _ _ _ . - _ - _ _ _ ._ _ .. i i l i trip unit or the multiplexer input module is an associated divisional circuit routed in conduit. De . i multiplexers send the digitized temperatures via an RS422 link to the existing IHA-CPI computer for processing and output to multiplexer driven annunciator displays. t Summary of Safety Evaluation: : i De use of auxiliary analog output feedmg an associated divisional ciremt mcluding the multiplexer l input module prevents the non-safety-related multiplexer from degrading the safety function of the ; I trip unit. De use of associated divisional circuits to feed multiplexers cannot increase the i probability of any evaluated accident. Therefore, this modification does not increase the ! j probability of an accident previously evaluated in the SAR. No single failure exists due to the ! interfaces added to the safety related trip units that can change the consequences of an accident. , The use of associated circuits as inputs to multiplexers cannot change the consequences of any l evaluated accident. Therefore, this modification does not increase the consequences of an accident i previously evaluated in the SAR. This modification does not create the possibility of a shorting or grounding of any of the affected trip units. He voltages that currently exist in the bay where the multiplexer is located are no greater than the voltages in the location of the trip unit, so no new l type of fault exists after these changes are implemented. There are restrictions on the location of ; l the multiplexer with associated circuit inputa so that no new type of fault is introduced and no { unanalyzed accident or failure mode is created. Herefore, this modification does not create the possibility of an accident different from any previously evaluated previously in the SAR. No faults ! e shorts greater than the isolation capability of the trip units can occur as a result of this ! modification. Failure of the annunciator system including the multiplexers does not impact the safety related isolation functions. Therefore, this modification does not increase the probability of
~
E a malfunction of any safety related equipment previously evaluated in the SAR. Use of the ; ! auxiliary inputs of the trip units to drive the annunciators and multiplexers does not change the ! consequences of a trip unit failure. Therefore, this modification does not increase the consequences of a malfunction of any safety related equipment. No new fault levels are added by this modification. Therefore, this modification does not create the possibility of a malfunction of any ,
- safety related equipment different from any previously evaluated in the SAR. The Technical !
! Specification Bases for isolation actuation instrumentation is not changed by this modification. ! l Therefore, this modification does not reduce the margin of safety as defined in the basis of any ,
- technical specification. For these reasons, this modification does not constitute an unresiewed i- safety question. ;
I i Channe Number /USAR Section: LCN 07.02-029 Table 8.3-7 ! l Figure 7.2-1 Sheet 1 l Description and Basis for Channe: i 1 i This modification installed a test access facility to monitor input and output of the electrical j protection assembly (EPA) breakers IC71*S003E, IC71*S003F, IC71*S003G, and IC71*S003H l connected to an alternate power source. These breakers have been spuriously tripping. The test equipment will help to determine the root cause of the breaker trips. His modification revises the f reactor protection system (RPS) circuitry to connect Non-Class IE test equipment to RPS Class IE EPA breakers. Divisional separation between the Class IE and Non-Class IE circuits has been , achieved using double fusing in the circuit. Use of double fusing in the test control circuit has been j shown not to degrade the Class IE circuit below an acceptable level. ; I
, Section 1 - Page 36 of187 ,
~, . . - - _ _ -. - . _ . . , . -
Summary of Safety Evaluation: Installation of monitoring circuits according to Class IE requirements and providing isolation of the test equipment from the EPA breakers ensures that no credible fault exists which would prevent
- the RPS trip system or the EPA breders from perfonning their safety related functions. This -
modification will not degrade safety related power supplies 1RPS*XRC10Al and IRPS*XRC1081 or other Class IE portions of the circuits in question. He monitoring circuits
. will only be used during those times in which monitoring is required to determine the root cause of
. spurious EPA breaker trips, Installation of monitormg circuits according to Class IE
. requirements, and us'c of these facilities according to existing plant operating and test procedures will ensure that the reliability of the RPS trip system, the alternate power supply, and the EPA breakers is not reduced. Derefore, this modification does not increase the probability of an accident or a malfunction of any safety related system, structure, or component (SSC) presiously evaluated in the S AR. He installation of monitoring circuits will not alter the consequences of a failure to trip by the RPS EPA breakers. Derefore, this modification will not increase the
_ consequences of an accident or a malfunction of any safety related SSC previously evaluated in the SAR. Any failure of the monitoring circuits will not affect the EPA breakers. No new accident initiators are added by this modification. Herefore, this modification does not create the possibility of an accident or a malfunction of any safety related SSC different from any presiously evaluated in the SAR. De RPS Power System is considered nca-essernial due to the fail safe of the RPS. Acrefore, the margin of safety as defined in the basis of any technical specification is not reduced. For these reasons, this modification does not constitute an unresiewed safety question. Channe Number /USAR Section: LCN 7.2-30 Page 8.3-20 i Description and Basis for Chanee: ' This modification will replace existing electrical protection assemblies (EPAs) that utilize General ; Electric (GE) logic cards to monitor the reactor protection system (RPS) power supplies for undervoltage, overvoltage, and underfrequency conditions. De replacement EPAs will utilize solid l state relays to sense undervoltage, ove voltage, and underfrequency. An additional time delay relay t is used with the underfrequency device in order to obtain the enrrent time delay settings utilized for this trip condition. He use ofindividual relays in the new EPA units is considered to be more reliable based on the failures experienced by the GE logic cards over the history of the plant. He use of the relays will also improve the maintenance of the new EPA units. Summary of Safety Evaluation: ne RPS power distribution system does not perform a safety related function. Loss of the RPS power system is considered fail safe because loss of one division will initiate a half-scram which will place the plant in a position for the safety trip function to initiate, if required. Loss of the power supply will not prevent a safety function from occurring. He new EPAs use a manual reset molded case circuit breaker like the existing units. He existing testing prosisions will be applied
' to the new EPA units as well. De new EPAs are designated as Class 1E like the existing units.
he use of a dedicated timing relay that is qualified for this application is not considered to increase the probability of a malfunction of any EPA unit. He solid sate relays in the new EPAs
. used to sense abnormal voltage and frequency conditions are Class IE qualified relays that are Section i Page 37 of 187
- - . _ - . - - - _ - . - _ ~ - - - - - ~ - - . ~ -
P t certified by the original manufacturer. Wrefore, this modification does not increase the [ probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. h replec~= ant of the EPA breakers will not change the fail safe failure mode of the ! RPS. The RPS safety systems will continue to perform their required functions to place the plant - l in a safe condition and minumze the release of rmaation to the public. Herefore, this modification j does not increase the consequences of an accident or a malfunction of equipment important to -l safety previously evaluated in the SAR. h new EPA enclosures have louvers to provide j ventilation for the car.persts inside. He EPAs are not located in an area where general area air j flow is restricted. The new EPAs will use an internal 48Vdc power supply to provide control i power to operate the sensing relays and trip the breaker. De existing EPA units utilize a 12Vdc internally generated power source for the same functions. The loss of the 48Vdc control power will not cause the relays to change state, but the undervoltage trip function of the undervoltage coil on the molded case breaker will cause the breaker to trip when the voltage supplied to the coil ! reaches a minimum value. If 24Vdc control power is lost to the auxiliary time delay relay used for the added time delay on the underfrequency trip, the relay will de-energize after a time delay to trip j the EPA unit. His will caum: the EPA unit to fail in a safe position. The malfunction of the new j EPA units will have the same impact as the malfunction of the existing EPA units. De new EPA , units meet all the existing setpoints and tolerance bands for all parameters. Wrefore, this : modification does not reduce the margin of safety as defined in the basis of any technical : specification, For these reasons, this modification does not constitute an unresiewed safety [ question. j l Channe Number /USAR Section: LCN 7.3-135 Figures: 7.2-1 Sheet 3 ; 7.3-1 Sheet 1 , 7.3-2 Sheets 2 l 3 7.3-2 Sheet 4 7.3-2 Sheet 7A L I 7.7-4 Sheet 1 Descriptica and Basis for Channe: - i The proposed modification will add signals from the transmitters to ERIS. This will allow j enhanced monitoring of the reactor water level, reactor pressure, and reactor steam dome pressure. ! The change will enable the station to better determine the root cause of any future reactor scrams i or other reactor vessel level and/or pressure anomalies. l 4 i Summary of Safety Evaluation: ! - r t The transmitters and electronic trip units do not cause an accident identified in the SAR but are ! part of the instrument loops used to mitigate the consequences of an accident. Wrefore, the l 1 addition of the ERIS signal from the transmitters will not increase the probability of an accident , 1 previously evaluated in the SAR. The Class IE signals from the trip units associated with the i transmitters listed above are isolated from the non-divisional circuits of ERIS sia optical isolators.
. De optical isolators are part of the ERIS Data Acquisition System (DAS) Remote input . !
- Modules (RIM). The RIMS are qualified for safety-related applications. W optical isolators l prevent propagation of faults, short-circuits or grounding the signal within ERIS into the Class IE
, level or pressure instrument loops. Herefore, the instrument loops will continue their safety ; related functions and the consequences of an accident presiously evaluated in the SAR are j unchanged. The addition of the ERIS points from the transmitters are consistent with the ! configuration of the existing ERIS points. Therefore, the possibility of an accident which is l Section i Page 38 of187 l i f
-- - _ -~ . ,.
. - - . _ . - - - - - - . ~ . - - . - . ,
i l l l
;- i ;- t i di& rent from any previously evaluated in the SAR is unchanged. He Class 1E input signals are l 1
isolated from the non-divisional ERIS. His prevents malfunction in ERIS from propagating into l 4 the Class lE circuit and causing a malfunction in the safety-related instrument loops. Herefore, l ~ 4 the probability of a malfunction of the safety-related instmment loops associated with the i transmitters is not increased. He consequences of a malfunction of the safety-related instrument ! loops are not increased. The ability of the existing transmitters and their associated instmment l loops to perform their safety function will not be changed by this minor modification. Therefore, i the consequences of a malfunction of the safety-related instrument loops are not increased. The j i possibility of a malfunction of a safety-related structure, system, or component will not be created [ any different than that already stated in the SAR. De isolation between the existing safety-related j instmment loops and the non divisional ERIS equipment will ensure that the function of the ; j existing safety-related instrumentation will not be degraded. Therefore, the margin of safety as j j defined in the basis to the Technical Specifications wili not be reduced. For these reasons, this i modification does not constitute an unreviewed safety question. ; m-I i
- Channe Number /USAR Section
- LCN 7.3-136 Figure 7.3-2 Sheet 1 l !
) Description and Basis for Channe: This modification changes the open torque switch bypass switch (OTSBS) from a 10% setting to a l 50% setting on Valves 1033*MOVF001 and IG33-MOVF004. This is due to crud build-up that ! i is believed to have extended the time that the maximum unseating force existed after the OTSBS ! j opened This change ensures the effects of unscating and di&rential pressure have subsided. l
- i
! Summary of Safety Evaluation: ! v . l The safety function of the subject valves is to close to provide containment isolation. They do not 7 have a safety function to open. Increasing the OTSBS setting will ensure that the unscating and 1 i any di&rential pressure effect will not interfere with opening the valve. Bypassing the open torque j switch for approximately 50% of the valve travel will not a&ct the operability of the subject j valves to perfonn their safety function as long as each valve is maintamed in a satisfactory l condition. Bypassing a gate valve for approximately 50% ofits travel will ensure that both ! unseating and differential pressure forces will not inadvertently de-energize the open motor starter. ! His modification has no impact on any of the initiating events for any of the accident scenarios !~ evaluated in the SAR. Therefore, this modification does not increase the probability of an accident ! or a malfunction of equipment important to safety previously evaluated in the SAR. He subject i valves will remain capable of performing their safety related function. To ensure that static
- friction coefficients and scaling forces between the disc and seat surfaces do not interfere with the l reliability of the subject valves, an increase in the OTSBS setting is required. This modification will not change, degrade, or prevent actions described or assamed in any accident scenario L discussed in the SAR. The consequences of a malfunction of the subject valves will remain the
- same. Herefore, this modification does not increase the consequences of an accident or a
? malfunction of equipment impo: tant to safety previously evaluated in the SAR. Increasing the
- OTSBS setting will improve the operational reliability of the subject valves. His modification i
' does not involve any di& rent or new type of system, component, or functional requirement. This
- modification will not prevent the subject valves from performing any function that it is capable of, nor does it contradict its original design basis. Increasing the OTSBS setting for the subject valves
- - will not create any mode that could result in a malfunction of equipment important to safety.
Therefore, this modification does not create the possibility of an accident or a malfunction of l
- Section 1 Page 39 ofI87 1 '
s v , . . - - -. - m. r--,, ,--u , * , - - . _ _ _ _ - - - - - - -
Y j l' ' equipment important to safety different from'any previously evaluated in the SAR. His change
- does not alter the design bases, functions, or operations of any equipment required for safe j
; shutdown and does not adversely affect any other equipment important to safety. Both )
- IG33*MOVF001 and IG33*MOVF004 are capable of performing their design basis function j without exceeding any actuator, valve, or motor design parameters, Derefore, this modification )
does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. j l Channe Number /USAR Section: LCN 7.3137 Page 9.2-65 f Table 8.3-7 Sheet 2 j Figures: 7.3-14 Sheet 1 ; 7.3-14 Sheet 2 : 7.3-14 Sheet 3 i 7.3-14 Sheet 4 I 7.3-14 Sheet 6 ; 7.314 Sheet 12 i l 7.3-14 Sheet 13 l
- i. 7.3-14 Sheet 14 -!
- 7.3-14 Sheet 15
- i 1
Description and Basis for Channe: ! i 1 This modification installed a new data acquisition system (DAS) to monito: the status of the l control building ventilation chilled water (HVK) chiller safety controls, to monitor heating, j ventilation, and air conditioning (HVAC) air flow, and to record the operating history of chilled ! water pumps and chiller condenser pumps. His required the installation of a computer in the ! ERIS Computer Room, 68 field inputs, and several fuses to isolate the system from Class-1E l circuits. Four new safety related, wall mounted panels were also installed to house the fuses. The l
- . information provided by the new DAS will be used to aid Operations, System Engineering, and l Maintenance in trouble-shooting chiller trips. In addition, recorded pump operating history will aid ,
l in developing maintenance requirements, and obtaining pump history and trending data ; ' Summary of Safety Evaluation l l i
- _ This modification does not cause the control building chilled water system to operate outside any l i design or testing limits, or change the accuracy or response characteristics of any equipment l important to safety. Herefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This modification does not affect any fission product barriers and does not affect any equipment which l plays a direct or indirect role in mitigating the radiological consequences of any accident described
- in the SAR. He single failure criterion is still met as required. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety.
Electrical isolation between the DAS and the Class lE field input circuits is accomplished by installation of two fuses in series in each circuit. This will prevent a malfunction in the DAS from i
- propagating into the Class 1E circuits. De four new panels added by this modification provide physical separation between divisions and between disisional and non-disisional wiring and equipment. Therefore, this modification does not create an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. His modification is not
, , affected by and does not affect any technical specification or the basis of any technical !
. Section 1 - Page 40 of 187 i
specification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chanee Number /USAR Section: LCN 7.3-139 Figures: 7.3-1, Sheet 1 7.3-2, Sheet 7A 7.3-4, Sheet 1 7.7-4, Sheet 3 7.7-4, Sheet 5 7.7-4, Sheet 6 Description and Basis for Channe: This modification provides additional signals to the emergency response information system (ERIS) to allow for enhanced mwitoring of various parameters including reactor water levels, restor pressures, reactor steam dome pressure and several others to enable the station to better determine the root cause of any future reactor herams or other reactor vessel level and/or pressure anomalies. This modification also adds four computer points to the performance monitoring [ system (PMS) computer so that the reactor steam dome pressure and recirculation pump suction ; temperature can be monitored. Revision 3 of this modification provides for installation of a new conduit between power generation control complex cabinets for future ERIS points and disables four of the LFMG/ERIS points installed per the original modification. Syinraarv of Safety Evaluation: The instruments and electronic trip units associated with this modification cannot cause an accident identified in the SAR but are part of the instrument loops used to mitigate the consequences of an l accident. The equipment instruments (pressure transmitters and RTDs) are non-safety-related with all wiring adequately separated from safety related wiring. There are no interfaces with any other l equipment imponant to safety. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The j changes implemented by this modification only provide ERIS and PMS monitoring points. There i is no change to the function or operation of any equipment monitored by these ERIS points or the ; PMS points. Here are no interfaces with any other equipment important to safety. The new I cables in the control room are routed in conduits and maintain the required separation from other divisional and nondivisional cables. The conduits are seismically mounted so as not to have an ; adverse impact on an any safety-related S5C. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. The isolation is identical to the existing isolation between the Class IE signals and the nondivisional ERIS. These instruments are non-safety-related. The ability of the l existing instruments and their associated instrument loops to perform their safety function will not l be changed by this modification nor does it change the control or operation of any equipment imponant to safety. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment imponant to safety different from any previously evaluated in the SAR. The isolation between the existing safety-related instrument loops and the nondivisional ERIS equipment will ensure that the function of the existing safety related instrumentation will not be degraded. His modification does not change any setpoints or process parameters. Therefore, the margin of safety as defined in the basis to the Technical Specifications will not be reduced. Therefore, this modification does not reduce the margin of safety as defined in the basis of any Section 1 Page 41 of I87
- - - . - - ~ , , . . . . . - . - - .- - -- - - - -. . ,- , b i
technical specification. For these reasons, this modification does not constitute an unreviewed
; safety question.
Channe Number /USAR Section: LCN 7.3-140 ; Pages: 7.3-8 , [ 7.3 11 l Figures: 7.3-3 Sheet 1 - 7.3-4 Sheet 1 5 Description and Basis for Channe: ; i' ! p This modification will change the control logic for the low pressure core spray (LPCS) and low ! j pressure core injection (LPCI) injection valves. During anticipated transit without scram (A'IWS) ! conditions in the level / power control of emergency operating procedures (EOPs), the operator is { directed to terminate and prevent LPCS and LPCI injection to the reactor pressure vessel in part by l
, disabling the automatic opening logic of system injection valves. The disabling of these circuits is j currently accomplished by the physical removal of electrical relays from the control circuit of each !
valve. De current method of preventing automatic opemng of the injection valve is to physically remove relays from the control circuits. This action has proved impractical and, at times, could be l i untimely. In order to eliminate the problems associated with the removal of the relay, this I j modification will change the control circuit to allow the operators to manually lockout the valve auto open feature prior to actual valve opening by closing switch S2 after the LPCS initiation and
- power available interlocks are established. It will also inhibit potential reactor pressure vessel j
'(RPV) pressure transients from breaking the lockout scal-in. The relay contacts in the control circuits of the LPCI injection valves will be similarly modified.
i
- Summary of Safety Evaluation:
nis modification will not adversely affect the ability of the operator to manually control the l position of the injection valve subsequent to a system initiation signal. For accident analysis ; e purposes. the SAR considered the LPCI system to be an operating mode of the RHR system. l
- Changes to the operations of the injection valves are not part of the RHR shutdown cooling system. !
The control circuit change on these valves does not interact with any RHR shutdown cooling mode l l component or operating logic. Therefore, the circuit changes made by this modification will not j
- affect the probability of these type accidents. This modification which involves a change in the !
l operating logic for the LPCl/LPCS injection valves, will not impact any of the existing mechanical j properties (flow, pressure, fluid density, temperature, etc.) of the LPCS or LPCI systems. This ! E change will not add, alter, or remove any component of any piping system designed to protect the ; public from radiological hazards. As such, this modification will not affect the probability of a l loss of coolant accident (LOCA) as addressed in the S AR. Therefore, this modification does not i increase the probability of an accident previour!v evaluated in the SAR. Changes to the control logic for LPCS AND LPCI will not alter, change or degrade the fission product barriers designed
- to protect the public from radiological hazards The LPCS AND LPCI systems are part of the .
emergency core cooling system (ECCS) which is designed to mitigate the effects of a LOCA. The ! changes made to the LPCS AND LPCI injection valve control logic will not affect the automatic l operation of the ECCS as currently defined in the SAR. This modification will not change what is !
- required by the EOPs (preventing automatic opening of the injection valves), only the actions taken by the operator to accomplish this goal will be resised. This modification will enhance the ability of the operator to mitigate the nTects of an ATWS event in a timely manner. Therefore, this :
modification does not increase the consequences of an accident presiously evaluated in the SAR. { Section l' Page 42 of I87 j i i
The accidents created by malfunction of these systems have been previously evaluated in the SAR.
%c wiring materials and installation methods used to perform this modification are equal in quality to the existing wiring installations. Herefore, there are no new failure scenarios attributable to this change. No new systems or radiological sources that could create an accident of a different type than previously analyzed will be created by this modification. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. & LPCS AND LPCI systems, as modified by this change, will continue to automatically start the associated pumps, line up the required valves, and inject water into the RPV upon receipt of a system initiation signal. The operator will retain manual control of the injection valves after manual controls have been initiated regardless of the status of the auto open RPV pressure interlock. Therefore, this modification does not increase the probability of a 'nalfunction of equipment important to safety previously evahuted in the SAR.
I Changing the control logic for LPCS AND LPCI injection valves will not add any new radiologicsl sources to the plant or add, alter, or remove any existing radiological barriers designed ta protect tne public. Failure of a LPCS AND LPCI system has been addressed in the SAR and its rad.iological consequences evaluated. The consequence of a failure of an injection valve control circuit due to this modification would be no different than the failure of any other critical system components that would result in a system malfunction. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The change to the control logic for LPCS AND LPCI system injection valves will allow continuation of the automatically start furctions of the system as described above. This modification is consistent with the mau%ments of SAR Section 15.8, " Anticipated Transients Without Scram (ABVS)." He plant licensing basis will not be altered nor will the operation of required safety systems be degraded by this design change. No equipment is moved or relocated to any area where new or increased hazards exist. The new wiring to be installed will be equal in quality to the wire currently in use in the circuits. Installation and inspection of the revised wiring will be in accordance with existing plant requirements for Class I E equipment. Herefore, this modification does not create a malfunction of equipment to safety different from any presiously evaluated in the SAR. This modification will not change the operating characteristics of the initiating sequence of events for automatic opening of the LPCI/LPCS injection valves. The availability of the LPCI/LPCS systems to perform their intended functions will not be adversely affected by this change. No change to any margin of safety applicable to the LPC1/LPCS systems was identified. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Section I Page 43 of 187
I l Chanee Number /USAR Section: LCN 7.3-141 Figures: 7.3-21 Sheet 3 I 9.4-2b l 1 Description and Basis for Charee: 1 This modification replaces the roll filter timer associated with the fuel building air supply heating, ventilation, and air conditioning (HVAC) unit. He HVAC unit roll filter is designed to keep a fresh piece of filter curtain over the air intake. This is achieved by one of three methods: a timer that will initiate the roll filter to automatically start after a set period of time, a differential pressure switch that will start the roll filter when there is high differential pressure across the filter media, cr a manual switch. The existing timer does not have a sufficient time adjustment to permit maximizing the efficient use of the filter curtain. This causes premature run-out of the filter curtain which results in nuisance alarms in the control room. The intent of this modification is to replace the existing timer control unit with one that will provide a more flexible time-out adjustment and to optimize the filter media. Summary of Safety Evaluation: None of the accidents evaluated in the SAR have any information or description of the fuel building air supply or the HVAC unit roll filter timer. The fuel building roll filter timer is a non-safety-related component and the changes being made by this modification do not impact the original system design intent. The replacement of the roll filter timer does not adversely affect any equipment important to safety. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Since the roll filter timer and the portion of the fuel building air handling system impacted by this modification ac non-safety-related, any failure of the timer or support materials that are being modified by this modification vvili have no etTect on any of the accidents described in the SAR. This modification will therefoie have no a%t on the consequences of any accident described in the SAR and will have no effect on the off-site doses previously evaluated in the SAR. Therefore, this modification does not increwhe consequences of an accident or a malfunction of equipment important to safety prev;ously evaluated in the SAR. None of the activities resulting from implementution of this modification will create the possibility of an accident or a malfunction of equipment important to safety that is different from any previously evaluated in the SAR. This modification does not impact any technical specification. Therefore, this modification does not reduce the margin of safety as defined by the basis of any technical specification. For these reasonc, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: MR 95-0023 and LCN 7.3-142 Figures: 7.3-4 Sheet 4 7.4-1 Sheet 5 Description and Basis for Chance: This modification installs de-dc power supplies to replai e five Topaz inverters and seven GE power supplies presently operating in various panels in the main control room and control building. The existing power supplies and inverters will be removed. The Topaz inverters are obsolete and there are shelflife and reliability concerns associated with ti.cm. The de-dc power supplies ofter less heat, improved performance, small size, and reduced maintenance. Also two annunciators are renamed to clarify their function as a result of this modification. Section 1 Page 44 of 187
J ', i
^
l 1 ! l' Sunimary of Safety Evaluation: i j - The new de-dc power supply assemblies will be qualified for this Class IE application. The new j j power supplies will be mounted the same as the existing power supplies All ternunal boards and j l de relays are to be seismically mounted and are qualified for use in this Class 1E application. The j { new assemblies will be tested to ensure proper operation under maximum control room ! j temperatures that may occur with a station blackout event. The new power supply assemblics will offer less heat, improved electrical performance, smaller size, and reduced maintenance The new j j power supply assemblies therefore increase the reliability of the power supply for the affected !
- circuits. De functions of the existing Class IE circuits which are affected by this modification j
- remain unchanged. De new power supply assemblies are as a minimum specified to meet or ;
i exceed the performance requirements of the existing power supplies. The new power supply ; j assemblics are seismically qualified components and are designed to handle the expected electrical l
; transients that may impact the new assemblies. Each of the new power supply assemblies will bc l 4 qualified for the life of the plant. Therefore, this modification does not increase the probability of j
' an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. l
- The power supply configuration that is being modified is not mentioned in the accident evaluations l l for any system. Failure of a single power supply assembly will be the same as the loss of one of !
J the existing power supply configurations. The coracquences of a loss of each of the affected power ! supply configurations have been addressed by plant procedures. The affected safety systems will i continue to perform their required functions to place the plant in a safety condition and minimize : ! the release of radiation to the public. There have been special tests specified for the new power ! j- supply assemblics to ensure that anticipated transients can be adequately suppressed to ensure the i uninterrupted power. Therefore, this modification does not increase the consequences of an 4 accident or a malfunction of equipment imponant to safety previously evaluated in the SAR. The l i power supply assemblies contain no moving parts and will utilize solid-state components. Any { l unique tests required to suppon the Class lE qualification will be performed prior to receipt from j l . the vendor. New de annunciator relays will be used to replace ac annunciator relays because the ac , portion of the power supply circuit will be deleted. No new cables or conduits are to be installed. j ~ The engraving of two annunciator windows will be changed. The function of the annunciators will l; i- remain the same and the corresponding operator actions will be the same. The input to these new power supply assemblics will use the same input fuses currently installed for the existing inverters : 4 and power supplies. The new power supply assemblics will be designed for a wider range ofinput ;
- voltages than the existing inveners to preclude the kind of trips that have occurred during r equalization charges of the batteries. The replacement of existing power supplies will not affect !
the function of the existing Class IE circuits that will be powered from these new desices. ! Therefore, this modification does not create the possibility of an accident or a malfunction of i equipment important to safety d'frerent from any previously evaluated in the SAR. The Technical i Specifications and improved Tee.hnical Specifications do not specifically address any area of the ! design affected by this modifica tion. Requirements for the instrumentation powered from these . i new devices are not affected. The only two inverters addressed by the improved Technical I i Specifications and the improvut Technical Specif' cations Bases are not affected by this
- modification. Therefore, this modifi= tion does not reduce the margin of safety as defined in any ,
technical specification or the be. sis of any technical specificction. For these reasons, this i modification does not constnute an unreviewed safety queraon. t i w hn1- Page 45 of187 j 4 a .. , ' -
, Channe Number /USAR Section: LCN 7.3-144 Pages: 8.9-90 8.9-91 Figure 7.3-23 Sheet 13 Description and Basis for Channe:
The 125Vdc normal battery charger, located in the circulating water switchgear building, had failed and required replacement. The charger and its associated alarms are mentioned in several places in the SAR. NUREG 0800 and Regulatory Guide 1.70 do not require this level of detail to be included in the SAR. The alarm logic diagram for battery chargers BYS-CHGR1C and BYS-CHGR04 is being removed on this basis. The description of the individual alarms associated with these battery chargers is to be deleted also. , Summary of Safety Evaluation: The battery chargers affected by this change are non safety related pieces of equipment that are not supplied by a safety related power source. These chargers also do not supply any safety related loads. The loss of these chargers would have no impact on any accident. This change will in no way affect the function of these chargers or their associated alarms. This change represents no modification to the plant. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR. Based on their power sources, these chargers are assumed to fail during an accident. Therefore, this change does not increase the consequences of any accident previously evaluated in the SAR. The battery chargers and their alarms will continue to function in the same manner and will be governed by the same procedures. Therefore, this change does not create the possibility of an accident different from any previously evaluated in the SAR. All alarms involved in this change will still annunciate in the main control room as part of a common alarm associated with their respective buildings or functions. They have no affect on any safety related structures, systems, or components (SSC). Therefore, this change does not increase the probability of a malfunction of safety related SSC previously evaluated in the SAR. This change does not increase the consequences of a malfunction of any safety related SSC presiously eva!uated in the SAR. This change does not create the possibility of a malfunction of a safety related SSC different from any previously evaluated in the SAR. These battery chargers and their alarms are not addressed in any technical specification. None of the power supplies used by these components are described in any technical specification. Therefore, this change does not reduce the margin of safety defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed nfety question. Change Number /USAR Section: LCN 7.3-147 Figures 7.3-3 Sheet 2 7.3-4 Sheet 2 Description and Basis for Change: Three low pressure coolant injection (LPCI) check valves and one low pressure core spray (LPCS) . check valve are modified to delete all electrical controls and actuator indication. A kit provided by Atwood-Morrill will be installed to climinate the packing / stuffing box arrangement on the actuator side of each valve to minimize leaks which contribute to unidentified drywell leakage.
. Section I Page 46 of 187
1
)
e I i i Summary of Safety Evaluation: ] l i
! De modified valves are classified as containment isolation, pressure isolation, and emergency core ,
i
- cooling system (ECCS) injection valves. He air testable portions of these valves perform no l i safety related functions. The air testable components were installed to meet the ASME Section XI
- inservice test requirements. The ASME Section XI requirements are met in accordance with Pump
; and Valve In-service Testing Plan Valve Request for Relief No. I by manually stroking these j valves during cold shutdown. He removal of the remote air testable portions of the valve do not j j impact the functions of these valves. The electrical portions affected by this modification are not -
postulated to cause any accidents described in the SAR and do not have any impact on any
. equipment important to safety. He design change does not alter the design requirements or operation of the valves nor do they adversely affect any equipment important to safety. Therefore, j j this modification does not increase the probability of an accident or a malfunction of equipment to ,
i safety previously evaluated in the SAR. This modification does not affect the offsite dose to the l public. Therefore, this modification does not increase the consequences of an accident or a r malfunction of equipment important to safety previously evaluated in the SAR. Here are no new i failure modes, no new system interfaces, and no new material types associated with the final check l . valve configuration. The removal of me electrical components will not cause any new types of ~ failures or possibilities of an accident. No new or different modes are added and no new or ! , different system interfaces are added that could result in the possibility of a malfunction of : equipment important to safety of a different type than previously evaluated. The check valve ;
- configuration will remain a swing check valve that has less friction and no interference problems. l Therefore, this modification doel not increase the consequences of an accident or a malfunction of j equipment important to safety previously evaluated in the SAR. This modification does not impact
- the operability, function or performance requirements of the LPCS system. The modified valves are not functionally or operationally impacted and will be required to meet the same surveillance .
requirements. Valves which prevent the leakage of primary coolant outside containment are , e required to be subject to preventive maintenance, periodic visual inspection, and integrated leak i i rate testing. An inservice testing program is to be established. The LPCI and LPCS valves will
- still be governed by these requirements. These requirements are unaffected by this modification.
Therefore, this modification does not reduce the margin of safety as defined in any technical specification or the basis of any technical specification. For these reasons, this modification does ; i - not constitute an unreviewed safety question. t Channe Number /USAR Section: LCN 7.4-20 Page 5.4 28 Table 6.2-35 '
'7.5-1
, 7.5-2 Figure 7.3-1 , i
- 7.4-1 4
Description and Basis for Channe: > MR 94-0127 removes instmmentation and valve disc position components from the reactor core !
- isolation cooling (RCIC) injection valves 1E51*AOVF065 & 1E51*AOVF066 and high pressure ,
core spray (HPCS) injection valve E22*AOVF005. The actuators used to test the valves were i previously disconnected. The associated control room indication will also be removed in L Section 1 Page 47 of187 !
accordance with current human engineering practices. The design of these valves does not require position indication or the ability to remotely cycle the valve in response to any event. Additional changes to the valve shaft, bearing and packing will allow the valve to reach the closed position more easily thereby enhancing the reliability ofisolation. Summary of Safety Evaluation: The function of the valves remains as designed. This function is to allow injection into the reactor vessel from IIPCS or RCIC and to isolate the primary containment. The removal of the position indication and test actuators does not prevent the valves from performing the design functions. Alternate means remain in the Technical Specifications to verify the valve operability including the insenice flow test (for opening) and the 10CFR50 Appendix J local leak rate test (for closure). The changes to the shaft and bearing assemblics in this modification will not change the flow and pressure drop characteristics of the valves and the ability of the valve to close will be increased. Therefore, this modification does not increase the probability of an accident or a malf,nction of equipment important to safety previously evaluated in the SAR. The valves continue to have the , same flow characteristics and ability to open for injectico to the vesse! and to close as required for j containment isolation. The modification will not change a functional or operational abihty from original design. The valves continue to be in conformance with the sune codes and standards of original design. The valves will still be required to pass existing leakage and operability testing requirements currently specified. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The valves remain qualified as currently required by the SAR requirements. Sun'eillance requirements in the Technical Specifications will be continued. No new failure modes, system interfaces or material types have been mtroduced and all stresses are below ASME allowables. He electrical components removed will not cause any new types of failures. The system configuration remains as evaluated in the SAR. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment imponant to safety different from any ' previously evaluated in the SAR. Technical Specifications require valve leak testing per ASME code and 10CFR50 Appendix A and J, however, these requirements are unaffected by this change. Other requirements applied to these valves are unaffected by this change. Herefore, this . modification does not reduce the margin of safety as defined ia the basis of any technical I specification. For these reasons, this modification does not constitute an unrniewed safety question. 1 l Chance Number /USAR Section: LCN 7.6-39 Table 8.3-7 Sheet 2a Figure 7.6-1 Sheet 2 Description and Basis for Chance: This modification replaced two recorders in the leak detection system. The two recorders were I suffering repeated failures due to age and the outdated technology used in their design. I Summary of Safety Evaluation: Measures were taken to ensure that all potential failures were analyzed and appropriately considered. Testing and analysis had been perfonned on these recorders and appropriate design measures were implemented to ensure that the essential power source and nearby cables will not be Section I Page 48 of 187
Y 1 I impacted or degraded by this modification. Herefore, this modification did not increase the probability of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. This modification did not reduce the effectiveness of any safety related system to mitigate the consequences of postulated accidents and transient events. Therefore, this modification did not increase the consequences of an accident or a malfunction of equipment important to safety. Appropriate measures were taken during the design phase to ensure that no new accident scenarios were created and to ensure that no malfunction of the recorders would propagate to any safety related equipment. Therefore, this modification did not create the possibility of an accident or a malfunction of equipment important to safety different from any presiously evaluated in the SAR. His modification did not affect any technical specification and did not affect the basis of any technical specification. Herefore, this modification did not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification did not constitute an unresiewed safety question. Channe Number /USAR Section: LCN 7.7-035 Figure 7.7-4 Description and Basis for Channe: The nomenclatures on the Alarm Numbers 2094,2103,2108, and 2110 in the panel P680-04A in the main control room do not reficct the actual operating conditions. The Alann Number 2094 (2103) "RECIRC MOTOR A (B) LOCKOUT UNDERVOLTAGE"is initiated by the electric current actuated protective relays which has no relation to undenoltage. This alarm would be accurate ifit is read as "RECIRC MOTOR A (B) LOCKOtflTfRIP". He Alarm Number 2108 (2110) "RECIRC MOTOR A (B) OVERLOAD / LOCKOUT"is initiated by numerous desices, including the trip control switch which has no relation to an " OVERLOAD / LOCKOUT." his alarm would be accurate ifit is read as "RECIRC MOTOR A (B) TRIP" By reviewing the i design documents, there is no need for undervoltage protections on the recirculation pump motors, only overcurrent protection is appropriate. (Ref. ESk-5RCS04, SRCS06, SRCS07 sheet 2, i 5RCS08 sheet 2, SNPS25 sheet I and sheet 2, CR 95-0933). Summary of Safety Evaluation: These alarms are provided to operators for the purpose ofindicating the condition of the reactor recirculation pump motors. Changing the nomenclature on these alarms to reflect the actual plant operating condition does not change the function of the reactor recirculation pump or system. Therefore, this change does not increase th: probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Also this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR or reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an
.. unreviewed safety question.
Section 1 - Page 49 of187
t i ! i ! i l Channe Number /USAR Section: LCN 7.7-36 Pages: 7.7-34 ! 7.7-36 ! I' ! ! Description and Basis for Channe: j i His modification replaced the outdated Pl core performance calculation computer program with j
, the 3D Monicore program A new computer system was also added to run the 3D Monicore i ! program. Wiring and cables were installed to make the 3D Monicore program accessible to the l main control room: Dese new components allowed the outdated P1 code to be removed from the !
process computer, j ; Summary of Safety Evaluation: l
; 1 i The 3d Monicore program does not perform any automatic operation or control any equipment in ! the plant. De new computer program only provides calculational information to the operator.
4 The only changes made to the plant were the removal of the old computer program from the , t
' process computer, addition of a new computer system, addition of new software, and installation of l wires and cables in the main control room. No other plant equipment was affected by this i j modification. He 3D Monicore program was validated through an extensive test program as l desenbod in plant procedures. The 3D Monicore program met or exceeded all the functions of the j i P1 program. Therefore, this modification did not increase the probability of an accident or a i malfunction of any equipment important to safety previously evaluated in the SAR. Since the 3D l Monicore program is no worse than the existing computer program, the consequences of any ;
4 accident will be no worse than if the P1 program was still being used. Therefore, this modification ! did not increase the consequences of an accident or a malfunction of equipment important to safety j l previously evaluated in the SAR. The installation of the new 3D Monicore software and j ._ associated components did not create any new operator actions and did not affect any automatic j equipment actions. Therefore, this modification did not create the possibility of an accident or a - l malfunction of equipment important to safety different from any previously evaluated in the SAR. , i- This modification did not affect any technical specification and did not affect the basis of any l
- technical specification. Therefore, this modification did not reduce the margin of safety as defined ;
in the basis of any technical specification. For these reasons, this modification did not constitute ; an unreviewed safety question. j i ! t Channe Number /USAR Section: LCN 7.7-037 ' Figure 7.7-6 0 ' Description and Basis for Channe: j This modification provides additional signals to the ERIS System to allow for enhanced monitoring of various parameters including reactor water levels, reactor pressures, reactor steam dome pressure and several others to enable the station to better determine the root cause of any future j mactor scrams or other reactor vessel level and/or pressure anomalies. This change also adds a spare cable between two panels in the main control room for future usage. ! ,' i l Summary of Safety Evaluation: l The instruments and electronic trip units associated with this modification cannot cause an accident
, identified in the SAR; The instruments are non-safety-related with all wiring adequately separated j
'l Section 1 Page 50 of187 ;
I _ - .)
i
- i from safety related wiring. The ERIS points added by this modification perform a passive function {
4 and do not provide any manual controls or automatic actuations of plant equipment. There is no 1 L change to the function or operation of any equipment monitored by these ERIS points or the PMS points. Here are no interfaces with any other equipment important to safety. Therefore, this ; modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. He changes implemented by this : modification only provide ERIS monitoring points. There is no change to the function or operation I of any equipment monitored by these ERIS points. There are no interfaces with any other ! equipment important to safety. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The addition of the ERIS points from the instruments is consistent with the configuration of , existing ERIS points. The instruments and their associated safety-related instrumentation are unchanged by the addition of the ERIS signals. The isolation is identical to the existing isolation i between the Class 1E signals and the non-divisional ERIS. The new analog input modules are l identical to the existing ones. Dese instnunents are non-safety-related. The ability of the existing l instruments and their associated instrument loops to perform their safety function will not be l changed by this modification nor does it change the control or operation of any equipment l important to safety. Therefore, this modification does not create the possibility of an accident or a ) malfunction of equipment important to safety different from any presiously evaluated in the SAR. { The isolation between the existing safety-related instrument loops and the non-disisional ERIS equipment will ensure that the function of the existing safety-related instrumentation will not be ; degraded This modification does not change any setpoints or process parameters. Therefore, the l margin of safety as defined in the basis to the Technical Specifications will not be reduced. ; Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Channe Number /USAR Section: LCN 7.7-038 Figure 7.7-6 Description and Basis for Channe: This modification provides additional signals to the ERIS System to allow for enhanced monitoring of various parameters including reactor water levels, reactor presseres, reactor steam dome pressure and several others to enable the station to better determine the root cause of any future , reactor scrams or other reactor vessel level and/or pressure anomalies. This change also adds four l computer points to the Performance Monitoring System (PMS) computer so that the reactor steam dome pressure and recirculation pump suction temperature can be monitored. Summary of Safety Evaluation: De instruments and electronic trip units associated with this modification cannot cause an accident identified in the SAR. The instruments are non-safety-related with all wiring adequately separated I from safety related wiring. He ERIS and PMS points added by this modification perform a ] passive function and do not provide any manual controls or automatic actuations of plant i equipment. The changes only provide ERIS and PMS monitoring points. There is no change to l the function or operation of any equipaent monitored by these ERIS points or the PMS points. There are no interfaces with any other equipment important to safety. Therefore, this modification does not ' increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The changes implemented by this modification enly proside Section 1 Page 51 of I87 __ _=
i ERIS and PMS monitoring points. There is no change to the function or operation of any j
.cquipment monitored by these ERIS points or the PMS points. There are no interfaces with any j other equipment important to safety. Therefore, this modification does not increase the !
consequences of an accident or a malfunction of equipment important to safety previously . evaluated in the SAR. The addition of the ERIS points from the instruments is consistent with the l configuration of existing ERIS points. The instruments and their associated safety-related j instrumentation are unchanged by the addition of the ERIS signals. The isolation is identical to the j existing isolation between the Class IE signals and the non-divisional ERIS. The new analog input l modules are identical to the existing ones. The modification will add PMS computer points from j the pressure transmitters for the reactor dome pressure and from the recirculation pump suction i temperature RTDs. These instruments are non-safety-related. The new PMS computer points are i for monitoring purposes only and do not provide any control function. The ability of the existing } instruments and their associated instrument loops to perform their safety function will not be ! changed by this modification nor does it change the control or operation of any equipment important to safety. Therefore, this modification does not create the possibility of an accident or a ! malfunction of equipment important to safety different from any previously evaluated in the SAR. l r The isolation between the existing safety-related instrument loops and the non-divisional ERIS equipment will ensure that the function of the existing safety-related instrumentation will not be : degraded. This modification does not change any setpoints or process parameters. Therefore, the margin of safety as defmed in the basis to the Technical Specifications will not be reduced. l Therefore, this modification does not reduce the margin of safety as defined in the basis of any - technical specification. For these reasons, this modification does not constitute an unreviewed { safety question. ; i I Channe Number /USAR Section: LCN 8.3-043 CRs 93-0342 & 92-0162A Pages 8.3-73 ; 8.3-74 Description and Basis for Channe: The technical change required is the exception to the use of cables with red or blue colored jackets in direct burial cable installations for applications where there are only non-safety-related circuits. In particular, when changes were being made to the sanitary waste and disposal system (wastdscwage treatment plant), a one time deviation from Specification 248.000 was issued to use { Category I color coded cable for a Category III application. J l Summary of Safety Evaluation: The waste / sewage treatment plant (WSTP) is a Category III non-safety-related system located outside the protected area. It is completely isolated from all safety-related systems; it's only connection to plant operations is through it's main power supply. Even there, the motor control center supplying the power has no other loads which are safety related. A resiew of the SAR indicates that the WSTP has no impact on any accident evaluated in the SAR. Therefore, this change does not increase the probability of an accident or a malfunction of equipment impostant to safety previously evaluated in the SAR. Also, for the same reasons, this change does not increase the consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR; The WSTP has no affect on the syst:m designed to maintain the integd!y of the physical barriers designed to contain radiation. It has no relationship to any safety-related system. Prior to this change, there was no mention of the WSTP in the SAR. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to Section I Page 52 ofI87
i
-l safety different from any previously evaluated in the SAR. De WSTP is not mentioned in the
. Technical Specifications. Derefore, this change does not reduce the margin of safety as defmed in .
the basis of any technical specification. For these reasons, this change does not constitute an l
- _ unreviewed safety question.
l i 1 Channe Number /USAR Section: LCN 8.3-49 Table 8.3-7 Sheet 2a ; - Description and Basis for Channe: a .
- His modification adds another power source for the contamment airlocks 1JRB*DRAl and IJRB'DRA2. In the current configuration, the airlocks are operated with a non-Class IE power ;
, supply. His modification establishes a Class 1E power source as a backup which will permit ' l
; . reliable power to be available under accident conditions should normal power fail. The Class IE i power source is isolated from the non-Class IE circuits by double fusing. He alternate power l source is to ensure the local airlock lights will be f-tianal post accident to enable operatmg l
- personnel to know condition of each door. As such, this is an enhan=nant to the system. Electric l
! ; .,wer to the airlock is not required to operate the airlocks, but is needed to ensure door conditions ;
- are known and to help airlock users make certain only one airlock door is opened at a time. -
< Sununary of Safety Evaluation: i
- During normal and accident conditions, this modification does not alter the originally decionad !
- operation of the contamment airlock. His modification does not reduce the reliability of the !
reactor building systems. The non-safety-related portion of the modification is adequately isolated l
- 3. from the safety related portion so that faults will not be propagated through the system and affect j the emergency diesel generators. De normal power requirements for the containment air-lock door !
circuits are insignificant relative to the other loads fed from the diesel generators and have been l . detennined to be within the diesel generators' load limits. Herefore, this modification does not j increase the probability of an accident or a malfunction of equipment important to safety ( previously evaluated in the SAR. The changes made by this modification will not reduce the ;
- cffectiveness of any engineered safety feature system to mitigate the consequences of postulated i lL accidents and transient events. All electrical components have been seismically mounted to preclude interaction with other safety related equipment in the area. nerefore, this modification l
- does not increase the consequences of an accident or a malfunction of equipment important to l safety previously evaluated in the SAR. This modification does not create any new accident i
, scenarios and does not create any new failure modes which result in an unanalyzed condition. Therefore, this modification does not create the possibility of an accident or a malfunction of ; 4 equipment important to safety different from any previously evaluated in the SAR. The electrical j portion of the containment door system affected by this modification is not necessary for the doors ! i to perform their safety function of providing contamment integrity. This modification has no . ! } impact on the diesel generators or the power supply to any Class 1E loads, and therefore, has no l L unpact on any technical specification. Herefore, this modification does not reduce the margin of. i
. safety as defmed in the basis of any technical specification. For these reasons, this' modification
.l
, does not constitute an unresiewed safety question. j l a L l
- Section I Page 53 of187 t
a'. Channe Number /USAR Sectioni LCN 8.3-54 Table 8.3-7 Sheet 2A Figures: 7.2-2 ) 4.6-5c 1 l l Description and Basis for Channe: l This modification replaced temperature recorder 1C11-TRR018. His modification also added a !
- new fuse panel IRDS*PNL1 to hold the fuses necessary to provide isolation from its safety related power supply, he new panel required the installation of two new safety related conduits. The recorder was replaced because the old recorder had a long history of repetitive failures and was i obsolete. l 1
Summary of Safety Evaluation: j 1 The non-safety-related control rod drive (CRD) temperature recorder, the power supply, and its related raceway were not postr. sated to cause any accident described in the SAR. Therefore, this
- modification did not increase se probability of an accident previously evaluated in the SAR. Since the recorder and its supporting electrical equipment were seismically mounted and electrically !
isolated, safety related equipment will not be affected. Therefore, this modification did not increase the probability of a malfunction of any safety related system, structure, or component (SSC) ; previously evaluated in the SAR. The CRD temperature recorder and its associated circuits were i not used for, or accredited with, any accident mitigating capabilities. Therefore, this modification did not increase the consequences of an accident or a malfunction of any safety related SSC , previously evaluated in the SAR. This modification did not create any new system interfaces and 1 did not add any new failure mechanisms. Failure of the new recorder will not result in a loss of any safety function or degrade any safety related SSC. Therefore, this modification did not create the possibility of an accident or a malfunction of any safety related SSC different from any previously evaluated in the S AR. The CRD temperature recorder, its associated circuits, and the l information provided were not addressed in any technical specification or in the basis of any l technical specification. Herefore, this modification did not reduce the margin of safety as defined ; in the basis of any technical specification. For these reasons, this modification did not constitute ; an unreviewed safety question. ; Channe Number /SAR Section: LCN 8.3-57 Page 8.3-21 l i Description and Basis for Channe: l 1 he reactor protection system (RPS) power system consists of two high inertia motor generator ] (MG) sets and distribution equipment. Normally, when the MG sets are started, the generators will j automatically build up voltage due to the residual magnetism in the exciter field. However, the vendor confirmed that the residual magnetism will decrease as the MG sets age. The application of an external source of power was required to flash the generator exciter field. His nxxiification installed a permanent field flash power supply to allow the RPS MG sets to build up voltage during start-up. The new field flash power supply consists of a rectifier that supplies SVdc to the exciter field upon depression of the " motor on" push-button (1PB). The new rectifier was field mounted inside the existing MG set control boxes. Existing push-button IPB was replaced with a new i push-button that provides the required normally open contact. ) l i Section 1 Page 54 of187
.\
Summary of Safety Evaluation: The RPS MG sets are the normal power supply for the reactor protection system trip logic system. However, this function requires that the MG sets are already in operation and at rated voltage and frequency before the RPS logic circuitry is energized. This modification only affects the start-up of the RPS MG sets. He start-up or coast-down of the MG sets is not relevant to the designed functionality of the reactor protection system power supplies, and further, is not addressed in the SAR. Therefore, this modification does not increase the probability of an occurrence of an accident previously evaluated in the SAR. The RPS trip logic system will continue to function as described in the SAR, so the consequences of a previously evaluated accident will not be increased. This modification only affected the RPS MG sets at start-up. The safety function of the RPS power supplies when the plant is in operation is not affected. Therefore, the possibility of an accident which is different than any evaluated previously in the SAR is not created. The RPS power supply is classified as non-essential because failure of this power supply causes a reactor scram and isolation. Changing the method in which the RPS MG sets are staned neither changes the way the RPS functions during operation of the plant nor affects the safe operation of the plant. Therefore, this modification neither increased the probability of a malfunction nor the consequences of a malfunction of any safety related structure, system, nor component (SSC) as previously evaluated in the SAR. This modification does not introduce any activity that affects the fail-safe design of the RP S power supplies. Therefore, the possibility of a malfunction of any safety related SSC different from what is evaluated in the SAR is not created. The RBS Technical Specifications and the Improved Technical Specdcations do not address any area of the design affected by this modification. Thus, the margin oisafety as defined in the basis for the RBS Technical Specifications is not reduced. For thes.. reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCF 8.3-58 Pages: 8.3-13 8.3-16 8.3-51 Table 8.3-6 Figures: 7.3-23 Sheet 1 7.3-23 Sheet 15 7.3-23 Sheet 17 7.3-23 Sheet 18 7.3-23 Sheet 21 8.3-3 Sheet 1 8.3-4 8.3-12 Sheet 2 8.3-12 Sheet 3A 8.3-12 Sheet 4 8.3-13 Lescription and Basis for Change: This modification will provide new synchronous check relays to each of the Division I, II, and III diesel generator synchronizing control circuits. Rese relays will be used to verify that the voltages on either side of the supply circuit breakers to the Division buses are in synchronism. This modification will prevent breaker manual closure if the two voltages are out of phase, but will not prevent breaker closing on a de-energized bus. This modification will also provide blown fuse Section I Page 55 of 187
i i indication for the closing circuits of the circuit breakers on the high pressure core spray (HPCS) 4.16 KV bus. This modification will add local indication lights and tie into a common trouble alarm for the switchgear. Summary of Safety Evaluation: The diesel generators are not postulated to cause any of the accidents described in the SAR. The additional loading of the potential transformer supplying the new synchronous check relays will have a negligible afTect on other components supplied by the potential transformers. A short in the new indicating lights will not result in a blown fuse and hence a loss of 125Vdc control power. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. During emergency conditions, the output relay contacts are bypassed for Divisions I and II, thus ensuring that the safety buses will be supplied by the emergency diesel generators even if the synchronous check relay fails. The Division ill diesel generator check contact is closed either by satisfactory synchronism or loss of power. In addition, the existing analysis assumes the loss of one diesel generator and the availability of the other. He addition of the blown fuse indicator actually makes the system more reliable by alerting the operators to the blown fuse condition before the need to operate the breakers arises. Existing separation is maintained since all new wiring is internal to the divisional control panels and switchgear. No pressure boundaries are affected and there is no change in the radiological ccnsequences at the site boundary. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. The new components are qualified for their intended use. He use of synchronous check relays in breaker closing circuits and the use of blown fuse indication in control circuits are consistent with methods already in use at River Bend. There are no new missiles created by this nodification. This modification will not afTect the failure modes of existing equipment. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This modification does not afTect any technical specification or the basis of any technical specification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 8.3-059 (CR 95-0887) Page 8.3-57 Tables: 8.3-3 9.4-4 Figure 8.3-3 Description and Basis for Chance: His modification replaces the fan motor assembly for the high pressure core spray (HPCS) pump room unit cooler. The fan motor assembly consists of a motor, a shroud and an impeller. The original Westinghouse 40 HP single speed motor is replaced with a Reliance 50 HP single speed motor. He impeller consists of a hub and five blades. The original fan motor assembly sufTered a catastrophic failure. The replacement parts come from a spare turbine building unit cooler and we:e upgraded to Class 1E standards. Section I Page 56 of I87
4 i Summary of Safety Evaluation: i
' The additional heat load due to the 50 HP motor is still below the design cooling capacity of the l HPCS unit cooler, and it has been determined that the temperature in the HPCS pump room will !
remain at or below the design temperature, here structural and seismic integrity of the fan motor . assembly has been demonstrated by test. The HPCS pump room unit cooler is not directly i involved in an accident event described in the SAR. He replacement of the fan motor assembly does not affect the performance any equipment important to safety. Existing electrical , independence and physical separation are maintamed Therefore, this modification does not ~ increase the probability of an accident or a malfunction of equipment important to safety : previously evaluated in the SAR. While the HPCS unit cooler is not designed to directly mitigate ; l any accident, it will continue to maintain design environmental conditions in the HPCS room. [ Therefore, this modification does not increase the consequences of an accident or a malfunction of , t equipment important to safety previously evaluated in the SAR. No new failure modes are , introduced by this modification. His modification does not create any new internally generated l missiles. Through redundancy and separation, failure of the HPCS room unit cooler will not affect the other division to perform its safety function. Therefore, this modification does not create the ! ] possibility of an accident or a malfunction of equipment important to safety different from any l [ previously evaluated in the SAR. He new fan motor assembly is capable of maintaining the required temperature as required by the improved Technical Specifications, herefore, this i
}
1
- modification does not reduce the margin of safety as defined in the basis of any technical i specification. For these reasons, this modification does not constitute an unreviewed safety !
! question. ! Channe Number /USAR Section: LCN 8.3-60 Tables: 8.3-1 Sheet 5 1 , 8.3-1 Sheet 6 ! i 8.3-1 Sheet 11 ! 8.3-2a Sheet 3 I 8.3-2b Sheet 3 l 9A.2-1 Sheet 6 i l 9A.2-2 Sheet 3 l 9.4-9 Sheet 2 l i Figures: 8.3-14a I
- 8.3-14b l 8.3-15 Description and Basis for Channe:
i' his e.odification performs work on dnwell unit coolers DRS-UCl A, DRS-UClB, DRS-UCIC, DRS-UClD, DRS-UCIE. AND DRS-UCIF. Changes generic to all unit coolers include replacing the existing drywell unit cooler motor with a 60/30 HP Reliance motor, mounting capability for 4
. vibration monitoring equipment on the motors, and painting the floor panels. Unit cooler DRS-UCIE will also have the existing fan shroud replaced with a new fan shroud assembly and will have the slow speed function used for testing reinstated. This modification will also proside the ability to adjust fan blade pitch in the future and guidelines to establish limits within which the fan blade pitch may be adjusted without impacting the design basis of the dnwell cooling system.
Section I Page 57 of187
m __ _ . ._. _ - _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _.._ _ . _ _ _ + Summary of Safety Evaluation: a. His modification will result in the drywell cooling system being returned to its original Mg=1
. and installed configuration and operation. This modification does not change the design basis of 4
1 the dnwell cooling system. Therefore, this modification does not increase the probability of an ] accident or a malfunction of equipment important to safety previously evaluated in the SAR. j i Failure of the drywell cooling system or any component associated with the drywell cooling system i
! will not impact any existing accident analysis. Therefore, this modification does not increase the )
] consequences of an accident or a malfunction of equipment to safety previously evaluated in the ]
- SAR. No new or different failure modes are added and no new or different system interfaces are i
<- added that could result in any new, unanalyzed accident. Therefore, this modification does not ! create the possibility of an accident or a malfunction of equipment important to safety different -j
- from any previously evaluated in the SAR. His modification does not impact any technical )
- specification or the basis of any technical specification. Herefore, this modification does not !
- reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. .
i l Chanae Number /USAR Section: LCN 8.6-001 Page 9.5-10 i Tables: 3.6A-41a ! 3.6A-47a 3.6A-38a l L Description and Basis for Che,gte: j 5 His modification replaces the existing cables for all six hydrogen igniters located near the dnwell ) ceiling with a new type of Class 1E stainless steeljacketed, "hard wall," (SiO2) insulated cable. I t ne existing cables' qualification is limited to 78 weeks of operation in the dowell environment. j
- He new cable, designed to be routed outside of conduit, is emironmentally qualified, has a 40 year i radiation and thermal life, meets 10CFR50 Appendix R requirements, and is loss of coolant j
! accident (LOCA) qualified for BWR installations. The new cables will be routed along the route !
l points of the existing cables and attached to existing conduit supports. If existing supports cannot ;
- accommodate or exceed seismic maximum span requirements for the new cables, new supports will i be installed.
4, i i Summary of Safety Evaluation: 1 i The function and operation of the hydrogen igniters will remain the same. The cable route used for
- routing power to the drywell ceiling igniters well be divisionally separated. His cable will bc l
. seismically installed to meet the design basis event requirements. His cable is designed to meet the )
j design conditions for cables used in the dowell and containment. Evaluations documented in the S AR indicate that the hydrogen control system will still be able to perform its safety-related design function underjet impingement even with the loss of all dowell ceiling igniters. -Therefore, this modification does not increase the probability of an accident presiously evaluated in the SAR. 1 Once terminated and connected, the cable will require no maintenance actisities. This cable
' installation will be subject to the same previously analyzed events as the existing conduits to the
. drywell ceiling igniters. Therefore this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the S AR. The design standards used to qualify the metaljacketed "hard wall" cable are the same standards used to qualify the existing cables used in the drywell and containment. The consequences ofloss of Section 1 Page 58 ofI87 f
-..c--
i i ; I i 4 drywell ceiling igniters will be the same with the new cables as with the old cables. Herefore, this j 3 modification does not increase the consequences of an accident'or a malfunction of equipment ( important to safety previously evaluated in the SAR. This cable will be assembled in sections and j use environmentally qualified connectors to ensure that the cable continuity is maintamed under i i design accident conditions other thanjet impingement. The new cable is the same copper ! conductor size as the original cable installation and is more than sufficient to carry the load. The existing protective devices used to protect the existing cables will be used. He initiation of the ! hydrogen bum is a manual operation that is performed by an operator based on hydrogen - ; concentrations in the drywell. His modification will not impact the point at which this manual ! l action is initiated and it does not impact any device the is used to decide when to initiated this j manual action. All terminations will be performed, documented, and inspected to verify proper i connections for a safety related application. Therefore, this modification does not create the . possibility of an accident or a malfunction of equipment to safety different from any previously j evaluated in the SAR. Technical specifications require the hydrogen control system to be operable ; in modes one and two. If a design basis event such as jet impingement climinates all the igniters on [ the drywell ceiling, The non-impacted igniters in the drywell at low elevations will still function to i perform their required safety related function. De safety related power sources used to supply . j these igniters will not be impacted by this change. Herefore, this modification does not reduce the ; margin of safety as defined in the basis of any technical specification. For these reasons, this ; modification does not constitute an unreviewed safety question. l i
. Channe Number /USAR Section: LCN 9.1-038 Pages: 9.1-26 l 9.1-32 Table 9.1-5 ;
Figures: 7.6-7 Sheet 1 9.1-23a ! 9.4-2b ! I i Description and Basis for Channe: l De potential oflosing fuel pool cooling due to the loss of pump room ventilation during a fire caused the cooling systems to be reanalyzed. The proposed modification trips the fuel pool cooling l pump when high room temperature is attained which indicates a loss of ventilation. His action protects the pump, enabling it to be restarted after the ventilation problem is corrected. This ; ensures that the pump is available for cooling the fuel pool. ; Summary of Safety Evaluation: !
. De automatic shut off of the fuel pool cooling pumps occurs when the room temperature of the l control room exceeds 131*F. The maximum abnormal temperature for the pump room is 132*F ;
which is the temperature the pumps are qualified. His situation will probably occur when an ! Appendix R fire occurs in the control room or the fuel building. His would cause a cold shutdown j of the plant. The fuel handling accident is not affected by the increase in temperature because the ; temperature rise would only occur during a fire when no fuel is being moved. Without a fuel I handling accident the radiation levels in the exhaust train are negligible, thus charcoal filters are : not required. Therefore, the probability of an accident previously evaluated in the S AR is not j increased The HVC emergency system will function to remove heat and maintain the pump room ( at or below 131*F. The ambient temperature in the charcoal filtration fan rooms would be at ! 123*F maximum with the fan running. The pool temperature would be a maximum of 170*F. This l Section I Page 59 of I87 i
i i is within the bounding post accident operating temperature of 124*F. Herefore, the consequences l of an accident previously evaluated in the SAR is not increased. Tripping the pump protects the ! pump and enables it to be restarted after the loss of ventilation is corrected. He control circuitry ) is indanandent for each train, therefore no single failure can affect both pumps. His ensures that , the pump is available for cooling the fuel pool. With this modification the possibility of an !
- accident which is different than that previously evaluated in the SAR will not be created. The l action of tripping the pump will not increase the probability of a malfunction. It has been analyzed j that the effects due to the loss of cooling in the fuel pool are within design requirements for the !
expected time the pump is to be out of service. It was determined that this new maximum j temperature was bounded by the existing design and evaluated to be acceptable. He changes to the fuel pool cooling pump circuit does not change the consequences of any malfunction of a safety-related SSC previously evaluated in the SAR. For these same reasons the possibility of a ; malfunction of a safety-related SSC will not be created differently than that presiously evaluated in ! the SAR. He proposed modification does not affect the basis to any Technical Specification. l Therefore, the margin of safety as defined in the basis to any technical specification will not be ! reduced. For these reasons this change does not constitute an unreviewed safety question. . I Channe Number /USAR Section: LCN 9.1-45 Page 9.1-64 ; Description and Basis for Channe: l The SAR specifies that six reactor vessel studs are removed in line with the fuel transfer canal. This is being changed to specifyjust that the studs in line with the fuel transfer canal are removed. Only five studs are required to be removed to allow for fuel movement. Removal of the sixth stud is not desired because this impacts the installation of the fuel transfer canal shield. Summary of Safety Evaluation: ! The sixth stud is used to support the fuel transfer canal shield. Since the fuel transfer canal shiel'd } is required to be in place for irradiated fuel movement, the additional stud does not increase the , probability or potential for fuel movement interference. Therefore, this change does not increase i the probability or an accident previously evaluated in the SAR. The removal or lack of removal of l a reactor vessel head stud has no effect on the amount of fission products released in any of the ! evaluated refueling accident scenarios. Therefore, this change does not increase the consequences ! of an accident previously evaluated in the S AR. The reactor vessel studs do not perform any safety function in the refueling mode and do not affect any equipment important to safety. - Therefore, this change does not increase the probability or the consequences of a malfunction of ! equipment important to safety previously evaluated in the SAR. The intent of removal of studs l from the fuel transfer canal is to provide a pathway for fuel movement without interference so that ! a fuel assembly is not damaged or dropped during handling. The additional vessel stud is used to , anchor the fuel transfer shield. With the fuel transfer shield in place, the extra stud does not create any additional interfaces to fuel movements. Therefore, this change does not create the possibility i of an accident or a malfunction of equipment important to safety different from any presiously [ evaluated in the SAR. This change does not affect any technical specification or the basis of any i technical specification. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. ; i Section I Page 60 of187 j _____n___
l l Channe Number /USAR Section: LCN 9.1-128 B Pages: 7.7-34 9.1-42 9.1-55 9.1-56 9.1-60 9.1-61 9.1-62 9.1-63 9.1-64 9.1-66 9.1-76 9.1-79 Description and Basis for Change: This change clarified refueling procedures concerning the operation of the refueling platform, the fuel handling platform, the fue' preparation machine, the movement of fuel, the receipt and inspection and channeling of nov fuel. Summary of Safety Evaluation: The most limiting accident scenario concerning refueling operations is the dropping of a spent fuel bundle onto unchanneled spent fuel. This modification enhances and clarifies the procedures for refueling operations and does not affect any of the assumptions or inputs to the fuel handling accident analysis. This change does not increase the probability or consequences of an accident ; previously evaluated in the SAR because the probability of an accident is bounded by accident ; analyses already described in the SAR. This change has no affect on the design of any refueling ] equipment. The platforms are Safety Class 2 and Seismic Category I and all portions of the hoist : are designed to have a safety factor of 5 based on the ultimate strength of the material. The lifting of fuel has previously been evaluated in the SAR, therefore no new or different accidents are created as a rese't of this change. No physical modifications to plant equipment were performed as a result of this change. All of the interlocks on the platforms that are discussed in the Technical Specifications remain unchanged, therefore there is no decrease to the margin of safety as described in the basis to any technical specification as a result of this change. This change does not create an unreviewed safety question. 1 1 Change Number /USAR Section: LCN 9.2-141 Table 9.2-3 Figures: 5.4-15a 9.2-2b Description and Basis for Change: This modification installs larger capacity reactor water cleanup (RWCU) seal coolers on RWCU pumps A and B. Temperature indicators are added to each of the RWCU pumps. Plant drawings are revised to reflect the addition of component numbers for RWCU skid mounted coolers. I Section I Page 61 of 187
Summary of Safety Evaluation: The work performed under this modification has been evaluated to ensure that the additional loading due to the new heavier coolers, flanges, and resistance temperature detectors (RTDs) will not adversely affect the seismic qualifications of the affected piping. Required ASME and other code specifications for the equipment are met. He new RWCU seal cooler has a 100% larger heat removal capability at rated conditions, and will pennit lower RWCU seal temperatures, longer seal life, and less equipment downtime. Herefore, this change does not increase the probability of an accident previously evaluated in the SAR. Any possible failures of equipment affected by this change will not affect any equipment that is required to mitigate an accident. Therefore, this change does not increase the consequences of an accident presiously evaluated in the SAR or create the possibility of an accident different from any previously evaluated in the SAR. Safety related equipment affected by this change is associated with the RWCU pressure boundary, the RWCU seal flush piping, and the new cooler. He new piping configuration meets all , specifications. Therefore, this change does not increase the probability of a malfunction of ! equipment important to safety previously evaluated in the SAR. The consequences of a ! malfunction of a safety related structure, system, or component are not increased because l radiological releases could be isolated in the event of pipe failure. He possibility of a malfunction i of a safety related system other than those evaluated in the SAR are not increased by this change. l The margin of safety as described in any technical specification is not reduced. For these reasons, ; this modification does not constitute an unreviewed safety question. l 1 Channe Number /USAR Section: LCN 9.2-245a Figures: 9.2-la 9.2-1 b 9.2-8a 9.2-8g Descrintion and Basis for Chanee: i i This modification replaces the ventilation chilled water system (HVN) purge units with self- l contained and environmentally sound purge units. The new purge units eliminate the need for ! cooling water lines, purge compressors and associated instrumentation. The new purge units l utilize a CFC-free, integrated refrigeration system and have a microprocessor based control system l which enables superior refrigeration protection and significant energy sasings. Summary of Safety Evaluation-The new purge units are functionally equivalent to the old purge units and do not reduce tlx: reliability of the HVN system. Also, the HVN system and associated purge units are not required for safe shutdown or post-accident mitigation. This modification does not affect any equipment whose malfunction is postulated in the S AR to initiate an accident. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The HVN system is isolated in the event of a loss of coolant accident (LOCA) and the safety related equipment is transferred to the standby service water system. The change of cooling water supply to the safety related equipment is not affected by the modification. Emissions from the new purge units are much lower compared with the old units. All purge units are powered by non-nuclear safety related distribution panels from the normal 480V load centers. Herefore, this modification does not increase the consequences of an accident Section I - Page 62 of I87 l
or a malfunction of equipment important to safety previously evaluated in the SAR. The HVN system is isolated from other stmetures, systems, and components during accident conditions and the purge units were replaced with functionally equivalent units. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The technical specifications do not contain a requirement for the operability of the HVN chillers or associated purge units. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification l For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 9.2-248 Figures: 5.4-!5a 9.2-2b Description and Basis for Channe: This modification installs one seal oil drain valve each for reactor water clean-up (RWCU) pumps A and B. It also adds one seal cooler vent valve per RWCU seal cooler for RWCU pumps A and B. The addition of these components supports a larger effort to replace the mechanical seal on RWCU pumps A and B. Summary of Safety Evaluation: The work executed by this modification has been evaluated to ensure that the additional loading does not adversely impact the seismic qualification of the affected piping as well as ASME and other code specifications. Therefore, the probability of an accident not previously evaluated in the SAR is not increased. In addition, the new RWCU has a larger heat removal capability and permits lower seal temperatures, longer seal life, and less equipment down time. Therefore, the consequences of an accident previously examined in the SAR are not increased. All possible accidents and malfunctions that could be initiated by the new equipment are bounded by analysis already incorporated in the SAR. The probability or consequences of a malfunction of a safety-related structure, system or component are not increased. The margin of safety = Mmed in any basis to the technical specifications is not reduced because the affected portion of the RWCU system is not specifically addressed in the Technical Specifications. For these reasons, this alteration does not constitute an unresiewed safety question. Channe Number /USAR Section: LCN 9.2-250 Page 9.2-36 Description and Basis for Channe: The required temperature limit for standby cooling tower basin in the Tecimical Specifications is 88'F. This change aised the annunciator setpoint of the basin temperature from 80 F to 84'F. The new setpoint infonns the operators of an increasing basin temperature before the technical
' specification limit is reached and also prevents a nuisance alarm during the summer months. The setpoint takes into account the instrument loop inaccuracy ofil.8 F.
Section I Page 63 ofI87
i Sununary of Safety Evaluation: i The change is limited to temperature monitoring of the standby cooling tower (SCT). Failure or malfunction of the_ monitoring system was not postulated to initiate any of the accidents previously j evaluated in the SAR. The SCT temperature monitoring system is Q-Class 2 and the purpose is to j alert the operators of an upward trend in the SCT basin temperature. The annunciator on the ; monitoring system does not initiate any automatic actions and the only operator action is to run the .
' SCT fans and pumps as necessary to lower the basin temperature. The ability of the SCT to i perform its DBA function was unaffected by the modification because the setpoint is below the :
SCT basin temperature of 88'F as stated in the Technical Specifications. Therefore, the l consequences of an accident previously evaluated in the SAR is not increased. The annunciator j setpoint modification did not change, add, or impact the monitoring system interaction with safety ; related equipment associated with Operator actions, hrefore, the possibility of an accident which - is different than any previously evaluated in the SAR is not created. "Ihis change will remain as a monitoring function and is not related to any safety related structure, system, or component (SSC). - Therefore, the probability of a malfunction of a safety related SSC previously evaluated in the ! SAR is not increased. h SCT temperature monitoring system is a Q-Class 2 system and is not l required to be operable following a DBA. The consequences of a malfunction of a safety related l SSC previously evaluated in the SAR is not increased. Since, the SCT basin temperature ; annunciator does not interface or support the function of any safety related SSC then the possibility -l of a malfunction different than that previously evaluated in the SAR is not created. The technical ; specification monitoring requirements for the SCT basin was not impacted by changing the ' annunciator setpoint. Therefore, the margin of safety as defined in the basis to any technical specification was not reduced. For these reasons, this change does not constitute an unresiewed j safety question. i Channe Number /USAR Section: LCN 9.2-251 Figures: 9.2-24b l 9.3-16c q Description and Basis for Channe: i p This modification upgrades the sump drain system on elevation (-)l5'-0" of the intake stmeture. l The discharge capacity of the drain pumps 1DFM-PSA and IDFM-PSB is increased. The ; downstream lines of the drain pumps are rerouted to discharge the sump into the makeup water ' 1 system (MWS) line rather than the river. A 3" ball valve is installed and used to perform a single hot tap into the MWS cross line, upstream of the MWS pumps IMWS-P4A and IMWS-P4B. l Two check valves are installed in series in the sump discharge line. The trip coils and overload heaters associated with sump pumps 1DFM-PSA and IDFM-P5B are replaced and the motors are re-rated. A 2" globe valve IMWS-V3169 is installed upstream of the ball valve to tiuottle the sump pump discharge flow. Summary of Safety Evaluation: The insertion of the 2" globe valve is not detrimental to the system because even ifit shuts while the pumps are operating, the pumps will shut down automatically after reaching the dead head of
- 110 feet. The replacement of the trip coils and overload heaters as well as re-rating the pump motors conform to station, utility, and industry codes and standards. These changes will have no adverse impact on the plant electrical distribution system or its ability to perform its design
- Section 1 Page 64 of187
function. Herefore, this modification does not increase the probability of an accident presiously evaluated in the SAR. He sump drain system in the intake structure and the MWS are in no way related by function or physical proximity to any safety related system, structure, or component (SSC). Therefore, this modification does not increase the probability of a malfunction of any safety related SSC. The sump drain system of the intake structure and the MWS have no role in
]
mitigating the consequences of an accident. Therefore, this modification does not increase the l l consequences of an accident or a malfunction of any SSC previously evaluated in the SAR.
- Considering the above discussion, this modification does not create the possibility of an accident or i a malfunction of any safety related SSC different from any previously evaluated in the SAR. The !
sump drain system of the intake structure and the MWS system are not included in any technical i specification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 9.2-255 Figure 9.2-7c , Descrintion and Basis for Channe: , He purpose of this modification is to relocate the main feedwater pump mechanical seal heat exchangers. This change promotes thermal syphon cooling of the seals while a pump is in " hot" 4 standby. This reduces and possibly eliminates the potential for a seal premature failure due to the effect of thermal shock. The change entails the reversal of the arrangement of the inlet and outlet ports on the seal cartridges. Summary of Safety Evaluation: All changes associated with this modification are confined to non-safety-related components located in a non-seismic area of the turbine building. The proposed changes do not alter the function or method of performing the function of the reactor feed pump seal cooler. Rather, the modification improves the ability of the pump seal cooler to perform its function, thus improsing seal reliability. All permanent changes authorized by this modification are designed, constructed i and tested according to approved specifications, procedures. existing codes and license conditions. His modification was implemented when the affected portions of the system were not required to be operable. Therefore, this change does not increase the probability of an accident previously evaluated in he SAR. The affected equipment neither functions to mitigate the consequences of any accident nor supports any other structures, systems, or components (SSC) in mitigating accident consequences. Herefore, this change does not increase the consequences of an accident presiously evaluated in the SAR. This modification has been designed to ensure that all existing and new components added by the proposed modification will maintain the existing design basis requirements. Since the additional components are passive, all component functions (passive pressure boundary or passive mechanical support) remain essentially unchanged such that no new - failure modes are introduced by the proposed change. Therefore, this change does create the possibility of an accident of a different type than any evaluated previously in the SAR. This change is limited to non-safety-related components located in a non-seismic area of the turbine building containing only non-safety-related equipment. This change does not add or change any system functions. This change was constructed and tested according to approved site specifications, procedures and license conditions. Therefore, this change does not increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. The equipment in the surrounding area neither functions to mitigate the consequences of any accident ; Section i Page 65 of 187 l
nor supports any SSC responsible for mitigating accident consequences %erefore, this change does not increase the consequences of a malfunction of a safety related SSC pre 5iously evaluated in the SAR. He nature of this modification is that any failure resulting from the modification will only result in failure of the reactor feed pumps or other equipment not required for safe shutdown of the plant. Derefore, this modification does not create the possibility of a malfunction of safety related SSC different from any previously evaluated in the SAR.- There are no technical specifications applicable to the non-safety-related components that are modified or potentially impacted. This change dces not change system operation or functional requirements. Therefore, this change dccs not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 9.2-256 Figure 9.2-24a Description and Basis for Channe: His modification extends an existing heat trace on clarifier line IWTL-004-60-4. This modification also adds heat tracing to clarifier line IWTL-004-65-4 and clarifier valves IWTL-V92 and IWTL-V93. He new circuits will make use of the same non-safety-related 120V ac power source as the existing heat trace by tapping into its power connection kit. His will prevent the lines and valve from freezing. Summary of Safety Evaluation: The clarifier and makeup water systems are non-safety-related. They are not postulated to cause any of the accidents evaluated in the SAR. Malfunctions of these systems are not esaluated in the SAR. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. The clarifier and makeup water systems are not used to mitigate the consequences of any accident evaluated in the SAR. Herefore, this modification does not increase the consequences of any accident previously evaluated in the SAR. This method of freeze protection is consistent with the method in use for other lines in the same system. It makes use of the same non-safety-related power supply, so no new power cables are added. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The clarifier and makeup water systems are independent of and safety related components and systems. They are not required for safe shutdown of the reactor. Therefore, this modification does not increase the probability or consequences of a malfunction of any equipment important to safety previously evaluated in the SAR. This modification does not create the possibility of a malfunction of equipment important to safety previously evaluated in the SAR. His modification does not affect any technical specification and does not affect the basis of any technical specification. Therefore, this modification does not reduce the margin of safety as defmed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l 1 n
. Section I Page 66 ofI87-
?
l l l l Channe Number /USAR Section: LCN 9.2-257a Figure 9.2-24c I i Description and Basis for Channe: This change improves the pump seals on make up water system (MWS) pumps, IMWS-P4A and j 1MWS-P4C, Dese make up water pumps are located in the intake structure which is located on i the east bank of the Mississippi River, about 2.5 miles from the plant. Hey had been leakmg due i to the presence of grit and silt in the river water that is presently lubricating the pump seals. This ; modification provides water from two new wells drilled outside the intake structure to lubricate the ! pump seals. Submersible pumps, a water storage tank, two fresh water pumps, several pipes, two new control panels, and a telephone system with sound isolation booths are to be installed, also. In addition the well water will be used in the vacuum priming system (VPS) and for bearing lubrication of the make up water pumps.
. I Summary of Safety Evaluation: !
None of the accidents in the SAR have any relation to the MWS, VPS, and telephone systems that are modified. He MWS, VPS, and telephone systems are non-safety-related systems. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. : Should any or all the MWS, VPS, or telephone components that are modified experience a failure, f it will not have any effect on any of the accidents evaluated in the SAR. Herefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. ; The MWS, VPS, and telephone systems are required for operation of the plant but are completely [ inha-lent and are not required for safe shutdown of the plant. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The geographical distance between the modified systems and the plant's safety related systems ( precludes inflicted damage due to missiles in case of seismic events or gross failure of a i component. Ecrefore, this modification does not increase the probability of a malfunction of any ; safety related system, stmeture, or component (SSC) previously evaluated in the SAR. The MWS, l , VPS, and telephone systems are not required for safe shutdown of the plant. Therefore, this , modification does not increase the consequences of a malfunction of any safety related SSC , , previously evaluated in the SAR. The MWS, VPS, and telephone systems do not interface with i ! any safety related SSC. Therefore, this modification does not create the possibility of a l malfunction of any safety related SSC different from any previously evaluated in the SAR. The ; MWS, VPS, and telephone systems are not addressed by any technical specification and do not [ ! have any affect on SSCs that are addressed by technical specifications. Therefore, this modification does not reduce the margin of safety as defined in the basis for any technical , specification. For these reasons, this modification does not constitute an unresiewed safety question. f 1 i i I Section I ~ Page 67 ofI87 f [
4 ? i i
~
4 h 1 Channe Number /USAR Section: LCN 9.2-257b -. Figure 9.2-24c ' 1 Descriotion and' Basis for Channe:
] ' > his modification involves the addition of a one-inch gate valve,1MWS-V3170, in the cross line : }
between the two redundant seal make-up water system (MWS) water pumps, IMWS-P8A and ; 1MWS-P8B. His will allow full autonomy of each train. This modification also involves minor i i changes to hardware to acen==~b*a the field condition and addition of pipe supports and the i l operational test of the " Refurbished" make up pump IMWS-P4C will be done while 1MWS-P4A still has the old packing seal and is providing make up water to the plant. Summary of Safety Evaluation: j l None of the accidents in the SAR have any relation to the portion of the MWS system that is l 1 modified. De MWS system is a non-safety-related system. Herefore, this modification does not ; increase the probability of an accident previously evaluated in the SAR. Should any of the modified j MWS components experience a failure, it will not have any affect on any of the accidents l
- evaluatei in the SAR. Therefore, *.his modification does not increase the consequences of an i accident previously evaluated in the SAR. The MWS system is required for operation of the plant ,
j .. but is not required for safe shutdown of the plant. Herefore, this modification does not create the j possibility of an accident different from any previously evaluated in the SAR. He geographical distance between the modified systems and the plant's safety related systems precludes inflicted damage due to missile in case of seismic events or gross failure of a component. Therefore, this 3' modification does not increase the probability of a malfunction of any safety related system, a' structure, or component (SSC) previously evaluated in the SAR. The MWS system is not required q for safe shutdown of the plant. Therefore, this modification does not increase the consequences of j a malfunction of any safety related SSC previously evaluated in the SAR. The MWS system does ! not interface with any safety related SSC. Therefore, this modification does not create the l possibility of a malfunction of any safety related SSC different from any previously evaluated in j the SAR. - ne MWS system is not addressed by any technical specification and does not have any i effect on SSCs that are addressed by any technical specification. Therefore, this modification does
. not reduce the margin of safety as defined in the basis for any technical specification. For these reasons, this modification does not constitute an unresiewed safety question.
.a Change Number /USAR Section: LCN 9.2-257c Figure 9.2-24c Description and Basis for Channe: his modification involves the installation and deletion of several valves. Eight one-halfinch ; needle valves were installed downstream of the flow indicators for fresh water supply to the make- i up water (MWS) system pumps, IMWS-P4A and IMWS-P4C. The new valves will prevent the
, backup river water from pressurizing the flow indicators during the calibration of the flow I i indicatorsf Pressure Relief Valves MWS-RV166 and MWS-RV167 will be deleted to allow the !
seal water pumps to operate under all operating conditions without interruption. He added valves will be used for partial draining of the system and for venting of the pumps. The drain lines for the
' deleted relief valves will be isolated with two one-inch gate valves, MWS-V3180 and MWS- l V3181; j i
. l
- Section I Page 68 ofI87 l l
L,.. . . _ , . . . .,_,__...,.m , .i _ . $
_ _ . _. _ _ _. _ _ . _ . . _ _ _ . _ _ .~ _ _ _ _ _ . _ _ -. - i i ! l- l ) Summary of Safety Evaluation: ! F None of the accidents in the SAR have any relation to the MWS system that is modified. He ! t MWS system is a non-safety-related system Herefore, this modification does not increase the - l
- probability of an accident previously evaluated in the SAR. Should any of the modified MWS ;
; componer,.s cxperience a failure, it will not have any effect on any of the accidents evaluated in the !
- i. - SAR. Derefore, this modification does not increase the consequences of an accident previously j
- - evaluated in the SAR. The MWS system is required for operation of the plant but is not required ;
i for safe shutdown of the plant. Derefore, this modification does not create the possibility of an ! I accident different from any previously evaluated in the SAR. De geographical distance between !
- the modified systems and the plant's safety related systems precludes inflicted damage due to !
i missile in case of seismic events or gross failure of a component. Derefore, this modification does ! i not increase the probability of a malfunction of any safety related system, structure, or component I (SSC) previously evaluated in the SAR. The MWS system is not required for safe shutdown of the i 1- plant. Therefore, this modification does not increase the consequences of a malfunction of any i safety related SSC previously evaluated in the SAR. He MWS system does not interface with any l
- ' safety related SSC. Derefore, this modification does not create the possibility of a malfunction of {
any safety related SSC different from any previously evaluated in the SAR. The MWS system is ! l not addressed by any technical specification and does not have any effect on SSCs that are l 2- addressed by any technical specification. Therefore, this modification does not reduce the margin j l of safety as defmed in the basis for any technical specification. For these reasons, this ! ! modification does not constitute an unreviewed safety question. j i ! L l Channe Number /USAR Section: LCN 09.02-258a Figure 9.2-2b l Description and Basis for Channe: b This modification replaces the existing reactor plant component cooling water (CCP) pump
- discharge swing check valves 1CCP-V25, V37 and V46 with Mannesmann Demag nozzle check valves. The reason for the replacement for the existing valves is that they have required excessive 4 maintenance in the past. De nozzle check valves are better able to survive in the turbulent flow ,
L conditions at the discharge pump. Due to the design of the nozzle check valves, they will be flanged into the piping system rather than welded in as the existing valves are. 4 j Summary of Safety Evaluation: } , Based on a review of Chapter 9, Sections 3,6,9 and 15, of the SAR the CCP system does not initiate any of the accidents evaluated in the S AR. For this reason, the replacement of the CCP l , pump discharge check valves with a different type will not increase the probability of an accident l previously evaluated in the SAR. De only CCP components required to operate following an l accident are the contamment isolation valves which are required to close on a containment isolation signal. Due to the design differences between the existing swing check valves and the replacement , o nozzle check valves, the pressure drop across the nozzle check valves is approximately 1.2 ft. i higher than the swing check valve at maximum system flows. This small pressure drop increase is j insignificant and will not i.npact the dynamic stroking of the CCP containment isolation valves. j Derefore, this modification does not increase the consequences of an accident presiously evaluated j in the SAR. The replacement check valves are of different design than the existing swing check l valves, however, they are built to the same codes and standards as the existing check valves. No - 'l i Section i Page 69 of187 l l l u ~ ___ ._!
new or difTerent failure modes are added and no new or different system interfaces are added that could result in the possibility of an accident of a different type than already stated in the SAR. The CCP system is a non-safety-related system that provides cooling water to various safety and non-safety-related systems in the containment, auxiliary, and fuel building. The safety related loads on the CCP system are the fuel pool cooling heat exchangers and the RHR pump bearing coolers. In addition to these two safety related loads, there are various safety related motor operated valves and check valves within the CCP system. The primary impact that the replacement valves could have on these components would be due to a decrease in flow rate. The decrease in the flow rate and pressure is negligible and will have no impact on the operation of the MOVs, check valves, or on cooling of the safety related loads. Here are also two sets ofpressure switches in the CCP system that causes automatic actions to occur. These pressure switches initiate the safety related SSW system as backup cooling for the CCP. Therefore, the probability of a malfunction of a safety related structure, system, or component (SSC) previously evaluated in the SAR will not be increased. Since this modification does not impact the operation of the safety related MOVs then it does not increase the consequences of a malfunction of a safety related SSC previously evaluated in the SAR. The replacement check valves are of different de::ign but are built to meet the same codes and standards therefore, the possibility of a malfunction of a safety related SSC different than any previously evaluated in the SAR can not be created. The Technical Specifications and Bases have been reviewed for any impact due to the changes made by this modification. It was found that the margin of safety as defined in the basis to any technical specifications was not reduced. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 9.2-268 Figure 9.2-24b Description and Basis for Change: This change reflects the addition of a small submersible sump pump in the makeup water flow control valve chamber located near the makeup water pump Structure at the Mississippi River. It includes the installation of the new pump, associated valves and piping to tie the pump discharge into the existing piping (MWS-016-287-4) which contains make up water system (MWS) surge anticipator valves (MWS-Vl9A,B) which discharge to the nyer. Summary of Safety Evaluation: , The makeup water system (MWS) including the makeup water control valve chamber sump pump and discharge piping are non-safety-related. This equipment does not interface with any safety structures, systems or components (SSCs) that are postulated to cause an accident in the SAR. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The new sump pump does not change the operation of any equipment in the MWS nor in any other plant system and its failure does not adversely impact any safety-related SSC. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The MWS is not mentioned in the Technical Specifications, including the Improved Technical Specification (ITS). The make-up water system is not relied upon for any margin of safety defined in the TS Bases, SAR or SER. Therefore, this , change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l Section 1 Page 70 ofI87 l
Change Number /USAR Section: LCN 9.2-269 Figure 9.2-1k Description and Hasis for Change: This change deletes the scan function for Tamaris computer point ISWC-T24C. The computer point was giving erratic temperature indications and creating a nuisance alarm in the main control room. His was caused by a shorted thermocouple in pump motor ISWC-PIC. The hardware will remain in place so that if the motor is rewound or replaced, the thermocouple can be replaced and monitored again. Summary of Safety Evaluation: The senice water cooling (SWC) system is not evaluated in any of the accident scenarios in the . SAR. There are no postulated scenarios whereas the SWC system is relied upon to mitigate the consequences of accidents. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the S AR. The pump will continue to function as intended after the change is made. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This change does not impact any technical specification or the basis of any technical specification. Therefore, this change does not reduce the margin of safety as defined in the basis of any tecimical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 9.2-271 Figures: 9.2-24A 9.2-24b 9.3-l b Description and Hasis for Chance This modification revises the makeup water system by adding new clarifier level transmitters, clarifier level indication in the auxiliary control room (ACR), and adds senice water cooling (SWC) tower basin level indication in the ACR (signal provided by existing SWC level instrument) and a new air operated actuator valve MWS-A0V267 with control from the ACR. The purpose of the changes is to provide operators with clarifier collector fiume level indication and remote operation of the actuator. The indications of the clarifier collector flume level, senice water cooling tower basin flume level, and circulating water cooling tower basin fiume level (existing indication), with remote control of Valve MWS-AOV267, allows the operator to prevent overflow of the operating clarifier and assures that it does not reach levels which cause spillage in the clarifier not in operation. Summary of Safety Evaluation: He changes to the cooling tower makeup water system (MWS) do not affect or alter operation of any safety related equipment nor do they affect any statements in the SAR since there are no accidents evaluated ia the SAR which are related to this activity. The changes provide additional means for the operator to perform tasks that are already performed. The changes will not delete or modify any safety system's protection features, will not downgrade any safety system's performance necessary for reliable operation of equipment importart to safety, will not reduce any safety system redundancy or independence, will not increase frequency of challenges to equipment '.section 1 Page 71 of 187
i I i important to safety, nor will it impose increased or more severe testing requirements on equipment j imponant to safety. Herefore, this modification does not increase the probability of an accident or j ~ the probability of a malfunction of equipment important to safety previously evaluated in the SAR. l The MWS is not'a safety related system, it does not interface with a safety related system, so, it is ! not required to mitigate accidents nor provide for plant safe shutdown. Derefore, this j modification does not increase the consequences of an accident or the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. This modification j does not add any new functions to the MWS or SWC systems. He piping, tubing, conduit and ; instruments that were installed have been analyzed, qualified and supported in accordance with the same criteria as the existing equipment. The changes to the MWS do not introduce any new failure ; mechanism for any safety related system which would lead to a failure mode of a different type than any previously evaluated in the SAR. Therefore, this modification does not create the ,
' possibility of an accident or the possibility of a malfunction of equipment important to safety .
.j different from any previously evahed in the SAR. Changes to the .MWS will not affect the mode '
of operation of any safety related system, will not create a system configuration or operating condition such that the Technica! Specification (TS) limiting condition for operation or surveillance ' requirements are no longer adequate, nor will it bypass or invalidate automatic actuation features l _ required to be operable by TS. Therefore, this modification does not reduce the margin of safety as . d defined in the basis of any technical specification. For these reasons, this modification does not - constitute an unreviewed safety question.
- t. i i
! Channe Number /USAR Section: LCN 9.2-278 Figure 9.2 !
- i
- Description and Basis for Channe
- l nis change results in the addition of a surge check valve to the cooling tower make-up water j system (MWS). The installation of this valve is part of an overall plan to improve the existing :
. MWS design and operational performance. Summary of Safety Evaluation: !
- The proposed change is a minor change to the MWS system intended to improve the reliability of !
the system. The MWS system is non-safety-related, and is used to convey non-radioactive water ! , from the Mississippi river, to the clarifiers and ultimately to the circulating water system, where l' the treated water is used to cool the main turbine condenser. A malfunction or failure of the MWS system is not analyzed in Chapter 15 of the SAR. Therefore, this modification does not increase j the probability of an accident previously evaluated in the SAR. The proposed change is a minor ; change to the MWS system intended to improve the reliability of the system. The MWS system is non-safety-related, and is not relied on to mitigate any accident scenarios. Therefore, this i modification does not increase the consequences of an accident previously evaluated in the S AfL l The proposed change is a minor change te the MWS system intended to improve the reliability of l the system. ' The MWS system is non-safety-related. None of the failure modes associated with ; this modification will adversely affect safety. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The MWS system and the proposed activity are non-safety-related, and do not interface with safety related equipment i e or equipment important to safety. Therefore, this modification does not increase the probability of -
- ; a malfunction of equipment important to safety previously evaluated in the SAR. The MWS j system and the prcposed activity are non-safety-related, and do not interface with safety related ;
. ' equipment or equipment immrtant to safety. Therefore, this modification does not increase the
. Section I Page 72 ofI87 l
_. __. ._. _= . . . _ . . _ . . _ . _ _ _ . . __._ _ _ _ _m , _ _
- i t
w consequences of a malfunction of equipment important to safety previously evaluated in the SAR. i ne proposed change is a minor change to the MWS system intended to improve the reliability of
- the system. He MWS system is non-safety-related. None of the failure modes associated with this modification will adversely affect safety; Therefore, this modification does not create a {
; malfunction of equipment important to safety different from any presiously evaluated in the SAR. !
; . No margins of safety defined or established by analysis are Aqwadent on the operation of the i
- MWS system. Herefore, this modification does not reduce the margin of safety as defmed in the i j basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
I i E Channe Number /USAR Section: LCN 9.2 284 Figure 9.2-27 i Description and Basis for Channe: This modification restores the standby cooling tower (SCT) chemical feed system. This modification entails replacing a small section of carbon steel piping with PVC, installing a new :
; check valve, and addition of an alternate feed pump suction connection. It will also allow the use of chemicals other than sodium hypochlorite in the ultimate heat sink (UHS). ;
Summary of Safety Evaluation: l . The feed system for the SCT does not perform any safety functions and therefore does not function 1 to prevent the occurrence of any previously evaluated accidents. He changes to the chemical feed , . system only impact the SCT from a standpoint ofincreased reliability. He only other system or component potentially impacted is the standby service water (SSW) system and the UHS. Neither i the SSW nor the UHS function to prevent the occurrence of any evaluated accidents. He portion of the chemical feed system being modified is in a non-seismic area and failure of the piping or any j components would not result in a malfunction of any portions of the SSW or the UHS. Therefore, j this modification does not increase the probability of an accident or a malfimetion of equipment i important to safety previously evaluated in the SAR. The SCT chemical &cd system does not l function to mitigate the consequences of any evaluated accidents. Failure of the SCT chemical ! feed system for any reason will not adversely affect the SSW or the UHS. His modification does l ! not impact the capability of the SSW or the UHS to function. Therefore, this modification do(s
;. not increase the consequences of an accident or a malfunction of equipment important to safet) l previously evaluated in the SAR. The SCT chemical feed system does not introduce any new failure modes for the SSW or the UHS. Therefore, this modification does not create the possibility J ! of an accident or a malfunction of equipment important to safety different from any presioup.y l cvaluated in the SAR. This modification does not affect any technical specification or the tasis of
- any technical specification. Therefore, this modification does not reduce the margin of safety as
, defined in the basis of any technical specification. For these reasons, this modification does not
' constitute an unreviewed safety question.
1 4 Section 1' Page 73 ofI87
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l l l l 1
.I Channe Number /USAR Section: LCN 9.2 285 Figure 9.2-le l
' Description and Basis for Channe: I i This modification will provide a method to reject excessive water from the standby cooling tower. .j to the circulating water system (CWS) fiume. He new section of piping will originate at the , standby service water (SSW) tower and, using an electric pump, route water from SSW to the l CWS flume. The modification will allow Operations to have a permanent and reliable method to reject excess water from the standby cooling tower. Summary of Safety Evaluation: l l De new equipment will not be required to provide any safety related function. The design of the
. proposed system does not represent any changes to the existing safety related functions provided by :
the standby cooling tower. The initiating events of this modification and the plant equipment .; response to previously analyzed accidents will not be altered as a result of this change. The source ; of electrical power will not threaten, or introduce additional risks or loads to the electrical t distribution system used for safety related equipment or components. The modification will not affect the ability of the standby cooling tower or other safety related equipment to properly respond to previously analyzed transient or accident scenarios. By using existing piping sections: no new penetrations are required; makes it impossible to lower the standby cooling tower basin below TS minimum requirements; and no piping enters the plant area to cause re-evaluation of seismic or other accident scenario analysis. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. Since it will not be possible to decrease the basin water level below Technical Specification . minimum requirements, the change will not alter the performance of the standby cooling tower nor will the modification affect the plant's response to previously analyzed transients or accident , scenarios. Also since the transfer of water is from a non-radioactive system to another non-radLactive system, failure of the equipment installed by this modification will not introduce new pathways for radioactive effluents, nor will there be an increase in the presiously analyzed l radioactive effluent release pathways. Therefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. The modification is confined to areas outside the plant and, as such, the modification does not introduce any high pressure systems, represent equipment or services which will be used for the I generation, transportation, or storage of radioactive materials or fluids, nor will it be required to provide a safety related function. Based on this and above stated information, the proposed activity will not alter the plant's response to previously analyzed accident or transient scenarios, nor will the proposed activity create the possibility of an accident of a different type than presiously , evaluated in the USA. Therefore, this modification does not create the possibility of an accident i different from any previously evaluated in the SAR. I
%c modification introduces no new risks or loads to the electrical distribution system supplying
. safety related equipment or components. The pipe routing has been designed to eliminate the z introduction of potential flooding scenarios. %e proposed pump-down system will not affect the probability of a malfunction of the standby cooling tower, since there is no impact to water inventory in the basin. De operating parameters for the proposed equipment have been selected such that no new high energy systems or conditions are introduced. As such, installation, Section 1- Page 74 of 187 1 J
operation, or complete failure of the equipment installed by the proposed activity will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated and documented in the SAR. The equipment being installed by this modification will not be required to provide a safety related function, nor will the new equipment be required to function in order to mitigate the consequences of other evaluated equipment malfunctions or failures. Complete failure of the equipment being installed by the proposed activity will not increase, or otherwise alter, the quantity or concentration of radioactive effluents released from RBS. Based on this and above stated information, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the S AR. The proposed activity has several features which will minimize the potential failute of the new equipment, or minimize potential adverse affects experienced should the new equipment fail while in service. The existing suction piping was installed using seismically qualified support to ensure that piping will not fall into the standby cooling tower and affect the operation of the standby cooling tower pumps. The seismic interaction between the existing standby cooling tower wall and new pump slab has been evaluated and there is no impact to the seismic qualification of Standby Cooling tower wall. The qualification calculation for existing supports has been resised to included the addition of this modification and all supports remain qualified. Based on this and the information provided above, the proposed activity will not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated and documented in the USA. The new equipment installed by this modification will not reduce the margins of safety previously established for and associated with maintaining a certain water level in the standby cooling tower basin. The modification, being routed through areas other than in-plant areas, does not represent any new scenarios for in-plant flooding events. Therefore, the margins to safety associated with safety related equipment operability or availability as a result ofin-plant flo > ding are not affected by this modification. Therefore, this modification does not reduce the marg n of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 9.3-171 Tables: 3.2-1 9A.2-36 uf Figures: 1.2-2 1.2-50 9.3-I f Description and Basis for Chance: Installation of a permanent diesel-driven air compressor, air dryer, after cooler, and air receiving tank have been connected to the IAS and SAS headers in the turbine building. These components were originally temporarily installed to provide back-up air supply whenever the IAS and SAS air compressors are not available. Permanent installation includes a battery charger to keep the engine battery charged at all times for starting the engine. Section i Page 75 of 187
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Summary of Safety Evaluation: 1 l he existing air compressors, air dryers and filters are designed as "other equipment" in the SAR and are shown as non-safety-related and non seismic. There are no evaluated accidents in the SAR which are caused by the failure of the IAS system or equipment. The diesel driven air compressor I provides an alternate source ofinstrument air to the plant systems during the unavailability of clectric driven air compressors or during the loss of offsite power. Since there are no evaluated accidents caused by failure of the 1AS or SAS systems, the portions of those systems that perform i safety related functions are not impacted and there is no adverse electrical impacts postulated. He i proposed modification does not increase the probability of occurrence of an accident previously evaluated in the SAR. He diesel driven air compressor has no function to mitigate the consequences of an accident, the only IAS components rdied upon for mitigating the consequences of an accident are safety related air accumulators, air tetes, isolation valves and piping located in the control, auxiliary and fuct buildings. He proposed modification has no impact on the operation of the safety related air accumulators or components. Therefore, there are no increase in the consequences of an accident evaluated previously in the SAR. He dietel air compressor is connected to the IAS and SAS system headers through check valves. This arrangement prevents the depressurization of the systems in case or a line break in the proposed installation. Therefore, ! the proposed activity does not increase the nossibility of an accident that is different than any previously evaluated in the SAR. The safety related components will continue to receive quality l j l air and therefore the proposed change will not impact their function, reliability, operability, and availability. So, the probability of a malfunction of a safety related SSC previously evaluated in the SAR will not be increased. The modification is limited to the non-safety portion of the system l and provides back-up only. Herefore, the consequences of a malfunction of a safety related SSC previously evaluated in the SAR are not increased. Since this modification provides a permanent > back-up to the non-safety portion to the IAS/SAS system it does not create the possibility of a malfunction of a safety related SSC different than that already stated in the SAR. There are no technical specifications affected by 6is modification and therefore, the margin of safety to any technical specification is not reduced. Therefore, this change does not constitute an unresiewed l l safety question. Figures 9.3-7a l Cnance Number /USAR Section: LCN 9.3-194 9.3-7k Description and Basis for Chance: q The existing impellers are replaced on nine auxiliary building floor drain sump pumps, four of which are classified as safety related sump pumps. He existing two HP motors on two of the non-safety-related auxiliary building sump pumps are replaced with five llP motors. Flow restricting orifices are installed on the discharge lines of two containment floor drain sump pumps, two pedestal floor drain sump pumps, and four auxiliary building floor drain sump pumps. These changes are made as the result of a design adequacy review which identified ways of getting the j pumps more effective, j i Summary of Safety Evaluation: j There is no change in the operating mode of the reactor plant floor drain (DFR) sump pumps. The l DFR sump pump sizing and flow balance calculation was reviewed considering the design changes Section I Page 76 of187 I
)
i and concluded that the intended function of the DFR sump pumps will not be affected. The existing pipe routing the instrumentation, and the controls for the subject sump pumps wi!! not be altered and does not adversely affect any safety related structure, system, or component (SSC). I ne original design intent is unchanged. He safety related sump pumps that are modified are ( designed to handle the impeller size change. While the other affected pumps are not classified as safety related, the DFR system does perform some important functions such as room flooding ! mitigation and reactor coolant pressure boundary leakage detection. The changes to the sump ; pumps improve system performance and have no impact on cubicle integrity, cubicle to cubicle ; leakage, or sump level indication. Therefore, these changes do not increase the probability of an j
; accident or a malfunction of any safety related SSC previously evaluated in the SAR. The increase in pump impeller size will increase the pump discharge head and enhance the ability of the pumps ,
- to perform their function. This change will increase the design margin for these pumps. Normal
operation of the pumps will remain unchanged. He failure of a pump will result in the same , consequence regardless of these changes. Normal and abnormal operation of the DFR system i remains unchanged. Therefore, these changes do not increase the consequences of an accident or a malfunction of any safety related SSC previously evaluated in the SAR. He upgrade in motor size ; for two of the auxiliary building floor drain sump pumps to a 5 IIP,6.8 full load amps will not ; i have an impact of the cable and not change in cable size is required. New overload heaters and trip coils sized for the 5 HP motor will be installed. The replacement impellers are the same material ! and design as the existing impellers. The replacement impellers will allow the subject pumps to , overcome the additional initial friction head from the spring loaded piston check valves. The flow j restricting orifices are designed to the codes and standards for the existing piping system. Increases in size and weight have been evaluated and found to be acceptable. No new or different failure modes are added and no new or different system interfaces are added by changing the i subject pumps. Therefore, these changes do not create the possibility of an accident or a malfunction of any safety related SSC different from any previously evaluated in the SAR. These l changes have been evaluated and will not impact the performance of the sump level monitoring ! system. These changes will provide additional margin for the surveillance requirement for pump capacity. Therefore, these changes do not reduce the margin of safety as defined in any technical specification or in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. 3 Change Number /USAR Section: LCN 09.03-197 Pages: 1.2-16 1.2 18 9.3-1 9.3-3 : 9.3-4 9.3-6 i Tables: 3.2-1 9.2-17 9A.2-36 Figures: 1.2-2 1.2-4 1.2-33 1.2-37 ! 1.2-49 9.2-7d 9.3-l a , 9.3-1 g Section I Page 77 of 187 : i
9.3-1h i, 9.3-2a 9.3-2e 9.3-2f 9.3 4a
'9.3-6j 9.3-16e 9A.2-2A 9A.2-2B Description and Basis for Channe:
- his change modifies the instrument air system (IAS) and the service air system (SAS) to provide a more reliable system of providmg the plant's compressed air needs ne modification involves constructing a new building adjacent to the side of the turbine building. . He building will be an open, steel structure, hm=d~i by the turbine building on one side and open air on the other three sides. The new building houses six new air cooled rotary screw air compressors with trim coolers
' and moister separators, two new air dryers with prefilters and after filters. The six air compressors
. will be divided up evenly to provide compressed air to the SAS and IAS with the ability to switch more compressors over to the IAS when needed ne new air compressor piping enters the turbine building and connects to the existing IAS and SAS headers. In addition four new 480-volt circuit breakers will be installed to provide power to four air compressors, and two exiting circuit breakers will be converted to 300 AMP breakers to provide power to the remaining two compressors.
Summary of Safety Evaluation: The function of the IAS is to provide clean, dry air for plant instrumentation and controls, and the SAS is daig =I to provide clean air for plant services. This modification which provides new IAS and SAS air compressors and associated equipment does not effect safety-related structure, system, or component; and failure of the equipment or system has not been evaluated in the SAR as a postulated design basis accident or operation transient. The implementation of this modification in four phases ensures that the IAS and SAS systems remain operational during the construction period, therefore this modification does not increase the probability ofloosing instrument air which
~may or may not affect accidents already addressed in the SAR. The instrument air system prosides ASME Section III Safety Class 3 air accumulation tanks in the control building to maintain and ensure control room habitability for an indefinite period for the safe-shutdown of the plant in case of LOCA. De new system will provide air compressors from the SAS, in the event IAS losses header pressure. There are also no accidents evaluated in the SAR that rely on IAS or SAS for accident mitigation therefore, there will be no increase in the consequences nor probability of an accident previously evaluated in the SAR. There is also no possibility of this modification introducing the possibility of a malfunction of a safety-related structure, system, or component
- different than any previously evaluate in the SAR. His design change does not reduce the systems' performance, capacity but increases reliability. There are no technical specifications that are affected by this modification to the IAS and SAS. Therefore, the margins of safety as defined in the basis of any techmcal specification are not affected.
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- Section I Page 78 ofI87
Channe Number /USAR Section: LCN 9.3-205 Figure 9.3-6j Description and Basis for Channe: His modification joins the 1.5" with the 2" drain line from the lube oil purifier PURI which is occasionally contaminated with oil, and it reroutes the 2" line into the nearby floor drain hub from where a 4" oil drain line leads to the waste oil sump. His involves deletion of a small bore support, addition of a small bore support, and rerouting a .75" skid-mounted pipe out of the way of the 2"line. His modification effectively prevents purifier oil from entering into the water sump. Summary of Safety Evaluation: The affected floor drains were not involved in the assessment of any accident. Therefore, this modification does not increase the probability or the consequences of an accident presiously evaluated in the SAR. There is no safety related equipment in the affected area. Therefore, this modification does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the SAR. His modification eliminates a water contamination problem in the water system and adds water the oil sump which contains some water already from other sources. There is no shutoff valve downstream of the changed piping, so that the water can not back up into the piping. This modification will not result in any new interfaces with safety related equipment. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the S AR. He improved technical specifications do not address the water and oil carried by the affected lines because the lines do not carry anything that is radiologically contaminated. , Therefore, this modification does not reduce the margin of safety as defined in the basis of any l technical specification. For these reasons, this modification does not constitute an unresiewed ' safety question. Channe Number /USAR Section: LCN 9.3-206 Pages: 10.4-11 ; 10.4-13 I Figures: 9.3-8d 10.3-Ic 10.4-7k Description and Basis for Change: This change isolates the radwaste reboiler from all interconnecting systems in order to prevent any future unwanted pressure buildup in the reboiler. All drain and steam outlet lines have been isolated, llowever there will be overpressure protection for both the shell and tube side of the reboiler provided by two relief valves. His will prevent the reboiler equipment from adversely afrecting the operation of the plant. Summary of Safety Evaluation: The radwaste reboiler was designed to provide steam to radwaste equipment and an alternate i source of gland sealing steam. In isolating the reboiler, blinds and restrictive orifices will be installed and the closing of valves will be used to decommission the reboiler. Upon decommission, this alternate source of scaling steam will be removed. However, Section 10.4.3.3 of the SAR evaluates the complete loss of gland scaling, and determines that a relatively small amount of steam Section i Page 79 of I87 l
I I l would leak out of the high pressure gland before air in leakage through the low-pressure turbme j glands, would cause a turbine trip, and MSIV isolation due to high condenser pressure. The turbine trip would preclude any substantial release of radwaste to the environment. Both turbine trip and MSIV isolation are evaluated in the SAR and this modification does not increase the , probability or consequence of these and previous SAR evaluated accidents. With respect to l radwaste equipment the isolation of the radwaste reboiler will not create any accidents which have : not been evaluated in the SAR because, the reboiler was designed to provide steam to radwaste equipment that is not in operation. The gland scaling system has been evaluated in Section , 10.4.3.3 of the SAR and thus the isolation of the reboiler would not create any accident that has j not been evaluated in the SAR. The reboiler is located in the turbine building away from safety l related structures, systems, or components and it's primary function is to back up a non-safety-related system, which has been evaluated for its worst case failure mode in the SAR. Therefore, ; this modification will not increase the consequences or possibility of a malfunction to a safety related structure, system, or component. In the event gland seals were unable to prevent air leakage in to the condenser through the low pressure turbine or steam leakage to the turbine , building through the high pressure turbine, air leakage in to the main condenser will be monitored i in the off gas system in accordance with Technical Specification 3.11.2.4. Steam leakage in to the 4 turbine building would be monitored in accordance with Technical Specification 3.3.7.11. This . modification does not require changing any existing River Bent technical specification. For these ! reasons, this modification does not constitute an unreviewed safety question. ; Change Number /USAR Section: LCN 09.03-207 Page 8.3-72 a i Figures: 4.6-5A 9.2-1 B 9.3-4B 9.3-10C 9.5-2 D 4.6-5C 9.2-1C 9.3-6A 9.3-11 9.5-11 5.1-3A 9.2-1D 9.3-6E 9.3-16A 10.3 l A 5.1-3B 9.2-1E 9.3-6F 9.3-16B 10.3-1 B . 5.1-3C 9.2-1 F 9.3-6G 9.3-16D 10.3-1C 1- 5.4-2A 9.2-2A 9.3-7A 9.41A 10.3-1 D 5.4-2B 9.2-2 B 9.3-7B 9.4-1 B 10.4-1 5.4-8 9.2-7A 9.3-7C 9.4-1 C 10.4-2 ; 5.4-12A 9.2-8A 9.3-7D 9.4-2A 10.4.3A 5.412B 9.2-8D 9.3-7E 9.4-2 B 10.4.3C 5.4-12C 9.2-8 F 9.3-7F 9.4-3A 10.4-6A
- 5.4-15A 9.2-8G 9.3-7G 9.4-5 10.4-6B l 5.415B 9.2-811 9.3-711 9.4-6A 10.4-7A l 6.2-58 9.2-8J 9.3-7J 9.4-6C 10.4-7B 6.2-66 9.2-21 A 9.3-7K 9.4-6D 10.4-7C 6.2-73A 9.2-21C 9.3-7L 9.4-7A 10.4-7D 6.2-73B 9.2-3 B 9.3-7M 9.4-78 10.4-7E 6.3-1 9.3-1B 9.3-7N 9.4-7C 10.4-7F 6.3-4 9.3-1 C 9.3-8A 9.4-7D 10.4-7G 6.7-IA 9.3-1 D 9.3-8B 9.4-7E 10.4-7H 6.7-1 B 9.3-1 E 9.3-8C 9.4-8 10.4-7J 6.7-1C 9.3-2C 9.3-8E 9.5-1C 10.4-7K 6.7-1 D 9.3-3D 9.3-9 9.5-2A 9.1-23A 9.3-3E 9.3-10A 9.5-2B :
9.2-IA 9.3-4A 9.3-10B 9.5-2C j l Section I Page 80 of I87
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Description and Basis for Channe: This change revises the process for numbering components due to the creation of the component database (CDB) which will replace the Q-list. The asterisk (*) is being removed from all existing and all future component numbers. In the past, the asterisk had been used to identify nuclear i safety related components. De asterisk will be changed to a dash (-) and will only be used as a { field separator, his change is being made for several reasons The asterisk has not been used l consistently in many databases and documents. The asterisk has been used incorrectly to indicate , l components inspection requirements instead of quality class. This change will also force personnel j to use the CDB to find the correct quality classification of a component. i Summary of Safety Evaluation- ! l This is a procedure change and does not add or delete equipment in the field nor does it affect the ! ability of equipment to perform its intended function. The identification of components without an i asterisk to indicate safety-related requires people to go to the CDB and not to depend on equipment tag which may be inconsistent in the use of the asterisk and the dash. The replacement of the dash j for the asterisk does not create the duplication of mark numbers because the Q-list did not allow ; the use of mark numbers which were identical in all respects but the asterisk or the dash. The ; i CDB will contain a field called " ops' description"_which will provide further clarification for operators when the CDB's description is different from those shown on the tag and in a procedure. Pmviously used mark numbers will be cross-referenced with the currently used component ] numbers in the CDB. Therefore, this change does not increase the probability of an accident or a ! malfunction of equipment to safety previously evaluated in the SAR. Equipment that was required j to be safety-related will continue to be safety related. Therefore, this change does not increase the j consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. This change does not create the possibility of an accident or a malfunction l of equipment important to safety different from any previously evaluated in the SAR. This i procedure change does not add or delete equipment in the field nor does it affect the ability of j equipment to perform its intended function. Any components appearing the technical specifications i by the old mark format will be cross-referenced to is corresponding component number in the l CDB. Therefore, this change does not reduce the margin of safety as defined by any technical specification or the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. j 1 l Channe Number /USAR Section: LCN 9.3-210 Figures: 10.3-la 10.4-2 5.4-12c 9.3-7e 10.3-Ic 6.3-1 9.3-8b 9.3-7j 10.4-7c 6.3-4 9.3-8c 9.3-7k 10.4-7d 5.4-8 9.3-8d 9.3-7n 10.4-7j 5.4-12a 9.3-8e 7.5-11 10.4-7k 5.4-12b 9.3-7d 7.6-1 1 I Section 1 Page 81 of187
l Summary of Safety Evaluation: His modification does not affect any equipment postulated to cause any accident. He leakoff lines will be either capped or plugged according to plant procedures using appropriate materials for the applicable valve. This
- vill maintain the integrity of the pressure boundary within the design I
pressure and temperature rating. Cutting and capping or removing the leakofflines and all associated components will not afTect th: fit, form, or function of the valves, or any other equipment important to safety. The equipment and components to be deleted are not safety related i and not required for the safe shutdown of the plant. Therefore, this modification will not increase l the probability of an occurrence of an accident or a malfunction of equipment important to safety evaluated previously in the SAR. Any future valve packing leakage will be defmed as unidentified l leakage and will be collected in the drywell floor drain sumps. This modification will not affect any leakage rate limits. The unidentified leakage rate is sufficiently low that corrective action could be taken before the integrity of the barrier would be threatened. Therefore, this modification will not increase the consequences of an accident or a malfunction of equipment to safety evaluated ; previously in the SAR. He GE faceplates that will replace the lights and switches on panel lil3*P870 will maintain the structural integrity and seismic qualification of the panel. His , modification removes unnecessary, non-safety-related equipment and instrumentation according to existing plant procedures and specifications. The equipment and instrumentation are not safety related and are not required for the safe shutdown of the plant. Therefore, this modification will not create the possibility of an accident or a malfunction of equipment important to safety of a i difTerent type than evaluated previously in the S AR. This modification does not alter the leakage rates in any technical specification. He deleted equipment is not included in the basis for any technical specification. Therefore, this modification will not reduce the margin of safety as defined in the basis for any technical specification. For these reasons, this modificatica does not constitute an unreviewed safety question. : Chance Number /USAR Section: LCN 09.03-211 Page 5.2-45 Figure 9.3-7f Description and Basis for Chance: j i Inadequate breaker protection for the electrical containment penetration was discovered during a l review of Calculation E-190 and MR 87-0837 FCN 1. This modification will temporarily spare the power cables to E31-FYN021-1 in SCV-PNL2Bl. l Summary of Safety Evaluation: l l Redundant protective devices are currently installed to ensure that the Conax penetration conductor ! i is not damaged under fault conditions. The temporary sparing of the electrical containment penetration for E31-FTN021 power will ensure no possibility for damage by a fault condition. Sparing the power cable in the SCV panel exceeds the SAR requirements. Therefore, the probability of an accident previously evaluated in the SAR is not increased. Primary containment ) penetration conductor overcurrent protection devices ensure the pressure integrity of the l containment penetration. With failure of the device it is postulated that the wire insulation will j degrade, resulting in a containment leak path during a LOCA. The protective devices are not i meant to provide circuit continuity under fault or overcurrent conditions. They are designed to protect the circuit conductors against damage or failure due to ove current heating effects and Section i Page 82 of187 j
subsequent penetration failure. The only safety function of the protective devices is to proside the overcurrent protection to preclude containment penetration degradation. Herefore, the consequences of an accident previously evaluated in the SAR is not increased. The 120V ac electrical circuits pass through low voltage electrical penetrations that form part of the containment boundary. De-energization is an allowable method of penetration protection. The lifted leads will ensure the penetration is de-energized. There is no failure mode created by this repair that will result in a condition which has not already been analyzed. Herefore, the possibility of an accident which is difTerent than any previously esaluated in the SAR is not created. Since the function of the overcurrent protection devices is performed by isolating the device to prevent degradation of the wire insulation, no containment leak path will occur. Sparing of the power connections and electrical penetration for an existing inoperative instrument does not increase the probability of a malftmetion of a safety related structure, system, or component (SSC) previously evaluated in the S AR. The radiation effects at the site boundary will not be changed by this repair. Therefore, the consequences of a tr.alfunction of a safety related SSC previously evaluated in the SAR will not be increased. The possibdity of a malfunction of a safety related SSC different than any previously evaluated in the SAR is not created. The proposed repair does not afTect the basis to any technical specification. Therefore, the margin of safety as defined in the basis to any technical specification is not reduced. For these reasons, the modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 9.3-214 Figures: 9.3-8c 10.3-1 c Descrir tion and Hasis for Change: This change is to reflect the modifications made by MR 95-0033. This MR modifies MSR high load and low load valves (1 MSS-PVRSHL Vl/V2 and 1 MSS-PVRSLL Vl/V2) by retrofitting existing valves with components designed for severe service applications. A retrofit kit supplied by Control Components incorporated (CCl) will be installed in these valves. The retrofit kit includes the valve intemals, bonnet, and valve actuator. Additionally, in accordance with the RBS valve stem packing improvement program, the new valve bonns are being supplied without leak-ofTline connections. Existing leak-offlines are being physically removed up to their respective drain header, and components contained in these lines will be removed. The header is to be cut toward the condenser beyond the point where the drain lines connect and are then capped. This is in lieu of replacing the header pipe with FAC resistant material, i Summary of Safety Evaluation: The purpose of this modification is to retrofit the existing non-safety-related MSR high and low ; load valves so they can perfonn their intended function under severe conditions without failure. j This modification meets or exceeds the system design specifications. The function of the valves remains unchanged and the leak-offlines are no longer required due to the new valve packing arrangement. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety presiously evaluated in the SAR. This modification will not afTect the operations or the reliability of any safety related valves or systems required for safe shutdown of the plant. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated Section I Page 83 of 187 i
. . . _ -_ _ _ . . _. . _ _ _ _ _ _.. _ _ _ _ . _ _ . _ ~
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in the SAR; nor will it reduce the margin of safety as defmed in the basis of any technical L specification. For these reasons, this modification does not constitute an unreviewed safety question. 4 Channe Number /USAR Section: LCN 9.3-218 Page 9.3-25 Description and Basis for Channe:
, High conductivity alarms from' three turbine building sump pumps continuously masked the j high/ low sump condition. High conductivity is a nonnal, rather than an alarm, condition for the '
turbine sumps. Under a previous modification, the leads from the associated conductivity indicating transmitter switches were lifted and taped. Dese wires were tagged as " lifted leads" at ! that time. This modification is performed to ensure that three wires that were lifted and taped are i retagged to indicate they are permanently " lifted and taped" to inhibit the high conductisity alarms. f Summary of Safety Evaluation: I i { l De turbine building sump high conductivity alarm condition is not evaluated in the SAR, with 4 respect to causing an accident. Also, by removing these alarms the probability of flooding of the ! ( turbine building sumps will be decreased. Therefore, this modification does not increase the l l probability of an accident or a malfunction of equipment important to safety previously evaluated { in'the SAR. There is no previously evaluated accident in the SAR with reference to the turbine l building sump high conductivity, and there is no equipment involved in the turbine building sump , j conductivity circuits that is safety related or important to safety. Therefore, this modification does' { not increase the consequences of an accident or a malfunction of equipment important to safety l ,' previously evaluated in the SAR. This modification has lessened the potential for sump overflow. l l Also, the alarm circuit is independent of equipment operation in accordance with Reg. Guide 1.75 l t and will not cause equipment to operate or not operate due to the designed electrical isolation. l' Derefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The sump
- conductivity alarms are not defined, nor are they discussed as part of any technical specification or j
- technical requirement, due to their design conformance to Reg. Guide 1.75. Therefore, this i modification does not reduce the margin of safety as defined in the basis of any technical j specification. For these reasons, this modification does not constitute an unresiewed safety 1 question. l 4
n Channe Number /USAR Section: LCN 9.4104 Figure 9.4-7e Description and Basis for Channe: This change added thennocouples 1LDS-T/C037A,1LDS-T/C037B, ILDS-T/C040A, and 1LDS-1 T/C040B to reactor water cleanup (RWCU) pump rooms 1 and 2. Dese thermocouples are not
- safety related. These heat sensors provide independent verification of the accuracy of temperature i signals from safety related leak detection thermocouples ILDS-T/CNO37A, ILDS-T/CNO37B, ILDS-T/CN040A, and 1LDS-T/CN040B.
i Section i Page 84 of187 7
i Summary of Safety Evaluation: , i To prevent movement and keep from being a missile hazard during a seismic event, the new ! thermocouples were properly installed and secured. This addition of these thermocouples does not j increase the probability or consequences of an accident previously evaluated in the SAR. The possibility of an accident not previously evaluated in the SAR will not be increased because this modification does not functionally change any plant operating systems, pressure boundaries, or j barriers to the release of fission products. Malfunctions other than those presiously evaluated will j not occur because the thermocouples are not safety related and do not affect any plant processes. l These thermocouples do not affect the ability of the existing instrument channels used for leak detection. Therefore, the margin of safety as defined in the basis of any technical specification is j not reduced. Consequently, this change does not constitute an unreviewed safety question. ; i Channe Number /USAR Section: LCN 9.4-123 Pages 9.4-77 i 9.4-83 Table 9.4-10 sh. 4 Figures 9.4-6b 1.2-45 i 1.2-47 l Description and Basis for Channe: ! l This change removes the split cooling unit condenser (lHVY-CURL) associated with the makeup l water intake pump house battery room air conditioning unit (lHVY-ACUI). Deletion of this : condensing unit climinates the air conditioning for the battery room in the makeup intake water structure. The battery (IBXY-BAT 01) located in that room provides control and switching power for equipment in the make up water intake pump house. The cooling coils will remain in place. Only the rooftop condenser is being removed. The primary reason for the deletion of this l condenser is that it has been unreliable and a continual maintenance problem, apparently due to i oversizing. All the equipment affected by this change is non-safety-related. - Summary of Safety Evaluation: ! This change only affects the makeup water intake pump house battery room cooling, which will impact only the service life of the non-safety-related battery. - The decreased battery life will not l warrant any additional change out of the battery over the life of the plant. The battery's capacity I will not be decreased as a result of the temperature exceeding the design temperature for the room. ; Therefore, this change does not increase the probability of an accident presiously evaluated in the SAR. He battery is not required for mitigating the consequences of any evaluated accident , condition. Therefore, this change will not increase the consequences of an accident presiously evaluated in the SAR. His change does not have any effect on the operation of any other equipment. Therefore, this change does not create the possibility of an accident different from any
. previously evaluated in the SAR. The battery involved is important to plant reliability, but is not ,
important to plant safety. His change does not involve or impact any safety related system,
- structure, or component (SSC). Therefore, this change does not increase the probability of a malfunction of any safi:ty related SSC. This change does not increase the consequences of a ;
malfunction of any safety related SSC. His change does not create the possibility of a malfunction of any safety related SSC. This change does not affect the margin of safety for any
'Section 1 Page 85 of187 l
. - -- - . - . - - . . =-
i technical specification. Therefore, this change does not reduce the margin of safety as defined in l any technical specification. For these reasons, this modification does not constitute an unreviewed ; safety question. , Char.ne Number /USAR Section: LCN 9.4-125 Page 9.4-78 Tables: 9.4-10 Sheet 4 9.4-10 Sheet 5 Figure 9.4-6a Description and Basis for Channe: j He electrical and piping tunnel ventilation system supplies the electrical and piping tunnels with fresh outside air to remove the heat generated by cables and piping. The ventilation system also mainains design ambient temperatures. No filtration system has been provided for any of these ventilation systems and there have been problems due to the accumulation of dust over time, which includes housekeeping and personnel safety problems. This change added inlet filters to seven of the eight fans for the electrical and piping tunnels. . Summary of Safety Evaluation: q , The addition of these inlet filters has the potential to slightly increase tunnel temperatures. This increase in temperature will not decrease the qualified life of the electrical cable insulation in the tunnels as design temperatures are not exceeded. Consequently, this change does not result in any effect on the operation of any other equipment. All new equipment is adequately supported to preclude seismic interactions with safety related equipment. Therefore, this change does not increase the probability of occurrence of an accident evaluated previously in the SAR. The electrical and piping tunnel ventilation system is not required for mitigating the consequences of any evaluated accident condition. The slight increase in temperatures does not result in any affect on the operation of other equipment in the electrical and piping tunnels. Therefore this change does not increase the consequences of an accident evaluated previously in the SAR. Considering the discussion above, this change will not create the possibility of an accident different from any previously evaluated in the SAR. This change has no effect on the operation or capacity of any safety related system, structure, or component. This new equipment is either located away from safety related equipment or restrained to prevent contact during a seismic event. Therefore, this change does not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. His change does not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change does ) not create the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. The technical specifications were reviewed in detail and it was determined that the electrical and piping tunnel ventilation systems are not addressed in the technical specifications. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l Section i Page 86 of 187
Change Number /USAR Section: LCN 9.4-126 Table 9.4-4 Description and Basis for Chance: The existing cooling capacities, located in the auxiliary building, listed in Table 9.4-4 are replaced with the required design cooling capacities. The SAR values were extracted from the original Purchase Specification 215 252 which reflects a larger margin than is required. The revised design cooling capacity includes the actual calculated design requirements with adequate design margin per Design Calculation PB-30. There is no physical change and cooler capacities are not affected. Summary of Safety Evaluation: The unit coolers function to remove heat and maintain design ambient temperature in various areas in the auxiliary building during normal operation, loss of offsite power and DBA. The values stated in the SAR reflects a larger margin than is required and does not reflect the required design cooling capacities of the unit coolers. The replacement of the required design cooling capacity will not increase the probability of an accident previously evaluated in the SAR. The cooling capacity and other design parameters for unit coolers are not affected by this change. The resision of Table 9.4-4 with the required design cooling capacity, in lieu of existing cooler capacity, will not increase the consequences of an accident previously evaluated in the SAR. The existing cooler capacity values are being replaced in the table with required cooling capacity which is more than adequate to maintain design ambient temperature. Therefore, the possibility of an accident which is different than any previously evaluated in the SAR is not created by this change. This change only replaces the required design value calculated heat load plus the design margin in lieu of manufactured equipment capacity. The described changes will not increase the probability of a malfunction of a safety-related structure, system, or component previously evaluated in the SAR. This change does not impact the safety-related equipment in the area. Therefore, the consequences of malfunction of
]
a safety-related SSC are bounded by the existing SAR accident analysis and are not impacted by 1 this change. The proposed changes to the table does not create the possibility of a malfunction of a safety-related SSC different from that previously analyzed in the SAR. No technical specification or bases directly applies to the operation of the auxiliaiy building unit coolers. Therefore, the magin of safety as defined in the bases to any Technical Specification will not be reduced by the I proposed changes. For these reasons, this alteration does not constitute an unreviewed safety I question. I l Chance Nmnher/USAR Section: LCN 9.4-127 Table 9.4-1 Description and Hasis for Chance: Calculation G13.18.2.l*058, Revision 0, shows that 1) the range of 20% to 90% relative humidity is expected for the emironment in the fuel building outside of the Fuct Pool area, and 2) no safety-related or non-safety-related components in the fuel building outside the fuel pool area are susceptible to condensation in this environment. SAR Table 9.4-1, which previously showed a humidity range of 40% to 55%, was revised to reflect this new humidity range. Section I I Page 87 of 187
l i l l Summary of Safety Evaluation: The fuel building HVAC system is an accident mitigator, not and accident initiator. Thus, any change to the system would only potentially affect the system's ability to mitigate currently analyzed accidents and would not affect the occurrence of an accident. Also, no physical plant changes were performed by this modification. The relative humidity range in the fuel building was revised based upon new calculated values. Neither the operation of the system nor the method of operation of the system was altered. The modification does not effect the design basis analysis for the fuel building HVAC system to mitigate the consequences of either a LOCA or fuel handling accident in the fuel building. This modification does not affect the operation of any safety-related equipment or equipment important to safety in the fuel building and does nothing that would affect the radiological consequences of an accident at the site boundary. No new accident is introduced by the modification. No margin of safety as defined in the basis for any Technical Specification is affected by this modification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 9.4-128 Figure 9.4-4a Description and Basis for Chance: Turbine building ventilation fan motor winding temperature switches have caused the masking of a i common trouble alarm. This can permit a fan trip to go unnoticed by the operators because the ! alarm circuit does not have re-flash capability. Reworking a failed temperature switch requires motor replacement since the motor windings encase the temperature switches. In two previous 1 cases, these types of temperature switches failed almost immediately after motor replacement. This modification decommissioned seven turbine building fan motor winding temperature switches and , the temperature switch inputs for the common alarm circuit. l Summary of Safety Evaluation l The turbine building ventilation system is non-safety-related, non-seismic, and is not postulated to cause any accident described in the SAR. Therefore, there is no increase in the probability of any j accident previously evaluated in the SAR. This system and the components affected by this ! modification are not used to mitigate any accidents described in the SAR. Therefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. I The alann circuits being modified and the pieces of equipment they monitor are not safety related ! and are not seismically qualified. The deletion of these circuits does not create any situations or failures that can be an initiator of an accident. Therefore, there is no creation of the possibility of an occurrence of an accident of a different type than any evaluated in the SAR. The equipment affected by this modification is in an area that precludes it from affecting equipment important to ! safety. Therefore, there is no increase in the probability of an occurrence of a malfunction of I equipment important to safety. The presence or absence of the motor winding temperature alarm circuits have no effect on equipment important to safety. Therefore, there is no increase in the consequences of a malfunction of equipment to safety These changes will have no affect on the operation of any installed equipment. Therefore, there is no creation of a malfunction of a different type than any evaluated in the SAR. The equipment affected by this modification is neither addressed in any technical specification nor is it tested by any surveillance requirements in any technical specification. Therefore, there is no reduction of the margin of safety as defined in the Section I Page 88 of 187 i
basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. . Change Number /USAR Section: LCN 9.4-99b Figure 9.4-6a Descriotion and Basis for Change: This change completes the modification to better maintain the temperature above 40'F in the standby cooling tower pump and switchgear rooms. Specifically, this part of the modification deals with back draft dampers. Two intake backdraft dampers for the standby service water pumphouse switchgear room A and two intake backdraft dampers for the standby service water pumphouse switchgear for room B are designated as normally closed. Two permanent back draft dampers are added in the exhaust air opening of the standby service water cooling tower pump room. Each of these dampers will only open when it associated fan is in operation, thereby not j letting heat escape in the winter months. Summary of Safety Evaluation: This change was made to ensure sufficient temperatures in these rooms will be maintained during periods of extreme cold. The system safety function was not altered. Therefore, the probability and consequences of a previously evaluated accident or malfunction are not increased. The overall system configuration was not changed by this modification and no new accident scenarios are introduced. No margin of safety of any RBS technical specification is reduced. Hence, this change does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 9.5-65a Figure 9.3-lb J Description and Basis for Change: This change provides dry air from Quality Class 2 instrument air system (IAS) components to the standby dicsci air start dryers, so the desiccant towers will be continually dried when the associated compressors are not running. This prevents moisture accumulation in the desiccant towers which results from inadequate operational purging due to repeated short cycle dryer operation. Summary of Safety Evaluation: i Loss of diesel starting air is only postulated to result from the loss of an air receiver or other safety related component. The dryers are part of a non-safety-related support system for the air receivers. ! Failure of a dryer alone would not affect an air receiver. All parts of this modification are built to the same structural, system and electrical requirements as all of the other seismic, non-safety- i related components in the area. Therefore, this change does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Since l this modification does not interface directly with either diesel generator, it does not change the l consequences of a loss of a diesel generator. Loss of power and subsequent loss ofIAS air supply ) results in plant wide occurrences whose scope and magnitude are not changed by this modification. 1 I Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety praiously evaluated in the SAR. This system does not make any cross connection into safety related system. A break in the supply tubing would produce the same Section I Page 89 of187 I
results as any other loss ofIAS and could be less due to the localized nature of this modification. , No addition failure modes affecting safety related functions can be introduced since no safety related components are affected. Therefore, this modification does not create to possibility of an accident different from any previously evaluated in the S AR. No new failure modes are created , relative to a loss ofIAS air supply. This modification is designed so that multiple means are provided to ensure that the IAS is not subjected to too much pressure. This modification is also designed to prevent any single failure from compromising or attempting to compromise the IAS. Herefore, this modification does not create the possibility of a malfunction of equipment important j to safety different from any previously evaluated in the SAR. The technical specifications and their bases do not address non-safety-related components of the IAS or the diesel air starting system. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed , safety question. Change Number /USAR Section: LCN 9.5-076 Pages: 9.5-50 9.5-51 Figure 9.5-2c Description and Basis for Change: This modification adds a check valve in the manual prelube line for the Divisions. I and II diesel generator (DG) turbocharger to prevent the lube oil keepwarm "Before & after" engine-run pump from spinning backward when the manual prelube shutoff valve is opened just prior to shutting i down the engine. This modification also revises the pump and alarm's on/off set points which are 4 associated with the three DG fuel day tanks. This modification also revises the set point data sheet , (documentation) for the lube oil pressure regulating valve and the jacket water thermostat. Summary of Safety Evaluation: An accident as defined in NSAC 125 is a nuclear accident involving a release of radioactisity in excess of regulatory requirements, or invohing a potential release. The standby diesel generator's (DG) function is to prevent such accidents by providing reliable electric power to safety-related components. The design of the check valve effectively prevents a failure to open when needed during pre-start or post-stop manual turbo prelubrication. The design and prescribed operation of the check valve ensures that its installation will not adversely affect the reliability of the diesel turbocharger or lube oil system and, in turn, not decrease the reliability of the diesel in carrying out its part of the functions of the plant's safety systems in preventing an analyzed accident. Failure of the check valve to close after the dicsci is started will not adversely affect the diesel turbo because l I the check valve is not required for retention of the turbo oil pressure boundary. The on/off set points for the pump and alarm are moved away from each other so that instrument accuracy drift l cannot cause the set points to cross. The entire operating range has been moved upward (higher tank oil level), to provide more margin against inadvertently letting the oil level drop below Technical SpeciGcation minimum. This modincation did not cause any changes to the existing DG operating procedures. Therefore, installation of the check valve cannot increase the probability of I any previously evaluated accident. 1 The DGs function is to mitigate or minimize the consequences of any accident in the same manner j they function to prevent such an accident, in order for any modification to potentially increase the consequences of an accident, it would have to decrease diesel reliability or affect areas in the plant Section I Page 90 of I87 1 I
outside the DG envelopes. This modification is fully contained within the diesel's system and does not directly affect any SSC outside the diesel building. Each diesel's lube oil system is independent or separate, and the three DGs have no connection to other plant buildings except for normal loads, alarm or control circuits to the control room. His modification changes only instrument set points at devices within the DG building, or in circuits dedicated to each diescl. No changes are made in i any other building or circuit outside the DG systems. Because diesel reliabi'ity is maintained (or improved), the DGs will interact with and support other plant systems as designed. Therefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. To create the possibility of an unanalyzed accident would require this modification to change the way the diesels interact with the systems they support, create a new path for raiioactive release, or create a potential common mode failure. This modification does none of these. The modification makes no changes to the interface between the diesels and any other plant SSC anil cannot create a new path for radioactive release. Separation and redundancy are maintained and the change is fully contained within the DG systems. Therefore, this modification does not create the possibility of an accident different from any presiously evaluated in the SAR. Because this modification is isolated within the diesci systems, the SSCs requiring evaluation can be limited to the DGs, systems directly supporting the DGs or interfacing with them, and the DG building itself. The malfunctions previously evaluated include failure of the diesel to start, run, carry load, or failure of the oil tank level instrumentation circuitry, a structural or seismic failure within the equipment or building, and failure of support systems such as standby sersice water (SSW). The SSW can be ruled out because there is no connection or interface of the two. Other than the tubing, the check valve has no interface with any other part of the diesel building structure (except for the foundation anchor bolts) so, stmeturally, the seismic integrity of the diesel oil system is demonstrated adequate. The change in set points make no physical or circuit changes; , the controls and devices will function as before with improvements in how the operators will use them. None of the changes under this modification can adversely affect the diesel's load carrying capability. Because the manual prelube valve is still required to be closed as before, under operating procedures, positive isolation and the oil pressure boundary will be maintained. In fact, turbo reliability should be increased. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. As discussed above, this modification is isolated to within the diesel systems, and will not create any common failure scenario between redundant diesels or diesel systems. Because this modification does not create any new interface between the diesels and other systems, or change an existing interface, it cannot affect the way the DGs affects other systems. Herefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. A review of the bases proves that the only way to decrease the margin of safety would be to cause a decrease in the reliability of the required components thereby potentially invalidating the associated basis. This modification, however, will not cause a decrease in diesel or diesel system reliability, and will not violate redundancy requirements. It will also not affect any SSC outside the immediate diesel system cnvelopes. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Section I Page 91 of 187
l l l I i I Change Number /USAR Section: LCN 9.5-94 Pages: 9.5-24a i 9.5-26 l Figures: 9.S-2c ; 9.3-8a Description and Basis for Change: This modification adds a level indicator to the Divisions I and 11 standby diesel generatorjacket water standpipes at a location where it can be read by an operator standing on the floor adjacent to I the standpipe. This new indicator will provide water level measurement over the full range of the standpipe. The existing indicator is deleted by this modification. Drain connections are added to the standpipes to provide a means of obtaining periodic water samples for analysis. The new level indicators 1EGT-L124A and 1EGT-L124B will be tapped off of the existing process valves. The process valves will become nonnally open and standpipe water sampling capability will be provided from the instrument tubing drain valves at the instrument stand. Summary of Safety Evaluation: he standby diesel generators and thejacket water standpipe are not postulated to cause any accident. The sole purpose of the new level indicators is to provide operations and maintenance personnel with a means for easily determining water levels in the diesel generator jacket water standpipe. Therefore, this modification does not increase the probability of an accident presiously evaluated in the SAR. The diesel generators' ability to minimize the consequences of an accident is not changed or challenged. Therefore, this modification does not increase the consequences of any accident previously evaluated in the SAR. This modification meets or exceeds the origmal system design requirements. There are no new failure modes created. No new system interfaces or initiating events are introduced. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. The new indicators are seismically qualified and mounted for mechanical pressure boundary integrity. This modification does not cause interaction with other safety related component or equipment in the area. The structural integrity of the jacket water system is maintained. Therefore, this modification does not increase the probability of a malfunction of any safety related system, structure, or coraponent (SSC) previously evaluated in the SAR. This modification will not affect the ability of any safety related system in perfonning its function to mitigate the consequences of postulated accidents and transient events. This modification does not impact the performance of any barriers used to mitigate the consequences of a malfunction of any safety related SSC. Therefore, this modification does not increase the consequences of a malfunction of any safety related SSC previously evaluated in the SAR. His modification does not create any new failure modes. Herefore, this madification does not create the possibility of a malfunction of any safety related SSC different from rny previously evaluated in the SAR. There are no technical specifications that address thejacket water standpipes. Therefore, this modification does not reduce the margin of safety as defit ed in the basis of any technical specification. For these reasons, this modification does not comtitute an unreviewed safety question. Section i Page 92 ofI87
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i i l 4 .I i i Channe Number /USAR Section: LCN 9.5-090 Figure 9.5-lb ! Description and Basis for Channe: i
- i His modification removes the existing fire protection supply and electrical power feed to the l
- . material release facility and adds a permanent fire protection supply and electrical power feed to i the relocated material alcase facility. Due to the location of the new instnament air system and l
. senice air system compressor building, the material release facility is being moved from south to l j north of the oil storage building. He new fire protection supply to the material release facility is :
an extension of the existing fire protection header to the oil storage building. : ] Summary of Safety Evaluation: . 4 he fire protection system (FPS) is non-safety-related and the oil storage building and materials , i release facility do not contain or support any safety related equipment. De change is limited to the ; l fire protection piping serving the oil storage building and material release facility which contains l
- non-safety-related equipinent only. Failure or malfunction of the fire protection piping sening -
I these areas is not postulated to initiate any accidents previously evaluated in the SAR. Since the j j modification only extends an existing suppression system in a non-safety-related area, the failure ; j probability of the affected FPS and surrounding non-safety-related equipment is unchanged by the ! , modification. Herefore, this modification does not increase the probability of an accident or the l ! probability of a malfunction of equipment important to safety previously evaluated in the SAR l
!. He FPS piping sening the oil storage building and material release facility which contains non- i
, safety-related equipment does not function to mitigate the consequences of any accident nor does it support any other structures, systems or components in mitigating accident consequences. ! 4 Inadvertent operation of the suppression system will not adversely impact any safety equipment. ! Herefore, this modification does not increase the consequences of an accident or the consequences ! of a malfunction of equipment important to safety previously evaluated in the SAR. The FPS ; i- supply to the material release facility adds passive components to the fire protection / suppression ; , system in a non-seismic area containing non-safety-related components. He suppression system ; i does not introduce any new failure modes. Here is no significant increase in failure probability
- due to the addition of these components. Herefore, this modification does not create the l l possibility of an accident or the possibility of a malfunction of equipment important to safety !
l different from any previously evaluated in the SAR. He margin of safety of the technical i specifications are not affected by this change as tho ;1roposed modification only adds components i
- that have a passive safety function and does not affect system operation or modify setpoints.
l i Therefore, this modification does not reduce the m argin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. 4 Channe Number /USAR Section: LCN 9.5-94 Pages: 9.5-24a 9.5-26 l Figures: 9.5-2c 1 9.3-8a Section I 'I Page 93 of187
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t i= ! Description and Basis for Channe: 4 l i nis modification adds a level indicator to the Divisions I and 11 standby diesel generatorjacket l' j water standpipes at a location where it can be read by an operator standmg on the floor adjacent to the standpipe; His new indicator will provide water Icvel measurement over the full range of the - l
- standpipe. He existing indicator is deleted b'y this modification. Drain connections are added to . -l
! the standpipes to provide a means of obtaining periodic water samples for analysis. He new level :
1 - indicators 1EGT-L124A and 1EGT-L124B will be tapped off of the existing process valves. The l process valves will become normally open and standpipe water sampling capability will be j provided from the instrument tubing drain valves at the instrument stand. j h Summary of Safety Evaluation: The standby diesel generators and thejacket water standpipe are not postulated to cause any !
- accident. The sole purpose of the new level indicators is to proside operations and maintenance i personnel with a means for easily determining water levels in the diesel generatorjacket water 3 l standpipe. %crefore, this modification does not increase the probability of an accident previously !
E evaluated in the SAR. The diesel generators' ability to minimize the consequences of an accident is ] not changed or challenged. Therefore, this modification does not increase the consequences of any j i accident previously evaluated in the SAR. His modification meets or exceeds the original system ' ! design requirements. There are no new failure modes created. No new system interfaces or initiating events are introduced. Herefore, this modification does not create the possibility of an ! accident different from any previously evaluated in the SAR. The new indicators are seismically 2' qualified and mounted for mechanical pressure boundary integrity, his modification does not , 3 cause interaction with other safety related component or equipment in the area. The structural l [ integrity of thejacket water system is maintained. Therefore, this modification does not increase the probability of a malfunction of any safety related system, structure, or component (SSC) { j l previously evaluated in the SAR. This modification will not affect the ability of any safety related ! system in performing its function to mitigate the consequences of postulated accidents and transient ; ' events, his modification does not impact the performance of any barriers used to mitigate the consequences of a malfunction of any safety related SSC. There# ore, this modification does not : increase the consequences of a malfunction of any safety related SSC presiously evaluated in the : 1 SAR. This modification does not create any new failure modes. Herefore, this modifiestion does : not create the possibility of a malfunction of any safety related SSC different from any previously , evaluated in the SAR. There are no technical specifications that address thejacket water ) standpipes. Derefore, this modification does not reduce the margin of safety as defined in the l . basis of any technical specification. For these reasons, this modification does not constitute an . unreviewed safety question. F , Channe Number /USAR Section: LCN 9.5-% Page 9.5-4 Figures: 9.5-l a 9A.2-6 Description and Basis for Channe: , This modification supplies fire protection utility services for a new low level radw3ste (LLRW) storage facility. De new facility is required to store LLRW generated at Entergy's River Bend Station. De new building is located southwest of the' clarifiers. Fire hydrants are required in the Secten I Page 94 of187
-ru.a w 4 - , , , - . , . , --
~- . - - _ . _ - - .. . - - - . - - --. - - . - - - . . - - -.
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1
- l 1
general area to satisfy fire protection requirements A new 8"-diameter fire protection line will be tied into existing fire protection underground water loop at valve IFPW-V28. This line will be ! composed of PVC, ; 4 Summary of Safety Evaluation: I t , j- , .. l
- De PVC piping will be UL listed and/or FM approved for fire sersice. Ac piping will be l installed per the manufacturer's instructions, existing plant specification and NFPA 24 " Private
~
Fire Service Mains and Deir Appurtenances " :He installation will meet the same acceptance criteria as the existing underground piping. The fire protection wv.er loop being modified is
- neither an initiator of nor related to any of the accidents analyzed in the SAR. Herefore, there is ,
l no increase in the probability of an occurrence of an accident evaluated previously in the SAR. l The fire protection line is not used to mitigate the consequences of any accident. He consequences ; j of a failure of the new piping would be no worse than the consequences of an existing 8"-line i break. Acrefore, this does not increase the consequences of an accident evaluated previously in the SAR. The LLRW facility is not required for plant operation or for safe shutdown of the plant. ; Any failure of the new piping would be isolated from the yard loop just like any failure to occur in j , . any other existing piping being fed from the yard loop. Therefore, this does not create the i possibility of an accident of a different type than any evaluated presiously in the SAR. The : I ' installation only affects the fire protection underground piping, which is all exterior to safety i related structures. The new fire protection line does not interact by either function or physical l . proximity to any safety related components or to any components that are enaW to safety related systems. Therefore, this does not increase the probability of occurrence of a malfunc+n of l equipment important to safety evaluated previously in the SAR. The new piping is not reqmrod for ; safe shutdown of the plant. Therefore, this does not increase the consequence of a malfunction of l i ! equipment important to safety evaluated previously in the SAR. His line has no interface with any safety related systems, structures or components. Therefore, this change does not create the , possibility of a malfunction of equipment important to safety of a different type than any evaluated l
; previously in the SAR. The stmetures, systems, and components affected, introduced, or both are j not referenced in any technical specification. ney do not any affect on interface components, i systems, and structures that are addressed in the technical specifications. Therefore, there is no reduction of the margin of safety as defined in the basis of any technical specification. For these i reasons, this modification does not constitute an unreviewed safety question.
r { i l Channe Number /USAR Section: LCN 9.5-100 Pages: 8.3-16 ! 8.3-17 i 9.5-42 ! 9.5-43 f 9.5-45 l-9.5-46 ) 9.5-47 l Table 3.2 1 ] Figures: 1.2 28 ) 8.3-11 ) 8.3-12 9.5-2d Section ! Page 95 ofI87 L ) l
Description and Basis for Change: This modification replaces the high pressure core spray (HPCS) diesel generator (DG) non-safety-related starting air compressors and air dryers. The purpose of the replacement is to increase the reliability, availability, maintainability, and performance of the operation of the starting air system that supplies air to the safety related starting air receivers. The existing system contains two air start trains, each with an air compressor and refrigerant type air dryer. One train has an electric motor driven air compressor and on train has a diesel engine driven air compressor. Both air compressors are being replaced with electric motor driven air compressors rated at 500 psig. Both refrigerant type air dryers are being replaced with membrane type air dryers. Summary of Safety Evaluation: None of the affected equipment initiates accidents previously evaluated in the SAR. The changes to the IIPCS DG starting air sys'em do not alter operation of any safety related equipment. The new air compressors are comparable to the existing compressors in performance and are rated for 500 psig versus the 250 psig of the existing compressors The new membrane type air dryers are designed to provide -10 F dewpoint air compared to +35 F dewpoint air supplied by the existing ref !gerant type air dryers. New filters are being installed to provide si micron filtration compared to the 15-20 micron filtration of the existing air compressor fdter. These changes will improve the quality of the air supply. The change to the llPCS DG starting air system will not delete or modify the system's protection features, does not affect equipment used to mitigate an accident, and will not downgrade they system's performance necessary for reliable operation of equipment important to safety. This modification will not reduce system redundancy or independence, and will not increase frequency of challenges to equipment important to safety. This modification will not impose increased or more severe testing requirements on equipment. The installation of new equipment will meet or exceed the same design specification as the existing equipment and will be analyzed for seismic considerations. Therefore, this modification does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the S AR. The changes to the IIPCS DG starting air system do not alter the system's response to an accident that the llPCS DG is required to mitigate. All equipment will function as assumed in the accident analysis. This modification does not change the way the llPCS DG interacts with the plant to mitigate accident consequences. The complete loss of the HPCS DG has been analyzed and changes do not affect the redundancy or independence of the electrical power system. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety. The IIPCS DG system itself does not prevent an accident. The function of the electrical power system is to provide sufficient capacity and capability to ensure that the equipment important to safety performs as intended. The reliability of the diesel generator affects the reliability of the electrical power system but not that of the offsite power source. The changes to the llPCS DG s*.arting air system air compressors and air dryers do not change the design of the diesel generater logic or affect the diesel generator operation. The changes to the llPCS DG starting air sysiem do not introduce any new failure mechanisms for the llPCS DG which would lead to failure modes of a different type than any previously analyzed in the SAR. Loss of air to the air receivers would not prevent the diesel generator from responding to an autostart signal since the air receivers store suflicient air to start the diesel five times without recharging them. Loss of a diesel generator is part of the design basis. Additionally, a total loss of ac power has been analyzed and addressed. Therefore, this modification does not create the possibility of an accident or malfunction of equipment important to safety different from any previously evaluated in the SAR. Implementation of this modification to the 11PCS DG starting air system will not affect the Section i Page 96 of I87
f i mode of operation not any of the diesel generators, will not create a system configuration or ! operating condition such that the technical specification limiting condition for operation or ! surveillance requirements are no longer adequate, not will it bypass or invalidate automatic j actuation features required to be operable by the technical specifications. He ability of the HPCS l DG air stardng system to provide sufficient air for five starts is unaffected by this change. j Therefore, this modification does not reduce the margin of safety as defined in the basis of any : technical specification. For these reasons, this modification does not constitute an unresiewed i safety question. t Channe Number /USAR Section: LCN 09.05-102 Figures: 9.5-l a 9A.2-6 i Description and Basis for Change: This modification provides the design details and implementation requirements necessary to modify l a portion of the existing fire protection water supply underground piping to accommodate the l l construction of, as well as provide water for exterior fire protection, the new administration complex designated as the generation support building (GSB). A section of the existing underground piping will be abandoned in place. A new section will be installed to compensate for 4 that being abandoned. In addition, a new fire hydrant will be installed which will replace one being : f removei Summary of Safety Evaluation: All actions required for this modification will occur outside the protected area, away from any safety related structure, system, or component (SSC). The piping to be installed will be UL listed , and/or FM approved for the fire service as is required by code. The installation will meet the same acceptance criteria as the existing underground piping. Therefore, the probability of an accident previously evaluated in the SAR will not be increased. The consequence of failure of the new piping will be no worse than the consequences of failure of an existing underground fire protection line. Therefore, the consequences of an accident previously evaluated in the SAR is not increased. Should failure occur in the newly installed piping, the fault would be isolated from the yard loop in the same manner as if a failure occurred in any other piping being fed from the loop. The fire suppression water system would remain operable even in the event of a failure of the newly re-routed piping. Therefore, the possibility of an accident which is different than any previously evaluated in the SAR is not created. The installation affects only the portion of the fire protection 2 underground piping, which is exterior to all safety related structures. The present piping configuration provides a dual feed to suppression systems in the control building, diesel generator building, auxiliary building, and reactor building. After rerouting the system, the complete installation will retum the dual feed capability to the systems. Therefore, the probability of a malfunction of a safety related SSC previously evaluated in the SAR will not be increased. Since the piping is independent of, and isolated from, safety related SSCs then, the consequences of a malfunction of a safety related SSC is not increased. The fire protection water line being added by this modification is of the same material and construction as the existing lines. The line will not be required for safe shutdown of the plant. Therefore, the possibility of a malfunction of a safety related SSC different than any previously evaluated in the SAR in not created. The piping and fire hydrant being affected are not subject to the requirements of these technical specifications. The installation will however, meet the same design, installation, and acceptance criteria as did that part of the underground system that is subject to the technical specification requirements. Therefore, Section I Page 97 ofI87
the margin of safety as dermed in the basis to any technical specification is not reduced. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 9.5-104 Figures 9.5-2c 7.3-17 Description and Basis for Chance: This change will provide new level switches to replace the existing level switches on the lube oil sump for the diesel generator, Divisions I and II. The current mark numbers represent high and low set points for the same level switch nerefore, it was decided that only one mark number will be used. Summary of Safety Evaluation: This modification will provide new level switches to replace the existing level switches on the lube oil sump for the diesel generator lube oil system. He current switch has a high and low level setpoint but it does not take into account the status of the diesel engine when the engine is running and therefore it provides meaningless information and nuisance alarms. This modification will modify the circuit such that the high and Icw level setpoints and associated annunciations will corre::;sond to the running status of the diesel engine. In addition, the pneumatic oil sump level indicator does not work correctly. The dipstick located nearby is adequate for determining the sump oil level. The level switches perform non safety-related functions. Therefore, this modification will not increase the probability of occurrence of an accident evaluated previously in the SAR. The level switches proposed to be replaced, the pneumatic sump level indicator and the associated push-button to be removed are non-safety-related and are not mentioned in the SAR. The panel holes will be properly covered. Only the level switches are required for normal plant operation, but are completely independent from and are not required for safe shutdown of the plant. Herefore, the replacement of the level switches will not increase the consequences of an accident previously evaluated in the SAR. Failure or malfunctions of these level switches or the removal of the sump level indicator and associated push-button will not result in any accident previously stated in the SAR. The level switches are non-safety-related, however it will be ensured that the level switch assembly and its mounting is scismically qualified. Therefore, the probability of a malfunction of a safety-related structure, system, or component previously evaluated in the SAR will not be increased. The changes proposed by this modification will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the S AR. Since the level switches are not required for the operation of the diesel generator or the safe shutdown of the reactor the modification will not create the possibility of a malfunction of a safety-related SSC previously evaluated in the SAR. He level switches are non safety-related and therefore, the changes will not reduce the margin of safety as defined in the basis for any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Section 1 Page 98 of187
Chance Number / USAR Section: LCN 9.5-105 9.5.-16 Description and Basis for Change: This change allows short lengths of communications and 120-volt power supply cable in the power block to be routed outside of conduit and electrical raceways. Temporary communication cable of longer length may be installed outside of conduit for one operating cycle or less. Summary of Safety Evaluation: The short cables to be installed outside of conduit are constructed to prevent the transmission of noise outside of the cable. He cables allowed by this change to be outside conduit are either coaxial or twisted pair. Shielding is provided in coaxial cable and the twisted pair cable construction prevents noise transmission. The cables will also be physically separated from safety-related systems and components in accordance with plant criteria. Cable construction and separation will climinate the potential for noise to affect safety related signals. Cable installation will be restricted to keep fire loading from exposed cables within combustible loading limits. Therefore, there is no increase in the probability of the occurrence of an equipment malfunction or an accident previously evaluated in the SAR. Noise propagation and combustible loading from cables outside conduit has been minimized to accepted limits. This minimization will prevent bare cable from effecting the ability of safety-related equipment to perform accident mitigation. Therefore, there is no potential to increase the consequences of any accident or malfunction of equipment important to safety previously evaluated in the SAR. By installing cable outside conduit but with shiciding and physical separation from safety related plant equipment, the possibility of a different type of accident or equipment malfunction than previously described in the S AR will not be created. Cable outside conduit in the power block will not have any adverse effect on any technical specifications. Therefore, this change does not reduce the margin of safety as defined in a technical specification bases. For these reasons, this change does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 9A.2 25 Pages: 9A.2-17 9A.218 1 Tables: 9A.2-9 Sheet 1 9A.2-9 Sheet 2 9A.2-10 Sheet 2 9A.2-35 Sheet 2 Figure 9A.2-1 Description and Hasis for Chance: his modification installs a fire door for diesel generator IB cable vault and establishes a new fire I area (DG-7) including extension of the fire detection system. Summary of Safety Evaluation: l The new fire detection equipment ensures that any fires in this area are more quickly detected. There is no accident analysis which takes credit for the newly installed fire door. Based on this information, the probability of an accident previously evaluated in the S AR is not increased. This change has no affect the fire protection design basis for the control room and ensures separation Section i Pag: 99 of I87
i l between safe shutdown fire areas in accordance with 10CFR50 Appendix R, Section Ill.G. ]
; Derefore, the consequences of an accident not evaluated in the SAR are not increased. Since the l
[ ; doorway cannot create a new accident and the new fire detectors are not tied to any fire . . l i l suppression systems, the possibility of an accident which is different than any previously evaluated in the SAR is not created. All safe shutdown cables in the new fire are in conduit and have ! negligible heat gain which does not adversely affect the area temperature. Hence, the probability J of a malfunction of a safety-related structure, system, or component (SSC) is not increased. The i O added door and fire detectors cannot increase the consequences or create the possibility of any !
- malfunction of a safety-related SSC because they proside added separation and detection that does l j not adversely affect any item that is safety-related or supports a safety-related function. De !
margin of safety as defined in the basis to any RBS technical specification is not reduced. For .I j these masons, this alteration does not constitute an unreviewed safety question. j Channe Number /USAR Section: LCN 09A.2-29 Page 9A.2-8 -! I Description and Basis for Channe: l 4 i 3- This modification installs two closed circuit television cameras (CCTVCs) inside the drywell to ! ' allow video viewing of the recirculation pump shaft seals at approximate elevation 95', from i t outside of the drywell when access to the drywell is restricted. The equipment is electrically l powered from existing 120V ac receptacles and will transmit signals through existing l communication (telephone) wiring. De cameras are also capable of panning their respective areas l
- - for viewing around the drywell as necessary or desirable. There is no safety function provided by l this equipment. The video images provided by the cameras are for information only. j Summary of Safety Evaluation
- l The added CCTVC equipment will provide a method for obtaining additional information
,. regarding the conditions inside the drywell during times when access by personnel is restricted. No_ credible failure modes which initiate accident scenarios previously evaluated in the SAR will be ; ! increased. All new instrument cables are being installed in accordance with the separation '
- requirements of Regulatory Guide 1.75. There are no electrical interconnections between the new cables and any other plant system except for AC outlets and communications which are non-safety-1 related. Therefore, the consequences of an accident previously evaluated in the SAR is not !
; increased. The new cable for the CCTVCs will be routed in the drywell without conduit. . The !' amount of combustible loading added in accordance with this modification is negligible. He l possibility of an accident which is different than any previously evaluated in the SAR is not <
- created. De equipment is installed without contact to plant operating equipment and does not i affect the plant reactor coolant pressure boundary. The equipment and cables are tied off to plant l j structures and supports of sufYicient capacity. ' Since none of the equipment is safety related, the l probability of a malfunction of a safety related stmeture, system, or component (SSC) presiously I evaluated in the SAR will not be increased. The equipment added is completely independent of the
- functions of both active and passive drywell components. All potential interactions between the i new equipment and components or systems important to safety have been addressed and been i determined to be acceptable, including where the new equipment has been supported off existing structures and component supports. Therefore, the modification does not increase the consequences of a malfunction of a safety related SSC previously evaluated in the SAR. The )
modification has no detrimental effects on the reactor coolant pressure boundary. Since the proposed equipment is independent of all SSCs then, the possibility of a malfunction of a safety ~ . i Section ! - Page 100 ofI87
i i t i- i 1-j related SSC di& rent from any previously evaluated in the SAR is not created. De cameras ! installed are not required to be operable by any technical specification. Also, the modification has l l no adverse impact on the operability'of any equipment which is required operable by the technical !
- specification. Derefore, the margin of safety as defined in the basis to any technical specification j is not reduced.- For these reasons, this modification does not constitute an unreviewat safety ;
question. l f ! Channe Number /USAR Section: LCN 9A.2-31 CR 95-0444 Figures 9A.2-6 f j 9.5-1 a Description and Basis for Channe: , 4
- This change removes a metal building from plant drawings and changes the fire protection piping i and valves associated with that building. Specifically, the service building fire protection header !
l inlet isolation valve is shown to be normally closed and a pipe is shown to be plugged at its j terminal end. The work was performed in the field under a previous modification. j Summary of Safety Evaluation: , a + 2 nis change will not a&ct fire protection system performance or reliability in a manner that could ; lead to the occurrence of an accident. This change will not affect the stmetural or hydraulic ; j integrity of the fire protection water system. No new or different fire hazards are introduced and ' j no new or different exposures to safety related components are created. The change does not j create ar.y common mode or common cause failures. Herefore, this change does not increase the 1 6 probalility of an accident or a malfunction of equipment important to safety previously evaluated ., l in the SAR. This thange will not cause any fire protection component to operate outside ofits ! 1 dcsign or test limits. This change does not affect any parameter which could alter radioactive . l isotope population, release rate or duration. This change does not create new release mechanisms 4 or impact radiation release barriers. Neither fire protection system capacity nor redundancy is i affected by this change. Therefore, this change does not increase the consequences of an accident j j or a malfunction of equipment important to safety previously evaluated in the SAR. This change g does not create new accident initiators or failure exposures during any plant operational mode. ,
- This change does not increase the probability of any accident or malfunction of equipment i important to safety previously evaluated for inclusion in the SAR and found to be not credible.
Modifications to the valve reduce the possibility for inadvertent operation as well as physical damage in the future. The subject valve is not required to operate since the piping is not connected to any fire protection system or component. The plug in the pipe precludes the entry of deleterious , !' materials and reduces the possibility ofphysical damage. This change does not introduce any new failure mechanism associated with the fire protection water system or any other plant system or , component. His change does not affect pressure, flow, instrumentation, control, physical protection, hazard warnings, or hazard severity of the fire protection water system. Additionally, ; the location, orientation, and construction of safety related facilities exposed by the fire protection l water system are not affected. Therefore, this change does not create the possibility of an accident : or a malfunction of equipment important to safety different from any previously evaluated in the !
- SAR. Operability of the fire suppression systems is not adversely affected by this change. Fire ,
, suppression system capability to minimize potential damage to safety related equipment is not ! affected. Therefore, the margin of se.fety as defined in the basis of any technical specification is l , not reduced. For ther reasons, tW modification does not constitute an unreviewed safety i question. ! u , I
- Section 1 Page 101 ofI87 t
I
- . . . - _ - . - .. . . .- ~-
f Channe Number /USAR Section: LCN 9A.2-32 Pages: 9A.2-29 9A.3-58 Description and Basis for Channe: This modification adds permanent maintenance platforms and walkways around the gearbox assembly of each fan in the circu!ation water towers CWS-TWRI A, CWS-TWRIB CWS-TWRIC, and CWS-TWR1D. In all,32 walkways and platforms will be installed. The platforms and walkways will be constructed from pressure-treated redwood structural members and , fiberglass grating. The platforms and walkways will include handrails, knee rails and toe boards made from pressure treated redwood. Open grating will be used to allow air to flow up through it. A 6" inspection port will be installed for each fan cylinder, in all,32 inspection ports will be installed. This modification will provide a safer environment for maintenance on the gearboxes by maintenance personnel. Summary of Safety Evaluation: None of the accidents in the S AR take credit for the cooling towers being one of the initiating mechanisms. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. The cooling towers are over 1000 feet away from any safety related system, structure, or component (SSC). A fire in any of the cooling towers would be contained in the cooling tower and would not affect any safety rehted SSC. Loss of one or more of the cooling towers results in a shut down of the plant using emergency systems. Therefore, this 1 modification does not increase the probability of a malfunction of any safety related SSC previously evaluated in the SAR. The cooling towers are not used to mitigate the consequences of any accident or malfunction of any safety related SSC. The cooling towers are not associated with any radiological impacts of any accident or malfunction of any safety related SSC. Therefore, this modification does not increase the consequences of any accident or malfunction of any safety related SSC previously evaluated in the SAR. It is unlikely that fire would be created in more than one cooling tower at a time as they are in no way connected. It is not probable that a fire would start with the cooling tower in operation, as a constant flow of air with heavy moisture will bc
- ' . passing over the platforms and walkways at all times when the fans are operating. The plant can respond to transients that would be similar to the transient that would be created by a loss of a cooling tower. Therefore, any accident that was already analyzed bounds the loss of one or more of the cooling towers. This modification does not introduce any new failure modes for any safety related SSC. Therefore, this modification does not create the possibility of an accident or a malfunction of any safety related SSC different from any previously evaluated in the SAR. There are no technical specifications governing cooling towers CWS-TWRI A, CWS-TWRIB CWS-TWRIC, and CWS-TWRID. Therefore, this modification does not reduce the margin of safety as dermed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
l Channe Number /US AR Section: LCN 9B.4-004 Page 9B.4-20 Description and Basis for Channe: This modification added floor hardware and the associated fire-rated ceiling access door in the floor of the 951 elevation in the auxiliary building (Fire Area AB-2, Zone 2). This door gives
. Section i Page 102 of187
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access to the D Tunnel in the auxiliary building (Fire Area AB-7). This change provides a fire ] rated assembly in an existing opening. l Summary of Safety Evaluation: 1 The operation or failure of the newly installed door does not increase the probability of any accident previously evaluated in the SAR because it is not used in any accident analysis. The installation of this access door decreasca the consequences of a fire by providing a three-hour fire barrier. The possibility of an accident which is different from r.ny prniously evaluated in the SAR is not created by the modification Since the fire door is not required and flood levels are not above the levels used in S AR analysis, the probability of consequences of a malfunction of a safety-related structure, system, or component (SSC) are not increased. The closing of new fire door A95 14 cannot cause the malfunction of a safety-related SSC. The possibility of a malfunction not evaluated in the SAR is not created. The margin of safety as defined in the basis to any RBS technical specification is not reduced because the new door meets the same requirements as other fire-rated assemblics in the plant. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 10.2-8 Pages: 10.2-8-11 10.2-18 3.5-7-16 3.5-23-24 Tables: 1.8-1 3.5-1 3.5 2 3.5-3 3.5-4 3.5-6 4 3.5-8 3.5-10 3.5-20 3.5-21 Figures: 3.5-1 3.5-2 3.5-3 3.5-4 3.5-5 10.1-3 10.1-4 Descrintion and Basis for Change: During the turbine rotor and dovetail inspection performed during RF-4, cracks were discovered in the wheel bore axial keyways and bucket dovetail areas adjacent to the notch blocks. This modification replaces the current " built-up" style low-pressure turbine rotors with the new "monoblock" style.
'Section 1 Page 103 of187
Summary of Safety Evaluation: he new turbine rotors will be installed and operated in accordance with the specifications and standards that meet the original system design requirements. He probability of an accident previously evaluated in the SAR will not be increased because the replacement of the low pressure turbine rotors does not affect any equipment whose malfunction is postulated in the SAR to initiate an accident. The installation of the new low pressure turbine rotors will not affect the operational characteristics of the system. Therefore, the consequences of an accident presiously evaluated in the SAR will not be increased. The replacement rotors are manufactured by the OEM to design specifications and standards that meet or exceed the original system design requirements. This modification does not change the operation, function, or design basis as described in the SAR. The possibility of an accident which is different than any previously evaluated in the SAR will not be created. Since the replacement turbine rotors have no affect on any safety related equipment, the probability of a malfunction of a safety related SSC is not increased by the replacement rotors that are previously evaluated in the SAR. Due to the reason stated above the consequences of a malfunction of a safety related SSC previously evaluated in the SAR is not increased. Since the equipment being replaced does not interact with any safety related equipment, the possibility of a malfunction of a safety related SSC different than any previously evaluated in the SAR is not created. The main turbine system is not described in the Technical Specifications. The new replacement rotors do not change the design basis, functions, or operations of any safety related l I equipment and do not adverrely affect any other safety related SSCs. Therefore, the margin of safety as defined in the basis for any technical specifications is not reduced. For these reasons, this alteration does not constitute an unreviewed safety question. 1 Chance Number /USAR Section: LCN 10.2-9 Pages: 10.2-4 ; 10.2-11 ! 3A-iia l l 3A.29-1 Figures: 10.4-7j 10.4-7k Descrintion and Hasis for Chance: This change incorporates the use of nozzle check valves in the extraction steam lines rather than l power assisted check valves. The use of nozzle check valves climinates all testing of the valves since there is no circuitry or extemal control of the valves. The power-assisted nozzle check valves close on a turbine trip or high water level in the feedwater heater. Summary of Safety Evaluation: Closure requirements are intended for those valves in lines with a serious potential for causing a , turbine overspeed. However, none of the feedwater heaters have enough available energy to result in a turbine overspeed. No accident previously analyzed in the SAR takes credit for the extraction ! steam nonreturn valves to mitigate the consequences of the accident. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. No new failure modes or accident possibilities are c cated by the change. The only safety-related structure, system or component (SSC) that is afTected are the control panels and the wiring. The controls for the power-assisted valves are being removed from the control boards, and the cables are spared and do not change the Section 1 Page 104 of 187 i
analysis of any cable tray or conduit. It is not possible to postulate any credible malfunction of a safety related SSC different than previously evaluated as a result of this change. The associated . technical specification does not apply to the extraction steam nonretum valves, herefore, this ! 4 change does not create the possibility of an accident or a malfunction of equipment important to i safety different from any previously evaluated in the SAR. This change does not impact any technical specification or the basis of any technical specification. Therefore, this change does not ! i reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 10.2-10 Page 10.2-6 : Description and Basis for Change: This modification changes the electrical trip solenoid valve (ETSV) assembly from a two coil design to a single coil design. The current design results in failures due to the heat generated by the solenoid. The turbine backup overspeed trip system meets the requirement of GDC 4 with respect , to the prevention of the generation of turbine missiles. I Summary of Safety Evaluation: Developmental work has led to an improved finned single-coil design with electrical and mechanical characteristics equivalent to the original design. Prototypes have been laboratory i tested and installed on an operating unit. Following three years of operating experience, the new finned, single-coil ETSV successfully met the reliability improvement criteria. Based on this information, the probability or consequences of an accident previously evaluated in the S AR are not increased. Installation of the single-coil ETSV does not change the function or operation of the turbine backup overspeed trip system; and therefore, does not increase the probability or ; consequences of a malfunction of a safety-related structure, system, or component (SSC). The ! replacement of the dual-coil ETSV with the new single-coil ETSV does not introduce additional l equipment or equipment interfaces, and the improved finned design is equivalent to (or exceeds) the original design in characteristics and reliability. He possibility of a malfunction of a safety- . related SSC is not created by the change. The margin of safety as defined in the basis to any RBS j technical specification is not reduced because the new single-coil ETSV does not change the design , basis, function, or operation of any safety-related equipment and does not adversely affect any other safety-related SSCs. For these reasons, this alteration does not constitute an unresiewed safety question. Channe Number /USAR Section: LCN 10.211 Pages: 10.2-13 10.2-15 4 15.2-9 Description and Basis for Channe: I This modification added a switch that will ensure that no inadvertent turbine trips occur durir.g testing of the turbine trip system. This modification also changes the alarm for turbine se;,ervisory instruments (TSI) trip voltage. The existing circuit for the alann was modified by ad6ng an auxiliary relay, The existing alarm will also provide input to a new alarm to show that the turbine trip function is enabled. The engraving of the window of the existing alarm will be resised to read Section i Page 105 of187 i
i l i " Card out of File or TSI Power Supply Failure" and the window for the new alarm will read ' Turbine Vibration Trip Enabled." This modification eliminates a nuisance alarm. ; l Summary of Safety Evaluation: l l The modified alarm circuits are not postulated to cause any accident identified in the SAR. This l modification to the alarm circuits does not increase the probability of a turbine trip. There is no increase in either the probability of occurrence or the consequences of a turbine trip accident l because the alarms have no interface with the turbine trip logic. Therefore, this modification does l not increase the probability or the consequences of an accident previously evaluated in the SAR. j He turbine generator and its components are not safety related. The modified alarm circuits have l no interface with the control or operation of any safety related equipment. He modified alarm circuits are not related to any radiological consequences since the modified alarm circuits do not i have any interface with the turbine controls. Herefore, this modification does not increase the probability or consequences of a malfunction of equipment important to safety presiously evaluated in the SAR. A credible accident of a different type that would be caused by the alarm ) logic change could not be identified. There are no new failure modes of the alarm circuit that can be postulated by this modification. Herefore, this modification does not create the possibility of l an accident or a malfunction of equipment important to safety different from any presiously ) evaluated in the SAR. This modification does not affect any technical specification or the basis of ! any technical specification. Therefore, this modification does not reduce the margin of safety as l defined in the basis of any technical specification. For these reasons, this modification does not I constitute an unreviewed safety question. Chance Number /USAR Section: LCN 10.3-103 Figure 10.3-ld Description and Basis for Chance: This modification provides justification and instructions for replacement of the existing Worthington Model CV air compressors with Ingersoll-Rand Model 7C3 compressors as well as replacement of the air dryers, differential pressure switches, and carbon steel piping downstream of 1 the after-filters. This MR also addresses other componetto associated with the replacement of the compressors. The new compressors and dryers have higher output ratings than the current equipment; however, system pressure is controlled to satisfy system design requirements for minimum and maximum pressure. Summary of Safety Evaluation: l I I This modification does not change the design basis or the function of the air supply system (SVV) for automatic depressurization System (ADS) and non-ADS safety relief valves. The replacement compressors and dryers satisfy all current plant design requirement for flow rate, operating pressure, and air quality. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR. His MR does not change the severity of any accident l described in the SAR and does not affect postulated release paths. Therefore, this modification ! does not increase the consequences of an accident previously evaluated in the SAR. The compressor / receiver and dryer assemblics are installed at the same locations as existing equipment and are powered by the sarne divisional electrical power source. The modified design continues to satisfy the design, material, and construction standards for the SRV air supply system. Therefore, this modification does not create the possibility of an accident different from any presiously Section i Page 106 of 187
evaluated in the SAR. There is essentially no change to the interfacing system design or operation. J All rei acement components maintain existing design requirements for both safety-related and non-safety ulated portions of the piping system. The proposed changes do not increase the operation frequency ofimportant-to-safety equipment or impose increased or more severe testing requirements on such equipment. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. The postulated accident scenarios described in the SAR are not affected and the ADS system function described in Chapters 5,6, 7, and 15 is not impacted by this change. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. Equipment in non-safety-related portions of the SVV system is replaced with like equipment. This modification provides additional assurance that the safety-related ADS and non-ADS SRV accumulators and suppression pool level instruments are able to perform their functions, thus reducing the possibility of a malfimetion of safety-related structures, systems, or components. Therefore, this modification does not create a malfunction of equipment to safety different from any previously evaluated in the SAR. The number of operable safety relief valves and low-low set function as well as ADS trip systems and ADS function are not changed. ECCS actuation instrumentation or setpoint are not affected Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. I Chance Number /USAR Section: LCN 10.03-106 Figure 10.3-IC , l Description and Basis for Chance: Steam provides motive force for the steam jet air ejectors (SJAE) to create vacuum and remove non-condensable gases from the condenser during normal plant operation. Steam is supplied to the SJAE from main steam via the pressure regulating valve. The relief valve is provided for protection against overpressurization. This modification is designed to increase the relief valve setting from the current 275 psig to 350 psig. By providing a higher margin between the pressure required, to ensure dilution steam flow requirement, and the relief pressure, this will eliminate frequent lifting of the relief valve and crosion of the seat. On further investigation it was found that the existing relief valve cannot be set higher than 285 psig. Therefore, the current valve has to be replaced with a relief .alve that can be set at 350 psig. Crosby model 3K4-JBS-36E will replace the current relief valve without any modification of the piping due to its identical dimensions and only one additional pound in weight. Summary of Safety Evaluation: A review of the affected sections of the SAR reveals that the " Loss of Condenser Vacuum"is one of the accidents considered in the SAR. This modification revises the relief valve setpoint within the B31.1 allowable stress limits. Replacement of the pressure valve will allow steam flow to the SJ AEs within the original specified operational limits for vacuum as well as for hydrogen dilution at plant startup without operator help. The offgas equipment and piping is designed to contain any hydrogen-oxygen detonation, which has a reasonable probability ofoccurring. Analysis of the offgas system demonstrates that the equipment failure results in doses which are well within guidelines of 10CFR100. Therefore, this modification will not increase the probability of occurrence of an accident evaluated previously in the SAR. No credit is taken for the condenser air removal system to mitigate the consequences of an accident previously evaluated in the SA.R. This modification changes the setpoint of the relief valve to preclude the inadvertent lifting of the safety Section i Page 107 of 187
valve. He pressure valve is of fait close design therefore, its failure to full open is highly unlikely. l
' nerefore, the possibility of a malfunction of a safety related structure, system, or component l (SSC) presiously evaluated in the SAR is not increased. The equipment to be modified is non- ,
safety-related, Q Class 2 lu:ated in the turbine building. Changes due to the relief valve and , pressure valves are within stress limits. Therefore, this change does not increase the probability of occurrence of a malfunction of a SSC previously evaluated in the SAR. This modification enhances the operation of the system in that the frequent lifting of the safety relief valve will be eliminated. A controlled amount of steam will be available to the second stage of SJAEs at low startup pressure without operator assistance. The existing warning and system isolation signals are
; still available. Therefore, the consequences of a malfunction of a SSC previously evaluated in the SAR is not increased. The function of the condenser removal system is not modified by this :
modification. The new equipment is designed and procured to the same codes and standards as the existing equipment. Herefore, none of the equipment important to safety is changed and this modification does not create the possibility of a malfunction of a SSC different than any presiously evaluated in the SAR. This modification is not included or associated with Technical Specifications. Therefore, the margin of safety as defined in the basis to any technical J- specification in not reduced. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 10.4-119 Page 10.4-10 10.4-12 , i 10.4-13 ). Description and Basis for Change: l This change revises the SAR to reflect the current operation of the turbine gland seal system I (TGSS). This modification states that the steam seal evaporator (SSE) supplies steam to the TGSS without a backup system. This change also deletes descriptions of the automatic function of 1 MSS-MOVISS, which is no longer needed to function. Summary of Safety Evaluation: The only credible incidents included in the SAR that are affected by this change are loss of main condenser vacuum and automatic MSIV closure. Any increased probability of occurrence for an automatic MSIV closure is bounded by the loss of condenser vacuum. A low vacuum : rip is the precursor for the automatic closure of the MSIVs. The as-designed supply of steam to the TGSS is from the SSE while utilizing the radwaste reboiler and the auxiliary reboiler as backups; however, the plant currently uses only the SSE with no backups. He Engineering Analysis group calculated that using the SSE only, without backups, increases the probability of an accident previonly evaluated in the SAR by a negligible amount,0.4 occurrences per year. In the event of an accident in which there is a loss of seal steam, a loss of condenser vacuum with anticipatory MSIV closure and turbine trip will occur. When steam flow is shut off by MSIV closure, no radioactive steam is released to the gland exhaust system or the main stack. Therefore, the consequences of an accident previously evaluated in the SAR is not increased. Equipment and operator actions that occur after an initiating event are unaffected by this change and do not create the possibility of an accident different than those previously evaluated in the SAR. This change does not increase the probability of a malfunction of a safety-related structure, system, or component previously evaluated in the SAR. The TGSS performs no safety related function described in the SAR and hence does not increase the consequences of a malfunction of a safety-related system as previously evaluated in the SAR. Here are no safety-related systems, structures, Section 1 Page 108 of187 I i l
or components within the TGSS area and no major components have been physically moved in the plant. Thus, this change does not create the possibility of a malfunction of any safety-related items different from those previously evaluated in the SAR. His change does not reduce the margin of safety as defined in the basis of technical specifications. Herefore, this change does not constitute an unreviewed safety question. l l Chance Number /USAR Section: LCN 10.4-139 Figure 10.4-4 Description and Basis for Chance: This modification addresses sulfuric acid leakage from pumps IWTA-P2A and IWTA P28. Flexible hose was installed to replace rigid stainless steel piping on both the suction and discharge sides of these pumps. The replacement of these sections of piping with acid resistant flexible hose is an enhancement to the existing design which should eliminate the potential for sulfuric acid leaks at the pump connections due mainly to thermal expansion of the stainless steel piping. Additionally, this change will minimize acid induced corrosion on the existing structure and grating below these connections. Summary of Safety Evaluation: The purpose of the sulfuric acid injection system is to mitigate the scaling, corrosion and biological fouling in the circulating wates system. None of the accidents evaluated in the SAR in any way relate to the acid injection system. Therefore, this change does not increase the probability of an occurrence of an accident evaluated previously in the SAR. The acid injection system is not required for safe shutdown of the plant. Therefore, this change does not increase the consequences of an accident evaluated previously in the SAR. Based on the above discussion, this change does not create the possibility of an accident that is different from any previously evaluated in the SAR. This change is confined to the suction and discharge sides of the chemical pumps which are not in direct or indirect ca.itact with any equipment important to safety. Therefore, this change does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. This change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. This change does not create the possibility of a malfunction of equipment important to safety previously evaluated in the SAR. The sulfuric acid system is not referenced in the technical specifications as verified by reviewing the index and definition sections of the technical specifications. Therefore, this change does not reduce the margin of safety as defined in the basis of any tecimical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 10.4-141 Figure 10.4-3a Description and Basis for Chance: A thennocouple has a shorted connection inside the circulating water system (CWS) pump motor iB and can not be removed from inside the stator windings. This condition is gising a nuisance alann in the main control room when the shorted connection gives a value above the alarm setpoint. To eliminate this nuisance alarm, the " SCAN" function on the computer is to be eliminated so that it will simply ignore the erroneous alarm. By climinating the computer " SCAN" function for this Section i Page 109 of 187
~
one computer point, the hardware will remain installed in the system and can be easily reactivated if the motor is ever rewound and the defective thermocouple is replaced. Summary of Safety Evaluation: There are six thermocouples in the motor as a redundant means of monitoring the motor winding temperature. The loss of one or more of the thermocouples will not affect the operability of the ; circulating water pump motor or the plant (i.e., the operability of the motor is not dependent upon the functionality of the thermocouples). Herefore, this change does not increase the probability of an occurrence of an accident evaluated previously in the SAR. The purpose of the thermocouples in monitoring the motor winding temperature is to detect a deterioration of the winding insulation and a possible need for rewinding the motor. The consequences of a failure of one of the thermocouples were not evaluated in the SAR since they are not a safety related piece of equipment [ and there are redundant thermocouples in the windings. The loss of one of the motor winding temperature thermocouples will not prevent the temperature monitoring and will not have any effect on the plant operation. Therefore, this change does not increase the consequences of an accident previously evaluated in the SAR. The removal of the computer scanpoint for the subject thermocouple will not affect the operation of any component at the plant. Their sole function is to monitor the temperature of the rr.otor windings. Therefore, this change does not create the possibility of an accident of a different type than any evaluated in the SAR. The CWS neither
- contains nor contacts any safety related components. The affected thermoc6uple which this analysis addresses does not contain any trip functions and will not affect the operation of the CWS i system in any manner. Therefore, this change does not increase the probability of an occurrence of a malfunction of equipment important to safety as previously evaluated in the SAR. The e information provided by the thermocouples is helpful in determining if the pump motor will require future maintenance, but it is not necessary for the operation of the circulation water pump motor or any other equipment important to safety. Therefore, this change does not increase the consequences of a malfunction of equipment important to safety as evaluated previously in the -
SAR. Data received from the thermocouple is strictly for infonnational purposes and cannot affect the performance of any equipment important to safety. Therefore, this change does not increase the possibility of a malfunction of equipment important to safety evaluated previously in the SAR. Neither the circulating water pump motor nor the thermocouples enveloped by it are mentioned in the Technical Specifications. The affected thermocouple does not generate a signal to any desice that can affect the operation of the CWS. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this change does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 10.04-143 Figure 10.4-3b Description and Basis for Change: his modification provides installation of pulsation dampers (or surge suppressers) on the discharge side of dechlorination pumps CWS-P20 and P21. The pumps are located in the yard
, south and outside of the protected area, and are non-safety-related. There is no particular regulatory basis for the current configuration and, as such, is no change to the regulatory basis.
. Section I Page 110 ofI87
Summary of Safety Evaluation: The circulating water system (CWS) system is Category II, serves no safety function and is classed as non-seismic. Malfunction or failure of a component of the system does not affect the intended function of any safety-related system or components. This modification does not directly or indirectly affect the circulation water pumps or any other component that may result in the loss of condenser vacuum. These new surge suppressers and the affected area"of the CWS system do not interface with equipment important to safety and thus, has no impact on the off-site radiological conditions. Therefore, this modification does not increase the probability of an accident or the probability of a malfunction of equipment important to safety previously evaluated in the SAR. This modification will not affect safety related systems or components to mitigate a design basis accident If a design basis accident or a loss of station power occurs, the CWS system is not relied upon to be operable. Malfunction or failure of a component of the system will not affect the intended function of a safety-related system or components to mitigate a design basis accident since during such an event the system is not relied upon for operation. This modification does not affect the operafian of the CWS system and does not affect any system that contains, processes or stores radioactive materials. The affected areas of the CWS system do not interact with equipment important to safety and has no impact on off-site radiological conditions Therefore, this modification does not increase the consequences of an accident or the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The CWS system will continue to function as before since the change is still within the original design basis of the CWS system. Malfunction or failure of a component of the dechlorination subsystem does not affect the intended function of any safety-related system or component. Therefore, this modification does not create the possibility of an accident or the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. The CWS does not function as a radiological barrier and is not required for shutdown cooling. The modification will not reduce the margin of safety as defined in the basis for any technical specification. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Chance Number /USAR Section: LCN 10.4-145a Pages: 9.2-13 9.2-59 9.4-69 9A.2-8 10.2-13 j 10.4-23 ' 10.4-38 l 10.4-39 ) Figure 5.4-2a l Description and Basis for Change: This modification replaces the currently existing vibration monitoring system with a new sibration monitoring system. The purpose of both vibration monitoring systems is to provide information on the operating condition of the rotating machinery. The information provided by the new system will be used to predict when maintenance should be performed. Section I Page 111 of 187 l I
Summary of Safety Evaluation: His modification will not cause the equipment or system to operate outside any design or testing limits. His modification will not change the accuracy or response characteristics of any instrument or system important to safety. No new, credible failure modes which initiate accident scenarios evaluated in the SAR are created by this modification. Therefore, this change does not , increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. He new vibration sensors will be surface mounted on selected motors, pumps, and gearboxes in a non-intrusive manner. This modification will have no effect on the reactor coolant pressure boundary. All wall sleeves breached by this modification will be j rescaled. The added weight of the new instruments on the machinery is negligible. All new equipment is Category II and none of the equipment scheduled to be monitored is safety related. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. This modification will not change, degrade, or prevent any actions described or assumed in the accident analysis. Failure of the new system's components would, at worst, cause loss of diagnostic capability of the vibration monitoring system or parts thereof. Short circuits in the system will not damage any penetration conductors. This modification does not affect any fission product barriers and does not affect any systems, structures, or components which play a direct or indirect role in mitigating the radiological consequences of accidents described in the SAR. No new, credible failure modes can be identified with this modification that could contribute to the consequences of a malfunction of safety related equipment. Therefore, this modification does not increase the consequences of an accident or the consequences of a malfunction of equipment important to safety previously evaluated in the S AR. This modification does not add any new accident scenario not previously evaluated in the SAR. All the equipment installed by this modification is independent of safety related equipment. Therefore, this modification does not create the possibility of an accident or the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. Neither the old nor the new system is required to be operable by any technical specification. This modification will have no adverse impact on the operability of any equipment which is required to be opecable by the technical specifications. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 10.4-79C Pages: 10.4-018 10.4-19 10.4-21 10.4-23 Figures: 1.7-2c 10.4-3c ;
)
Description and Basis for Chance: , I This modification to the circulating water system is part of the sponge ball condenser tube cleaning ! system used to maintain cleanliness within the system. This modification installs ball strainers at i cach condenser outlet waterbox and ball injectors into each inlet waterbox. The interconnecting piping and a ball recirculation pump are connected with distributor sight glasses. This i modification also replaces some pipe caps with blind flanges. ! l l l Section i Page 112 ofI87 I
I Summary of Safety Evaluation: i All equipment and pipmg connections, mcluding sightglasses, are located under the condenser in the turbine building where they can be isolated upon failure or leakage. De circulating water ; system is not safety-related and does not interface with any safety-related equipment. Therefore, i the probability of an accident previously evaluated in the SAR is not increased his change does ! not increase the consequences of the SAR evaluated accident of pit floodmg due to a pipe break - i which causes the pump to fail and shut down. Inadvertent backwash may cause loss of balls to the l cooling tower and pump suction fiume, but it has been determmed that the circulating water flow is ; not affected. Thus, the possibility of an accident different from those evaluated in the SAR is not ; created. This addition does not affect any safety-related structure, system, or component (SSC); - l therefore, neither the probability nor the consequences of a malfunction of a safety-related SSC are > not increased. Also, the possibility of malfunction of a safety-related SSC is not created. His -i change does not cause the operational setpoints of any related systems to be exceeded Therefore, l the margin of safety as defined in technical specifications is not reduced. Consequently, this change does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 11.2-43 Page 11.2-8 Figures: 9.2-21b . 9.3-Ic { 11.2-Id ; 11.2-2 Description and Basis for Channe: The waste sludge tank IWSS-TK14, pump IWSS-P14, associated equipment, and instrumentation j were designed to collect resins and sludge from the phase separator and backwash tanks, mix them ; into a homogeneous mixture and, after sampling and analysis, deliver them for processing. During j preoperational testing, the system completion was delayed and temporary bypass capability was I installed in order to start-up the plant. It was reported that the sludge tank pump was never , satisfactorily tested. He objective of this modification is to make the tank functional as a mixing ] ! tank with sampling capabilities. The final retest for this modification will include preoperational testing of the sludge tank and its associated equipment. ; i Summary of Safety Evaluation:
- - The changes to the waste systems only affect the internal flow paths and components. The liquid and radwaste inventory are not changed in either quality or quantity. De system functions and l equipment classifications are not revised. He radioactive effluents released to the environment are l
- not affected by this modification. Herefore, this modification does not increase the probability of l an accident previously evaluated in the SAR. He modified system is bounded by the existing accident analysis. The added equipment meets or exceeds quality and classification standards described in the SAR. Therefore, this modification does not increase the consequences of any accident previously evaluated in the SAR. Considering the above reasons, this change does not .
1
' create the possibility of an accident different from any previously evaluated in the SAR. His )
entire system and associated components are contamed entirely in the radwaste building. The j radwaste building does not contain any safety related system, structure, or component (SSC) i interfaces or any safety related SSCs. His modification does not affect the function of any safety ' l
= Section I Page 113 of I87 4
,s.- - - - - - , - . . - - - - - - . - . - - b-
related SSC. Herefore, this modification does not increase the probability of a malfunction of any safety related SSC previously evaluated in the SAR. His modification does not increase the consequences of a malfunction of any safety related SSC previously evaluated in the SAR. His modification does not create the possibility of a malfunction of any safety related SSC different from any previously evaluated in the SAR. Here is no margin of safety associated with the radwaste system described in the technical specifications. Therefore, this modification does not reduce the margin of safety as dermed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 11.2-046 Pages 9.2-30 9.2-31 9.2-44 I1.2-15 11.2-16 Figures: 9.2-l a 9.2-l e 11.2-Ig Description and Hasis for Chance:
~
This modification deleted the alternate dilution from the standby senice water system to the circulating water system and converted the line into a liquid radwaste discharge line to the circulating water system. This new liquid radwaste discharge line supplements the existing discharge line to support increased liquid radwaste discharge flow. This modification also relocated the sample inlet for radiation monitor 1RMS-RE108 from its previous point on the sample line to the point on the blowdown line currently used as the sample chamber purge. Administrative controls have been implemented to ensure that 1RMS-RE108 will not be relied upon to detect the presence of radioactive material in the circulating water blowdown in the event the circulating water bypass line is being used without the normal flowpath also in use. Summary of Safety Evaluation: Standby senice water alternate dilution is not relied upon to prevent any accidents from occurring. It is an alternate source of dilution water which has never been used. The only initiating event in applicable accident scenarios is liquid radwaste system piping failure. Since all piping and support changes were per accepted codes and standards, the possibility of a piping failure was not increased. Relocation of the sample inlet for IRMS-RE108 does not interact with any components postulated to fail in any previously analyzed event. Therefore, this modification does not increase the probability of an accident or the probability of a malfunction of equipment important to safety previously evaluated in the SAR. Performance of barriers that serve to mitigate the consequences of analyzed accidents are not impacted since the release occurs outside the containment building. The only factor potentially impacted by this modification was the concentration ofisotopes in the discharge, which is affected by the amount of dilution water. Engineering and administrative controls are relied upon to ensure that adequate dilution is provided so that all liquid radwaste system discharges are within 10CFR20 limits. His modification did not impact any of these controls. Therefore, this modification does not increase the consequences of an accident or the consequences of a malfunction of equipment important to safety previously evaluated in tlc SAR. The only initiating event in applicable accident scenarios is liquid radwaste system piping failure. Section i Page 114 of187
he modification involves only the re-routing of piping, so no new failures were introduced. Since relocation of the 1RMS-RE108 sample point does not change the ability of the monitor to detect or alarm the presence of radioactive material in a nonnally non-radioactive stream, and there is no change to the monitor operation or function, there are no new accidents crcated. Therefore, this modification does not create the possibility of an accident or the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. No technical specification margins of safety are affected since the proposed modification does not impact the capability of the liquid radwaste system or the radioactive effluent monitoring instrumentation to ensure concentrations of radioactive material released in liquid effluents is within 10CFR20 limits. Hus the capability for monitoring both the actual and postulated liquid release pathways is maintained. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number /USAR Section: LCN 11.3-30 Page 11.3-7 Description and Basis for Chance: This change corrects the hydrogen analyzer high alarm setpoint of 1% as given in the SAR. The correct value for the hydrogen analyzer high alarm setpoint is 2%. Summary of Safety Evaluation: As the revised 2% hydrogen concentration high alarm value is well below the flammability concentration of 4%, there is no significant change to the ability of operations personnel to take ; compensatory action before reaching the flanunability concentration limit. The ability of other l system features provided to maintain the hydrogen concentration well below the 4% flammability limit are unaffected by this setpoint change. He change to reflect the correct setpoint value does I not increase the probability of the events that could cause an offgas system leak or failure, thus there is no corresponding increase in the probability of an offgas system leak or failure occurring. He analyzers are not postulated to initiate any malfunction of equipment important to safety either during normal analyzer operation or as a result of analyzer malfunction. Therefore, this change does not increase the probability of occurrence of an accident or the probability of a malfunction of safety related equipment evaluated previously in the SAR. The quantity and composition of radioactive material in the offgas system or within any other equipment important to safety is unchanged. Herefore, this change does not increase the consequences of either previously evaluated accidents or previously evaluated malfunctions of equipment important to safety. The design and operation of the offgas system and its components are unchanged, thus there is no change to the postulated failure mechanisms. Therefore, this change does not introduce an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The revised setpoint value is well within the required 4% maximum which provides assurance that hydrogen detonation and the associated release of radioactive material will not l occur. Herefore, this change does not reduce the margin of safety as defined in any technical specification or the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Section i Page 115 of187
Channe Number /USAR Section: LCN 11.3-031 Figure 11.3-2a Description and Basis for Chanee: This modification replaces an existing non-safety-related differential pressure switch (PDS) with a ; flow indicating switch (FIS). His FIS will provide the input to control room alarm for offgas i pretreatment. This alarm presently receives its input from PDS OFG-PDSI17, which will be spared-in-place by this modification. Figure 11.3-2a will be revised to show the instrument valves , associated with the PDS as closed, and the differential pressure switch noted as " spared-in-place." Summary of Safetv Evaluation: The spared-in-place PDS, the replacement FIS, and the offgas combination pretreatment sample panel will not cause any accident described in the SAR. The replacement FIS and the spared PDS are non-safety-related components and have no interface with any important-to-safety system, , structure or component (SSC). Failure of the FIS and PDS would have no affect on any safety-related SSC. Therefore, this modification does not increase the probability of an accident or the l probability of a malfunction of equipment important to safety previously evaluated in the SAR. ! He differential pressure switch and the FIS are located in an offgas pretreatment sample panel; all these components are non-safety-related; none are used to mitigate the consequences of an accident; and none of them interface with any safety-related SSC. The PDS and FIS will not affect ! the operation of either the offgas pretreatment and post treatment radiation monitors or the , hydrogen analyzer. The replacement FIS, like the existing PDS, only initiates an alarm which does i not require any short tenn operator actions. There are no automatic actions associated with the alarm. Failure of the FIS and PDS will have no affect on any safety-related SSC and will not result in any change to radiological consequences at the site boundary. Therefore, this ! modification does not increase the consequences of an accident or increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The replacement FIS, like the item being replaced, is non-seismic. He instrument root valves upstream of the spared-in-place PDS will be closed to ensure no potential for leakage. However, the root valves i are also non-safety-related and nonseismic. The sparing in place of the existing pressure switch, ! the closing of the associated instrument root valves and the replacement of existing flow indicator with a FIS will not have an interface with any SSC. All work will be internal to the non-safety- ~ related, nonseismic offgas combination pretreatment sample panel. Therefore, this modification does not create the possibility of an accident or the possibility a malfunction of equipment important to safety different from any previously evaluated in the SAR. The differential pressure switch, the replacement flow indicating switch and the offgas combination pretreatment sample panel are non-safety-related components and are not used to maintain the margin of safety defined in the bases. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. i Section I Page 116 ofI87 i
l ~ Channe Number /USAR Section: LCN I1.414 Pages: 11.4-7 l I1.4-9 !
' 12.4-5 {
] 12.4-6 !
1.2-17 ; Tables: 3.2-1 i i 12.4-2 ! Figure 1.2-2 I i Description and Basis for Channe: t I his change will add a building to store low level radwaste generated by the plant. The structure r will include storage for dewatered plant resins and dry active waste (DAW). ne structure will be for storage only. All processing of the radwaste will be done before it is shipped to the facility. t The transport of containers will be accomplished entirely within the property boundary limits of the l power station. > 4 Sununary of Safety Evaluation: , s
- Storage oflow level radwaste does not create an initiating event for any of the accidents evaluated i in the SAR. Ac storage oflow level radwaste in the new remote structure does not involve or i
!- impact any of the systems or components of the gaseous and liquid waste systems subject to l
- failure. The waste to be stored will contain no gasses and only minuscule amounts ofliquids. l i Vulnerability to tomadoes is not increased because wastes stored in the original radwaste building i are not tomado protected. Susceptibility to flood is not increased because the highest elevation
! design flood water level is below 100 feet. The top of the slab for the new facility is at 110.5 feet. . The new facility is not designed to withstand the effects of an earthquake. Since River Bend ! I
- Station is located in an carthquake zone with one of the lowest frequencies of occurrences in the
[ country, the increase in probability due to an earthquake is negligible. Operational accidents are ! l no more likely to occur because the usage of Se new facility is much less than the operation of the
- existing radwaste building. Therefore, this change does not increase the probability of an accident previously evaluated in the S AR. There are no radioactive gasses or liquids available for release to the environment except for small amounts of water (1%) that can not be removed from the dewatered resins. Ecrefore, this change does not increase the consequences of an accident 1 evaluated previously in the SAR. In the cases of the accidents postulated for the new facility, possible releases result in doses at the station boundary no greater than only a small fraction of the 10CFR100 limitations for the duration of an accident. Therefore, this change does not create the possibility of an accident different from any previously evaluated in the SAR. None of the
- functions or operations of plant systems or components are altered as a result of the construction !
!- and operation of the new low level radwaste storage facility. He new facility is fully equipped and ) adequately controlled to handle the storage of the wastes. Therefore, this change does not increase ) l the probability of an occurrence of a malfunction of equipment important to safety evaluated previously in the SAR.' He maximum exposures from the accidents postulated for the new facility ) l are no worse than those evaluated in the SAR. Therefore, this change does not increase the l consequences of a malfunction of equipment important to safety presiously evaluated in the SAR. l Here are no new failure modes created by the construction and operation of the new facility. l l Therefore, this change does not create the possibility of a malfunction of equipment important to safety different from any previously evaluated in the SAR. Technical specification dose limits are j based on liquid and gaseous releases and not from radiation fields emanating from the facility. The !
- Section 1: Page i17 of187 1 l
4-i
dose rate at the nearest point of unrestricted access to the facility is 6 mR/lu;. which is well below the station limit of 2.0 mR/hr for such areas. Since the facility is more than 1000 feet away from the nearest station boundary line, the dose rate at the station boundary is minuscule. Herefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 12.03-011 Pages: 12.2-7 12.3-17 12.4-6 Table 12.4-2 Description and Hasis for Chance: This change adds two modular fiberglass reinforced plastic turbine rotor enclosures. These enclosures will be located outside the protected area approximately 800 feet southwest of the turbine building. Each enclosure is 40 feet long,17 feet wide, and 16 feet high and will be located on concrete slabs. These enclosures will facilitate the on-site storage of two low pressure turbine rotors and four 8-stage turbine diaphragms. These components have been surveyed for radioactivity and have been classified as fixed contamination low-level dry active waste. There are no utilities provided to the enclosures. Summary of Safety Evaluation: I l He enclosures will not be used for or house any systems or equipment used for safe operation or l safe shutdown of the plant. The enclosures have no stmetural component that is located closer that j 1100 feet from any plant site seismic category 1 safety related system, structure, or component (SSC). The presence of the enclosures will have minimal effect on the wind loading of any safety l related SSC. Any missiles detached from the enclosures or their stored components all not ! jeopardize the structural integrity of any safety related SSC. The enclosures and associated l concrete slabs are not in the flooding envelope and are located above the flood elevation. There is . sufficient separation between the enclosures and safety related SSCs to preclude any adverse fire I interaction. Therefore, this change will not increase the probability of an accident previously evaluated in the SAR. The stored components contain low-level fixed contamination and have a radiation field isolated to the turbine rotors, diaphragms, and localized surrounding areas which are ; not affected by any postulated event occurring inside or outside of the plant. There will be no ) consequential airbome dispersion because all radioactive contamination is fixed. The enclosures ; will not adversely affect emironmental windage conditions associated with the calculation of on- l site or off-site normal, cmergency and accident dose rates. Therefore, this change does not merease l the consequences of an accident previously evaluated in the SAR. The modular enclosures, i associated concrete slabs and stored components do not invalidate existing seismic and tornado missile design criteria, probable maximum precipitation flooding analysis conclusions, or pose a serious fire hazard to any surrounding plant SSC. Therefore, this change does not create the possibility of an accident or the possibility of a malfunction of any safety related SSC different from any previously evaluated in the SAR. There are no safety related structures sufficiently close to the enclosures to be affected by any seismic response of the enclosures or any soil / structure interaction problems. Therefore, this change does not increase the probability of a malfunction of any safety related SSC previously evaluated in the SAR. He addition of the enclosures and the associated concrete slabs will not cause any essential plant systems and equipment to function Section i Page 118 of 187 I i
differently than assumed in any accident analysis. Therefore, this change does not increase the consequences of a malfunction of any safety related SSC previously evaluated in the SAR. The enclosures, their associated concrete slabs, and the components to be stored are not the subject of any technical specification. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number /USAR Section: LCN 12.5-8 Pages: 12.5-6 12.5-15 12.5-16 Tables: 12.5-2 12.5-3 12.5-4 Sheet 1 12.5-1 Description and Basis for Change: This change allows different types and quantities of portable instruments used for personal monitoring. This change also allows calibration ofinstruments to be performed by other than plant personnel. New instrument types make use ofimproved technology and meet the standard criteria for selection ofinstruments. The changes in quantities of each type ofinstrument meet the criteria established for quantity ofinstruments. Counting systems may be calibrated by vendors or by Entergy Operation's central calibration facility. Calibrations will also be permitted to be p:rformed by other facilities. Summary of Safety Evaluation: The change to type, quantity, or calibration of portable instrumentation used for personal monitoring does not affect any plant system, structure, or component (SSC). Therefore, this change neither increases the probability of an accident nor the probability of a malfunction of equipment important to safety previously evaluated in the SAR. This change neither increases the consequences of an accident nor the consequences of a malfunction of equipment to safety previously evaluated in the SAR. This change neither creates the possibility of an accident nor creates the possibility of a malfunction of equipment important to safety difTerent from any previously evaluated in the SAR. This change does not reduce the margin of safety as defined in the bases of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Channe Number /USAR Section: LCN 13.01-33 Page 13.1-20 Description and Basis for Channe: This enange to the SAR is being made to remove reference to the welding program from the responsibiihy of the Superintendent-Plant Services under the Maintenance Section 13.1.2.1.2. This change is being made due to administrative changes that will place the administrative responsibility for the welding pogram under the Director-Design Engineering. Section I Page 119 of 187
Summary of Safety Evaluation: A change in responsibility between the Director-Site Design Engineering and the Manager-Maintenance will not cause an increase in the probability or consequences of an accident previously evaluated in the SAR because the welding program standards that provide technical requirements of the code will not change. The change in responsibility from the Director-Site Design Engineering to the Manager-Maintenance will not create the possibility of an accident which is different than any previously evaluated in the SAR because the welding program standards that provide technical requirements of the code will not change. He probability and consequences of a malfunction ofimportant safety equipment, as previously evaluated in the SAR,
. will not increase with the change in responsibility between the Director-Site Design Engineering and the Manager-Maintenance because, the welding program standards that provide technical requirements of the code will not change. The margin of safety, as defined in the basis for any technical specification, does not reduce with the change in responsibility between the Director-Site Design Engineering and the Manager-Maintenance because, the welding program standards that provide technical requirements of the code will not change.
Channe Number /USAR Section: LCN 13.1-34 Pages: 12.5-2 13.1-18 13.1-19 13.1-23 13.2-15 13.2-34 Figures:13.1-5 Sheet 2 13.1-6 Description and Basis for Channe: He Radwaste Department was reorganized. The solid radwaste function and personnel were moved into the radiation protection department and the liquid radwaste function and personnel were moved into the chemistry department. The description of the radiological engineering group is being deleted from the SAR. The description of training requirements for the radiological engineering group is being changed to describe appropriate training for the reorganized radiation protection staff. Summary of Safety Evaluation: This is purely an administrative change. All the functions of the radwaste department are being preserved. It will neither increase the probability of an accident occurring, increase the consequences of an accident, nor create the possibility of the occurrence of a different type of accident as previously evaluated in the SAR. It will neither increase the probability of a malfunction of equipment important to safety, increase the consequences of a malfunction of equipment important to safety, nor will it create the possibility of a malfunction different from any other one evaluated in the SAR. This change does not affect any of the technical specifications. Herefore, it does not decrease the margin of safety as defmed in the basis for any technical specification. For these reasons, this modification does not constitute an unresiewed safety . question. Section I Page 120 of I87
Change Number /USAR Section: LCN 13.5-22 Pages 13.5-2 13.5-15 13.5-21 a i- Description and Basis for Channe: l
- He procedure that governs the general requirements for control and use of procedures was i i' streamlined. He procedure was changed to allow either the General Manager Plant Operations 4
- (GMPO) or the mangers reporting to the GMPO to approve plant administrative procedures as
; long as those procedures are owned by that manager's department. Superintendents in the
- maintenance department were allowed to approved procedures specific to their discipline. :
Directors and managers who directly report to the Vice President-Operations were allowed to j approve all but one of the upper tier procederes. l t Summary of Safety Evaluation: ! l Procedures are not directly identified as potential initiators of accidents in the SAR This I procedure was revised to provide the appropriate verification and validation processes to ensure ! that procedures are appropriately reviewed to minimize the potential for operator errors or j
^
malfunctions of equipment important to safety. The approval levels were revised such that ownership of the procedure would be felt at the level where accuracy and effectiveness can best be 3 determined and implemented. Therefore, this change did not increase the probability or : L consequences of an accident or malfunction of equipment important to safety previously evaluated ! in the SAR. This procedure was revised to preclude the creation ofnew operator errors or new ; malfunctions of equipment to safety. Therefore, this change did not create the possibility of an ; accident or a malfunction of equipment important to safety different from any previously evaluated , in the SAR. This change did not affect any technical specification and did not affect the basis of i any technical specification. For these reasons, this modification did not constitute an unreviewed l
- - safety question. [
f I v l i l l i o l l < i l r. Section I Page 121 of 187
, -r n
i SECTION II LICENSE AMENDMENT REQUESTS Section 11 Page 122 ofI87
i l i Channe Number: License Amendment Request (LAR) 93-12 (MR93-0006) j l l. Description and Basis for Channe: The frequency of turbine steam valve testing is being changed from weekly to quarterly for main l stop valves and combined intermediate valves. Also, the frequency of testing for control valves is ' l being changed from monthly to quarterly. This is due to the installation of new low pressure (LP) ! turbine rotors. De analysis of the new rotors provided by the manufacturer recv.. ..cr.ded the ! i quarterly testing interval for the aforementioned valves. j J Summary of Safety Evaluation: ; l The quarterly steam valve testing interval is the interval recommended by the manufacturer to l' j remain within the predictive capabilities of the overspeed analysis. He new missile analysis stated that the monoblock stress levels are very low, the keyway cracking mechanism is not present for a ; monoblock rotor, and the probability of missiles being generated for the monoblock rotor is not ! present. The established missile probability limits, which only address the LP rotor missile l l~ generation, are maintained. The changing of the steam valve testing interval to quarterly does not l } affect any equipment designed to mitigate the consequences of an accident. Derefore, there is no j [ increase in the consequences of an accident previously evaluated in the SAR. The replacement of j i the LP rotors with monoblock rotors made the change in the valve testing interval possible. The i
- rotor was manufactured to design specifications and standards that meet or exceed the original
- design requirements. Even in the event a destructive overspeed level was reached, the LP rotors ;
I are not expected to generate missiles. %crefore, there is no creation of the possibility of an ! l accident which is different than any previously evaluated in the SAR. Here are no overspeed l' ! levels that can be achieved by the main turbine that could cause the monoblock rotors to generate I missi!cs. %crefore there is no increase in the probability of a malfunction of equipment important to safety previously evaluated in the SAR. There is no increase in the consequences of a i malfunction of equipment important to safety previously evaluated in the SAR. Nothing is added 3 j to increase the possibility of a malfunction of equipment important to safety different from any !
- evaluated in the SAR. The main turbine system is not described or mentioned in any of the j technical specifications. Therefore, there is no reduction of the margin of safety as defined in the l 4 basis for any technical specification. For these reasons, this does not constitute an unresiewed
- safety question.
- i Channe Number: LAR 94-06
- Description and Basis for Chanre:
This change adds more primary contairrnent penetration conductor overcurrent protection devices l to be tested in accordance with the surveillance requirements in the Technical Requirements . Manual (TRM). Primary containment penetration conductor overcurrent protection desices ensure I the pressure integrity for the containment penetration they protect. Postulated failure of these devices is that the wire insulation could degrade, where it passes through the containment penetration, resulting in a containment leak during a loss of coolant accident. Adding the additional protective devices helps ensure they meet this function. l I Section 11 Page 123 of187
u . 4 d l'
- Summary of Safety Evaluation
i i i Redundant protective devices are currently installed to ensure that the Conax penetration conductor is not damaged under fault conditions. His change ensures all the proper suneillances are performed on the added protective devices. He containment electrical penetration assemblies are j
- MyA to withstand, without loss of mechanical integrity, the maximum fault current versus time !
condition, which could occur from a single random failure of circuit overload protective devices. ! j None of the affected components are the initiator for any accidents. Herefore, this change does j 1 not increase the probability of an accident previously evaluated in the SAR. Primary centainment i penetration conductor overcurrent protection devices ensure the pressure integrity of containment j penetration. With failure of the device, it is postulated that the wire insulation will degrade } resulting in a containment lead path during a LOCA. The protective devices are designed to j protect the circuit conductors against damage or failure due to overcurrent heating effects and ; subsequent penetration failure. De only safety function of the protective devices is to provide the !
! overcurrent penetration to preclude containment penetration degradation. This change has been l 1 shown to not have any effect on accident mitigation. Therefore, the consequences of an accident j i previously evaluated in the SAR are not increased. It has been determined that the use of an ;
t appropriately sized overcurrent protection device is bounded by the existing design and evaluated !
- i. to be acceptable. No accident scenarios will be created as a result ofimplementing this change. !
Therefore, the possibility of an accident which is different than any previously evaluated in the
- SAR is not created. Since the function of the overcurrent protection devices is performed by
- isolating the device to prevent degradation of the wire insulation, no containment leak path will .
!- occur. Therefore, the probability of a malfunction of a safety related structure, system or j i component (SSC) previously evaluated in the SAR is not increased. This change will not affect the ! i
- ability of any safety related system in performing its function to mitigate the consequences of
- postulated accidents or transients. Therefore, the consequences of a malfunction of a safety related f SSC previously evaluated in the SAR is not increased. This change brings the surveillance
- j. requirements for overcurrent protection desices in the TRM into compliance with its design basis l and will not impact any other safety related SSC. Therefore, the possibility of a malfunction of a j safety related SSC different than any previously evaluated in the SAR is not created. The revision f i to the TRM brings into compliance the affected table with the surveillance requirements for l containment penetration overcurrent protection devices. There is no impact on any other safety [
, related system to perform its function, and this TRM revision has no functional impact on the ! 4 Technical Specifications. Therefore, the margin of safety as defined in the basis for any technical l specification is not reduced. For these reasons, this modification does not constitute an unresiewed ' safety question. j l Channe Number: LAR 95-17 ! Description and Basis for Channe: [ River Bend safety-related setpoints were originally incorporated into the Technical Specifications (TS) without taking into consideration the applicable setpoint data sheets (SPDS) " upper" and !
" lower" limits. This resulted in the SPDS " nominal" setpoint being identified as the TS setpoint !
l without the allowable tolerance. Removing the < or > symbol for all instrument trip setpoints l 4
. identified in the Technical Requirements Manual (TRM) ensures that the TRM and the applicable .
design basis documents are consistent in identifying the trip setpoint as a nonunal value rather than j a bounding limit. ! r i Section 11 Page 124 of I87 i i
?
i l i Summary of Safety Evaluation: l De recalibration of the setpoints to the design basis trip values will not degrade the performance l of a safety system in the accident analysis. Derefore, this modification does not increase the - .j probability of an accident or a malfunction of equipment important to safety previously evaluated ! in the SAR. The ability of the existing instruments and associated loops to perform their safety functions is unaffected by this change since the trip settings are within their design basis settmgs.- l 1 Berefore, this modification does not increase the probability of an accident or a malfunction of .l' equipment important to safety previously evaluated in the SAR. No new equipment was installed nor was the operation of wy existing equipment impacted by this modification. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The margin of safety is the difference between the design or failure limit and the acceptance limit. All instrumentation setpoints are set so there is no impact to the system operational limit, acceptance limits, limiting safety system settings or design / safety limits (process safety limits). This practice ensures that no . change to the acceptance limit occurs and subsequently no change to the margin of safety. ! hrefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. i f Channe Number: LAR 95-19 Part 1 l Description and Basis for Channe: l l This proposed change makes the neutron monitoring system control rod block function surveillance ! frequencies and applicabilities in the technical requirements manual (TRM) consistent with ; Improved Technical Specification (TS) requirements for these same instruments and the retained j 1 control rod block functions.- Summary of Safety Evaluation: ! The screening identified functions required to prevent an evaluated accident and required them to l be retained in the TS. He SER identified the neutron monitoring functions as only providing j redundant capability to the TS required functions. Therefore, this modification does not increase - the probability of an accident previously evaluated in the SAR. The neutron monitoring control ; rod block fimetion does act to terminate the worsening of some conditions resulting from rod withdrawal errors but this function is not credited with any reduction of accident monitoring : consequences. Therefore, this modification does not increase the consequences of an accident i previously evaluated in the SAR. The control rod blocks affected by this change only function to terminate rod withdrawals. The possible accidents associated with rod withdrawal are essluated in the SAR. Herefore, this modification does not create the possibility of an accident different from any previously evaluated in the S AR. This change applies the previously reviewed and approved frequencies and applicabilities of the same instruments used for the safety functions in TS to the control rod block function of these instruments. %crefore, this modification does not increase the probability of a malfunction of equipment important to safety presiously essluated in the SAR. The consequences of the failure of these instruments are esaluated as part of the basis for their safety functions' frequencies and applicabilities. Loss of neutron monitoring control rod block functions were identified in the application of selection criteria to the River Bend Technical Section 11 Page 125 of 187 t
.._._...~_m -
Specifications as being non-significant risk contributors to core damage frequency of offsite releases. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. This is a change in procedural frequency and applicability only and does not affect procedure performance or system operation. Therefore, this modification does not create a malfunction of equipment important to safety different from any previously evaluated in the SAR. His change is procedural frequency and applicability only for features not required by the TS or their bases. Thereforc, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Change Number: LAR 95-19 Part 2 Description and Basis for Change: Limiting condition of operation (LCO) 3.0.3 directs plant shutdown when a technical specification LCO is not met and either the associated actions are not met, associated actions are not prosided, or if directed by the associated actions. Plant shutdown is the appropriate compensaton action for failure to meet the requirements for the systems and equipment selected for retention in the Technical Specifications (TS)in accordance with the screening criteria established in the NRC fmal policy technical specification improvements. It is not necessarily the appropriate action when a relocated technical requirements manual (TRM) requirement which was screened as not meeting the safety criteria retained in the Technical Specifications is not met just as it is not always appropriate when other SAR requirements are not met. Because the previous TS required actions for equipment were, in general, retained in the TRM without performance of specific evaluations to alter or climinate the actions and provide for replacement of LCO 3.0.3 in each indisidual specification, a consistent process for performance of that evaluation on an as needed basis was developed as TLCO 3.0.3. The TLCO is consistent with Generic Letter 91-18 guidance for justification of continued operation which requires that a reasonable assurance of safety determination be made. l Summary of Safety Evaluation: Since none of the relocated items are in the primary accident prevention or mitigation success paths, their removal from Technical Specifications (TS) places them in the same categoiy as all l existing important to safety design features required by the SAR which do not meet the criteria to l be included in the TS. Application of the existing guidance for safety determinations assures that this change does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Relocation of the equipment operability requirements from the TS to the SAR also removed the specific plant shutdown requirement of LCO 3.0.3. The TLCO is consistent with Generic Letter 91-18 guidance forjustification of continued operation which requires that a reasonable assurance of safety determination be made. Application of the existing guidance for safety determinations assures this change does not increase the consequences of an accident or a malfunction of equipment important to safety presiously i evaluated in the SAR. The compensatory actions authorized by TLCO 3.0.3 use previously approved plant procedures for performance of actions. The Station procedure control and issue process has already determined that these procedures do not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the S AR. Herefore, this modification does not create the possibility of an accident or a malfunction of , I i i Section 11 Page 126 of 187
-~ .. . .- . ~-- -. - _ - - - . .. - - - - -
s i equipment important to safety different from any previously evaluated in the SAR. The , [ requirements affected by TLCO 3.0.3 are no longer contamed in TS. Herefore, this modification l does not reduce the margin of safety as defined in the basis of any technical specification. For ' i these reasons, this modification does not constitute an unreviewed safety question. l I Channe Number: LAR 95-20 !
- Description and Basis for Channe
- l l !
The average airspace temperature affects the calculated response to design basis accidents. His ! primary containment air temperature limit is an initial condition for the safety analyses. Technical .! Requirements Manual (TRM) surveillance requirement (SR) 3.6.1.5.1, " Containment Atmosphere ! I ! Temperature," includes six temperature detectors (CMS *RTD42D, E, F, H, j ud K) which j provide input to four control room recorders (CMS *TRX42A&B and CMSTRY42A&B). One of i the temperature detectors (CMS *RTD42D) is shown in the TRM as being located at azimuth 15 i l degrees, elevation 119 feet it is actually located inside the main steam tunnel between the drywell ; i- wall and the containment wall. As a result, it consistently indicates approximately 20 degrees F j higher than any of the other five temperature indicators. The equipment configuration in the ! i vicinity of the actual location of the temperature detector does not necessitate the need for air temperature monitoring. Since the containment ventilation system provides sufficient air flow in l
- . the containment to ensure that adequate air mixing occurs and that the bulk air temperature is l relatively uniform for a specific elevation, the remaining five containment air temperature detectors l j are adequate. Therefore, temperature detector CMS *RTD42D was deleted from TRM SR ;
3.6.1.5.1. He existing detector was not altered, the recorder point was removed from the recorder. j i
. Summary of Safety Evaluation-
! l ! The containment air temperature detectors only provide inputs to control room recorders. They do l not have any interfaces or interlocks with any other equipment important to safety and do not ! provide any automatic actuation signal to any equipment. Therefore, this modification does not l increase the probability of an accident or a malfunction of equipment important to safety , previously evaluated in the SAR. Containment air temperature is not a variable tint is required to ! meet post-accident monitoring per Regulatory Guide 1.97. This modification did not increase the radiation sources or affect release paths or accident mitigation equipment. Therefore, this t modification does not increase the probability of an accident or a malfunction of equipment i important to safety previously evaluated in the SAR. No new equipment was installed nor was the l l operation of any existing equipment impacted by this modification. Therefore, this modification i does not create the possibility of an accident or a malfunction of equipment important to safety l different from any previously evaluated in the SAR. Deletion of the subject temperature detector .i: input to the average containment air temperature still provides the operators with a valid indication
;; of the containment air temperature. This allows the operators to verify that the contamment air !
a temperature does not exceed the initial conditions for the design basis accident. Therefore, this j , modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. t l I I i Secten 11 Page 127 of I87 l l i . . _ _ a,
i l < 1 Channe Number: LAR 95-24 e Descriotion and Basis for Channe: This change places a containment air temperature monitor actually located at the 122 foot level m i the same group as containment air temperature monitors for the 119 foot elevation in the technical requirements manual (TRM). He subject monitor was moved up slightly due to interference in its ; intended location. His relocation will avoid unnecessary entry into a limiting condition of : operation. l Summary of Safety Evaluation: ;
- De containment monitoring system does not cause or contribute to an accident but it is used to mitigate the consequences of an accident. The existing detector and recorder are safety related but i have no interface with any equipment postulated to cause any accidents in the SAR. No new .
equipment is being added nor is the existing equipment being altered. Therefore, this change does ! not increase the probability of an accident or a malfunction of equipment important to safety . previously evaluated in the SAR. A review of RBS licensing and design basis determined that there is no specialjustification for mounting the detector at elevation 122 feet other than the - physical interference. There no unique reason for the number of detectors other than good design ; practice. In addition, the TRM shows the elevations as approximate, as denoted by the "~" symbol : i preceding the elevation. Existing detectors do not provide any automatic function. Herefore, this change does not increase the consequences of an accident or a malfunction of equipment important
- to safety previously evaluated in the S AR. This change does not involve any physical modifications to the plant. This change does not alter the operation of any equipment in the ;
containment monitoring system or in any other plant system. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different , from any previously evaluated in the SAR. The average air space temperature affects the i calculated response to postulated design basis accidents (DBAs). Combining of the detector at the l . 122 foot elevation with the others at the 119 foot elevation will continue to allow operators to { verify that the containment air temperature does not exceed the initial conditions for the DBA, thereby not changing the margin of safety. Therefore, this modification does not reduce the margin ! of safety as defined in the basis of any technical specification. For these reasons, this modification ; i does not constitute an unreviewed safety question. ' Channe Number: LAR 95 27 Description and Basis for Channe:
)
In order to maintain contamment integrity in modes 4 and 5 during actisities that necessitate a controlled configuration volume be maintained, an isolation barrier must be established for each ! primary containment penetration. A column of suppression pool water over submerged lines can , be considered an equivalent isolation barrier since it can proside a leak tight barrier to limit potential fission product escape to the emironment in modes 4 and 5. In modes 4 and 5 the probability and consequences of a loss of coolant accident are reduced due to the low pressure and i temperature. To allow the column of water in the suppression pool over submerged discharge lines
- to be considered an equivalent isolation barrier, a revision is required to the Technical :
j Section 11 Page 128 of I87 ] f l
Requirements Manual. To provide this isolation function, however, the column must be maintained at a minimum of 18 feet. Summary of Safety Evaluation: The accident of concern during plant shutdown or refueling mode is a fuel handling accident. The analyzed fuel handling accident in the fuel building bounds a fuel handling accident inside primary
< containment. Use of the suppression pool as a containment isolation barrier does not affect the method nor the equipment used in handling the fuel. Here is no change to the plant and there is no change to any system function. His change utilizes an existing condition to satisfy a Technical Specification requirement to maintain primary containment during certain shutdown actisities.
Administrative controls will be in place to eliminate the potential actions which would drain suppression pool water level below the required amount. Therefore, this change does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Primary containment performs no active function in response to this event. Ilowever, atmosphere leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage rates assumed in the accident analysis. Loss of shutdown cooling during modes 4 and 5 is not specifically addressed as an accident in the SAR. Losing shutdown cooling for a period of time long enough to pressurize containment is highly unlikely. Acrefore, this change does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. No new accident initiators are introduced by this change. Administrative controls will be placed in the surveillance test procedure entitled " Primary Containment Shutdown Verification" to require the suppression , pool level remains at or above the minimum required level of 18 feet, and the associated containment isolation valves are closed. Therefore, this change does not create the possibility of an accident or malfunction of equipment important to safety. The assumed containment pressure of 0.3 psig plus instrumentation inaccuracies utilized to determine the minimum required column of suppression pool water is conservative since the bases uses a containment pressure of 0.367 inches of water differential pressure in determining the maximum containment leakage through vent and drain lines during a fuel handling accident. Any applicable containment isolation valves that could impact containment integrity will be isolated. The isolation of those valves could potentially impact the reactor core isolation system (RCIC) and the reactor heat removal system (RHR) vacuum relief. Ilowever, lack of steam supply makes RCIC inoperable when the reactor is shutdown and RHR vacuum reliefis not necessary since the reactor is in Mode 4 or 5 with a coolant temperature below 200 F. Accordingly, these systems are not affected by this change. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Channe Number: LAR 95-31 Description and Basis for Channe: This change revises Section 3.3.4.1 of the Technical Specification Bases, entitled "End of cycle Recirculation Pump Trip (EOC-RPT) Instrumentation," so that the original intent can be utilized for breaker configuration. He purpose of this instrumentation is to initiate a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients, thereby providing additional margin to core thermal minimum critical Section 11 Page 129 of I87
1 power ratio (MCPR) safety limits. Brough a misunderstandmg of plant configuration, the present
. bases wording does not reflect the true breaker configuration %e specific changes are as follows:
- 1) change the definition of breaker response time to "He time from application of voltage to the i trip coil until the ' arcing' contacts separate." he definition previously stated until the main contacts separate; 2) arc suppression is changed similarly to the breaker response time and is now
' defmed as the time from arcing contact separation until the complete suppression of the electrical ; are across the open contacts; and 3) change wording from " Breaker response shall be verified by testing and added to the manufacturer's design are suppression time of 140 ms to determinc breaker interruption time. The breaker arc suppression time shall be validated by the performance I of periodic' contact gap measurements and high potential tests on the breaker vacuum interrupters i
in accordance with plant procedures" to " Breaker response shall be verified by testing to be within j the manufacturer's design response time. Testing of the breaker response time verifies the design l j intermption time to be 5 five cycles (83.3 ms). Breaker are suppression shall be in validated l
; during testing by visual observation of puffer performance and insulation testing of the breaker arc j chutes."
- I Summary of Safety Evaluation
- :
i Changing the method of verifying breaker interruption time can only affect the response to the 3 accident and not the probability of an accident occurring. The proposed change does not modify any system interfaces in a way that would increase the likelihood of an accident. Therefore, no j 1 new failure modes are created. Also, the probability of occurrence of an RPT accident as described in the SAR will not be increased since these changes only pertain to the response time for ! the RPT and do not affect the actual maintenance performed on the breaker or the actual operation of the breaker. Therefore, these changes do not increase the probability of an accident presiously evaluated in the SAR. These changes utilize a more conservative value for recirculation pump ! 1 breaker arc suppression time by utilizing the manufacturer stated full value of breaker interruption i time. The changes do not alter any assumption previously made in evaluating the radiological l consequences and do not increase the consequences of an accident previously evaluated in the : SAR. The changes only affect the testing methodology used in determining EOC-RPT breaker l interruption time and do not change the breaker design in any way. No new failure mechanisms are ! introduced, no additional failure modes are created and no system functions or interfaces are impacted. Therefore, these changes do not create the possibility of an accident different from any j previously evaluated in the S AR. These changes only affect the method for testing breaker interruption time. Total system response time is still limited to s 140 ms. Breaker interruption ! will continue to be used in evaluating total system response time. He bicaker will still perform its design function within the allowable time restraints and no assumptions have been altered. Therefore, these changes do not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. Since no new failure mechanisms are introduced, no l additional failure modes are created and no system functions or interfaces are impacted, these l changes do not increase the consequences of a malfunction of equipment important to safety . ) previously evaluated in the SAR. The changes affect only the testing methodology. Breaker design 1 and safety function are the same. Berefore, these changes do not create a malfunction of equipment important to safety different from any previously evaluated in the S AR. Total system
- response time is still limited to s 140 ms and a more conservative value is being used for recirculation pump breaker are suppression time. Therefore, these changes do not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, these
, changes do not constitute an unreviewed safety question. ; t Section 11 Page 130 of 187
1 I 1 l l Channe Number: LAR %-007 Description and Basis for Channe: ) Move Note 'C' from the ' Sampling Frequency' to the ' Type of Analysis' on principal gamma emitters and within Note 'C,' move the "One hour" from following the sample and analysis to following 15% of rated thermal power, Relocating the note will only require the sampling and analysis on the principal gamma emitters from which the information is needed. He first change will also include numbering the conditions which need to be evaluated for sampling. Changing the location of the one-hour time limit will remove the time restriction from the sampling and analysis allowing time to complete the analysis. The location will also limit the sampling to those power changes which should be reviewed for possible fuel leakage. The second change will remove the requirement for the tritium analysis upon startup, shutdown, or power changes. The final draft of the RBS low power operating license submitted for NRC review contained wording consistent with standard technical specifications (STS) guidance documents. However, when the low power license was issued, the wording had changed to require sampling and analysis within one hour. Several cases were identified where the submitted wording had been changed in the NRC approved version. No evidence was found that changes were requested and it appears that they were made during an editorial revision. Review indicates no SAR impact by this inconsistency. A resiew of other similar BWR operating plants specifications indicates that the requirement for sampling and analysis within one hour is not included. The other plants were consistent with the guidance within NUREG-0133 and within the STS. The delay in sample analysis is neither a non-compliance with regulatory programs, nor does it result in any loss of monitoring. Our ODCM (offsite dose calculation manual) follows the guidance in NUREG-0133 and in 10CFR50, Appendix I. He monitoring and the associated calculation establishes compliance with 10CFR20 requirements. Summary of Safety Evaluation: The design basis complies with 10CFR20. This increase in time will not change the releases. Therefore, the efTects on the public will not change and the change does not increase the probability of an accident previously evaluated in the SAR Tritium concentrations in reactor coolant are primarily due to temary fission neutron activation processes. These concentrations build to an equilibrium and do not change appreciably during a plant startup, shutdown, or thermal power change. No actions are associated with the change and all on-line monitoring instrumentation will function as required to limit gaseous effluents. Therefore, the change to the tritium sampling frequency will not significantly change the actual effluent and does not change or affect the consequences of an accident previously evaluated in the SAR. This change does not increase the possibility of an accident different from any previously evaluated in the SAR. Continuous on-line monitoring systems are credited in the accident analysis. With the monitoring system in place, event response will remain within the bounds of the previous analysis. The change in tritium sampling frequency is consistent with the generation of tritium in the reactor coolant. He change to the sample requirements to 15% rated power increases over a one-hour period is consistent with the purpose to sample under conditions which could result in fuel failures and is consistent with technical specifications. Regulatory Guide 1.21 addresses corrections associated with the decay of radionuclides. The delay in obtaining and analyzing a sample will not significantly affect the ' determmation of the radionuclides in the effluent and does not result in a loss of monitoring or non-compliance with regulatory programs. Therefore, this change does not create the possibility of a new malfunction of equipment imponant to safety er increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. Other monitoring corroboration Section 11 Page 131 of 187
is performed more often than monthly with samples being taken at the most probable source of tritium production. The change to sampling at 15% of rated thermal power is consistent with the generation of tritium in the reactor and with the pu pose of Technical Specifications (to sample under conditions which could result in fuel failures). Therefore, this change does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The requirements of 10CFR50, Appendix I, have not been revised ad the confirmation of offsite efiluent limits has been maintained. Therefore, the margin of safety related to sampling is maintained. For these reasons, this change does not constitute an unreviewed safety question. Channe Number: LAR 96-08 Description and Basis for Change: This change revised a Technical Requirements Manual (TRM) surveillance requirement (TSR) and a TRM limiting condition of operation (TLCO). The TSR (3.7.9.1.15) was revised to perform fire pump diesel engine inspection., at any time instead of being prohibited during plant operating Modes 1,2 and 3. Also, the fire pump diesel engine inspection frequency was changed from 18 months to 24 months. The TLCO (3.7.9.4) was revised to climinate the required action to proside gated wye (s) on the nearest operable hose station anJ clarify ins *: actions to provide a sign with directions for the proper hose to use. This will eliminate an .annecessary practice that is already provided by the plant hose standpipe design. Summary of Nety Evaluation: ne change to the TRM does not cause any fire protec. ion component to operate outside ofits design or tes* limits and will not affect any system hterface in a way that could lead to an accident. Fire protection system capacity nor redundancy is affected by this change. Therefore, this change does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This change does not affect any parameter which could alter radionuclide population, release rate or duration, create new release mechanisms or impact radiation release barriers. Therefore, this change does not increase the consequ:nces of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. No new or different fire hazards are introduced and no new or different exposures to safety-related components are created. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. Fire suppression system capability to minimize potential damage to safety-related equipment, by confining and extinguishing fires occurring in any portion of the facility where safety-related equipment is located, is not affected by t'.as change. Herefore, this modification does not reduce the margin of safety as defined in the t, . sis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Chance Number: LAR 96-15 Description and Basis for Chance: This change removes the requirement for the facility review committee to condact an annual resiew of the emergency plan and emergency implementing procedures (EIPs), when no changes are Section 11 Page 132 of I87
required, and to remove the requirement for the Director- Nuclear Safety and Regulatory Affairs, General Manager-Plant Operations, and Vice President to approve EIPs. Summary of Safety Evaluation: This change does not address any accident evaluated in the SAR. This change does not address any equipment important to safety. Therefore, this change does nct increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The emergency plan and EIPs provide the appropriate levels of review and approval to ensure that offsite does to the public as analyzed in the accident analysis is unchanged. Therefore, this change does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This change is restricted to administrative activity related to document reviews and approvals. This change does not address administrative controls related to the function or operation of any system, structure, or component. 'Iherefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This change does not address any margin of safety and does not address any safety related equipment or systems. Therefore, this change does not increase the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. L 6 Section 11 Page 133 of187
1 I l 1 i I i
)
i 1 I 1 l SECTION III l PROCEDURE CHANGES l l
)
l l l 1 l l l 1 Section 111 Page 134 of187 i
)
1 1 I Safety Evaluation Initiatine Document: Administrative Procedure (ADM) 0024 l Surveillance Test Procedure (STP) 050-3001 l Description and Basis for Channe:
- This program change redermes the average power range monito-(APRM) setpoints T-factor. The l modified APRM setpoints T-factor provides additional operating margin to achieve power distributions which allow fraction of core boiling boundary (FCBB) stability control without compromising linear heat generation rate (LHGR) protection for off-rated operations.
Summary of Safety Evaluation:
, The modified APRM setpoints T-factor allows power distributions which permit the application of ,
stability controls to increase stability margin, thus the probability for initiation of reactor , instability is significantly reduced. The proposed change in APRM setpoints T-factor maintains adequate oft-rated LHGR margin for all operating conditions. Therefore, this change does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. %e severity of a postulated reactor instability event is ' l significantly diminished since the initial reactor conditions are very stable. In addition, the modified APRM setpoints T-factor is confirmed to provide adequate LHGR protection at off-rated conditions for other anticipated events. Therefore, this change does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. De proposed changes do not involve any new modes of operation or any plant modifications. Therefore, this change does not create the possibility of an accident or a malfunction of equipment ; i important to safety different from any previously evaluated in the SAR. The higher power pealdng resulting from the APRM setpoints T-factor change is below applicable LHGR limits. The proposed change does not result in an increase of core damage frequency. Therefore, this change does not reduce the margin of safety as dermed in any technical specification or in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. , Safety Evaluation Initiatine Document: Annunciator Response Procedure 870 52 (SEN 95-0051) Description and Basis for Channe: Automatic closure of the steam jet air ejector (SJAE) suction valves on loss of motive steam supply to the second stagejet is designed to prevent offgas process flow from reaching flammable limits. One potential cause for main control room panel 870 alann number 522, " Steam To Air Ejector 1 A Extreme Low Flow," and alarm no. 524, " Steam To Air Ejector IB Extreme Low Flow,"is failure of the flow transmitter which provides the extreme low flow alarm and closure signal for I ARC-A0VI A and B which are SJAE A and B suction valves. This procedure was revised to add direction to the operators to confinn instrument failure using another available flow instmment. The procedure then directs removal of a relay to disable the malfunctioning signal to the valve and allow the valve to re-open. A warning statement was added to alert the operator to the fact that the valve must be manually closed if an extreme low steam flow condition occurs or hydrogen concentration exceeds 2%. Additional steps are added to increase monitoring of SJAE steam supply flow and offgas system hydrogen concentration at one hour intervals to heighten the Section III Page 135 of 16
l l l operator awareness as well as prevent accumulation of flammable gas mixtures in offgas and possible backflow through the offgas system during low steam flow conditions. Summary of Safety Evaluation: he additional actions in the procedure (i.e., confirmed instrument failure, increased monitoring of critical parameters, and warning to operators identifying automatic functions disabled and required contingency actions) provide adequate administrative controls to allow disabling the automatic function in the case of a failed flow instrument. The equipment and piping can sustain a hydrogen detonation. He defeating cf the automatic closure feature will not cause a SJAE failure. Acrefore, this modification ('oes r ot increase the probability of an accident or a malfunction of equipment important to safety proiously evaluated in the SAR. The consequences of a hydrogen detonation have been evaluated. Disabling the low SJAE inlet steam flow closure of the air ejector suction valves and associated alarms will not change the previously analyzed consequences. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the S AR. The equipment afTected by this change (SJ AE suction valves and associated alarms) is not classified as safety related or important to safety. The procedure change reduces the possibility of a plant transient imposed by a single instrument failure and provides time for the operators to place the altemate SJAE in service or reduce power in a controlled manner to remove the SJAE from service. In addition, hasing the SJAE available increases the availability of the main condenser as a heat sink to support other accident scenarios. Herefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety difTerent from any previously evaluated in the SAR. There are no technical specifications associated with the condenser air removal system. The specific technical specification for the offgas system is focused on the hydrogen analyzers and the radiation monitors. The hydrogen analyzers alann at 2% hydrogen concentration. This alann will alert operators to monitor second stage SJAE dilution flow and manually shut the SJAE suction valve if steam dilution flow is below the required value. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: Emergency implementing Procedure (EIPF2-001 Description and Basis for Chance: Emergency implementing Procedure EIP-2-001 entitled " Classification of Emergencies" was revised to improve the emergency classification scheme in accordance with NUMARC/NESP-007 guidance, to incorporate recommendations from a branch technical position paper dated July,11, 1994, and to make the procedure casier for the end user to implement. Summary of Safetv Evaluation: Nothing in this procedure or the revision to this procedure has any impact on plant operations during normal or ofTnormal events. Thcre is no effect either direct or indirect on the functionality of systems, structures or components (SSC) important to safety. The procedure and its changes do not contain any wording which could change, degrade, or prevent action described or assumed in an accident discussed in the SAR. Therefore, this change does not increase the probability of an accident or a malfunction of any SSC important to safety previously evaluated in the SAR. Both Section111 Page 136 of187
i this procedure and the changes to it contain no wording or directions which could alter any assumptions presiously made in evaluating the radiological consequences of an accident or affect any fission product barriers. While this procedure has no direct or indirect role in mitigating the radiological consequences of an accident, it does initiate actions which are intended to reduce or limit the radiological consequences to the general public. Therefore, this change does not increase the consequences of an accident or any SSC important to safety previously evaluated in the SAR. The procedure and its revision do not generate a new accident, do create a change to SSCs which adds a new failure mechanism, alter the probability of an accident or malfunction presiously , considered to be of such low probability that it was not included in the SAR, or direct activities which could result in an accident which is different from any previously evaluated in the SAR. No : activities are directed or implied in this procedure which could damage or alter any margin of safety, either implicitly or explicitly, as defined in the licensing basis documents. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiating Document: EIP-2-009 - Medical Emergencies Descrintion and Hasis for Change: EIP-2-009 was revised to make it a stand alone procedure independent of administrative procedure ADM-0060,"First Responder Emergency." This revision also incorporates title changes. This procedure provides guidance and direction in medical emergencies for the handling ofinjured or ill personnel who are contaminated or potentially contaminated and require transport to an offsite medical facility. Summary of Safety Evaluation: This procedure change does not impact the administrative controls required to ensure the perfonnance of physical barriers during anticipated operational occurrences or postulated accidents. There is no reference to any plant structure, system, or components (SSC). There is l neither a direct nor an indirect impact on the functionality of any SSC. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR. This procedure does not address interfacing with plant operations during nonnal or abnormal events. This procedure contains neither any reference to any SSC nor the functioning or operation of any plant equipment. Therefore, this change does not increase the consequences of any accident previously evaluated in the SAR. This procedure change does not create a change to any SSC which adds a new single failure mechanism. Therefore, this change does not create the possibility for an accident which is difTerent than any previously evaluated in the SAR. The are no references to any safety related SSC or to any administrative controls over the operation functioning of any safety related equipment. There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system performance is degraded below design basis. Therefore, this change does not increase the probability of a l malfunction of a safety related SSC previously evaluated in the SAR. This revised procedure does not increase the radiation doses to the public above the licensing limit. It does not change any i barrier performance such that there are increased radiation doses to the public. Therefore, this I change does not increase the consequences of a malfunction of a safety related SSC previously evaluated in the SAR. This procedure was found to contain no wording or create any actisities which could produce any failure modes not already analyzed. Therefore, this change does not , create the possibility of a malfunction of safety related SSC different than any previously evaluated l l Section 111 Page 137 of 187
in the SAR. This procedure'does not reference physical parameters any related to safety SSC not the administrative controls required for barrier performance as found in the Technical Specifications. Herefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: EIP-2-011 - Fire Emergencies Degription and Basis for Channe: EIP-2-01I was revised to make it a stand alone procedure indenradant of fire prdxtion procedure FPP-0010, " Fire Fighting Procedure." Tlus revision also incorporates title changes. This procedure provides guidance and direction in combating a fire in the owner controlled area. Summary of Safety Evaluation: This procedure change does not impact the administrative controls required to ensure the performance of physical barriers during anticipated operational occurrences or postulated accidents. There are no references to any plant structure, system, or component (SSC). Here is neither a direct nor an indirect impact on the functionality of any SSC. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR. His procedure does not address interfacing with plant operations during normal or abnormal events. His procedure contains neither any reference to any SSC nor the functioning or operation of any plant equipment. Therefore, this change does not increase the consequences of any accident presiously evaluated in the SAR. This procedure change does not create a change to any SSC which adds a new single failure mechanism. Therefore, this change does not create the possibility for an accident which is different than any previously evaluated in the SAR. The are no references to any safety related SSC or to any administrative controls over the operation functioning of any safety related equipment. There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system performance is degraded below design basis. %crefore, this change does not increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. This resised procedure does not increase the radiation doses to the public above the licensing limit. It does not change any barrier performance such that there are increased radiation doses to the public. Therefore, this change does not increase the consequences of a malfunction of a safety related SSC presiously evaluated in the S AR. This procedure was found to contain no wording or create any actisities which could produce any failure modes not already analyzed. Therefore, this change does not create the possibility of a malfunction of safety related SSC different than any previously evaluated in the SAR. His procedure does not reference physical parameters related to safety SSC not the administrative controls required for barrier performance as found in the Technical Specifications. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Section 111 Page 138 of187
Safety Evaluation Initiatine Document: EIP-2-023 - Joint Information Center Description and Basis for Change: This procedure provides guidance and direction in activation of the joint information center (JIC). It also provides direction for the operation of the facility following the declaration of an emergency. He JIC's sole function is to provide accurate and timely information concerning an event at RBS to the news media and the general public. This procedure was revised to reficct the changes of River Bend Station being brought into the Entergy system. Old terminology and the hierarchy of the JIC of were changed to bring the procedure up to date. Duties and responsibilities of JIC members were duplicated in the procedures for each of those individual positions. This information was removed and replaced by a reference to the individual position procedures. There were no specific procedures for the security oflicer of the JIC, so those duties and responsibilities were added to this procedure. Summary of Safety Evaluation: This procedure change does not impact the administrative controls required to ensure the perfonnance of physical barriers during anticipated operational occurrences or postulated accidents. There are no references to any plant structure, system, or component (SSC). There is neither a direct nor an indirect impact on the functionality of any SSC. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR. This procedure does not address interfacing with plant operations during normal or abnonnal events. This procedure contains neither any reference to any SSC nor the functioning or operation of any plant equipment is contained in this procedure. Therefore, this change does not increase the consequences of any accident previously evaluated in the SAR. This procedure change does not create a change to any SSC which adds a new single failure mechanism. Therefore, this change does not create the possibility for an accident which is different than any previously evaluated in the SAR. The are no references to any safety related SSC or to any administrative controls over the operation functioning of any safety related equipment. There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system performance is degraded below design basis. Therefore, this change does not increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. This revised procedure does not increase the radiation doses to the public above the licensing limit. It does not change any barrier performance such that there are increased radiation doses to the public. Herefore, this change does not increase the consequences of a malfunction of a safety related SSC previously evaluated in the SAR. This procedure was found to contain no wording or create any activities which could produce any failure modes not already analyzed. Therefore, this change does not create the possibility of a malfunction of safety related SSC different than any previously evaluated in the SAR. This procedure does not reference physical parameters related to safety SSC not the administrative controls required for barrier performance as found in the Technical Specifications. Herefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Section 111 Page 139 of187
Safety Evaluation Initiatine Document: EIP-2-028 - Recovery Description and Basis for Change: This procedure provides guidance and direction in establishing a recovery organization following a serious accident to assist in the clean-up and recovery effort. It was revised to reflect the addition of River Bend Station to the Entergy system. This revision entailed title changes, terminology changes, and some methodology changes. Summary of Safety Evaluation: This procedure change does not impact the administrative controls required to ensure the performance of physical barriers during anticipated operational occurrences or postulated accidents. There are no references to any plant structure, system, or component (SSC). There is neither a direct nor an indirect impact on the functionality of any SSC. Therefore, this change does not increase the probability of an accident previously evaluated in the SAR. This procedure does not address interfacing with plant operations during normal or abnormal events. This procedure contains neither any reference to any SSC nor the functioning or operation of any plant equipment. Therefore, this change does not increase the consequences of any accident previously evaluated in the SAR. This procedure change does not create a change to any SSC which adds a new single failure mechanism. Therefore, this change does not create the possibility for an accident which is different than any previously evaluated in the SAR. The are no references to any safety related SSC or to any administrative controls over the operation functioning of any safety related equipment. There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system perfornunce is degraded below design basis. Therefore, this change does not increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. This revised procedure does not increase the radiation doses to the public above the licensing limit. It does not change any barrier performance such that there are increased radiation doses to the public. Therefore, this change does not increase the consequences of a malfunction of a safety related SSC presiously evaluated in the SAR. His procedure was found to contain no wording or create any activities which could produce any failure modes not already analyzed. Therefore, this change does not create the possibility of a malfunction of safety related SSC different than any previously evaluated ; in the SAR. This procedure does not reference physical parameters related to safety SSC not the administrative controls required for barrier performance as found in the Technical Specifications. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l Safety Evaluation Initiatine Document: EIP-2-100 - Procedure Review, Revision, and Approval Description and Basis for Chance: I This procedure provides guidance and direction in the review, revision, and approval of procedures used by the Emergency Planning group. This revision entailed some title changes and a prosision i requiring the Facility Review Committee (FRC) to review these procedures annually. Section 111 Page 140 of 187
. = -
I i 1 1 Summary of Safety Evaluation: 1 1 This procedure change does not impact the administrative controls required to ensure the .l l performance of physical barriers during anticipated operational occurrences or postulated i accidents. There are no references to any plant structure, system, or component (SSC). There is ! i neither a direct nor an indirect impact on the functionality of any SSC. Derefore, this change does !
- j. not increase the probability of an accident previously evaluated in the SAR. This procedure does l not address interfacing with plant operations during normal or abnormal events. This procedure l l contains neither any reference to any SSC nor the functioning or operation of any plant equipment. :
Acrefore, this change does not increase the consequences of any accident previously evaluated in l the SAR. This procedure change does not create a change to any SSC which adds a new single ; failure mechanism. Therefore, this change does not create the possibility for an accident which is ' different than any previously evaluated in the SAR. The are no references to any safety related SSC or to any administrative controls over the operation functioning of any safety related ! 4 equipment. There are no direct or indirect references in this procedure which could degrade the l j performance of or increase challenges to safety systems such that the system perfonnance is l degraded below design basis. Herefore, this change does not increase the probability of a j malfunction of a safety related SSC previously evaluated in the SAR. This revised procedure does : not increase the radiation doses to the public above the licensing limit. It does not change any j barrier performance such that there are increased radiation doses to the public Therefore, this j l change does not increase the consequences of a malfunction of a safety related SSC previously I 3 evaluated in the SAR. This procedure was found to contain no wording or create any activities ;
- which could produce any failure modes not already analyzed. Therefore, this change does not
create the possibility of a malfunction of safety related SSC different than any presiously evaluated l
- - in the SAR. This procedure does not reference physical parameters related to safety SSC not the ;
- administrative controls required for barrier performance as found in the Technical Specifications. j j Therefore, this change does not reduce the margin of safety as defined in the basis of any technical l j specification. For these reasons, this modification does not constitute an unreviewed safety l question.
l c i Safety Evaluation initiatine Document: EIP-2-101 - Periodic Review of the Emergency Plan f i Description and Basis for Channe: $ This procedure provides the administrative functions necessary to review, revise, and approve the [ j Emergency Plan. It was revised to incorporate some title changes and make the resiew annual. ! ! Summary of Safety Evaluation: : 1 i This procedure change does not impact the administrative controls required to ensure the performance of physical barriers during anticipated operational occurrences or postulated ; accidents. There are no references to any plant structure, system, or component (SSC). There is !
- neither a direct nor an indirect impact on the functionality of any SSC. Therefore, this change does 4
not increase the probability of an accident previously evaluated in the SAR. This procedure does : not address interfacing with plana operation, during normal or abnormal events. This procedure ! contains neither any reference to any SSC nor the functioning or operation of any plant equipment.
' Herefore, this change does not increace the con:,equences of any accident previously evaluated in the SAR. This procedure change does rot create a change to any SSC which adds a new single Section 111 Page 141 of 187 f i
i 4
- ' failure mechanism. hrefore, this change does not create the possibility for an accident which is different than any previously evaluated in the SAR. The are no references to any safety related j 4
i SSC or to any administrative controls over the operation functioning of any safety related l , ,cquipment. Here are no direct or indirect references in this procedure which could degrade the l l ' performance of or increase challenges to safety systems such that the system performance is ' ; j degraded below design basis. Therefore, this change does not increase the probability of a
- malfunction of a safety related SSC previously evaluated in the SAR. This resised procedure does ,
not increase the radiation doses to the public above the licensing limit. It does not change any
, barrier performance such that there are increased radiation doses to the public. Therefore, this j change does not increase the consequences of a malfunction of a safety related SSC presiously i j' evaluated in the SAR. This procedure was found to contain no wording or create any activities j which could produce any failure modes not already analyzed. Therefore, this _ change does not t create the possibility of a malfunction of any safety related SSC different than any presiously ;
j evaluated in the SAR. This procedure does not reference physical parameters related to safety l ' SSCs not the administrative controls required for barrier performance as found in the Technical l Specifications. Herefore, this change does not reduce the margin of safety as defined in the basis i of any technical specification. For these reasons, this modification does not constitute an , 5 unreviewed safety question. l 1 ! Safety Evaluation initiatina Document: EIP-2-102 - Training, Drills, and Exercises ; r ! Description and Basis for Channe: l i EIP-2-102 provides the training requirements for personnel who perform emergency functions and [ the requirement for conducting drills and exercises. His revision changed the requalification l training for the Emergency Response Organization, changed the responsibilities of the Emergency i i Plan Trainer with respect to reporting to the Emergency Planning Department, and made title l changes. [ . - Summary of Safety Evaluation: l. . f . This procedure change does not impact the administrative controls required to ensure the l performance of physical barriers during anticipated operational occurrences or postulated ; accidents. There are no references to any plant structure, system, or component (SSC). There is , neither a direct nor an indirect impact on the functionality of any SSC. Herefore, this change does i j not increase the probability of an accident previously evaluated in the SAR. This procedure does . not address interfacing with plant operations during normal or abnormal events. This procedure l contains neither any reference to any SSC nor the functioning or operation of any plant equipment is contained in this procedure. Ecrefore, this change does not increase the consequences of any accident previously evaluated in the SAR. This procedure change does not create a change to any SSC which adds a new single failure mechanism. Therefore, this change does not create the ! f possibility for an accident which is different than any previously evaluated in the SAR. He are no references to any safety related SSC or to any admmistrative controls over the operation or j l functioning of any safety related equipment. There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system performance is degraded below design basis. Therefore, this change does not l increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. ! his revised procedure does not increase the radiation doses to the public above the licensing limit. l Secten Ill Page 142 of187 -. i
'r It does not change any barrier performance such that there are increased radiation doses to the j
- public. Herefore, this change does not increase the consequences of a malfunction of any safety j related SSC previously evaluated in the SAR. His procedure was found to contain no wording or l create any activities which could produce any failure modes not already analyzed. Therefore, this j i
change does not create the possibility of a malfunction of safety related SSC different than any i previously evaluated in the SAR. This procedure does not reference physical parameters related to j J safety SSC not the administrative controls required for barrier performance as found in the ; i' Technical Specifications. Therefore, this change does not reduce the margin of safety as dermed in l l the basis of any technical specification. For these reasons, this modification does not constitute an j unreviewed safety question. l c l Safety Evaluation Initiatine Document: ' EIP-2-105 - Control of Radiological Monitoring 11 Vehicles l Description and Basis for Channe: ! ! This procedure was revised to bring it up to date. Here were some title changes. The most
- i. important vehicle controlled by this procedure was renamed The descriptions of the radios were updated to match the new radios.
l i Summary of Safety Evaluation: l i . l His procedure change does not impact the administrative controls required to ensure the l performance of physical barriers during anticipated operational occurrences or postulated l accidents. There are no references to any plant structure, system, or component (SSC). There is neither a direct nor an indirect impact on the fMannlity of any SSC. Herefore, this change does not increase the probability of an accident previously evaluated in the SAR. His procedure does i i not address interfacing with plant operations during normal or abnormal events. This procedure contains neither any reference to any SSC nor the functioning or operation of any plant equipment.
- Therefore, this change does not increase the consequences of any accident previously evaluated in .
the SAR. This procedure change does not create a change to any SSC which adds a new single j - failure mechanism. Therefore, this change does not create the possibility for an accident which is l i different than any previously evaluated in the SAR. The are no references to safety related SSC or to any administrative controls over the operation functioning of any safety related equipment. l There are no direct or indirect references in this procedure which could degrade the performance of or increase challenges to safety systems such that the system performance is degraded below design basis. Therefore, this change does not increase the probability of a malfunction of a safety related SSC previously evaluated in the SAR. This revised procedure does not increase the radiation doses to the public above the licensing limit. It does not change any barrier performance such that there are increased radiation doses to the public. Therefore, this change does not increase the consequences of a malfunction of a safety related SSC previously evaluated in the SAR. His procedure was found to contain no wording or create any activities which could produce any failure modes not already analyzed. Herefore, this change does not create the possibility of a l' malfunction of any safety related SSC different than any previously evaluated in the SAR. His procedure does not reference physical parameters related to safety SSC not the administrative controls required for barrier performance as found in the Technical Specifications. Therefore, this change does not reduce the margin of safety as defmed in the basis of any technical specification. O For these reasons, this modification does not constitute an unreviewed safety question. i . I ' - Section 111' Page 143 of 187 - i
Channe Number /USAR Section: PEP-0042, Revision 7 (SEN 96-0017) Description and Basis for Channe: The reactor pressure vessel inservice leakage test is performed every refueling outage and only in
]
] operating mode 4. Procedure PEP-0042, "RPV Inservice Leakage Test," provides the l
; administrative guidance for this test. Previously, the procedure involved pressurizing the reactor l j vessel to 1025 psig out to the inboard containment isolation valves. The procedure was revised to l l extend the test boundary out to the boundary between ASME Class 1 and Class 2 components, L typically located at the outboard containment isolation valve. This revision allowed the test to be ;
used to satisfy the ASME Section XI ten year test. ! i ! Summary of Safety Evaluation: l 1 !
- Primary containment is required in mode 4 if there is a potential for draining the reactor vessel !
(OPDRV). The impacted valves were reviewed to verify that they could tolerate the test pressures. ! ! The revision does not affect the functional or operational aspects of the valves. He valves still l 2 meet their design requirements. Thejumpers andjumper locations were reviewed to determme if ; an OPDRV existed. Eachjumper connection is at a vent, drain, or test connection prosided with a root valve. The root valve allowed for isolation of the hose thereby isolating any leak. Based on the determination that an unisolable leak cannot be created, it is concluded that an OPDRV l condition is not be created. Therefore, this test does not increase the probability of an accident or a ; i malfunction of equipment important to safety previously evaluated in the SAR. He vessel head } must be on at the time of the test and all refueling activities in containment must be complete i before the test begins. The test is performed when the vessel is nearly water solid, and decay heat ! i values and reactor stored energy are very low. Therefore, this test does not increase the !
- consequences of an accident or a malfunction of equipment important to safety previously i evaluated in the SAR. Types of failures for existing equipment that could be affected by this ,
revision are pipe leaksforeaks and mechanical failures (e.g., check valves, snubbers). Jumper lines l 4 are rated for well in excess of the test pressure. Use ofjumpers introduces the possibility of a i
- jumper leak or bred which is not a new type of failure. Jumper use/ failure will not render !
- inoperable the < heck valus, containment isolation valves or the emergency core cooling systems.
Cross connection does not render either system unable to perform its design function. Therefore, l l this test does not create tbo possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This procedure change does not impact , , the operability, functional or performance requirements of the standby liquid control system, high ! ] pressure core spray system or low pressure core spray system. The test revision does not create an ; [ OPDRV and the test is not conducted while refueling activities are in progress. Therefore, this test l
- does not reduce the margin of safety as defined in the basis of any technical specification. For i these reasons, this modification does not constitute an unreviewed safety question. i k l t
d Safety Evaluation initiatine Document: PEPS 129,130,159,160, and 171 a Description and Basis for Channe: 2 i 2 . The purpose of this change is to temporarily allow another source of cooling water to be used for i the control rod drive (CRD) pump lube oil coolers during the motor operated valve (MOV) j signature testing of reactor plant component cooling (CCP) loop B inlet isolation MOV, the two l Section til Page 144 of187 l l
i CCP loop B outlet isolation MOVs, and the CCP loop B service water supply and return valves. ; The preferred temporary source of cooling water is demineralized water from the make-up water system.110 wever, water from the condensate makeup storage and transfer system (CNS) can be used as an altemate temporary source of cooling water in case a problem is encountered using the MWS supply. After the water is used, it is be directed to the equipment and floor drain system. i Flexible red rubber hose and associated fittings are used as temporary piping to connect MWS or CNS water to the CRD pump lube oil coolers. Summary of Safety Evaluation: ; All portions of the systems that are affected by this temporary line-up are classified nonsafety i related and do not initiate any of the accidents evaluated in the SAR. Loss of CRD pumps will not j increase the probability of occurrence of a malftmetion but rather necessitate the initiation of a l scram. The impact of the loss of both CRD pumps has been considered for all transients included I in the accident analysis. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety. All failures that can occur as a result of this temporary change are bounded by failures previously evaluated in the SAR. Any failure of the rubber hose that could cause flooding or spraying in the CRD pump room is bounded i by the failure of the existing nonsafety related CCP, MWS, and CNS piping. Overpressurization 4 protection of the lower pressure CCP piping from the higher pressure MWS piping is accomplished by the existing relief valves which are set at the proper pressure and within the CCP piping boundary supplied by this temporary change. Herefore, this change does not create the i possibility of an accident or a malfunction of equipment important to safety different from any ) previously evaluated in the SAR. The systems and components affected by this temporary change are non-safety related and are not defined in the basis for any technical specification. Therefore, this change does not reduce the margin of safety as defined by the basis for any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Safety Evaluation Initiatine Document: Reactor Engineering Procedure (REP) 0030 Description and Basis for Change: This revision represents a change to the plant computer that permits adjustment of the feedwater flow measurement used by the plant computer in the calculation of core thermal power. The adjustment is made by manually enterir.g a correction factor that accounts for biases in the feedwater flow measurement instrument loops. These biases include those present due to fouling of feedwater venturi nozzles as well as those introduced during routine instrument loop calibration or resulting from instrument loop drift The correction factor will be determined by comparing the feedwater flow rate as indicated by the plant computer to the feedwater flow rate as indicated by a j previously installed Caldon Inc. model 8300 leading edge flow meter (LEFM). The feedwater flow j control signal used to control reactor water level is not being modified by this change since th. l existing controls automatically compensate for any bias present in this signal.
)
Summary of Safety Evaluation: This revision only involves a change to a procedure to apply a manual correction factor to feedwater flow indication used by the computer for heat balance calculations. Core thermal power Section 111 Page 145 of187 1
is not a causal factor of any accident evaluated in the SAR. This change only increases the accuracy with which actual reactor power can be monitored. The determination of core thennal power through use of the LEFM provides greater accuracy and less uncertainty than the original Venturi instrumentation system. His change does not change the rated power level or allow plant operation in a region where it has not operated before. Therefore, this procedure revision does not increase the probability or consequences of an accident previously evaluated in the SAR This change does not alter the venturi nozzles, any coruponent included in the differential pressure measurement loops, or any component in the feedwater temperature loops. His change does not add or delete any plant hardware on safety related structures or components. Therefore, this procedure revision does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the SAR. Use of the LEFM to correct feedwater flow measurement used in the reactor thennal power calculation will not introduce any new or different failure modes to systems or equipment. Therefore, this procedure revision does not create an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. His procedure revision does not reduce the margin of safety as defmed in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: Temporary Procedure (TP) 94-0025 Description and Hasis for Change: His temporary procedure is written to address suspected packing leakage in the drywell. The leak is believed to be located on either the reactor water clean-up (RWCU) suction line throttle valve from the reactor bottom head drain, or the RWCU suction header from reactor recirculation loops l A and B isolation valve, or both of these valves. To attempt to stop the suspected packing leakage, it is desired to electrically back seat one or both of these valves from the motor control center. l I This temporary procedure will minimize the risk of any damage to the valves, actuators, or motors during the process. Both valves are normally open and have no active safety function. Power is supplied from a non-Class 1E source. Neither valve receives an automatic actuation signal. Since I both of the actuators have the capacity to produce torques and thrusts which could overstress both the valves and actuators under full voltage conditions, the input voltage to the valve motor l actuators will be lowered to prevent damage by a Variac device. At the specific voltages, neither ) the actuator thrust and torque limitations nor any valve ASME code allowables will be exceeded at motor stall output. Summary of Safety Evaluation: The only accident evaluated in the SAR the could potentially be impacted by this temporary l procedure is the large pipe break inside containment. Backscating these valves can only impact the l probability of this accident occurring if the backseating causes excessive stresses on the valve body ) resulting in a failure of the pressure boundary. The voltage to the motor will be reduced through I use of a Variac device which results in the motor stalling at a lower current and torque value than j it would under 100% voltage. This casures no portion of the valve or actuator will be overstressed. j Stress levels in the pressure retaining portions of the valves will remain below maximum allowable levels when stroked at reduced vohage levels. The probability of occurrence of the large pipe break inside containment is not changed provided the Variac device does not fail. Probability of the failure of the Variac device is approximately one tenth of one percent of the probability of Section III Page 146 of I87
_ _ _ - _ _ . . --___ ._ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _ ___m
; occurrence of a large pipe break. Addition of the frequency of failure of the Variac device over the l one minute period to the yearly frequency of occurrence of the large pipe break results in about a
- one tenth of one percent change in the frequency of occurrence of the large pipe break. His is a i negligible change in frequency which is below the level of accuracy of this type of analysis. Job !
briefings, training, and personnel awareness are critical items to ensure the calculated thrust limits j are maintained. Dese administrative controls provide a high level of assurance that the valves will l be protected. Therefore, this temporary procedure does not increase the probability of an accident
- previously evaluated in the SAR. The only equipment important to safety that could be impacted !
by performance of this procedure is the electrical penetration through which the power and control i j cable are routed. These electrical renetrations are provided with overcurrent protection to prevent damage to the penetrations in an overcurrent condition. He running current and stall current are !
- lower when stroking the valves at lower than normal voltages. Since voltages and currents will be ;
j at reduced levels during the backseating operation and since the electrical penetration overcurrent j i protection will remain operable, it is concluded that this evolution will not impact the electrical > penetration. Therefore, this temporary procedure does not increase the probability of a
- malfunction of equipment important to safety previously evaluated in the SAR. These RWCU j
- valves are not credited for mitigating the consequences of any accident or any malfunction of i equipment important to safety previously evaluated in the SAR. The SAR does evaluate the i
, consequences of steam jet impingement and pipe whip on various components in the area of the ; piping. Electrically backscating these valves does not impact this previous evaluation. There is no 4 change in piping supports or whip restraints, piping gecmetry or piping size and thus, there is no j impact on these previously evaluated consequences should a piping failure occur while the valves
- are backscated. Operation of these valves is not asumed to occur to replace failed or >
- malfunctioning equipment important to safety. Therefore, this temporary procedure does not ,
increase the consequences of an accident previously evaluated in the SAR. The valve and actuator l
- components in the load path due to backscating have been evaluated for structural integrity and l
, pressure retaining capability where appropriate. These evaluations demonstrate that no missiles j will be generated and no other systems in the arca will be affected. Electrically backscating these j valves with administrative controls does not create any new or different failure mode for the valves 1 i
or any other components near the valve. Therefore, this temporary procedure does not create the l possibility of an accident or a malfunction of equipment important to safety different from any . previously evaluated in the SAR. Backseating these valves neither affects water chemistry nor the [ limits on water chemistry imposed by the Technical Specifications. The margin of safety in the i i Technical Specifications governing reactor coolant leakage, structural integrity of ASME Class 1, 2, and 3 components, primary containment penetration conductor overcurrent protective devices, i and other overcurrent protective devices are not reduced by backseating of the subject valves. Unidentified leakage in the drywell is continuing to be tracked. Backseating these valves is
) '
expected to reduce packing leaks into the drywell and reduce the magnitude of unidentified leakage into the drywell. The administrative controls imposed by this temporary procedure ensure the l reactor coolant pressure boundary is not compromised. Derefore, this temporary procedure does ) 4 not reduce the margin of safety as defined in any tecimical specification or the basis of any 1 technical specification. For these reasons, this modification does not constitute an unresiewed safety question. ] [ 4
. Section Ill Page 147 of187 i
~_ _ _ _ _ __
i 1 4 v 1 Safety Evaluation Initiatine Document: TP 95-0009 , i Description and Basis for Channe: ! i j nis change is to install a temporary blowdown system for use during RF-6. This system will j employ two diesel-driven pumps and temporary piping to provide at least 2200 GPM of water from l the flume to the blowdown line so that liquid waste system (LWS) and waste water treatment l [ (WTW) discharges may continue to be performed when circulating water system (CWS) is ; secured. He diesel driven pumps will be located on the west side of the fiume, south of the sodium , i hypochlorite tank ICWS-TK-3. The pumps will take a suction in the fiume and discharge sia j j temporary piping to the blowdown line. The temporary procedure (TP) requires that the cooling water tower blowdown radiation monitor IRMS-RE-108 suction be temporarily relocated to assure that blowdown dilution is monitored for radioactivity. i 4 1 Summary of Safety Evaluation: I 1 . He proposed temporary modification is not safety related as well as the normal source of i blowdown (CWS system). CWS is not required to mitigate the consequences of any accident and does not affect any safety related structure, system or component. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The existing permanent plant CWS components which are the source of dilution water for the radwaste system are not safety related and do not interface with 1 any safety-related systems, components or structures. Therefore, this change does not create the ! possibility of an accident or a malfunction of equipment important to safety different from any ; previously evaluated in the SAR. This temporary procedure does not reduce the margin of safety
- l as defined in the basis of any technical specification. For these reasons, this modification does not 1 4 constitute an unreviewed safety question. l Safety Evaluation Initiatine Document
- TP 95-0010 i
Description and Basis for Channe: This temporary procedure allows for the use of alternate media (s) in the liquid radioactive waste i demineralizer vessels to improve liquid radioactive waste processing ability in order to obtain maximum recycling of water back to the condensate storage tank. Summary of Safety Evaluation:
- Le most limiting fault for the liquid radwaste system (LWS) is defined as an unexpected and uncontrolled release of radioactive liquid stored in all of the LWS tanks. Use of this temporary procedure will not result in this most limiting fault occurring. Therefore, this modification does not
- increase the probability of an accident previously evaluated in the SAR The use of alternative j media in LWS will not increase the consequences of an accident evaluated in the SAR. This is
- based on data provided in the accident analysis for the most limiting accident events associated with the LWS. If alternate media and/or current media were to leave the vessel, it would be contained within the radwaste building. Building engineering controls and operator actions would
! mitigate the consequences of a media spill in the radwaste building. Therefore, this modification does not increase the consequences of an accident previously es31uated in the SAR. The temporary Section ill Page 148 of187
----__w___- ----_e_ _ _ - _ _ m _ - _ * , , + - -w
_ - - -_ m.- - u. A e . --,.
procedure is only changing the media used within the LWS demineralizer vessels. No change to plant structures, systems or components results from the implementation of this temporary procedure. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. All of the equipment associated with the radioactive liquid waste management system, as described in the S AR, is evaluated and designed as equipment not necessary for safety. Herefore, this modification does not increase the probability or consequences of a malfunction of equipment important to safety previously evaluated in the SAR. This modification does not create a malfunction of equipment important to safety different from any previously evaluated in the SAR. This modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: Technical Section Procedure (TSP.) 0053 Description and Basis for Change: A temporary power supply is required for the Division I control rod and information system (RCIS) in order to maintain it functional while the Division I power supply is out of senice during outages. This power supply is non-safety related and therefore, constitutes a change to the SAR. He control rod indications are needed to support refueling activities. This procedure also temporarily removes the advanced core monitoring computer system (ACMCS) from senice. Summary of Safety Evaluation: The difTerence between the normal power source and the temporary source is the lack of emergency diesel generator backup. The fuel bridge is not backed by the diesel generators either. In case of a loss of offsite power, the refueling bridge equipment will fail safe. RCIS has two safety related functions (rod worth limiter and rod pattern controller) neither of which is required during shutdown and refueling. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR Neither of the two safety functions provided by RCIS are required during shutdown and refueling, and the ACMCS is not required to mitigate the consequences of an accident. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This modification will provide regulated power to the RCIS panel just as the original source did. This will insure that the RCIS equipment will not be damaged by power supply problems. Also, the refueling bridge will fail safe in case of loss of power. Therefore, this modification does not create the possibility of an ::ciim or a malfunction of equipment important to safety different from any previously evaluated in the SAR. RCIS has no safety function in mode 5, and ACMCS perfonns no required function during periods , of reactor shutdown. There are no explicit safety limits in the Technical Specifications on RCIS. ' Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. l Section III Page 149 of 187
SECTION IV MODIFICATIONS TO THE PLANT Section IV Page 150 of I87
Channe Number /USAR Section: EEAR 95 E0020 Description and Basis for Channe: 1 A presious modification provided keylock switches and indicating lights to assist operations in performing , the actions required by emergency operating procedures (EOPs). Prior to the installation of these switches and lights, the action associated with the EOPs were accomplished byjumpering and lifting leads within - control room cabinets. The switches disable the automatic actuation of the main steam isolation valve (MSIV) and main steam line (MSL) drain isolation, the RPS trip logic, the containment instrument air isolation, the drywell cooling isolation and the RPV low level 1 MSIV and MSL drain isolation. The unreviewed safety question determination (USQD) for the modification added a qualification in the form of a special requirement. The special requirement stated,"The switches being installed under MR 87-0713 shall be used for the performance of EOP 1 Enclosures 9,12,16,20 and 27 only. Any other use of these switches shall result in nullification of this USQD." Since these switches are used as bypasses in lieu ofjumpers, this evaluation was prepared to eliminate the special requirement. >
- Summary of Safety Evaluation
S AR Chapter 15A demonstrates that the RPS and containment and reactor vessel isolation control system , (CRVICS) are essential in the protection sequences. Technical Specification Section 3.3 prosides the minimum performance requirements for their instrumentation and trip logic. Whether a channel is bypassed with ke> lock switches orjumpers, the minimum performance requirements as specified in the Technical Specifications :nust still be met, or the appropriate action statement entered. Therefore, the use of these switches does not increase the probability of occurrence of an accident evaluated presiously in the SAR. Since the Technical Specification minimum levels of performance for RPS and CRVICS address the condition of bypassing or inoperability (regardless of the bypassing devices), any release of radioactive material to the emironment will be consistent with the assumptions used in the analyses for a DBA. Therefore, the use of keylock switches as bypassing devices does not represent an increase in the ; consequences of an accident evahiated previously in the SAR. The purpose of the switches is to bypass parameters in the protection system. The action is to aid in the implementation of responses to accident conditions. The use of these switches in place ofjumpers during plant operations regardless of the mode is governed by the Technical Specifications. The use of switches in lieu ofjumpers is bounded by the accidents in Chapter 15 of the SAR. Therefore, the use of switches does not create the possibility of an accident different from any previously evaluated in the SAR. The normal method of accomplishing bypassing of signals is thmugh the use ofjumpers. The installation ofjumpers versus the use of a keylock switch is far less desirable from a human factors standpoint due to the possibility of human error inside the cabinet. Therefore, the use of keylock switches does not increase the probability of a malfunction of equipment important to safety presiously evaluated in the S AR. Sections 7.2.2.2 and 7.3.2.1.2 of the SAR provide that a sensor or trip unit may be removed from senice under administrative control procedures. ,~ Since only one sensor or trip unit is removed from service at any given time, prmective action capability for RPS and CRVICS automatic initiation is maintained through the remaining redundant instrument channels. Therefore, the use of key lock switches does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. Since both the use of keylock switches and the use ofjumpers are both es aluated, the use of keylock switches in lieu ofjumpers does not create the possibility of a malfunction of equipment important to safety of a different type than any presiously evaluated in the SAR. The Technical Specifications proside the minimum performance requirements for the RPS and the CRVICS. The use of keylock switches in lieu ofjumpers does not change any limiting safety system settings or performance requirement and therefore does not reduce the margin of safety as defined in the basis for any technical specification. For these reasons, this modification does not ^ constitute an unreviewed safety question. Section IV Page 151 of187 t
i i i Channe Number /Initiatine Document: ER 96-0026 i Description and Basis for Evaluation: Operations requested the Design Engineering evaluate the acceptability of removing standby senice water (SSW) pump SWP-P2C from senice to perform STP-200403," Division III Remote Shutdown Panel , Test." ( Summary of Safety Evaluation: ! Since the reactor would be shut-down for 18 days prior to perfonning STP-200403 there would be a minimum heat load on plant components. Calculations concluded that the RHR heat exchanger would be ! able to remove the decay heat load at evahiated in the USAR with the equipment lineup required for STP- ; 200 4 03. Therefore, this condition does not increase the probability of an accident or a malfunction of { equipment important to safety presiously evaluated in the SAR. The calculation showed that one SWP l pump could provided the required flows. Also, the proposed change would have no affect on radiological source terms or other assumptions made in the USAR accident analysis. Therefore, this condition does , not increase the consequences of an accident or a malfunction of equipment important to safety presiously j evaluated in the SAR. The equipment lineup required for one SWP pump to provided design flow would ! not degrade the performance of the RIIR heat exchanger and other safety-related components, nor does it i affect the independence or redundancy of safety related equipment. Therefore, this condition does not { create the possibility of an accident or a malfunction of equipment important to safety different from any presiously evaluated in the SAR. Technical Specifications do not provide any requirements to how SSW must be operated during Modes 4 and 5. The Technical Requirements Manual indicates that in Modes 4 , and 5, the requirements of SSW are determined by the systems that they support. Therefore, this condition l does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this condition does not constitute an unresiewed safety question. t
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Safety Evaluation Initiatine Document: Minor Modification (MM) 93-0007 Description of Modification: " This modification rearranges the exit area of the primary access point (PAP) building, so that the
- current security officer post is no longer required at the key card surrender window. His change maintains the integrity of the protected area while ensuring that there is no unauthorized use of an unsurrendered keycard in the absence of the security officer. In order to achieve the presious objectives, the remodeling and relocation of monitors and full length turnstiles are required. Also the necessary software changes to the CAS and SAS security computers will be developed to include the changes made in the PAP remodeling.
Summary of Safety Evaluation: a This modification is an enhancement of the PAP building exit area and does reduce the present need for the security officer station at the badge drop off window. This change does not impact the perimeter alarms, cameras, lights, or any other security equipment that could reduce or lower the j effectiveness of the plant security system. The physical security plan and licensee commitment 05935 address the temporary degraded barriers that will exist during the implementation of this modification. Therefore, this is not a violation or degradation of security. Therefore, this condition did not constitute an unreviewed safety question. ; Section IV Page 152 of I87 l
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Safety Evaluation Initiatina Document: MM 95-0511 (SEN 95-0115) f !
;I
' Description and Basis for Channe: !
Recorder CMS-TR103, located on remote shutdown panel RSS-PNL 101, is a Texas Instruments Model TIGRAPH 100 recorder. It will be replaced with a Westronics 2400 Series recorder. Performance specifications for the new recorder are equal to or bet.cr than those of the previous recorder. Since this modification represents a digital upgrade of a safety-related instrument, the guidance of NRC Generic Letter 95-02 and Electric Power Research Institute Report TR-102348 were considered. There is no change to the SAR. - Sununary of Safety Evaluation: The replacement recorder was tested in accordance with Electric Power Research Institute Report TR-102323. In addition, the environment for panel RSS-PNL101 was not considered an electromagnetic interference (EMI) threat to the recorder based on several EMI limiting practices currently employed (e.g., radio exclusion zones, arc welding restrictions, grounding requirements, etc.). Finally, the remote shutdown panel is not postulated to cause any of the accidents identified in the S AR. Therefore, this modification does not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The replacement of the recorder does not affect any equipment used to mitigate the consequences of an accident. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previoush evaluated in the SAR. No new failure modes are introduced by this modification. There are no changes in the number or type of parameters monitored by the recorder. Also, the redundant panel (RSS-PNL102) contains instruments (indication only) which can be used in the event of failure of the recorder (CMS-TR103). These redundant instruments are a different design and, therefore, are not susceptible to software common mode failures or cycling in a continuous loop. Therefore, this modification does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The capability to shutdown and maintain hot shutdown from outside the control room will not be affected. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
. Section IV Page 153 ofI87
Safety Evaluation laitiatina Document: Modification Request (MR) 91-0081 Description and Basis for Channe: A permanent diesel-driven air compressor, air dryer, after cooler and air receiving tank have been installed and connected to the instrument air system (IAS) and service air system (SAS) headers in the turbine building. These components were originally temporarily installed to provide back-up air supply whenever the IAS and SAS air compressors were not available.' Permanent installation includes a battery charger to keep the engine battery charged at all times for startmg the engine. Summary of Safety Evaluation: he existing air compressors, air dryers and filters are designated as other equipment in the SAR
, and are shown as non-safety related and non scismic. Here are no evaluated accidents in the SAR
- which are caused by the failure of the IAS system or equipment. The diesel driven air compressor
; provides an alternate source ofinstrument air to the plant systems during the unavailability of
] clectric driven air compressors or during the loss of offsite power. Since there are no evaluated j accidents caused by failure of the IAS or SAS systems, the portions of those systems that perform safety related functions are not impacted and there are no adverse electrical impacts postulated. The proposed modification does not increase the probability of occurrence of an accident . previously evaluated in the SAR. De dicsci driven air compressor has no function to mitigate the
} consequences of an accident, the only IAS components relied upon for mitigating the consequences of an accident are safety related air accumulators, air bottles, isolation valves and piping located in the control, auxiliary and fuel buildings. Ec proposed modification has no impact on the operation of the safety related air accumulators or components. Therefore, there is no increase in the consequences of an accident evniuated previously in the SAR. The diesel air compressor is e connected to the IAS and SAS system headers through check valves. This arrangement prevents the depressurization of the systems in case of a line break in the proposed installation. Therefore,
! the proposed activity does not increase the possibility of an accident that is different than any previously evaluated in the SAR. The safety related components will continue to receive quality . air and therefore the change will not impact their function, reliability, operability and availability.
- So, the probability of a malfunction of a safety related SSC previously evaluated in the SAR will not be increased Ac modification is limited to the non-safety portion of the system and provides
- tack-up only. Therefore, the consequences of a malfunction of a safety related SSC presiously i evaluated in the SAR are not increased. Since this modification provides a permanent back-up to
- . the non-safety portion to the IAS and S AS, it does not create the possibility of a malfunction of a f safety related SSC different than that already stated in the SAR. There are no technical
- , specifications affected by this modification and therefore, the margin of safety to any technical E specification is not reduced. As such, this change does not constitute an unreviewed safety j' question.
i P i I i Section IV Page 154 of 187 -
i j Safety Evaluation Initiatine Document: MR 95-0013 ; 1 Descriotion and Basis for Channe: ! The shroud head / separator assembly is designed to be clamped to the top guide flange by 28 l shroud head studs (SHSs) and nuts. This modification is for removal of all GE-style shroud head ' l stud bolts and retainers, to install two additional Westinghouse SHS lock devices at selected j . locations and to move existing SHSs with Westinghouse locks to new locations to achieve a _ l symmetrical pattern. Specifically, this modification accomplished the following: 1) removal of the ; 9 remaining GE bolt /retamer assemblies,2) removal of the 14 SHSs that do not have l Westinghouse locks,3) relocated certain SHSs having Westinghouse locks to new locations, and 4) l installed two new SHS/ Westinghouse lock assemblies to complete a quadrant symmetric bolting j pattem. This modification eliminated the wear problem nasociated with the GE bolt / retainers for ; the SHS by eliminating the remaining bolt / retainers and atilizing the low profile Westinghouse j locks that are not subject to the flow induced wear problems. ;
-i Summary of Safety Evaluation:
l l The Westinghouse locking devices installed by this modification have been analyzed and shown to provide the same function as the GE locking devices they replace. He SHS assembly (including the locking device) is a passive, non-safety related component that is not factored in the initiation or mitigation of any analyzed accident described in the SAR. Calculations show that reducing the ; number of SliSs from 28 to 16 and positioning with a quadrant symmetric pattern exceeds , minimum pressure requirements. He fewer SHSs will also result in a small bypass flow that has been analyzed. The results show no concem for erosion. This modification has no impact to any l dose or radiological results. Therefore, this modification does not increase the probability or l consequences of an accident or a malfunction of equipment important to safety presiously ! evaluated in the SAR. The reactor has been previously operated with these devices installed and l have not introduced any increased potential for loose parts, it is designed to withstand loads under l all operating conditions as well as postulated transient and seismic events. Therefore, this change l did not create the possibility of an accident or a malfunction of equipment important to safety ; different fiom any previously evaluated in the SAR, or reduce the margin of safety as defined in the i basis of any technical specification. For these reasons, this modification does not constitute an ! unreviewed safety question. j i l t Section IV. Page 155 of 187 l
Safety Evaluation Initiatine Document: MR 95-0532 (SEN 95-0124) Description and Basis for Change: This modification replaced the obsolete Westronics T4E2 recorders on main control room panel H13-P870 inserts SSB and 56B (SWP-PR50A, SWP-PR50b, SWP-FR60A and SWP-FR608) with Westronics 2l00 series recorders. These recorders monitor various standby senice water system parameters; Performance specifications for the new recorder model are equal to or better than those of the previous recorder model. Since this modification represents a digital upgrade of a safety-related instrument, the guidance of NRC Generic Letter 95-02 and Electric Power Research Institute Report TR-102348 were considered. Although this modification does not involve an explicit change to the SAR, the process by which the recorder inputs are processed results in an implied change to the plant as described in the SAR. Summary of Safety Evaluation: The replacement recorders were tested in accordance with Electric Power Research Institute Report TR 102323. In addition, the emironment for main control room panel IH13-P870 was not considered an electrom:gnetic interference (EMI) threat to the recorder based on several EMI limiting practices currently employed (e.g., radio exclusion zones, arc welding restrictions, grounding requirements, etc.). The recorders were seismically mounted and electrical separation maintained. Therefore, this modification did not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR The replacement of the recorders did not affect any equipment used to mitigate the consequences of an accident. The operation of the standby senice water system was not impacted. Therefore, this modification did not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the S AR. No new system interfaces or interlocks nor any new failure modes were introduced by this modification. Therefore, this modification did not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The capability of the standby senice water cooling system to ; provide sufficient cooling capacity for the continued operation of safety-related equipment during nonna! and accident conditions was not affected by this modification. Therefore, this modification did not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification did not constitute an unreviewed safety question. l l 1 l l l Section IV Page 156 ofI87 l
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' Safety Evaluation Initiatina Document: Prompt Modification Request (PMR) 89-0006 j (CANCELED) .
i ,
}i Description and Basis for Channe:
l t j The modification provided another source of air to the instrument air system (IAS) and service air l (SAS) system while new permanent compressors were being installed. (His modification was ! j- canceled when the installation of the permanent air compressors was complete.) I Summary of Safety Evaluation: ! ) There are no evaluated accidents in the SAR which are caused by the failure of the IAS or its j equipment.- This modification is in the non-safety related portion of the IAS and is not connected to , 4 safety related power supplics. The quality ofinstrument air will meet applicable requirements for ! up to 18 hours following a loss of offsite power (LOP). Administrative controls were in place to insure the temporary diesel air compressor would not operate lo".2er than 18 hours without power
- to the air dryers. Eighteen hours was sufficient time to resume power supply to the air dryers Therefore, this change did not increase the probability of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The diesel driven air compressor ;
j did not function to mitigate the consequences of an accident. The safety related IAS components ! i are not affected by this modification. This modification has neither an impact on the safety i systems assumed to function in the accident analysis nor did it degrade their performance below the , j design basis. Therefore, this change did not increase the consequences of an accident or a ; { malfunction of equipment important to safety previously evaluated in the SAR. The diesel air i
- compressor is connected to the IAS and SAS headers through check valves. This arrangement prevents the depressurization of the IAS and SAS in case of a rubber hose break. Since there are
- no safety related components in the vicinity of the hose, and since check valves are provided to j prevent depressurization ofIAS should a break occur, there was no impact to the safety related .
- components should a hose break occur. This modification had no negative impact to safety related components served by the IAS. Therefore, this change did not create the possibility of an accident i
or a malfunction of equipment important to safety that is different than any previously evaluated in the SAR. There were no technical specifications affected by this modification and this design j change did not reduce the performance or capacity of the IAS. Therefore, this change did not , f reduce the margin of safety as defined by the basis of any technical specification. For these i ! reasons, this modification did not constitute an unreviewed safety question. 4 4 1 ^ i 4 l i l , I Section IV Page 157 of187 I
1 l Safety Evaluation Initiatine Document: PMR 94-0021 (CANCELED) ) Description and Basis for Channe: [ This modification installed a temporary drain line to provide a flow path to maintain water level in' I F an offgas sample chiller. The permanent drain path for the sample chiller passes insufficient flow, thus causing the chiller to fill with water, ne new temporary drain will allow drain flow from the chiller to go to the conden.er. The piping to the condenser is the normal flow path for the offgas ) l pre-treatment sample disposal. The design of the temporary drain line meets the requirement for , > cxplosion protection for offgas piping. ! l l s Summary of Safety Evaluation. , ! The offgas system was designated as non-nuclear safety related and is a non-seismic system. The I only conceivable event which could cause significant system' damage is a seismic event. The tubing for the temporary drain line was installed to the same design and installation requirements as the existing piping. The installation of a temporary drain path did not change the release path or increase the volume ofgaseous or liquid radwaste available for release. Postulated failure of the ; temporary drain line or the lines to which it is connected will not increase previously analyzed ; doses at the site boundary. Therefore, this modification did not increase the probability or the i j consequences of an accident previously evaluated in the SAR. The offgas system is not utilized for . ; the mitigation of any accidents nor did it interface with any systems used in the mitigation of i ! accidents. The tubing was routed on the floor and was a low operating pressure system, therefore, j there was no adverse impact on any other systems by the addition of this temporary drain. Therefore, this modification did not increase the probability or the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The tubing used for the
- temporary drain was designed and installed in accordance with the design and installation 4
requirements of the existing offgas piping and was sized to prevent rupture due to hydrogen j explosion. The temporary drain tubing was routed to an existing connection for offgas disposal to
- the condenser. The additional flow to the condenser was small and had no impact on any existing
- system. Ecrefore, this modification did not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. Sampling of j the pretreatment radiation and hydrogen levels was required by the technical specifications. This modification enhanced the operability of the sample by removing condensate from the flow. It did i not adversely impact the sample flow rate. Therefore, this modification did not reduce the margin of safety associated with any technical specification bases. For these reasons, this modification did not constitute an unreviewed safety question.
J i A 4 J W i 4 i c Section IV- Page 158 of187
SECTION V CONDITION REPORTS l I l l l l 1 i l 1 I Section V Page 159 ofI87
Safety Evaluating Initiation Document: Condition Report (CR) 86-0045A Description and Basis for Change: This documents that 1E22*MOVF015 (Suppression Pool Suction Valve) has three wires landed on one point on the limit switch rotor. This deviates from specification 248.000, which states that "Not more than two wires shall be connected to a single terminal." Fic!d inspection of IE22*MOVF015 revealed that there is good electrical contact between the three lugs and adequate thread engagement to secure the termination. Additional field inspection revealed that there is no electrical contact between the lugs of the subject terminal point and the adjacent terminal points. Therefore, this condition is being treated as a "Use-as-is, one time deviation." Summary of Safety Evaluation: Valve IE22*MOVF015 is located in the high pressure core spray (llPCS) system. HPCS is designed to mitigate accidents and is not analyzed for causing accidents. The function of the HPCS system is not altered by the as-built wiring of IE22*MOPF015 and is equivalent to the standard configuration. The deviation from the standard does not impact any other equipment. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The as-built wiring of 1E22*MOVF015 does not impact any other system or component and there is good electrical contact between the three lugs and thread engagement to secure the termination. The "As-built" wiring does not impact any margins of safety as defined in the Technical Specifications related to emergency core cooling systems (ECCS), including instrument trip setpoints for ECCS actuation, HPCS system capacity, suppression pool capacity, or containment /drywell isolation. Therefore, this change does not create the possibility of an accident or a malfunction of , equipment important to safety different from any previously evaluated in the SAR or reduce the margin of l safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. l Safety Evaluation Initiating Document: CR 92-0198 Description and Hasis for Chance: i in response to NRC Information Notice 92-20 " Inadequate Local Leak Rate Testing," additional local leak l rate tests were performed on containment penetration (expansion) bellows IKJB*Zl9 and Z20. The I additional testing confirmed the existence of cracking on the tangent ends of the bellows elements of both i penetrations. The original safety evaluation demonstrated that the requirements of 10CFR50, Appendix J would be met and that structural integrity of the containment boundary would be maintained through Cycle 5. Revision 2 of the safety evaluation demonstrated that structural integrity would be maintained through Cycle 6. This revision (Rev. 3) of the safety evaluation changes the CR disposition from ' Repair' ) to 'Usc-as-is' and demonstrates that structural integrity will be maintained through Cycle 7. The sum total of information collected in the last three refueling outages now provides a basis to assert that the , behavior of the bellows flaws is static, i.e., the cracks are not growing as initially contended. Existing ! analyses, based on assumed growth of the flaws over the next operating cycle, demonstrate no challenge to structural integrity or required leak tightness, and since the flaws have been determined to be static in behavior, it is acceptable to disposition this condition as 'Usc-as-is.' Bellows replacement is not required for these components at this time. Summary of Safety Evaluation: Even tlough flaws in the bellows elements were identified w hich represent deficiencies in component characteristic, the quality of the components was not rendered unacceptable nor indeterminate. The analyses include both the current and predicted conditions of the bellows' elements. The only previously : Section V Page 160 of187
i j evaluated accident in the SAR that could occur during the plant mode in question is a fuel handling i accident. Flaws in the bellows have no relation to the probability of this accident occurring. The effect of ! the cracks on the overall availability of open pathways for the release of radioactisity is clearly ; insignificant. The bellows are currently operable and capable of performing their function based on observation. The cracks are barley visible and do not adversely affect the structural integrity of the component at this time. The possibility of a new malfunction is, therefore, not created. Based on the facts , that end-of-cycle leakage through the bellows will remain within Technical Specifications limits and that + the components will remain structurally viable, the consequences of accidents presiously evaluated in the i SAR will not be increased. There is no condition of the bellows that would directly contribute to the , creation of a new accident; the safety function of the bellows is accomplished without any specific action i of the component, i.e., passively. Therefore, this CR disposition to use as is does not increase the i probability of an accident nor does it increase the consequences of an accident presiously evaluated in the - SAR. Therefore. this uodification does not create a malfunction of equipment important to safety , different from any previously evaluated in the SAR. New leakage pathways are not created; leakage of. ; failed bellows can only take place through the annulus. Leakage through the bellows has been assessed f and determined to be minuscule. The margins of safety are related to maintaining existing leakage ; i pathways and rates within the limits used in Station accident analyses and maintaining structural integrity of the bellows so that the components can withstand the maximum accident condition containment pressure. These bases are not compromised by start up and operation of the plant through the next fuel ! cycle. Therefore, the condition of the bellows does not create a malfunction of equipment to safety i different from any previously evaluated in the SAR. Therefore, this condition does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: CR 94-1298 l CR 94-1556 Description of Condition: Stainless steel piping located in the drywell was reported to be cracked. The root cause was established as trans-granular stress corrosion cracking caused by the presence of chlorides on the piping introduced by a spill ofinsulation adhesive. A thorough inspection was conducted by a significant event response team which identified all contaminated piping. All contaminated piping was evaluated and replaced. Summary of Safety Evaluation: All cracks were identified to be axial cracking that occurred as a single event. The cracks were identified on stainless steel piping or tubing with internal or emironmental temperatures in excess of 140'F or material stresses in excess of 10,000 psi. Based on these limits, systems with susceptibility were inspected for contamination by the insulation adhesive. There is high confidence that the extent of any misapplication or spillage of this adhesive has been identified and rectified. The SAR prmides an analysis which indicates that for pipe cracks in high pressure systems, through crack lengths of greater than four inches would be required to produce a leakage of five GPM. Since the adhesive does not cause crack propagation beyond its contact boundary and results only in small axial cracks, the potential presence of an unidentified location containing this material does not increase the probability of an accident previously analyzed in the SAR. Evidence that this problem is not a global issue and the probability of any susceptible piping being contaminated with the adhesive and not identified is considered to be unlikely. Therefore, the current situation does not increase the consequences of an accident presiously evaluated in the SAR. It is very unlikely that the adhesive has caused a crack of sufficient size to create a seriously degraded condition and remain undetected. Therefore, the current situation does not create the possibility of an accident not previously analyzed in the SAR. The potential possibility of an unidentified material being contaminated with this material does not increase the probability of a failure of a safety related system, structure, or component (SSC) previously analyzed in the SAR. The iruwhertent contamination of two redundant trains of susceptible equipment is not considered Section V Page 161 of187
l l l t likely due to separation criteria. Therefore, the potential possibility of a unidentified susceptible material being contaminated with this material does not increase the consequences of failure of a safety related SSC previously analy7;ed in the SAR. Since all susceptible piping has been identified and replaced, the ! possibility of creating a safety-related SSC malfunction not evaluated in the SAR is negligible; The low I probability for a crack of any magnitude makes the applicable technical specification still valid. The basis i for this technical specification is the Inservice Inspection program which is not impacted as a result of this l situation. Therefore, the margin of safety for any technical specification is not reduced. His change does l not constitute an unresiewed safety question. j 1 Safety Evaluation Initiatine Document: CR 95-0171 i I Description of Condition: -l l During the performance of the control rod operability surveillance, Control Rod 44-33 required a large i differential pressure increase for the insertion stroke. The withdrawal was difficult in that an automatic ; momentary insertion is required to lift the control rod off the collet fingers so that the control rod may ] withdraw. The momentary insert was difficult to achieve, but once motion began in the withdrawal , direction, movement time appeared to be as expected A full stroke accumulator scram will be performed i to ascertain the control rod scram times for Control Rod 44-33. If the full stroke scram does not clear the ! notch insertion problem with this control rod, compensatory measures must be implemented to ensure i control rod movement during the weekly notch insertion and withdrawal. First, the control rod will only J be used at notch insert and withdrawal speeds less than 3 inches per second. Second, an operator assisted l valve sequence procedure will be used instead of an automatic valve sequence for control rod withdrawal. ] These two proposed changes will only be implemented if the scram time of this control rod is acceptable. l Summary of Safety Evaluation: The accidents in the SAR are based on rapid continuous control rod withdrawal. Neither slow control rod l notch insertions nor withdrawals have any effect on any of the accidents analped in the SAR. The proposed method of operator assisted control rod notch insertion and withdrawal has the potential to cause , double nothing beyorel the intended position. However, this does not cause continuous movement of the control rod. Therefore, these changes do not increase the probability of an accident presiously evaluated in the SAR. None of the analyses of equipment malfunctions in the SAR are initiated by slow control rod l movement. Neither notch insertion nor notch withdrawal initiate a control rod drive failure. Therefore, . these changes do not increase the probability of a malfunction of any safety related structures, systems or [ components (SSC). The consequences of control rod accidents are based on the rate of energy deposition. ; The rate of energy deposition for the proposed changes is less than for control rods operated at normal l drive speeds. Adverse consequences of control rod withdrawal error above the low power setpoint are l prevented by the rod withdrawal limiter function of the rod pattern control system. The rod withdrawal i limiter confines movement of any control rod to either 2 or 4 notches depending on reactor power level. l Control rod insertion errors are not addressed in the SAR due to the benign effects of reactisity decrease '! on accident consequences. Therefore, these changes do not increase the consequences of an accident or a malfunction of any safety related system, structure, or component (SSC) previously evaluated in the SAR. Control rod notch snovements are for gross power adjustments. The rate of reactivity insertion (either j positive or negative) is less than for control rods operated at normal drive speeds. This reduces the rate of ! change of the local peaking, which is greatest at the tip of the control blade. This slower rate of change j results in less stress to the fuel. De rod withdrawal limiter limits the amount of positive reactisity ! inserted from a control rod withdrawal error. Control rod patterns are not sensitive to single notch l
~
deviations from the target pattern. These changes only affects the control rod movement method. For the ; control rod to remain in operation, it's scram time must remain acceptable. Therefore, this change does not create the possibility of an accident or a malfunction of any safety related SSC different from any ; previously evaluated in the SAR. These proposed changes do not affect the operability requirements found in any technical specification. These changes also do not affect the basis of any technical s Section V Page 162 of187
l I l specification. Therefore, these changes do not reduce the margin of safety defined in the basis of any , technical specification. For these reasons, this modification does not constitute an unreviewed safety l question. j j l Safety Evaluation Initiating Document: CR 95-0284 ' Description and Basis for Chanec: This change determines if the temperature control valves 1HVK*TV18A and B in the control building chilled water system can be gagged open during the interim until a pennanent modification can be implemented. These valves are currently gagged open and the interim disposition is " Repair." To support gagging open and eventual climination of these valves, Design Engineering provided an analysis widch evaluated the system affects if the outside ambient temperature is at the 25 F design limit and if the outside temperature drops below 25 F based on historical temperature data for the last 30 years. Elindnation of these valves become desirable due to (1) difficulty in accessing the valves w hich are located in the overhead with obstructions, (2) previous history of hydraulic leakage problems and (3) not essential for operation of the control building chilled water system. Summary of Safety Evaluation: Calculations evaluating the above control valves failing in the set /open position show that the emironmental design criterion room temperatures will be maintained. Conservative low outside air temperatmes were used based on local climatological data. Results show the control building air conditioning system was not impacted by this change. The air handling units are still capable of remming the heat loads under normal and accident conditions. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The design temperature limits as described in the SAR will still be maintained at outside ambient temperature design limit of 25 degrees F. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR or reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: CR 95-0304 Description of Condition: The second stage steam jet air ejector (SJAE) is designed to evacuate noncondensable gases from the j condenser air removal system intercooler, to provide dilution steam for the radiolytic hydrogen gas to ' maintain a concentration less that 4%, and to provide a motive force to drive these gases through the offgas system. This change will temporarily defeat the automatic closure on SJAE suction valve I ARC-AOVI A to trouble-shoot the intermittent second stage SJAE low steam flow and extreme low st:am flow alarms. The automatic closure feature is provided to prese ve condenser vacuum and to prevent further , buildup of detonable gases in the condenser air removal system following a loss of steam flow to the ; second stage SJAE. ; i Summary of Safety Evaluation: A scismic failure is the only conceivable event that could cause significant system damage. The equipment and piping are designed to contain any hydrogen-oxygen detonation which has a reasonable probability of occurring. Defeating the automatic closure of the valve may increase the possibility of a hydrogen detonation, but the equipment and piping are designed to sustain the detonation. Defeating the automatic closure feature does not increase the probability of a SJAE failure. If an actual low steam flow , i 1 1 Section V Page 163 of187
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I condition develops, the operator can still close the valve with the remote manual control. Therefore, this l change does not increase the probability of an occurrence of an accident presiously evaluated in the SAR. ; The vacuum breaker valve piping is of the same diameter as the condenser air removal piping and there is j
, a far longer run from the condenser to the main stack with numerous pressure restrictions. Consequently, ;
4 the loss of vacuum through the opened SJAE suction valve would be less severe that the analyzed accident ? of an opened vacuum breaker. Therefore, this change does not increase the consequences of an accident - previously evaluated in the SAR. The only two possible aa:idents which could occur from this activity ; have already been evaluated. Therefore, this change does not create the possibility of an accident of a : different type than any evaluated previously in the SAR. The condenser air removal and offgas systems ! are not safety related. The proposed change only affects these systems, Therefore, this change does not l 4 increase the probability or the consequences of a malfunction of any safety related structure, system, or ; component (SSC) previously evaluated in the SAR. This change does not create the possibility of a i malfunction of any SSC different from any presiously evaluated in the SAR. There are no technical j specifications associated with the condenser air removal system. The technical specification related to the j l offgas system are limited to the hydrogen analyzers and the radiation monitors. The hydrogen analyzers ; i alarm at 2% hydrogen concentration. This alarm will alert operators to monitor the second stage SJAE l dilution flow and manually shut the SJAE suction valve if steam dilution flow is below required value. : i This operator action replaces the automatic action being defeated and is allowed by the NRC Generic - Letter 91 18. Therefore, this change does not reduce the margin of safety as defined in the basis of any ; technical specification. For these reasons, this modification does not constitute an unrniewed safety ; i question. ; j i )
- Safety Evaluatine Initiation Document
- CR 95-0340 Description of Condition:
! A previous modification to the four control building chilled water (HVK) system chillers was designed to change their capacity from 165 tons to 189 tons and revise the inlet / outlet chilled water temperatures from l 60F/45F to 67.5F/52.5F. Inadequate post modification testing criteria and instructions were prosided in j the modification to verify the increased cooling capacity and control building chilled water system design
- conditions. The post modification testing performed to meet the requirements did not adequately j demonstrate the design. The interim recommended corrective action is to use the HVK chillers "As-is."
- Based on the review of the control building heat gain and airflow distribution and temperature verification
{l calculation, the required design cooling capacity is 171 tons. The shop testing of the HVK chillers show I an actual cooling capacity of 174 tons. The HVK system has been operating for several years and has
- consistently met the cooling load demand and maintained design conditions in the control building without any problems.
Fummary of Safety Evaluation: Each individual chiller is capable of handling design cooling capacity. No new single failures are introduced and the redundancy of the HVK system remains intact. Also, the condenser / evaporator tubes
- are inspected annually and cleaned as required to ensure that the design fouling is maintained. Using this F equipment as is will not affect the design criteria, and SAR accident analyses are not impacted.
, Therefore, this condition does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Indirect effects of the HVK system to initiate accidents by allowing increased temperatures to cause spurious equipment mMfunction are not , considered credible due to area temperature monitoring in spaces containing sensitive equipment and redundancy in these systems. These systems provide an emironmental support function and are not 4 considered potential accident initiatorx There is sufficient margin between the design basis cooling load and the factory tested capacity to preside reasonable assurance that the HVK chillers will perform their safety function. Therefore, this condition does not create the possibility of an accident or a malfunction of 1 equipment important to safety different from any previously evaluated in the SAR. There is sufficient t Section V Page 164 of 187
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P f margin between the design basis cooling load and the factory tested capacity to provide reasonable assurance that the llVK chillers will perform their safety function. Therefore, the condition does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: CR 95-0465 Description of Condition: During a resiew of the GE design specification for the control rod drive (CRD) system (GE document 22A5395, SDDF 0221.238-000-012P) it was discovered that the requirements of section 4.2.3.11.1 were not being met. Specifically, section 4.2.3.11.1 requires that scram discharge volume (SDV) vent and drain valves be adjusted such that the inboard veut and drain fully close 5.0 seconds before and start to open 5.0 seconds after the outboard valves. A review of data from previous full strode test per STP-052-6301 revealed that, while not specifically tested for this requirement, the valves do not appear to stroke within the requirements of this specification. Summary of Safety Evaluation: Failure of the SDV or the SDV vent and drain valves or lines is not an initiating event for any accident evaluated in the SAR. EOI calculation G13.18.10.2*129 revealed that all additional dynamic loads and stresses imposed on the vent and drain valves and lines are within the allowable loads and stresses of the ASME code. Therefore, there is no increase in the probability of the occurrence of an accident previously evaluated in the SAR. The components are directly related to the ability of the CRD hydraulic system to insert the control rods on a scram signal and isolating the SDV from the containment atmosphere to prevent reactor coolant from escaping into the containment following a scram. These functions are i required of the SDV, SDV vent and drain valves for all scrams regardless of the initiating event (e.g., a transient or an accident). The only requirement for operation of the vent and drain valves is that no single l active failure of the SDV vent and drain valves will result in an uncontrolled release of reactor coolant to the containment atmosphere. The result of calculation G13.18.10.2*129 states that the additional
- dynamic loads and stresses induced in the system as a result of not properly controlling the sequencing of the redundant valves are within ASME code allowable salues. Thus, they do not impact the ability of the SDV, SDV vent and drain valves to meet their design function. Therefore, there is no increase in the consequences of an accident evaluated previously in the SAR. The only consequence of failure to control the sequencing of the opening and closing of the SDV vent and drain valves is additional dynamic loads imposed on the piping, valves and supports associated with the SDV. Calculation G13.18.10.2*129 shows that the combined static and dynamic loads and stresses including those induced by a watr hammer are within the ASME code allowable values. Therefore, there is no creation of the possibility of an accident of a different type than any evaluated previously in the SAR. Since the SDV vent and drain valves and the piping are part of the reactor protection system (RPS) and take part in scrams, they are identified as equipment important to safety as defined in the SAR. The SAR requires the redundant SDV vent and drain vahes meet the S AR requirement that no single active failure of the SDV vent and drain valves result in an uncontrolled loss of reactor coolant. The addition of the 5.0 second delay was to reduce the effect of the water hammer event. Even without the delay, calculation G13.1810.2*129 shows that valves and piping can withstand the stresses induced from this lack of control. To ensure the SDV components work as they should, the instrumentation was installed in the right place. Prior to fuel load, the SDV instnunents were tapped into the 10" SDV line. This saved the instruments from being subjected to high flow rates and water hammer events which could potentially damage them. Since the SDV can withstand the pressures induced by scram and scram reset, and the instrumentation will function, thic condition does not increase the probability of a malfunction of equipment important to safety as presiously evaluated in the S AR. The SDV equipment associated with this condition is directly related to the ability -
to scram the plant and to stop the release of reactor coolant to the containment. However, calculation G13.18.10.2'129 shows that the condition does not impact the ability of the SDV and the SDV vent and drain lines and valves to perform their function. Additionally, the instrumentation is also not affected by Section V Page 165 of187
this condition due to its location. In the S AR, it is not given any credit in mitigation the consequences of any malfunction of equipment ; elated to safety. Therefore, this condition does not increase the consequence of a malfunction of equipment important to safety as previously evaluated in the SAR. The instruments are not connected to the vent and drain lines to shield them from these water loads. Therefore, this condition does not create the possibility of a malfunction of equipment important to safety of a difTerent type that any evaluated previously in the SAR. Technical specifications require the SDV instrumentation to trip the reactor when the volume had increased to a point to indicate that the volume is filling but before the teaming volume is less than required to accommodate the water from the movement of the control rods when they are tripped. Technical Specifications also require that the SDV vent and drain valm be operable so that it will be available, when needed, to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required. The SDV must be tested every 18 months to show the vent and drain valves will close within 30 seconds of receipt of a scram signal and open when the reactor is reset. Since the system meets these requirements, it does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this condition does is not an unreviewed safety question. Safety Evaluation Initiating Document: CR 95-0880 Descrintion of Condition: A small steam leak was discovered in the vicinity of the outboard reactor feedwater system (FWS) inlet header B motor operated isolation valve. The steam leak was found to be located at the penetration valve leakage control system (PVLCS) connection to the aforementioned valve. It was characterized as an open porejust upstream of the pipe side of the weld. A Furmanite enclosure is to be installed to control and maybe eliminate the leak. Summary of Safety Evaluation: The installation of a Furmanite enclosure, which is designed to system temperatures, pressures, and appropriate codes, will control the leakage rate and ensure that further degradation of the pressure ; boundary is not sustained. An evaluation of the socket wcld connection concludes that the weld has more j than 30% cxcess capacity with respect to allowable stress. In addition, with the structural integrity of the FWS is maintained within acceptable criteria and the leak under control by means of the Furmanite cnclosure, there is no increase in the probability or an accident initiating event. The worst case scenario, a main steam isolation valve closure u hich would initiate a reactor scrarn due to main steam tunnel , (MST) ambient temperature high or MST difTerential temperature high, is highly unlikely. Therefore, i this modification does not increase the probability of an accident previously evaluated in the SAR. A l worst case leakage through a sheared one inch schedule 80 PVLCS line during a design basis loss of l coolant accident will not exceed regulatory offsite dose limits. Assuming sudden failure or gross leakage through the subject line, the consequences associated with this event are bounded by failure of other lines in the MST Therefore, this modification does not increase the consequences of an accident presiously evaluated in the SAR. This modification does not have a negative impact on any plant system. Isolating the PVLCS supply line to the subject valve and maintaining PVLCS inoperable ensures that division of l the PVLCS system remains functional to the other containment penetrations served by this disision of I PVLCS by ensuring the PVLCS flow does not exceed the PVLCS setpoint should the line fail at the i connection to the subject valve. Based on the integrity of the Furmanite enclosure, no new accidents associated with missiles from pressurized equipment failure are credible. Therefore, this modification does not create the possibility of an accident different from any previously evahiated in the SAR. The line affected by this leak is not claaified as a high energy hne. Hence, the postulation for pipe whip is not required in this case. Flooding is not a concern due to the small amount ofleakage and the installation of the Furmanite enclosure to control or possibly eliminate the leak. No other safety systems are affected by this modification. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the S AR. There is an outboard reactor feedwater Section V Page 166 of I87
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- loop B containment isolation valve and an outboard reactor feedwater inlet header B automatically j operated check valve which provide containment atmosphere isolation outside of contaimnent for this ,
- process line. These valves are not affected by this change. Malfunction of the PVLCS isolation valve which supplies compressed air to the subject valve is not credible. Therefore, this modification does not .
4- increase the consequences of a malfunction of equipment important to safety previously evaluated in the .)
.. SAR. The only adverse effect of this condition is feedwater system inventory or PVLCS compressed air ;
j leakage in the MST. UW4ic leakage of feedwater inventory into the MST will be detected and j j appropriate plant response initiated. The small size of the weld defect, the _ integrity of thejoint and the . ] leak seal device will prmide assurance that large feedwater system inventory will not be lost. With the . . PVLCS isolated, the one inch line serves no safety function and at this point become similar to other class ] i four lines upstream of the subject valve. Therefore, this modification does not create the possibility of a j
- - malfunction of equipment important to safety different from any previously evaluated in the SAR. This ;
nodification does not affect the operability of the subject valve with respect to any required safety l [ j' function. In accordance with technical specification operability requirements, the PVLCS system was made inoperable by the isolation of the system from the subject valve and River Bend Station was placed ) in the associated 30-day action statement. Since the action statement was entered, the margin of safety of ! this particular technical specification is not reduced. Even though the subject valve is not subject any . I L structural integrity technical specification, an engineering evaluation found that the structural integrity of l - the ASME system is maintained. This modification does not affect the design basis, functions, or operation of the safety related equipment and does not adversely affect any other equipment important to ,
. safety. Therefore, this modification does not reduce the margin of safety as defined in any technical specification or the basis of any technical specification. For these reasons, this modification does not ;
} constitute an unreiewed safety question. ; } ! l Safety Evaluation Initiatine Document: CR 95-1094 j- , 4 Description and Basis for Channe: l ! I ! This CR provided for the installation of a temporary threaded cap to isolate line ISWP-750-562-4 until J p valve ISWP-V590 can be reinstalled. The line prmides normal senice water to the turbine building chillers. When a valve was removed to be relocated per a design change package, the isolation valves for
,- the line were found to be leaking. Since the V590 valve cannot be reinstalled until the isolation valves are q reworked, a temporary threaded cap was installed at the location of the valve. Once the isolation valves :
- are reworked, V590 will be reinstalled. !
4 ; Summary of Safety Evaluation: j ! Boundary integrity is still maintained with an acceptable method, but different than presently showri in ! the SAR. Since boundary integrity is not compromised, the probability of occurrence of an accident . l previously evaluated in the SAR is not increased The interim change will also not increase the l l consequence of any accident presiously evaluated in the SAR. The interim method of maintaining the
- boundary does not compromise integrity and does not create the possibility of an accident of a different ;
i type than previously evaluated in the SAR. This interim change has no interface with any equipment j important to safety and therefore, does not increase the probability of a malfunction occurrence or the ! . consequence of a malfunction of the equipment. Implementing this interim change will not impact the j i function of equipment important to safety and will therefore not create the possibility of a malfunction of a different type than any previously evaluated in the SAR. This interim change is on the normal senice c water system where it interfaces with turbine building chiller lHVN-CHIC. Operation of the turbine , i building chillers is not bound by any requirement in the technical specifications or the technical l requirements manual. The change does not impact the technical specifications. Because of the lack of l impact as noted above, no unreviewed safety question exists as a result of the temporary change. ] ! I l Section V Page 167 ofI87
Safety Evaluation Initiatine Document: CR 95-1221 Descriotion and Basis for Channe: In order to reduce a packing leak at valves B21-MOVF019 and B21-MOVF085, both of these valves as - l well as the associated inboard and outboard MSIV positive leakage control air supply header drain line , isolation valves and the main steam line warm-up header inboard containment isolation valve, will be ; closed. This action removes main steam-penetration valve leakage control system (MS-PLCS) from ! penetration KJBZ2 and also requires that B21-MOVF016 be shut to establish isolation of the penetration. The penetration is no longer supplied by MS-PLCS, and is considered not scaled by MS-PLCS. The
- combined leakage from the line is conservatively treated as part of the Type B & C leakage requirements.
The line will be returned to being scaled by MS-PLCS when the valves are repaired, prior to startup from refueling outage #6. The MS-PLCS is designed to control and minimize bypass leakage so that overall off-site dose rates do not exceed the guidelines of 10CFR100 following a design basis loss of coolant accident (LOCA). Once valve B21-MOVF016 is closed, leakage through the penetration is limited to leakage through this valve. Using the maximum path leakage rate obtained from the last LLRT shows that overall off-site leakage rates will not exceed the guidelines of 10CFR100 following a design basis LOCA, thereby allowing isolation of MS-PLCS to this penetratien. i Summary of Safety Evaluation: i Neither the structural integrity of the piping nor functional capability of the served valves are adversely impacted by the closure of the normally open inboard and outboard MSIV positive leakage contml air . supply header drain line isolation valves. Except for this change, the balance of MS-PLCS will remain l fully functional. These valves serve no support function and no system, structure or component (SSC) is ; adversely impacted by this change. The evaluated change will not degrade, change or prevent any action j or assumption described, made or assumed in the accident analyses evaluated in the SAR. The SAR does j describe the MS-PLCS in detail and takes credit for the MS-PLCS functions in off-site dose calculations ! for both the large and small break LOCA. A calculation was prepared which shows that the overall offsite ! dose rates will not exceed the guidelines of 10CFR100 following a design basis LOCA. No credible ! mechanism exists for which the closure of these valves could change the failure mechanisms that j contribute to increase the off-site dose of an accident prniously evaluated in the SAR. As the failure , mechanisms are unaffected, no consequence of an accident previously naluated in the SAR is impacted. > Therefore, this change does not increase the probability or consequences of an accident prmiously ! cvaluated in the SAR and it does not create the possibility of an accident of a different type than any ! evaluated previously in the SAR. The evaluated change will not cause the system to be operated outside of ! its design or testing limits. No system interface, other than that described in the change description is ! impacted. No credible direct or indirect effect exists which cou?d impact any SSC evaluated as being l j important to safety. No credible mechanism exists for which the closure of these valves could change the ; i failure mechanisms that contribute to the occurrence of the accidents evaluated prmiously in the SAR. As j the failure mechanisms are unaffected, no increase in the probability of occurrence of a malfunction of , equipment important to safety prniously evaluated in the SAR can exist. Therefore, this change does not ! i increase the probability of a malfunction or the consequences of a malfunction of equipment important to j safety prmiously evaluated in the SAR and it cannot create the possibility of a malfunction of equipment i ,.~ important to safety of a different type than any prmiously evaluated in the SAR. Although the Technical l 4 Specification Bases do not explicitly specify a margin of safety for the MS-PLCS system, the basis is i
)
clearly defined in the SAR. Maximum leakage rates from the last LLRT were used to determine that the overall off-site dose rates will not exceed the guidelines of 10CFR100 following a design basis LOCA. Therefore, this change does not reduce the margin of safety as defined in the basis of any tecimical specification. For these seasons, this modification does not constitute an unrniewed safety question. Section V Page 168 of187 1-
t Safety Evaluation Initiatine Document: CR 96-0127 (SEN 96-0016) Descrintion and Basis for Channe: . SAR Section 9.5.7.2 describes the Division 111 emergency diesel generator's lube oil systems. It explains f that the set point for the relief check valve in the circulating oil soakback pump loop (CSH-RV222B) is 30 l psig. Its primary function is to maintain a back pressure in the outlet piping from the AC and DC circulating oil pumps while the engine is in shutdown standby. This function is one of pressure j regulation. The other function of this valve is as a simple check valve to prevent reverse rotation of the circulating oil pump when the Division III diesel is operating. i Condition Report 96-0127 identified that the relief valve installed in the system actually relieves at 50 psig instead of 30 psig. Due to the unavailability of replacement parts, this evaluation addresses the , interim use, (use-as-is), of the 50 psig relief valve. The intention is to replace the relief valve during a j quarterly performance maintenance outage during fuel cycle 7; however, the evaluated "use-as-is" period l cxtends through cycle 7,just in case more time is needed l Summary of Safety Evaluation: i The relief check valve serves only the Division III emergency diesel generator and does not interface with any other plant structure, system or component (SSC). The potential effects of the higher than nominal { relief set point were evaluated and it was concluded that there is no adverse effect on the diesel. , Therefore, this modification does not increase the probability of an accident or a malfunction of l equipment important to safety previously evaluated in the SAR. For the proposed change to alter the 1 consequences of any analyzed accident or event, it would have to diminish the reliability of the diesel r generator to start and run, or change the way the diesel functions in support of other plant systems. Use of l the 50 psig nominal setpoint does neither. Therefore, this modification does not increase the i consequences of an accident or a malfunction of equipment important to safety previously evaluated in the , SAR. For the proposed change to create the possibility of an unanalyzed accident, it would have to I diminish the reliability of the diesel generator, or change the way the diesel functions in support of other l plant systems. Use of the 50 psig nominal setpoint does neither. Therefore, this mooilication does not j
. create the possibility of an accident or a malfunction of equipment important to safety different from any ,
previously evaluated in the SAR. The Technical Specification Bases discuss the diesel lube oil system m , terms of reserve volume and notes that standby conditions include lube oil being circulated at I temperatures consistent with manufacturer's recommendations. The capacity of the Disision III diesel's j , circulating oil system to function is not affected by this change. Therefore, this modification does not ' reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Safety Evaluation Initiatine Document: CR 96-0634 j i
- Description and Basis for Channe:
This change performs a partial weld repair to a cracked weld at the number eight cylinder exhaust pipe on the Division 11 diesel generator. The crack is at the base of the pipe where the pipe flange is bolted to the cylinder and extended through about 80% of the pipe circumference. The weld repair will fill all of the d
. crack that may be accessible. The weld repair is not intended to restore the nonconforming condition and will not restore the original weld such that it will meet all of the engineering function requirements.
4 Specifically, as the entire weld is not accessible, some of the exhaust leak will continue. Further, the root causes have not been fully evaluated and the action being taken is not intended to address the root cause. 4
' Section V Page 169 of 187 a- m %--1% d - -+ -we -- -. c- ,.,.- .--,w
Summary of Safety Evaluation: The crack does not affect the ability of the diesel generator to perform its specified safety function and its repair provides mitigating action to reduce the potential degradation and possible impact of the nonconforming condition only. The weld repair does not impact the design sources, operating characteristics, system functions, or system interrelationships. This change does not affect the manner which the station is operated and does not have any adverse impact on tie quantity or quality of any effluents. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. No automatic, protective, . or manual feature of any system, structure, or component is changed. As the senice water and diesel room environments can only be improved by this weld repair, no adverse system interaction is introduced by this change. No new failure modes are created by this weld repair. Therefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR or reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Safety Evaluation Initiatine Document: CR 96-0680 Description and Basis for Channe: This change will temporarily install 1%" schedule 40, A53 Grade B or Al20 hot dipped zinc coated pipe instead of 1 %" schedule 80, A106 Grade B pipe in the raw water supply to FPW-PIB fire pump diesel engine heat exchanger. The final CR disposition installs piping that meets vendor rcquirements and River Bend design requirements. Summary of Safety Evaluation: This change did not affect fire protection system performance or reliability or cause any fire protection component to operate outside ofits design or test limits and did not affect any system interface in a manner that could lead to the occurrence of an accident. No new fire protection system capacity nor redundancy was affected by this change. Therefore, this change did not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. The subject piping has an operating pressure of 57.7 psia @ 70*F and a design pressure of 100 psig @ 100"F ASTM A53,1981, Table X3 indicates a test pressure of 1,100 psi for Grade B Schedule 40 threaded pipe. He A53 pipe pro 5 ides substantial ; margin. Herefore, this change did not create the possibility of an accident or a malfunction of l equipment important to safety different from any previously evaluated in the SAR or reduce the . margin of safety as defined in the basis of any technical specification. For these reasons, this ! modification did not constitute an unreviewed safety question. l 1 l Section V Page 170 of187
SECTION VI TEMPORARY ALTERATIONS Section VI Page 171 of187
1 Safety Evaluation Initiatine Document: Temporary Alteration 95-0001 Description and Basis for Chanee: A temporary chemical feed system was installed to inject sodium hypochlorite (NaOCl) into the splitter box of the clarifiers. This change is needed because biological growth is occurring in the : clarifiers, adversely affecting their performance and then entering the service water cooling system l (SWC) and circulating water system (CWS). This material was found to be plugging the intake screen to a CWS water pump. Summary of Safety Evaluation: The proposed change would result in residual amounts of Na0Cl being admitted to those systems. Ilowever those systems are currently treated with NaOCI. Any residual biocide from the clarifiers would just reduce the Na0Cl demand in those systems. There are no adverse affects to CWS or SWC from excessive levels of biocide. Failure of these systems does not cause oc contribute to the occurrence of any evaluated accidents. Herefore, there is no increase of the probability of an accident previously evaluated in the SAR. None of the affected systems are relied upon to function during or following any previously evaluated accidents for the purpose of mitigating the consequences of an accident. Therefore, there is no increase in the consequences of an accident evaluated previously in the SAR. There are no new credible accidents created due to the addition of another NaOCl tank that could leak, because it has already been determined that any such leak, should it occur, would not endanger control room or plant personnel. There is no adverse impact to these systems since the only impact of the proposed change is to improve the water quality of all these systems. No adverse impact to the heat transfer capabilities to those systems is postulated since the water quality is no degraded. Therefore, there is no creation of the possibility of an accident which is different than any evaluated previously in the SAR. The CWS is the only system involved that could affect equipment important to safety. It has been evaluated for the impact on safety related equipment due to leakage from the CWS. There is no potential for flooding of safety related areas or components. Therefore, there is neither an increase in the probability of a malfunction or the consequences of a malfunction of equipment important to safety presiously evaluated in the SAR. Any malfunction of these systems would not be considered a malfunction of equipment important to safety. Since the only impact to those system that is postulated is a possible improvement in the water quality, there is not a failure mode created that would result in a new credible malfunction of safety related equipment. Therefore, there is no creation of the possibility of a malfunction of equipment important to safety different than any presiously evaluated in the SAR. For these reasons, this modification does not constitute an unresiewed safety question. Safety Evaluation Initiatine Document: Temporary Alteration 95-003 Description and Basis for Chanee: De low pressure alann setpoint for the high pressure core spray (HPCS) diesel generator air start system was increased from 160 psi to 220 psi. The original setpoint was intended to proside the plant operators with an alarm to warn them that the air pressure in the a ir start system has degraded to such a value that the ability of the diesel to start could bejeopardized. The new setpoint provides the alarm to warn that pressure in the air start system is degrading and the ability ; Section VI Page 172 of187 i
i I
-(
.i of the dicsci to crank five successive times without recharging the receivers may be impacted. The t revision to the setpoint provides the alarm much sooner than the original which will allow plant ;
operators to take corrective actions before the diesel generator is rendered inoperable on low air i start pressure. Summary of Safety Evaluation: ] The original setpoint,160 psi, warned the operators that a low pressure condition exists and the ability of the HPCS diesel generator to stan could bejeopardized. The new setpoint,220 psi, i prosides a warning to the plant operators that low pressure condition exists that couldjeopardize the ability of he HPCS diesel generators to crank five successive times without recharging the ; receivers. He new setpoint represents an increase in the conservative direction with respect to the minimum starting air pressure required to stan these generators. Therefore, there is no increased in . the probability of an occurrence of an accident evaluated previously in the SAR. The change in the { r setpoint will not change the ability of associated pressure switches to provided an alarm. Changing ;
- the setpoint will also not affect the way the HPCS diesel generator responds to mitigate the affects .
1 of a loss of offsite power, loss of coolant accident, or both. Therefore, there is no increase in the l
- consequences of an accident evaluated previously in the SAR. This alteration will not add new !
] equipment or change the physical configuration of any installed equipment that could affect the ! ability of the HPCS power supply system from performing its safety related function. Therefore, it !
- does not create the possibility of an accident of a different type that any evaluated previously in the SAR. Increasing the setpoint for the alarm that wams the plant operators of degrading conditions .
in the HPCS air start system will provide an earlier warning and allow prompt corrective action to i be taken. Therefore, there is neither an increase in the probability of occurrence of a malfunction i ! of equipment important to safety, an increase in the consequence of a malfunction of equipment ! important to safety, nor a creation of the possibility of a malfunction of equipment important to l safety of a different type than any evaluated previously in the SAR. The setpoint for the pressure , - switches associated with this modification are not specifically addressed in any of the technical specifications. Therefore, there is no reduction of the margin of safety as defined in the basis for any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. l l Safety Evaluation Initiatine Document: Temporary Alteration 95-004 and Changes to ITS Bases SR 3.6.1.10.1 Description and Basis for Channe: l This temporary alteration installs seals between the guardpipes and the containment liner to maintain the integrity of the primary containment boundary during the replacement of bellows expansion joint for two containment penetrations. The seals are comprised of rubber Link Seals mounted in steel plate holders welded to the contamment liner penetration sleeves. His temporary , alteration allows installation of the bellows expansion joints without restricting outage activities requiring containment to be closed. Implementation of this temporary alteration requires a resision to the Improved Technical Specifications (ITS) Bases Surveillance Requirements SR 3.6.1.10.1. Additional wording is added to the referenced section to allow the use of temporary scalants for closing containment during core alterations or movement ofirradiated fuel. His ITS change is based upon an NRC letter dated June 6,1993. i
' Section VI Page 173 of187
- . - - -- . - - - - . _ - - - - . - - _ . ~ - . . - . . . .-
i i Summary of Safety Evaluation: I I i he temporary seals do not contribute to the function of any system other than the containment ;
- liner. The accident of concern to the plant refueling mode is a fuel handling accident. His L temporary alteration does not affect the method of or equipment used for handling fuel. This 4
temporary alteration does not increase the probability of a fire. Welding during installation will be administratively controlled. He additional combustible fire loading represented by the non- ! ! metallic seals was determined to be acceptable. Therefore, this temporary modification does not l increase the probability of an accident previously evaluated in the SAR. Administrative controls j . will be in place to climinate activity in the vicirsty of the seals when contamment integrity is i required to preclude the possibility ofinadvertent damage to or tampering with the temporary scals. This limits the potential failure mechanism of the temporary components to those conditions i i already addressed in the testing and analysis. The susceptibility of the elastomer material to the 4 heat of welding required for the replacement bellows has been evaluated and found to be of no l consequence. The heat of welding will flow readily into the heat sink provided by the containment i liner. Hot gasses from welding will not build up near the seals. The welding processes that will be i used have been characterized as having little or no significant sparkmg The temporary seals do I not inhibit the function of other systems in the plant. He seal assemblies are qualified to i accommodate all relative displacements between the piping and the containment liner without loss 1 of seal integrity. The temporary assemblies do not come in contact with any equipment other than i i the guardpipes and the containment liner sleeves. Ecrefore, this temporary alteration does not > increase the probability of a malfunction of any safety related system, structure, or component j (SSC) previously evaluated in the SAR. The temporary seals have been designed to maintain containment integrity and appropriate leak tightness through the most severe conditions of plant , 1 modes 4 and 5 under which they may be required to perform. Release of dose offsite as a result of 4 an accident inside containment is bounded by the design basis accident of a release of dose offsite , as a result of a fuel handling accident inside the fuel building. The value of the containment liner !
- as a radiological barrier is not changed or reduced. The change to the passive containment vessel has been designed and tested to ensure that containment structural integrity for modes 4 and 5 is 3~
maintained. The rubber seal material has been shown to be unaffected by the radiological dose from a fuel drop inside containment. The materials of the proposed seals are of the same generic compounds as materials relied upon to provide the contamment pressure boundary during and
- following a design basis small break loss of coolant accident. Ecre is no change in consequences for a malfunction of the containment liner due to the proposed temporary alteration. Therefore, this temporary alteration does not increase the consequences of an accident or a malfunction of any safety related SSC previously evaluated in the SAR. No new accident initiator is introduced by the proposed change. The seals perform their function while allowing full lateral and axial
- displacements of the piping as applicable. His is assured through a combination of full scale ;
mockup testing, analysis, and dedication of the materials to be used. Placing the seals in a different ,
, . location was also considered. No other conditions need to be postulated since the temporary seals J will not be used in plant modes 1, 2, or 3. Therefore, this modification does not create the possibility of an accident or a malfunction of any safety related SSC different from any previously 4 .. evaluated in the S AR.' He NRC letter dated June 6,1993, is with regard to the Summer Plant. In l that letter, alternative containment barriers characterized as being leaktight but not necessarily pressure restraining are accepted for use. The wording of the ITS change does not conflict with the i letter. The bases for maintaining primary containment integrity during fuel handing operations is to ensure that the release of radioactive materials from the primary containment atmosphere will be )
restricted to those leakage paths and associated leak rates assumed in the accident analysis. Since I it will be demonstrated that there is no leakage across the boundary of the elastomer seal under Section VI Page 174 of 187
_.- . - -- - - -.-- -- ---- _~--. - -- i l 1
- 20 psig pressure prior to the seal being placed in service, it is reasonable to state that there is no ;
reduction in the margins of safety of the containment liner during fuel handling. Therefore, this . 1 i temporary alteration does not reduce the margin of safety as dermed in the basis of any technical i specification. For these reasons, this rnodification does not constitute an unresiewed safety question. (
)
l Safety Evaluation Initiatine Document: Temporary Alteration 95-0005 , l 4 ~ Description and Bssis for Channe: 4 r This temporary alteration physically removes the internals of the air supply air dryer skid A outlet : isolation valve. Remosing the internals leaves the valve open for all system conditions. He subject valve is a Target Rock solenoid that will be deleted from the plant as part of a later modification. This temporary alteration also revises a procedure and posts an operator aid to deal with non-functioning dual indicatior. in the control room. The functional capability of the main - l steam air supply system (SVV) train A is restored by this temporary alteration. i-Summary of Safety Evaluation: ' This temporary alteration does not change the design basis or the function of the SVV system for
- automatic depressurization system (ADS) and non ADS safety relief valves (SRV). The safety
, related pressure boundary interface was climinated by a previous modification which eliminates the ! safety function of the subject valve. The nonsafety related portions of the SVV system, including the subject valve, are isolated from the safety related portions. Backflow through the piping to the , ! air dryer is prevented by a check valve. He system will not be required to operate beyond its ! design and testing limits after the temporary alteration. The temporary alteration will not cause new vibration, water hammer, corrosion, thermal cycling or degradation of connected equipment
- important to safety that would exceed their design limits. Original pressure boundary integrity is ,
! preserved by the re-application of the seal weld to the valve bonnet cover. Herefore, this ! g temporary alteration does not increase the probability of an accident previously evaluated in the !' SAR. There is no change to the interfacing system design or operation. Separation and
- environmental criteria for safety related equipment in proximity to the subject valve are not impacted by the scope of this temporary alteration. This temporary alteration will not increase the frequency of operation of equipment important to safety or impose additional testing requirements !
on such equipment. The temporary nature of this change, the use of special tagging, and the use of operator aids are also important features that limit the possibility of human error. Therefore, this temporary alteration does not increase the probability of a malfunction of equipment impoitant to . l safety previously evaluated in the SAR. This temporary alteration does not change the severity of any accident described in the SAR. His temporary alteration does not affect any postulated ; effluent release path, does not affect the predicted quantity of radioactive effluent and does not
; impact any accident analyses described in the SAR. The system functional capability to maintain ,
SRV accumulator pressurized to the required limit in modes 1,2, and 3 is maintained. Therefore, t
; this change does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the S AR. His subject valve no longer has a safety ;
function in the SVV system. Therefore, this temporary alteration does not create the possibility of
- an accident different from any previously evaluated in the SAR. A comparison of the evaluated :
l malfunctions of equipment important to safety listed in the SAR with the malfunctions of equipment important to safety postulated to occur due to this temporary alteration shows that a ; I Section VI Page 175 ofI87 [
malfunction of equipment important to safety different from that previously evaluated in the SAR i is not created. The number of operable SRVs and the low low set function of the SRVs are i i unaffected by this temporary alteration. The ADS trip systems and ADS function are unchanged by this temporary alteration. Therefore, this temporary alteration does not reduce the margin of - safety as defined in the basis of any technical specification. For these reasons, this modification does not cor.stitute an unreviewed safety question. l l Safety Evaluation Initiating Document: Temporary Alteration 95-011 Description of Condition: l This temporary alteration closes the penetration valve leakage control system (PVLCS) air supply valve to the outboard reactor feedwater system (FWS) inlet header B motor operated isolation valve. A small steam leak was discovered at the PVLCS connection to the outboard reactor FWS inlet header B motor operated isolation vah e. Closing the air supply valve is a protective measure to regain operability of the balance of Division I PVLCS. Administrative controls are to placed on the PVLCS air supply valve to maintain its closed position until repair of the line during refueling outage 6. Summary of Safety Evaluation: The structural integrity of the piping has been previously evaluated and determined to be acceptable. Except for this temporary alteration, the balance of PVLCS Division I will remain operable. The PVLCS system is not an initiator of any accidents previously evaluated in the SAR. No conditions are created by closing the air supply valve that can be postulated to cause initiation of any accident in the SAR. Tnerefore, this temporary alteration does not increase the probability of an accident previously evaluated in the SAR. No other systems are affected by the proposed activity. Closing the air supply valve and declaring PVLCS Division I operable may cause an incremental increase in the dose and heat load in the main steam tunnel 25 minutes after a design basis loss of coolant accident (assuming failure of another valve). However, there is no equipment required to operate inside the main steam tunnel that would be adversely affected by the change in emironment. Leakage through the disc of the closed air supply valve during PVLCS operation could cause high air flow and cause the PVLCS compressor to shut off. However, the seating surface of the closed air supply valve is composed of stellite and the valve's ser ice is not one that would be expected to produce significant wear. Even at a seat leakage 10 times the original tested value at installation, there would be no significant affect on system flow rate. Herefore, this temporary alteration does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. He consequences ofisolation the PVLCS supply to the outboard reactor FWS inlet header B motor operated isolation valve have been evaluated. The evaluation clearly shows that the maximum offsite does for the plant licensing basis will not be exceeded in a design basis accident. Because actual as-left values from refueling outage 5 were used in the evaluation, the results only remain valid until the end of refueling outage 6. PVLCS is only assumed to function during a design basis loss of coolant accident. There are no other evaluated accidents that require PVLCS to operate. No changes in this temporary alteration would restrict access to vital areas or otherwise impede actions to mitigate the consequences of reactor accidents. A malfunction of the administratively tagged and locked closed air supply valve is judged not credible. Therefore, this temporary alteration does not increase the consequences of an Section VI Page 176 of187
I I r b accident or a malfunction of equipment important to safety previously evaluated in the SAR. This f temporary alteration does no have a negative impact on any plant systems or their functions that could cause an accident of a different type than previously evaluated. No new accidents associated ; with the creation of missiles from pressurized equipment failure are created since no new : appurtenances or equipment are added by this temporary alteration. No credible PVLCS leak path { to containment has been generated by this temporary alteration. There are no new system ! interactions or dependencies created by this temporary alteration. No new single failures are - , created by this change. The condition of a closed valve in PVLCS Division I is not explicitly j considered in the SAR, however, the acceptability of the condition has been analyzed. Therefore, this temporary alteration does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The technical specification limit on combined leakage through PVLCS penetrations will not be exceeded by this change, nor will the dose value associated with this rate be exceeded The overall design and i functional requirements for PVLCS are still being met. The Technical Specifications Bases . l specify that the purpose of the PVLCS system is to maintain the offsite dose within the accident analysis. Since the new calculated offsite does is within the original accident analysis, this basis is l maintained with the closed air supply valve. The margin of safety limits of 10CFR100 are i therefore implicitly maintained. Therefore, this temporary alteration does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unrevard safety question. Safety Evaluation Initiating Document: TA 95-015 Descrintion and Basis for Channe: This change provides a supply of compressed air to charge the starting air receivers for the high pressure core spray (HPCS) diesel generator while permanent changes are being made per modification request (MR) 88-0236. Both existing compressor /&yer/ piping systems are t disconnected at a point upstream of each receiver's inlet check valve and temporary stainless steel i tubing connected to the system. All modifications are on the nonsafety related side of the starting
- - air system. One of the 270 psig relief valves near the air dryers in the existing system will be
. relocated for the temporary alteration (TA) installation. It will be installed upstream of the l pressure regulator and is designed to protect the system in the event the compressor pressure switch fails to shut off the compressor. Summary of Safety Evaluation: l' This TA does not change the basic function of the safety-related parts of the starting system. Even though the typical receiver pressure will be 213 psig instead of 240, there is plenty of air for up to five start attempts of the HPCS diesel generator. In case ofoverpressure, the blowdown capacity F of the 270 psig relief valve exceeds the temporary compressor's rated flow. Temporary tubing and hoses will be securely supported so as not to damage safety-related components in the event of - , rupture or breakage. Temporary electrical power cables will be separated from safety-related electrical raceway and process piping, or fire wrapped as needed Weather protection enclosure is , i- provided for temporary equipment to prevent freezing of the dryer air passages and minimize the potential temperature effects. This TA will not increase the probability that the diesel will fail to start the first time on the emergency signal. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety l l Section VI Page 177 of I87 t E
, .- s e--.- --n- .
-f I
e ; { previously evaluated in the SAR. He temporary compressor and dryer are not inside the missile- l protected building and are thus susceptible to tornado-generated missiles. This is not a safety j concern since these are nonsafety related parts of the air start system. Herefore, this change does , not create the possibility of an accident or a malfunction of equipment important to safety different i from any previously evaluated in the SAR or reduce the margin of safety as defmed in the basis of : any technical specification. For these reasons, this modification does not constitute an unreviewed i safety question. ' , i 1 Chante Document: Temporary Alteration 96-0002 : 1 Description and Basis for Channe: , in order to support the inservice testing (IST) program during refueling outage 6, freeze seals were ; required on the discharge lines of relief valves E22-RVF014, F035, and F039 due to the fact that : there are no isolation valves for the relief valves. %ese relief valves are located in the auxiliary ! building and discharge into the high pressure core spray (HPCS) test retum line which discharges into the suppression pool. , Summary of Safety Evaluation: , The accident of concern related to plant cold shutdown is a fuel handling accident (FHA). This modification has no affect on the method or the equipment used in a fuel handling. Therefore, this _
~
modification does not increase the probability of an accident presiously evaluated in the SAR Re SAR evaluated accident scenario of a FHA in the fuel building bounds the FHA inside i contairiment. Also, the seal from the suppression pool would be adequate for containment isolation purposes and no atmospheric release path is present. Therefore, this modification does not increase the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. Since spraying water in the auxiliary building has been previously evaluated, no new initiator is introduced from the temporary alteration. Therefore, this ; modification does not create the possibility of an accident different from any presiously evaluated in the SAR. The freeze seal procedure (CMP-9186) includes adequate instructions to ensure that 1 the piping will not be damaged by the freeze seal and there are no electrical components in the area of the work which would be adversely affected by water spray. Therefore, this modification does ! not increase the probability of a malfunction of equipment important to safety presiously evaluated in the SAR. Implementation of the freeze seal in accordance with the aforementioned procedure will ensure that probability of failure of the discharge lines is not increased. Therefore, this modification does not create a malfunction of equipment important to safety different from any previously evaluated in the SAR. The HPCS system is not required in modes 4 and 5 during i
; refueling and implementation of the temporary alteration will ensure that containment integrity is i maintained. Herefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question.
Section VI Page 178 of187
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?
i e Safety Evaluation laitiating Document: Temporary Alteration %-003 l Description and Basis for Channe: This temporary alteration changes the control room annunciator for the reactor recirculation pump differential temperature (AT) interlock bypass switch. He contacts which control the annunciator j light will be reversed. Instead of the annunciator window being illuminated when the AT interlock l bypass switch is in the ' Bypass' position, it will be illuminated when the AT interlock b3pass l switch is in the ' Normal' position. De label for the annunciator window will be changed { accordingly. When the bypass switch is in the ' Normal' position, a detected AT between the reactor steam dome and the recirculation pump suction temperature will cause the recirculation i pump to downshift to 25% speed or to trip offif the other recirculation pump is at 100% speed. l' This temporary alteration changes the circuit configuration so that the annunciator window is energized when the bypass switch is in the ' Normal' position, i.e., the AT interlock is activated. {' The normal position of the switch will be reflected as an ' Abnormal' condition on the annunciator window. This change will enable the annunciator window to stay de-energized when operating - l outside the region of potential cavitation (expected to be the majority of the time), and only to . illuminate when the switch is placed in the normal position (i.e., activating the interlock). The l temporary alteration will only be implemented during two loop operation and will not affect the [ single loop operation. Once the plant enters into a region where cavitation may occur (as def'med l in the analysis), the operator will be instructed to change the bypass switch to the normal position, j activating the AT interlock and illuminating the annunciator window. Additional emergency response infonnation system (ERIS) points have been installed which will allow the actual inputs ; into the AT interlocks to be monitored. An audible alarm will sound to alert operations if the AT ! f setpoint is met to indicate potential cavitation. The changes still conform to tim design specifications. Herefore, this temporary alteration continues to maintain protection against , cavitation due to inadequate subcooling of downcomer flow for the reactor recirculation system i and provides an additional alarm via the plant process computer during all regions of plant i operation. [ Summary of Safety Evaluation: 1 None of the analyzed accident scenarios in the SAR describe the AT interlock bypass switch or ! take credit for it. The status of the AT interlock bypass switch does not contribute to any accidents ; described in the SAR. The changes in this temporary alteration are nonsafety related and do not degrade the performance of the recirculation pumps or any safety system in the accident analyses, i below the design basis. Herefore, this temporary alteration will not increase the possibility of an l accident previously evaluated in the SAR. He function of the AT interlock is nonsafety related l and is not required for the safe shutdown of the plant or to mitigate the consequences of an j accident.' These changes do not affect any other recirculation pump trips. Thus, no new accident i conditions are introduced that would result in an increase in the dose to the public as a consequence l of an accident. Therefore, this temporary alteration will not increase the consequences of an I accident previously evaluated in the SAR. The AT interlock bypass switch will continue to pro 5ide I the AT interlock while the plant is operating in a region where the possibility of casitation exists, but will be bypassed during normal operation when the possibility of casitation does not exist. The affected instrumentation is nonsafety related and is not required for the safe shutdown of the plant. His change does not create new interfaces with safety related equipment or require any new equipment to be added it will not affect the redundancy and separation of any equipment and will Section VI - Page 179 of 187 l
l l
; not a&ct the normal operation of the reactor recirculation pumps. Therefore, this temporary
- alteration will not create the possibility of a accident of a di& rent type than any previously
! ' evaluated in the SAR. He AT interlock serves to protect the recirculation components from cavitation when the subcooling margin is not sufficient. No changes are being made to any equipment or control circuits important to safety. There is no e&ct to the separation or redundancy of any circuits. Therefore, this temporary alteration will not increase the probability of a malfunction of equipment impostant to safety previously evaluated in the SAR. If a malfunction of the recirculation pumps, the flow control valves, or the jet pumps occurs for any reason, the ;, consequences of the malfunction cannot be a&cted by the status / operation of the AT interlock ,
t bypass switch or the associated instrumentation or the annunciator window his change is nonsafety related and cannot cause an increase in dose to the public. There is no impact on fission j !' product barriers, release path, or effluent processing equipment. Therefore, this temporary ~ l alteration will not increase the consequences of a malfunction of equipment important to safety ; j previously evaluated in the SAR. %c redundancy and separation of control circuits for any safety - l
. related equipment is not affected. Therefore, this temporary alteration will not create a malfunction of equipment important to safety different from any previously evaluated in the SAR. l
'%c AT interlock switch and the annunciator affected by this temporary alteration are not included i l~ in the basis for any technical specifications' margin of safety or in the Technical Requirements '
Manual. He specific technical specifications associated with the reactor recirculation pumps do ] not mention the AT interlock bypass switch or its operability. Therefore, this temporary alteration ! will not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this temporary alteration does not constitute an unreviewed safety question. i l Safety Evaluation Implementine Document: Temporary Alteration 96-004 l i
. Description and Basis for Channe:
i This temporary alteration is to temporarily use penetration WS561N01 of the fuel building wall for insenice inspection (ISI) cable routing through the fuel building to the reactor pressure vessel l (RPV). The cables will be used to control automated reactor vessel weld inspection equipment and i
! for communication during testing for refueling outage 6. i i !
Summary of Safety Evaluation: ! Temporarily breaching the fuel building wall and routing the of RPV ISI equipment cables through , , the penetration and the fuel building will have no affect on the hardware, controls, or operating i limits for fuel handling. Cable routing will be in accordance with ADM-0073, thus, no cable will be routed in close proximity to divisional cables, conduit, or cable trays. Also, fire watches will be j posted where cables pass through fire barriers. Herefore, this modification does not increase the ! probability of an accident previously evaluated in the SAR. Although the penetration will j temporarily provide an atmospheric release path, the area of the path is small and the fuel building j ventilation (HVF) system will still be able to mamtam a negative pressure of >0.25 in, w.g. as ) designed. Therefore, this modification does not increase the consequences of an accident I previously evaluated in the SAR. This modification does not create the possibility of an accident l different from any previously evaluated in the SAR, due to the existence of a fire watch and adherence to ADM-0073. The test equipment cables are routed per ADM-0073 and will not be , electrically can=M with any safety related or important to safety equipment in the plant. ! i l
' Section VI Page 180 ofI87 ]
i J
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Therefore, this modification doc-s not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. The breaching of the fuel building wall will not negate the HVF system's ability to maintain a negative pressure. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety presiously evaluated in the SAR. The proposed activity does not affect the operation, function, or control of and safety related or important to safety equipment in the plant. Also, fuel movement will be stopped per AOP-0029 on receipt of a storm waming. Therefore, this modification does not create a malfunction of equipment to safety different from any previously evaluated in the SAR. The HVF system is administratively turned on during fuel handling, therefore, the drawdown time is not important. He HVF system will be able to maintain the 0.25 in. w.g. negative pressure requirement. Herefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. Channe Document: Temporary Alteration %-005 Description and Basis for Channe: Temporary Alteration 96-005 provides for the breach and temporary use ofpenetration WS651N02 for a 1" diameter hose to perform hydrolazing activities at the refueling floor during refueling outage 6. The hose installation and routing requirements shall adhere to ADM-0073 requirements to ensure design separation and clearances for divisional cables, piping, tubing, and supports. Summary of Safety Evaluation: Temporarily breaching the fuel building wall penetration and routing of hose to the refueling floor will not increase the probability of an occurrence of a fuel handling accident (FHA) or a spent fuel cask drop accident as evaluated in the SAR. Also, the inventory from a break of the hose would end up in the suppression pool for the portions of the hose within containment. For the fuel handling, the inventory from a hose break would be bounded by other postulated events such as inadvertent fire suppression system actuation. Therefore, this modification does not increase the probability of an accident previously evaluated in the SAR The temporary modification will not affect the radioactive effluent release paths as the ventilation system will continue to maintain a negative pressure in the fuel building during penetration breach.. Therefore, this modification does not increase the consequences of an accident previously evaluated in the SAR. The proposed activity increases the ventilation air flow path through the fuel building penetration WS651N02 by approximately 2%. Based on past STP results, however, the fuel building will be able to maintain adequate negative pressure. Also, since the hose will be routed in accordance with ADM-0073, divisional cables, conduits, or raceways will not be damaged due to malfunction or failure of the hose. Therefore, this modification does not create the possibility of an accident different from any previously evaluated in the SAR. As stated above, the fuel building ventilation (HVF) system will be able to maintain its design requirements of 0.25 inches negative pressure. Also, vulnerability to tornadoes during the periods when the penetration is breached and before the seal material is entirely cured is minimized by the mandatory step to verify that a storm waming is not in effect. Finally, the hose will not be connected to safety related equipment or equipment important to l safety. Therefore, this modification does not increase the probability of a malfunction of , equipment important to safety previously evaluated in the SAR. HVF is a single failure proof and . Section VI Page 181 ofI87 4
has redundant charcoal trains. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. No new or different types of components are being added by the temp alt and the hose will not be connected to plant equipment. No new failure modes are being introduced and nothing is being done which would affect the assumptions of the current accident analysis. Therefore, this modification does not create a malfunction of equipment important to safety different from any presiously evaluated in the SAR. Technical Specifications Surveillance Requirement 3.6.4.1.5 requires the fuel building ventilation charcoal filtration system to draw down the fuel building to greater or equal to 0.25 inches w.g. in less than or equal to 12.5 seconds. Results from past surveillance test procedures indicate that these limits will be met. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unresiewed safety question. Safety Evaluation Initiatine Document: Temporary Alteration 96-007 Description and Basis for Chance: This temporary alteration fails an air-operated steam jet air ejector (SJAE) intercooler drain valve to the open position. The associated SJAE steam supply valve is leaking by its seat. The steam then condenses in the intercooler, creating a high level and extreme high level alarm once per shift. In accordance with site procedures, operations must then fail open this air-operated intercooler drain valve to drain the out-of-senice intercooler in order to prevent the masking of any actual high levels in the active SJAE. This temporary alteration will accomplish the same task by constantly draining the out-of-senice intercooler of condensed steam, with a greatly reduced impact on offgas flow. Once the air-operated valve is failed open, a manual drain valve is throttled open slightly to maintain the high and extreme high level alarms clear. Summary of Safety Evaluation: Section 15.2.5 of the SAR describes the loss of condenser vacuum event and identifies four possible causes for a loss of vacuum: failure of the SJAE, loss of gland sealing steam, opening of a vacuum breaker and loss of circulating water pumps. It has been shown that the practice of failing open the air-operated intercooler drain valve increases offgas flow by approximately 20 SCFM. Ilowever, no reverse flow is observed and no change in condenser is seen. The increased ' flow is due to evacuating the intercooler and it's drain line. By throttling open the manual drain valve slightly, most of the pressure drop will be felt across it, which will minimize the effect on offgas flow, in addition, the sequence for installing and removing this temporary alteration will prevent having both the air-operated and manual drain valves open at the same time. Since this temporary alteration is on the standby SJAE train, it will not effect condenser vacuum and cannot ; increase the possibility of an accident previously evaluated in the SAR. Section 15.2.5 of the SAR l describes the loss of condenser vacuum event. Five seconds into the event, the main steam isolation valves (MSIVs) close, isolating the main condenser and the drain lines which are open to the intercooler. Since the drain line is isolated from the reactor, this temporary alteration cannot increase the consequences of an accident previously evaluated in the SAR. Since this temporary ! alteration is on the standby SJAE train, it cannot effect condenser vacuum and cannot create the l possibility of an accident different from any previously evaluated in the SAR. The condenser air removal intercooler drain is not safety related, nor does it support or have the potential to efTect equipment import to safety. 'Ihis temporary alteration lines up equipment drains in a configuration Section VI Page 182 of 187 I I
which they would be in if the standby train were the running train. The condenser air removal system is not designed or required to mitigate the consequences of an accident. During the loss of condenser vacuum event described in section 15.2.5 of the SAR, the MSIVs are closed, which isolates the main condenser from the reactor, terminating any potential release. Therefore, this temporary alteration cannot increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. The condenser intercooler drain is located on the 95' elevation of the turbine building and has no interrelations or interactions with equipment important to safety, nor does it support equipment important to safety. Therefore, this temporary alteration cannot increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. He condenser intercooler drain is located on the 95' elevation of the turbine building and has no interrelations or interactions with equipment important to safety, nor does it support equipment important to safety. Therefore, this temporary alteration cannot create a i malfunction of equipment important to safety different from any previously evaluated in the SAR. Here are no technical specifications associated with the condenser air removal drains. Therefore, this temporary alteration cannot reduce the margin of safety as dermed in the basis of any technical specification. For these reasons, this temporary alteration does not constitute an unreviewed safety question. Safety Evaluation Initiating Document: Temporary Alteration 96-0009 Descrintion and Basis for Chance: This temporary alteration provides interim jumper connections around make-up water system (MWS) motor bearing cooling water strainer number six. This temporary alteration will provide cooling water flow to the running MWS pump motors 4A and 4C during maintenance to the aforementioned strainer. Jumper connections are provided to both pumps regardless of which pump is running in case a pump swap is required while work is in progress. Summary of Safety Evaluation: The purpose of the MWS is provide make-up water to the circulation water system and the service ; water cooling system fiumes. The MWS is not safety related and does not support the safety function of any other safety related system, structure, or component (SSC). Therefore, this change does not increase the probability of an accident or a malfunction of equipment important to safety ; previously evaluated in the SAR. The are no accident consequences tied to the MWS, its functions, or system supported by the MWS. Therefore, this change does not increase the ! consequences of an accident or a malfunction of equipment important to safety presiously . evaluated in the SAR. While this temporary alteration is installed, the MWS will still continue to function as described in the SAR. Therefore this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. The MWS is not safety related. Neither the Tecimical Specifications nor the Technical Requirements Manual have any limiting conditions or action tied to the operation of the system. Therefore, this change does not reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed ; safety question. I 4 i Section VI Page 183 of187
- Safety Evaluatine Initiation Document: Temporary Alteration 96-012 Description and Basis for Chanee: This temporary alteration is applicable to residual heat removal (RHR) trains A, B and C, and the low pressure core spray (LPCS) systems. His temporary alteration electrically disconnects the low pressure emergency core cooling system (ECCS) pressure permissive annunciators from the interlocking which allows manual injection valve operation when pressure is below design pressure of the upstream ECCS piping. This temporary alteration is installed on the nonsafety related annunciation portion of the low pressure injection systems to eliminate nuisance alarms to maintain a " Black board" philosophy in the control room. Use of this pressure permissive interlock for injection valve testing is not used at RBS and there are no procedures which utilize this. alarm. Summary of Safety Evaluation: Review of accidents described in SAR Chapter 15 found no accident for which the proposed temporary alteration would increase the probability of occurrence of an accident evaluated previously in the SAR. Elimination of the control room annunciation for the low pressure permissive interlock for manual stroking of the injection valve does not affect the interlock feature on the ECCS low pressure injection valves or inhibit its function. The interlock is still functional. Therefore, this change does not increase the probability or consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR. This temporary alteration does not affect any margin of safety as dermed in the basis. The lifted lead and annunciator are in a nonsafety related portion of the logic circuit and are not credited for any action. The annunciator is not required for accident mitigation. Herefore, this change does not create the possibility of an accident or a malfunction of equipment important to safety different from any previously evaluated in the SAR. This change does not reduce the margin of safety as ; defined in the basis of any technical specification. For these reasons, this modification does not l constitute an unreviewed safety question. i i I l Section VI Page 184 of187 l
SECTION VII MISCELLANEOUS EVALUATIONS i l l i l l I Section VI Page 185 of187 l 1
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c Safety Evaluation Initiatina Situation: Gell Fuel Receipt 4' Description and Basis for Safety Evaluation: The following evaluation was performed to address the receipt of new GElI fuel to be loaded for RBS cycle 7. gel 1 is a 9X9 fuel bundle and has not been loaded at RBS previously. Summary of Safety Evaluation: The new GElI bundics have similar weights, geometry, and handling requirements than GE 8X8 fuel bundles for which the fuel handling equipment and storage facilities were designed. Herefore, this modification does not increase the probability of an accident previously evaluated in the SAR. The design basis fuel handling accident considers the GElI fuel type. The gel 1 fuel meets acceptance criterion outlined in a Generic Environmental Assessment (GEA) which concluded that no significant radiological or non-radiological impacts are associated with the increased enrichment and exposure. Analyses were performed to confirm that the acceptance criteria are satisfied in the
- new and spent fuel storage racks. Therefore, this modification does not increase the consequences of an accident previously evaluated in the S AR. The new gel I bundles have similar weights, geometry, and handling requirements than GE 8X8 fuel bundles for which the fuel handling equipment and storage facilities were designed. He Gell fuel meets acceptance criterion outlined in a GEA which concluded that no significant radiological or non-radiological impacts are associated with the increased enrichment and exposure. Analyses were performed to confirm that the acceptance criteria are satisfied in the new and spent fuel storage racks. Therefore, this modification does not create the possibility of an accident different from any presiously evaluated in the SAR. The new GElI bundles have weights, germtry, and handling requirements similar to GE 8X8 fuel bundles for which the fuel handling equipment and storage facilities were designed.
The equipment required to be used for storage and handling is identical to equipment used for previous reloads. Therefore, this modification does not increase the probability of a malfunction of equipment important to safety previously evaluated in the SAR. He design basis fuel handling accident considers the GElI fuel type. Analyses were performed to confirm that the acceptance criteria are satisfied in the new and spent fuel storage racks. Therefore, this modification does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR. The new GElI bundles have similar weights, geometry, and handling requirements than GE 8X8 fuel bundles for which the fuel handling equipment and storage facilities were designed. The gel 1 fuel meets acceptance criterion outlined in a GEA which concluded that no significant radiological or non-radiological impacts are associated with the increased enrichment and exposure. Analysis were performed to confirm that the acceptance criteria are satisfied in the new and spent fuel storage racks. Therefore, this modification does not create a malfunction of equipment important to safety different from any previously evaluated in the SAR. He design basis fuel handling accident considers the GElI fuel type. Analyses were performed to confirm that the acceptance criteria are satisfied in the new and spent fuel storage racks. Therefore, this modification does not reduce the margin of safety as defined in the basis of any technical
- specification. For these reasons, this modification does not constitute an unreviewed safety question.
Secten Vil . . Page 186 of187
Safety Evaluation Number: SEN 95-0011 -. Description and Basis for Channe: This is a second summary report to the NRC based on a revised summary to this rafety evaluation. !
' Its purpose is to better explains the nature of the change. He safety evaluations md conclusions were not changed. j i
his change to the SAR is required to reflect the current " normal" make-up of the Fire Brigade. I Specific reference to Maintenance, Chemistry, or Radiation Protection technicians as part of the make-up was deleted. This change also removes the training requirement from the position-specific staff training matrix. There are no changes being made to the training or qualifications required of i fire brigade members. j Summary of Safety Evaluation: ' The change will accurately reflect the current make-up of the fire brigade, but will not change the training or qualifications required of fire brigade members. The change is administrative in nature ; and does not affect the number of members or the qualifications requirements of the Fire Brigade. The leader and operations members are not changed so decisions affecting plant operations will not i be affected by this change. Therefore, this change does not increase the probability or l consequences of an accident or a malfunction of equipment important to safety previously ) evaluated in the SAR. This change does not affect any margin of safety as defined in the basis. ! The change is administrative in nature. The training requirements and skill levels of the brigade members are not changed and therefore would not change the role of the fire brigade in mitigating the consequences of a fire. Therefore, this change does not create the possibility of an accident or j a malfunction of equipment important to safety different from any previously evaluated in the SAR l or reduce the margin of safety as defined in the basis of any technical specification. For these reasons, this modification does not constitute an unreviewed safety question. 1 I l l l l l I 4 Section VII Page 187 of 187 _ _ - .-. . ~}}