ML18017A277
| ML18017A277 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 12/28/1994 |
| From: | Byram R PENNSYLVANIA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PLA-4240, NUDOCS 9501050116 | |
| Download: ML18017A277 (327) | |
Text
PRIC)RI EY ACCELERATED I<IDS Pl<OCI'.SSliG)
A REGULATORY (INFORMATION DXSTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9501050116 DOC.DATE: 94/12/28 NOTARIZED: No DOCKET N FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION BYRAM,R.G. Pennsylvania Power & Light, Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
R
SUBJECT:
Forwards summary rept for safety evaluations approved during Jan-Dec -1993 for Susquehanna SES Units 1 6 2.
DISTRIBUTION CODE'OOID COPIES RECEIVED'LTR TITLE: OR Submittal: General Distribution NOTES Q ENCL l SIZE'cP 5+
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I RECIPIENT COPIES RECIPXENT COPIES XD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-2 LA 1 1 PD1-2 PD 1 1 POSLUSNY,C 1 1 XNTERNAL: ACRS 6 6 FI LE CENTER +0: 1 1 NRR/DRCH/HICB 1 1 NRR/D'S SA/S P LB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS 2 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 NOTES: 1 1 D
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N YOTE TO ALL "Rl DS" RECIPI E YTS:
PLEASE HELP US TO REDUCE iVKSTE! COYTACTTHE DOCL'ifEY,'I'COYTROL DESK, ROOh f P I -37 (rXT. 504-00S3 ) To I'.LINfl Y KTE YOL'R iAifL Pl<ohl Dls'rRlnu'rlo Y, Lls'I's I:oR DocnfEi"I's voL'oi "II I'.o!I TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 17
Pennsylvania Power 8 Light Company Two North Ninth Street ~ Allentown, PA 18101-1179 ~ 610/774-5151 Robert G. Byram Senior Vice President Nuclear 610l774-7502 Fax: 610/774-5019 Submitted pursuant to 10CFR50.59 DEC 28 1994 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Station P 1-137 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION 10CFR50.59
SUMMARY
REPORT - 1993 Docket Nos. 50-387 and 50-388
Dear Sir:
Pursuant to 10CFR50.59(b), enclosed please find a summary report of the safety evaluations approved during the period from January 1, 1993 to December 31, 1993 for Susquehanna SES Units 1 and 2.
The report format is as follows:
SER No.- Unique number for each safety evaluation.
Cross Reference- Reference to the document for which the safety evaluation was prepared.
Description C/tange- 2 brief description of the change made to procedures, equipment or tests.
Sununnry- 2 summary ofPP& I 's review of the three requirements for determining an unreviewed safety question as defined in 10CFR50.59(a)(2).
In accordance with the Final Rule "Reducing the Regulatory Burden on Nuclear Licensees'Ref.
57FR39353 pub. 8/31/92) we have chosen to change the frequency of submittal of the 10CFR50.59 Summary Report from annually to a Unit 1 refueling cycle basis. Beginning with the upcoming Unit 1 8th Refueling and Inspection Outage scheduled to begin in March 1995 and end in May 1995, we will submit our 10CFR50.59 Summary Report within 6 months aAer the end of the outage. A 10CFR50.59 Summary Report submittal covering both SSES units will thereafter be submitted within 6 months following closure of each Unit 1 refueling outage. This action will not affect the substance of our evaluation summaries or the documentation required for g 50.59 type changes.
i050lib 94i228 PDR ADOCK 05000587 PDR R
FILE R41-2A PLA-4240 Document Control Desk Ifyou have any questions, please contact J. B. Wesner at (610) 774-7911.
Very truly yours, R. G. r En osure cc: NRC Region I Mr. C. Poslusny, NRC Sr. Project Manager-OWFN Ms. M. Banerjee, NRC Sr. Resident Inspector - SSES
J'~
I 93-001 C SS EE E C DCP 93-9003Z, Unit Common D C C GE:
Add a transient suppressor on the contactor for Fuel Oil Transfer Pump (OP514A) and add flexible conduit on selected portions of circuitry in Panel OC512A to reduce the effects of induced fluctuation in the electronic governor power supply of Diesel Generator A.
$ 8ltlhBX:
No. The addition of a transient suppressor to the contractor coil of the Fuel Oil Transfer Pump OP514A and the addition of flexible conduit on select portions of circuitry in Panel OC521A does not affect any of the postulated initiating events identiTied in Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis of NUREG 0776.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. Thus, based on engineering judgment, the increase in probability is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations.
The consequences of a random single failure of the transient suppressor is no different than the existing consequences of a failure of the Diesel Generator A.
II~ No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed and did not identify the possibility of an accident or malfunction of a different type caused by the proposed action.
The addition of the transient suppressor on the contactor of the Fuel Oil Transfer Pump OP514A was dynamically qualified through dynamic qualification testing.
III. No. The proposed action does not affect the operability of Diesel Generator A (governed by Technical Specification 3/4.8.1) and therefore does not reduce the margin of safety as defined in the basis of any Technical Specification. This is based on a review of Technical Specification, FSAR, and NUREG 0776.
9501050ll6
MiZLZSEE: 93-002 CE: DCP 93-9003Z Rev. 1, Unit Common D C C Add a transient suppressor on the contactor for Fuel Oil Transfer Pump (OP514A) and a varistor on the 2FOT Relay coil and add flexible conduit on selected portions of circuitry in Panel OC512A to reduce the effects of induced fluctuation in the electronic governor power supply of Diesel Generator A.
QQJMJV~lA The addition of a transient suppressor to the contactor coil of the Fuel Oil Transfer Pump OP514A and a varistor on the 3FOT relay coil and the addition of flexible conduit on select portions of circuitry in Panel OC521A does not affect any of the postulated initiating events identified in Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis or NUREG 0776.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. Thus, based on engineering judgment, the increase in probability is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations.
The consequences of a random single failure of the transient suppressor is no different than the existing consequences of a failure of the Diesel Generator A.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed and did not identify the possibility of an accident or malfunction of a different type caused by the proposed action.
The addition of the transient suppressor on the contactor of the Fuel Oil Transfer Pump OP514A and a varistor on the 3FOT relay coil were dynamically qualified through dynamic qualification testing.
III. No. The proposed action does not affect the operability of Diesel Generator A (governed by Technical Specification 3/4.8.1) and therefore does not reduce the margin of safety as defined in the basis of any Technical Specification. This is based on a review of Technical Specifications, FSAR, and NUREG 0776.
REFERENCE:
NL 89-002, Revision 2, Unit Common DESCRIPTION OF CHANGE'valuation of the potential for Iow level radioactive effluent releases to the atmosphere during normal operation of the Second Sort Facility.
SUMMARY
No. Maintaining all material entering the Second Sort Facility below .5 mr/hr does not increase the probability of occurrence or consequences of an accident since this dose is bounded by calculations performed in previous safety evaluations. The calculated whole body dose of 470 millirem is below the site design criteria values provided in 10CFR100.
II. No. The Second Sort Facility has no safety related functions and does not interface with any safety related systems. The Second Sort Facility does not provide for any mechanism for routine radiological effluent dilution and/or release. No additional accidents or malfunctions of a different type than any evaluated previously in the FSAR have been identified. Also, due to normal operation of the Second Sort Facility ~
far below any of the limits specified in the Technical Specifications, any proposed accident will not reduce the margin of safety bounded by the FSAR as addressed previously.
P i. Ntv The Second Sort Facility operation has no bearing on limited safety system settings or reacto'r protection safe shutdown pursuant to Section 2.0 of Technical Specifications. The proposed operation of the Second Sort Facility is bounded within existing licensing documents. In addition, the facility does not interfere with any other system or component governed by the Technical Specifications.93-004 C: DCP 93-3010, Unit 1 DESC 0 C Install new electrically driven 24 ton trolley/hoist used for transporting Unit 1 containment equipment.
SlJJNJg~g:
No. The new 24ton trolley/hoist and special rigging devices have been designed to address all safety impact concerns as well as meet all the NUREG-0612 requirements for lifting heavy loads in nuclear plants. FSAR Chapter 6 and 15 have been reviewed, and based on the above, this modiTication - including in-plant usage of the hoist and rigging - will have no affect on the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification complies with all the PP8L Licensing commitments to NUREG-612 and NRC Generic Letter 85-11 for the lifting of heavy loads in nuclear plants and will be designed and tested so as to preclude any possibility of a dropped shield block that could damage safety related components. This modiTication does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Unit 1 Technical Specifications have been reviewed and the 24 ton trolley/hoist is not mentioned in any Technical Specification.
Since this modification has been designed to meet all the applicable requirements of NUREG-0612 and NRC Generic Letter 85-11 for the control of heavy loads at nuclear plants and seismically qualified for all safety impact concerns, the proposed actions taken for this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.93-005 E C: DCP 92-30208, G, J, K, L, M, N, Unit 1 DESC 0 C This change will make the necessary modifications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers located on selected SSES piping systems.
SUJHJRhKL:
No. A review of FSAR Chapter 15 was done and it was determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not increase with the implementation of the proposed action. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to be modified accordingly. The reanalysis is in accordance with the original design basis and will be performed in accordance with accepted industry codes. All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the design specifications and FSAR have been addressed in the reanalysis effort.
II. No. The accident events/causes in FSAR Chapter 15 that are applicable to this proposed action involve equipment malfunctions/failures and pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations.
The original allowable equipment interface loads are still being met and the pipe break criteria remains unchanged. Therefore, the proposed change will not increase the probability of an accident and a new unanalyzed type of malfunction or accident will not be created by the removal or replacement of snubbers.
III~ No. The Technical Specification bases related to these proposed modifications are:
3/4.1.5 Standby Liquid Control System, 3/4.4.1 Recirculation System, 3/4.4.7 Main Steam Line Isolation Valves, 3/4.4.8 Structural Integrity, 3/4.6.1 Primary Containment, 3/4.6.3 Primary Containment Isolation Valves, 3/4.7.3 Reactor Core Isolation Cooling System, and 3/4.7.4 Snubbers.
The bases, as defined in the Technical Specifications, deal with snubber test requirements and are not directly affected by this proposed action. Removing or replacing snubbers will not reduce the margin of safety as defined by the Technical Specification noted.
SE 93-006 C: DCP 92-9059, Unit 1 D SC This change relocates emergency lighting lamp assemblies to be in compliance with Specification E-50, FSAR 9.5.3 and 10CFR50, Appendix R, and ensures that adequate illumination levels are maintained.
SVJmhBY.:
No. The relocation of the emergency lighting lamps does not adversely affect the safety related systems or equipment. Electricai separation is maintained between the new location of lamps and the existing systems by adhering to the requirements of PP8L Specification E-1012.
II. No. The proposed action does not affect the plant electrical system which provides normal and emergency AC power to reactor protection and engineered safety feature equipment. Failure of the emergency lighting units will not prevent the plant electrical system from performing its design safety functions. Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15.
III. No. Technical Specifications 3/4.8.1, 3/4.8.2, 3/4.8.3, "AC Sources, DC Sources and Onsite Power Distribution were reviewed and determined that the relocation of lamps to maintain adequate illumination levels does not increase any voltage drop or reduce load ampacity. Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specifications.
II
~ ESU5LMl: 93-007 C: DCP 92-5007, Unit Common D C Replace 15 of the 21 manual line switches located in the Remote Data Analysis System (RDAS) peripheral cabinet in the EOF computer room with automatic line switches.
S~UMM~:
I. No. RDAS is not safety-related (FSAR Section 7.5c and Section 7.7.1.7.1.2). There are no safety-related systems in the EOF and the equipment does not interface with any safety-related equipment in the plant.
II. No. Since RDAS is not safety-related the failure modes after this modification remain the same, for the same systems, as that before the modification. The modification will in no way affect any safety-related systems.
III. No. RDAS is not addressed in the Technical Specifications. RDAS does not perform any safety functions and does not interface with any equipment which performs a safety-related function.
Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
BE: 93-008 C: DCP 93-3011, Unit 2 D C 0 C Install new electrically driven 24 ton trolley/hoist used for transporting Unit 2 Containment Equipment.
SVJHIHhBX:
No. The new 24ton trolley/hoist and special rigging devices have been designed to address all safety impact concerns as well as meet all the NUREG-0612 requirements for lifting heavy loads in nuclear plants. FSAR Chapters6 and 15 have been reviewed, and based on the above, this modification - including in-plant usage of the hoist and rigging - will have no affect on the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification compiies with all the PP&L Licensing commitments to NUREG-0612 and NRC Generic Letter 85-11 for the lifting of heavy loads in nuclear plants and will be designed and tested so as to preclude any possibility of a dropped shield block that could damage safety related components. This modification does not create a possibility of an accident or malfunction of a different type that any evaluated previously in the FSAR.
III. No. Unit 1 Technical Specifications have been reviewed and the 24 ton trolley/host is not mentioned in any Technical Specification. Since this modification has been designed to meet all the applicable requirements of NUREG-0612 and NRC Generic Letter 85-11 for the control of heavy loads at nuclear plants and seismically qualified for all safety impact concerns, the proposed actions taken for this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.93-009 E C: DCP 93-9006Z, Unit Common D C Add a transient suppressor on the contactor Fuel Oil Transfer Pump (OP514B) and varistor on the 3FOT relay coil and add flexible conduit on selected portions of circuitry in Panel OC512B to reduce the effects of induced fluctuation in the electronic governor power supply of Diesel Generator B.
SLlHIMhBX:
No. The addition of a transient suppressor to the contactor coil of the Fuel Oil Transfer Pump OP514B and the addition of flexible conduit on select portions of circuitry in Panel OC521 does not affect any of the postulated initiating events identified in Chapter6 and 15 of FSAR, the Design Assessment Report, the current Reload Analysis or NUREG-0776.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. Thus, based on engineering judgment, the increase in probability is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations.
The consequences of a random single failure of the transient suppressor is no different than the existing consequences of a failure of the Diesel Generator B.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed and did not identify the possibility of an accident or malfunction of a different type caused by the proposed action.
The addition of the transient suppressor on the contactor of the Fuel Oil Transfer Pump OP514B and a varistor on the 3FOT relay coil were dynamically qualified through dynamic qualification testing.
III. No. The proposed action does not affect the operability of Diesel Generator B (governed by Technical Specification 3/4.8.1) and therefore does not reduce the margin of safety as defined in the basis for any Technical Specification. This is based on a review of Technical Specifications, FSAR, and NUREG-0776.93-010 E E E CE: DCP 93-9004 and 93-9005, Units 1 and 2 D C The purpose of this modification is to upgrade the existing fire barrier enclosures for the 125 VDC distribution panels (1D624, 1D644, 2D624 and 2D644) to meet a one hour fire rating.
S~IElhBY:
No. FSAR Table 8.3-21, Failure Mode and Effect Analysis (FMEA) of the 125 VDC system was reviewed. The failure mechanisms discussed are loose connections, insulation failure, loose parts, dirt and failure of circuit breakers. The design and installation of the fire barrier enclosures do not increase the probability of an accident or the consequences of an accident or malfunction of equipment.
II. No. The modification upgrades the fire barrier enclosures surrounding the 125VDC distribution panels to meet as a minimum a one hour fire rating. This will ensure the operability of the 125 VDC distribution panels for Safe Shutdown Path 3 in the event of an Appendix R fire as evaluated in the Fire Protection Review Report.
The accident analyses in Chapter 15 and the FMEA in Chapter 8 of the FSAR were reviewed and this modification does not create a possibility of an accident or malfunction of a different type.
III. No. The upgrading of the fire barrier enclosures for the 125 VDC distribution panels to the required one hour fire rating does not effect the operability or safety function of the 124 VDC distribution system and it ensures the availability of the 125 VDC distribution panels during an Appendix R fire. Therefore, this modification ensures the margin of safety the 125 VDC distribution panels provide, as discussed in Technical Specification Bases 3/4.8.2 and 3/4.8.3.93-011 C: DCP 92-9020 Rev. 1, Unit Common DEC IP 0 C
- 1. Replace the 30 percent under voltage relays (ITE Type 27D) on MCC OB565 bus.
- 2. Add degraded voltage relays (ABB Type 27N) on MCCC OB565 bus.
- 3. Add undervoltage alarm circuit for MCC OB565.
- 4. Add a testability scheme to the MCC OB565 bus.
SUISRlhBY:
No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a new undervoltage protection scheme to the Diesel Generator E MCC 08565 bus does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine the addition of the new undervoltage protection scheme to the Diesel Generator E MCC 08565 bus does not create the possibility of an accident or malfunction of a different type.
III~ No. Based upon a review of the Technical Specifications, NUREG-0931 and NUREG-1042, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a new undervoltage protection scheme for the Diesel Generator E MCC 08565 bus, establishes the margin of safety for the scheme which is consistent with margin of safety for the Class 1E 4.16KV.
QE~~: 93-012 C C: DCP 92-9018Z, Unit 1 DE C 0 C The proposed action is to add redundancy to each Unit I 4.16kV bus degraded voltage timer reselect logic and to delete the Startup Transformer T10 and T20 voltage boost scheme.
I. No. The addition of redundancy to each Unit 1 4.16 kV bus degraded voltage timer reselect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme and the addition of two ESS transformers and the re-alignment of the offsite feeders to the Class 1E 4.16 kV buses in both units has assured that the Plant Electrical Auxiliary System can supply adequate voltage to all of the Class 1E equipment required to mitigate the consequences of an accident.
Thus, the change does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis or NUREG-0776. Thus, the proposed action does not increase the probability of occurrence of an accident.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of redundancy to each Unit 1 4.16 kV bus degraded voltage timer reselect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Based upon a review of the Technical Specification NUREG 0931 and NUREG 1042, the FSAR and NUREG 0776, Safety Evaluation Report for SSES, including Supplements 1 through 7, the addition of redundancy to each Unit 1 4.16 kV bus degraded voltage timer reselect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme does not reduce the margin of safety as defined in the basis for any Technical Specification.
C OSS E E C: DCP92-9038Z, Unit1 C G The proposed action is to add a redundant LOCA start signal to the Diesel Generator D PRIMARY Auto start logic.
I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a redundant LOCA start signal into the Diesel Generator D PRIMARY Auto Start logic does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. A random single failure in the Diesel Generator D PRIMARY Auto Start logic with a LOCA or a random single failure during a LOCNLOOP has the same impact as a malfunction with the existing Division I RHR LOCA logic or the Diesel Generator D PRIMARY Auto Start logic.
III. No. Based upon a review of the Technical Specification, the FSAR and NUREG 0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a redundant LOCA start signal into the Diesel Generator D PRIMARY Auto Start logic does not reduce the margin of safety as defined in the basis for any Technical Specification.
S~Q,: 93-014 E C: DCP 92-9036Z, Unit1 The proposed action is to add redundant LOCA start signal to the Diesel Generator C PRIMARY Auto start logic.
I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a redundant LOCA start signal into the Diesel Generator C PRIMARY Auto Start logic does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. A random single failure in the Diesel Generator C PRIMARY Auto Start logic with a LOCA or a random single failure during LOCNLOOP has the same impact as a malfunction with the existing Division I RHR LOCA logic or the Diesel Generator C PRIMARY Auto Start logic.
III. No. Based upon a review of the Technical Specification, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a redundant LOCA start signal into the Diesel Generator C PRIMARY Auto Start logic does not reduce the margin of safety as defined in the basis for any Technical Specification.
5gBJg): 93-015 C 0 E C: DCP92-9083, Unit18 2 E C The action is to replace the existing Standby Gas Treatment System (SGTS) fusible links which have a melting point rating of 160'F with new links rated for 212'F or above.
SUlgg~g:
I. No. The design, material, and construction standard of the replacement fusible links is the same as the original links. The only change is the material composition of the link's solder which determines its melting point temperature. The basic functions and performance of plant systems are not changed.
FSAR Sections 9.5.1 and 6.5.1, 6.2, 6.5 Appendix 9A 15.6, and 15.7 were reviewed.
II. No. The accidents of sections 6.2, 6.5, Appendix 9A, 15.6 and 15.7 were reviewed.
The proposed action does not create the possibility of an accident or malfunction of a different type than evaluated in the FSAR since a one-for-one replacement of equipment is being made with equipment that has the same failure mechanisms.
III. No. FSAR sections 6.2, 6.5, 9.5 Appendix 9A and Chapter 15 were reviewed for design basis requirements related to the actuation temperature rating of the fusible links that involve margin of safety. No margin of safety are defined in these documents which are affected by the scope of this modification since the Technical Specifications do not specifically address the actuation temperatures of the fusible links. Therefore, the replacement fusible links with an actuation temperature of 285'F meet this criteria for margin of safety.
~Egg): 93-016 C C: DCP 92-9016, Unit 1 E C C E:
This modification will add one new permissive to the logic which controls the automatic transfer of the HPCI pump suction source on high suppression pool level.
I. No. The evaluation considered failure of the F042 valve to open due to malfunctions of the new auxiliary relay as well as the limit switch which energizes the relay. In summary, calculations have determined that the increase in failure rate due to the addition of the relay falls within the error band of the predicted failure rate of the valves without the additional relay. FSAR sections 6.3, 7.3, 15.2.4 were reviewed.
II. No. The new auxiliary relay is identical to existing relays presently installed in the DC Control Center ID264. It is wired in accordance with the separation requirements of PP8L Specification E-1012. The change of the internal panel wiring in panel IC620 is in accordance with existing approved installation and termination procedures.
III. No. This modification adds a relay which will still permit the automatic transfer of pump suction when HPCI injection to the reactor is required. This modification does not change the originally designed ECCS function of the automatic HPCI suction transfer. Therefore, the addition of an auxiliary relay to provide the proposed revision to the HPCI suction transfer logic does not reduce the margin of safety as defined in the basis for any Technical Specification.
2 l I
S~E59: 93-017 C 0 EFE E CE: NCR90-0398, Unit1,28 Common E C Repair & Rework of Smoke Detector Tubing and Supports.
I. No. FSAR Section 9.5 describes the smoke detection feature of the fire protection system and requires operational testing to be performed. Since the smoke tubing and supports shall be inspected and modified to preclude any safety impact concerns, the proposed actions do not increase the probability or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in FSAR Chapters 6 and 15.
II. No. The proposed action, inspection and modification of tubing lines, will qualify the tubing and supports to preclude any safety impact concerns. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than evaluated previously in the FSAR.
III. No. Technical Specification Section 8 3/4 7.6 addresses the bases for fire suppression systems, but does not address the testing mechanisms for these systems. Since the tubing and supports shall be qualified with respect to safety impact criteria, the proposed actions do not reduce the margin of safety as defined in the basis of the Technical Specifications.
~E~: 93-018 C OSS E C: DCP92-9009, Unit1 This modification adds a 30 second time delay to the feedwater heater string high-high level isolation circuit to prevent the isolation of one or more feedwater heater strings as a result of a reactor scram.
I. No. The installation of an Agastat 30 second time delay relay into each of the three feedwater heater string isolation logics does not adversely affect safety related systems or equipment.
This modification does not interfere with the logic, control or operation of any safety-related plant system or components, nor does it adversely affect the ability to maintain core reactivity control during anticipated loss of feedwater transients. This modification does not increase the probability of occurrence or the consequences of the accidents described in FSAR Chapters 6 and 15.
II. No. The proposed action does not prevent the Feedwater Heater System nor its associated control logic from performing its design functions. Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15. This modification, which only adds a 30 second time delay on string isolation, will not cause a different type of accident scenario than those previously evaluated.
III. No. The bases for related or potentially influenced Technical Specifications, including 3/4.1.1, 3/4.2.3, 3/4.3.4, 3/4.3.9, 3/4.4.1 and 3/4.4.6, have been reviewed and do not discuss the feedwater heater string isolation. In addition, the 30 second time delay in each of the three feedwater heater isolation logics will not interfere with, nor adversely affect, the logic, control, or operation of any safety or Tech. Spec. related plant system or component. Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specification.
'I S~~O.: 93-019 E C: DCP 93-9003Z Rev. 1, Unit Common E C P FC The proposed action for Diesel Generator A is to add a transient suppressor on the contactor coils of the Fuel Oil Transfer Pump OP514A, the Air compressors OK507A1 and OK507A2 and a varistor on the 3FOT relay coil.
I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine if the proposed action has an effect on the spectrum of postulated initiating events for which transients or anticipated operational occurrences and accident conditions were analyzed.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. For the 3FOT relay, the addition of the varistor is, as indicated, a simple discrete component with high reliability. Thus, the addition of a transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514A, Air Compressors OK507A1 and OK505A2 and the addition of a varistor on the 3FOT relay coil does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514A, the Air Compressors OK507A1 and OK507A2 and the addition of a varistor on the 3FOT relay coil does not create a possibility for an accident or malfunction of a different type.
III. No. Based upon a review of the Technical Specification, NUREG-0931 and NUREG-1042, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514A, the Air Compressors OK507A1 and OK507A2 and the addition of a varistor on the 3FOT relay coil does not reduce the margin of safety as defined in the basis for any Technical Specification.
t5 C C: DCP 93-9015Z, Unit Common C 0 C The proposed action for Diesel Generator D is to add a transient suppressor on the contactor coils of the Fuel Oil Transfer Pump OP514D, the Air compressors OK507D1 and OK507D2 and a varistor on the 3FOT relay coil.
I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine if the proposed action has an effect on the spectrum of postulated initiating events for which transients or anticipated operational occurrences and accident conditions were analyzed.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. For the 3FOT relay, the addition of the varistor is, as indicated, a simple discrete component with high reliability. Thus, the addition of a.transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514D, the Air Compressors OK507D1 and OK507D2 and the addition of a varistor on the 3FOT relay coil does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514D, the Air Compressors OK507D1 and OK507D2 and the addition of a varistor on the 3FOT relay coil does not create a possibility for an accident or malfunction of a different type.
III. No. Based upon a review of the Technical Specification, NUREG-0931 and NUREG-1042, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a transient suppressor to the contactor coils of the Fuel Oil Transfer Pump OP514D, the Air Compressors OK507D1 and OK507D2 and the addition of a varistor on the 3FOT relay coil does not reduce the margin of safety as defined in the basis for any Technical Specification.
~ I 5KBJg): 93-021 C C: DCP 93-9017 Unit Common ESC 0 0 C This modification will install a Cheng Rotation Vane (CRV) in the discharge head of the River Water Makeup (RWMU) pump OP503B to reduce pump maintenance by suppressing destructive flow abnormalities.
I. No. The River Water Makeup System, in particular pump OP503B, serves no safety related functions. The only connection RWMU pump OP503B has with a safety k ski i>>
related system or component is as one source of makeup water supply to the spray fg igiig iliig pond. FSAR Section 9.2.7 "Ultimate Heat Sink", subsection 9.2.7.1, states "The gi the spray pond for at least 30 days...". FSAR Chapters 6 and 15 have been reviewed in relation to this modification, and based on the above, the proposed actions do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification will only enhance the performance of the RWMU pump OP503B by eliminating or reducing pump discharge flow abnormalities, thereby minimizing pump maintenance on the shaft bearings.
III. No. The bases for all potentially affected Technical Specifications, including 3/4.4.9, 3/4.7.1, 3/4.7.2 and in particular 3/4.7.3, have been reviewed, and do not address the River Water Makeup System nor the RWMU pumps. This modification shall enhance the performance and reliability of the RWMU pump by preventing the fluid separation at the discharge elbow that leads to turbulent flow and excessive stress on the shaft bearings. Consequently, the addition of a CRV to the discharge head of RWMU pump OP503B will not reduce any margin of safety as defined in the basis for any Technical Specification.
DCP 92-9060 Unit 1 DESC 0 0 C This modification will install an access door in the Zone I Reactor Building HVAC Supply Duct to the Unit I Primary Containment downstream of damper HD-17651 on Elev. 749, Area 29, of the Unit I Reactor Bldg.
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I. No. The new access door will be constructed and installed using existing approved PP&L engineering design details and procedures. FSAR Chapters 6 and 15 have been reviewed in relation to this modification and based on the above, the proposed actions do not increase the probability of occurrence of any accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. The duct modification will comply with the original construction codes and regulations, including existing PP8 L HVAC access door design details.
Furthermore, the location of the new access door is isolated from the Reactor Building Ventilation System by existing damper HD-17651 during normal operating and accident conditions. Therefore, the Reactor Building Ventilation System will continue to operate in accordance with its original design bases, and the proposed action does not create a possibility for an accident or malfunction of a different type than evaluated previously in the FSAR.
III. No. A review of Technical Specification 3/4.6.5, secondary containment, reveals no requirements or descriptions which would be affected by the proposed modification.
This modification, due to the design and construction requirements of the existing PP8L documents, does not adversely affect the duct, the Reactor Building Ventilation System, or the duct's structural integrity.
S~E~: 93-023 C OSS E C: SCP J92-1057 Unit1 C
Increase Unit 1 Reactor Water Cleanup (RWCU) Steam Leak Detection - Ambient Temperature Setpoints.
I. No. The FSAR does not analyze the size of the small leak on which the temperature setpoints are based. Other accidents which result in coolant leakage outside containment are analyzed in FSAR Section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System Piping Break Outside Containment). Both of these are assumed accidents with no causes identified. The analysis in Section 15.6 4 is the enveloping evaluation for pipe breaks outside containment. The 25 gpm design basis leak rate is well below the leakage corresponding to a catastrophic pipe failure for the RWCU System piping and does not significantly increase the risk of a break.
II. No. The proposed setpoint change does not affect any system other than the RWCU Steam Leak Detection Ambient Temperature switches The proposed setpoint change does not create a possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2 "Isolation Actuation Instrumentation", but the basis section does not discuss setpoint basis with respect to leakage rate or process conditions. The basis does however state "the setpoints ... are established at a level away from the normal operating range to prevent inadvertent actuation of the system involved." The proposed setpoints meet this basis. Temperature measurement is not discussed in the basis for Technical Specification 3.4.3 "Reactor Coolant System Leakage."
~SE NO.: 93-024 C C: SCP J92-1058 Unit 2 C G Increase Unit 2 Reactor Water Cleanup (RWCU) Steam Leak Detection - Ambient Temperature Setpoints.
I. No. The FSAR does not analyze the size of the small leak on which the temperature setpoints are based. Other accidents which result in coolant leakage outside containment are analyzed in FSAR Section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System Piping Break Outside Containment). Both of these are assumed accidents with no causes identified. The analysis in Section 15.6.4 is the enveloping evaluation for pipe breaks outside containment. The 25 gpm design basis leak rate is well below the leakage corresponding to a catastrophic pipe failure for the RWCU System piping and does not significantly increase the risk of a break.
II. No. The proposed setpoint change does not affect any system other than the RWCU Steam Leak Detection Ambient Temperature switches. The proposed setpoint change does not create a possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2 "Isolation Actuation Instrumentation", but the basis section does not discuss setpoint basis with respect to leakage rate or process conditions. The basis does however state "the setpoints ... are established at a level away from the normal operating range to prevent inadvertent actuation of the system involved." The proposed setpoints meet this basis. Temperature measurement is not discussed in the basis for Technical Specification 3.4.3 "Reactor Coolant System Leakage."
SKELbLQ: 93-025 C 0 S E E E C: SCP J92-1059, Unit1 Increase Unit 1 Reactor Water Cleanup (RWCU) Steam Leak Detection - Differential Temperature Setpoints.
No. The FSAR does not analyze the size of the small leak on which the temperature setpoints are based. Other accidents which result in coolant leakage outside containment are analyzed in FSAR Section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System Piping Break Outside Containment). Both of these are assumed accidents with no causes identified. The analysis in Section 15.6.4 is the enveloping evaluation for pipe breaks outside containment. The 25 gpm design basis leak rate is well below the leakage corresponding to a catastrophic pipe failure for the RWCU System piping and does not significantly increase the risk of a break.
II. No. The proposed setpoint change does not affect any system other than the RWCU Steam Leak Detection Differential Temperature switches. The proposed setpoint change does not create a possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2 "Isolation Actuation Instrumentation", but the basis section does not discuss setpoint basis with respect to leakage rate or process conditions. The basis does however state "the setpoints ... are established at a level away from the normal operating range to prevent inadvertent actuation of the system involved." The proposed setpoints meet this basis. Temperature measurement is not discussed in the basis for Technical Specification 3.4.3 "Reactor Coolant System Leakage."
S~~: 93-026 E CE: SCP J92-1060, Unit 2 C
Increase Unit 2 Reactor Water Cleanup (RWCU) Steam Leak Detection - Differential Temperature Setpoints.
No. The FSAR does not analyze the size of the small leak on which the temperature setpoints are based. Other accidents which result in coolant leakage outside containment are analyzed in FSAR Section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System Piping Break Outside Containment). Both of these are assumed accidents with no causes identified. The analysis in Section 15.6.4 is the enveloping evaluation for pipe breaks outside containment. The 25 gpm design basis leak rate is well below the leakage corresponding to a catastrophic pipe failure for the RWCU System piping and does not significantly increase the risk of a break.
II. No. The proposed setpoint change does not affect any system other than the RWCU Steam Leak Detection Differential Temperature switches. The proposed setpoint change does not create a possibility for an accident or malfunction of a different type than any previously evaluated.
III. No. The temperature switches and setpoints are listed in Technical Specification Section 3.3.2 "Isolation Actuation Instrumentation", but the basis section does not discuss setpoint basis with respect to leakage rate or process conditions. The basis does however state "the setpoints ... are established at a level away from the normal operating range to prevent inadvertent actuation of the system involved." The proposed setpoints meet this basis. Temperature measurement is not discussed in the basis for Technical Specification 3.4.3 "Reactor Coolant System Leakage."
Qgg Jg): 93-027 C S E E E C: NL93-005, Unit1 DEC 0C Temporary installation of three (3) delta-pressure transmitters will be attached to the pressure taps across the feedwater flow elements (FE-C32-INOOIA/8/C) in parallel to existing flow transmitters (FT-C32-IN002A/B/C) to provide an accurate indication of the delta-pressure across each flow element.
I. No. The temporary transmitters will be attached hydraulically to the plant during the performance of TP-145-016 and will operate at approximately 1100 psig. These transmitters will have a separate non-Q power supply and the output will be sent to a local data logger during performance of TP-145-011. The signal will not affect any control logic or control room indication. Other than its use during TP-145-011, the transmitters will not be used for any other purpose.
II. No. The FSAR has been reviewed. In particular section 7.7.1.4.1.2 states that feedwater control is a power generation system and is classiTied as not related to safety. Section 7.7.2.4.2 states that feedwater control is not required for safe shutdown of the plant, nor is it required during or after accident conditions. Based upon this review, it is concluded that any failure of the permanent transmitters is bounded by existing safety analyses.
III. No. The permanent feedwater flow transmitters will continue to provide all of the required functions, including input to the core thermal power heat balance throughout the period of time that this bypass is installed: The Technical Specifications and bases have been reviewed. It has been determined that the temporary installation of these feedwater flow transmitters does not reduce the margin of safety as defined in the bases for any Technical Specification.
Qgf~Q,: 93-028 C OSS E C: DCP 92-9044Z, Unit Common ESC C G Add a Class 1E 4.16 kV bus undervoltage signal into the Emergency Service Water (ESW) pump automatic start circuitry which resets the timer and restarts the pump following a trip caused by the Class 1E 4.16 kV bus undervoltage protection when the diesel generator is operating.
Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed. The inadvertent operation or malfunction of the 20 percent undervoltage relay could cause an inadvertent operation of the 27AX relay. However, the inadvertent operation of the 27AX relay under these conditions would have no impact. Thus, the addition of a Class 1E 4.16KV bus undervoltage signal into the ESW pump automatic start circuitry does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
II. No. A random single failure of the 27AX relay during a LOOP represents a random single failure for the corresponding Channel or Division of ECCS equipment. The impact of the random single failure of the 27AX relay during a LOOP is the same as the existing impact. Thus, the addition of a Class 1E 4.16 kV bus undervoltage signal into the ESW pump automatic start circuitry does not create a possibility for an accident or malfunction of a different type.
III. No. The bases for operability of the Class 1E 4.16 kV bus undervoltage protection scheme is to ensure that the Emergency Core Cooling System Actuation Instrumentation can provide the initiating actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. The proposed action does not affect the cooling capacity or redundancy of the Emergency Service Water System as governed by Technical Specification 3I4.7.1.2 and does not affect the initiating actions of the Emergency Core Cooling System Actuation Instrumentation. Thus, the addition of a Class 1E 4.16 kV bus undervoltage signal into the ESW pump automatic start circuitry does not reduce the margin of safety as defined in the basis for any Technical Specification.
5gBJg): 93-029 C SS E E E C: NL93-006, Unit1 The proposed action is to prevent automatic restart of the Condensate Pump 1D after a 13.8 KV Load Shed accomplished by opening a Sliding Link on each side of contact 62XI-10104 in the automatic closing control circuit for the Condensate Pump 1D Circuit Breaker 52-IAI0209.
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I. No. The opening of sliding links on each side of the 62X1-10104 contact in the automatic closing circuit for the Condensate Pump 1D does not increase the probability of occurrence of a malfunction of equipment because this action removes a component from the circuit which incrementally decreases the probability of malfunction. Chapter 6 and 15 of the FSAR and NUREG-0776 were reviewed.
II. No. Preventing automatic restart of the Condensate Pump 1D does not create the possibility of an initiating event which creates the possibility of an accident of a different type than identified in FSAR Chapter 6 and 15 and NUREG-0776.
The opening of sliding links on each side of the 62X1-10104 contact in the automatic closing circuit for the Condensate Pump 1D does not create the possibility of a malfunction of a different type.
III. No. The bases for operability of the Emergency Service Water System is to ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions as governed by Technical Specification 3/4.7.1.2.
Four ESW Pumps are available for the Main Steam Line Break and two ESW Pumps are available for the Recirculation Discharge Line Break. These described scenarios include the loss of one source of off-site power which places the plant in an LCO. The LCO is considered the single failure. One ESW Pump from each loop is sufficient to provide adequate cooling. Therefore, the proposed action does not reduce the margin of safety as defined in the bases for any Technical Specification.
Qgf~(~: 93-030 C OSS RE E C: NL93-007, Unit2 The proposed action is to prevent automatic restart of the Condensate Pump 2D after a 13.8 KV Load Shed accomplished by opening a Sliding Link on each side of contact 62XI-10104 in the automatic closing control circuit for the Condensate Pump 1D Circuit Breaker 52-2AI0209.
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I. No. The opening of sliding links on each side of the 62X1-10104 contact in the automatic closing circuit for the Condensate Pump 2D does not increase the probability of occurrence of a malfunction of equipment because this action removes a component from the circuit which incrementally decreases the probability of malfunction. Chapter 6 and 15 of the FSAR and NUREG-0776 were reviewed.
II. No. Preventing automatic restart of the Condensate Pump 2D does not create the possibility of an initiating event which creates the possibility of an accident of a different type than identified in FSAR Chapter 6 and 15 and NUREG-0776.
The opening of sliding links on each side of the 62X1-10104 contact in the automatic closing circuit for the Condensate Pump 2D does not create the possibility of a malfunction of a different type.
III. No. The bases for operability of the Emergency Service Water System is to ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions as governed by Technical Specification 3/4.7.1.2.
Four ESW Pumps are available for the Main Steam Line Break and two ESW Pumps are available for the Recirculation Discharge Line Break. One ESW Pump from each loop is sufficient to provide adequate cooling.,Therefore, the proposed action does not reduce the margin of safety as defined in the bases for any Technical Specification.
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S~~O: 93-031 C SS C: DCP 92-9045, Unit 1 The proposed action will enhance the existing design of the Cooling Tower Basin Level Instrumentation via a modification to provide protection for exposed instrument air lines and associated valving which supply the level instrumentation in panel IC557 near the cooling tower basin.
I. No. The Circulating Water System has no safety related function and is designed to remove the latent heat from the main condenser and sensible heat from the Service Water System, and to dissipate both in a hyperbolic natural draft cooling tower. This modification does not increase the probability of occurrence or the consequences of an accidents described in FSAR Chapter 6 and 15.
II. No. The modiTication does involve changes in system operation by providing for manual operator action in lieu of an automatic trip function but this difference has no impact on plant nuclear safety. Therefore, the proposed action does not create a possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The Circulating Water System is not specifically addressed in the Technical Specifications and, therefore, does not have formal limiting conditions for operation.
Therefore, the implementation of this change will not reduce the margin of safety as defined in the bases of the Technical Specifications.
~~iEE.50: 93-032 C OSS C: DCP 92-9014, Unit 1 C C GE:
The modification will install a new Reactor Cavity Upper Seal Ring, one that has been designed, fabricated and tested to provide a water-tight seal for the as-built seismic gap.
No. A review of the FSAR with particular attention to Chapters 6 8 15 has been performed and no mention of the Reactor Cavity Upper Seal Ring was identified.
The new seal will be of a very similar design to the existing seal, and meet all the design basis requirements.
Based on the above, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. The new seal will be virtually identical to the existing one in design, with the exception that it will be qualiTied to be leak-tight for the as-built seismic gaps.
Since this modification does not deviate from the original design basis, the modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The bases for all potentially affected Unit 1 Technical Specifications, including 3/4.6.5.3 "Standby Gas Treatment System", 3/4.9.8 "Water Level-Reactor Vessel" and 3/4.9.9 "Water Level -'pent Fuel Storage Pool", have been reviewed, and do not address the Reactor, Cavity Upper Seal Ring.
The wedge design shall guarantee no drain down of the water level in the Fuel Pool or Reactor Cavity during refueling operations. Consequently, the new Reactor Cavity Upper Seal Ring will not reduce any margin of safety as defined in the basis for any Technical Specification.
gijj.59: 93-033 C E C: NL 91-005, Unit Common SC 0 C G The action is to change the process of removing non-condensable gases from the reactor vessel before the reactor vessel head is removed.
No. If the reactor vessel is completely filled with water (level 343") and CRD flow is greater than 75 gpm, or either valves HV-141(241)-00I, or HV-141(241)-002 is closed, the vessel could pressurize to 98 psig which would isolate RHR shutdown cooling. The pressurization could be controlled by Reactor Water Cleanup (RWCU), which is to be operable and operated in the letdown mode. If RWCU were to fail, two SRVs would be functional to prevent any further pressurization.
The SRV will remain functional if water is vented through them. These two means of pressure control (allowing for a single failure), ensure that the design pressure of 1250 psig (FSAR Section 5.2.2.1.3) will not be reached.
II. No. A review of FSAR Chapters 4 and 5 indicates that the possible malfunctions that could occur due to the proposed action have been previously addressed and are still applicable. The method of currently purging the reactor vessel of non-condensable fission products is by the Standby Gas Treatment System. The proposed action would remove the fission products in the same manner, thus there is no different accident scenario or malfunctions that will cause a radioactive release.
III. No. Technical Specification 3/4.4.6 requires the reactor vessel to be maintained within certain pressure and temperature boundaries. With two functional SRVs and RWCU operable, the reactor vessel temperature and pressure will remain to the right of curve A which is shown in Figure 3.4.6.1-1 of the Technical Specifications. Thus, no reduction in the margin of safety will occur.
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SE~Q.: 93-034 C S RE C: DCP 92-9046, Unit 2 C
The proposed action will enhance the existing design of the Cooling Tower Basin Level Instrumentation via a modification to provide protection for exposed instrument air lines and associated valving which supply the level instrumentation in panel 2C557 near the cooling tower basin.
I. No. The Circulating Water System has no safety related function and is designed to remove the latent heat from the main condenser and sensible heat from the Service Water System, and to dissipate both in a hyperbolic natural draft cooling tower. This modification does not increase the probability of occurrence or the consequences of the accidents described in FSAR Chapters 6 and 15.
II. i,, No. The modification does involve changes in system operation by providing for manual operator action in lieu of an automatic trip function but this difference has no impact on plant nuclear safety. Therefore, the proposed action does not create a possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The Circulating Water System is not specifically addressed in the Technical Specifications and, therefore, does not have formal limiting conditions for operation.
Therefore, the implementation of this change will not reduce the margin of safety as defined in the bases of the Technical Specifications.
s4k SEEK): 93-035 C S E E C: NL93-004, Unit1 This Setpoint Change Package will adjust the Main Transformer tap settings from 230.0 KV to 235.75 KV.
No. Two accidents described in the FSAR involve the Main and Aux Transformer that are being affected by this Setpoint Change Package. FSAR Chapter 15 Section 2 identifies accident scenarios that could result in increasing reactor pressure.
These scenarios include Section 15.2.2 Generator Load Reject, and Section 15.2.6 Loss of AC Power. The tap setting changes do not increase the likelihood of the these accidents occurring nor will the tap setting changes increase the consequences of these accidents.
II. No. The proposed setpoint change does not change the intended function of the equipment involved, nor does it add any new equipment that could result in an accident or malfunction of a different type than any previously evaluated in the FSAR.
III. No. The tap setting changes to the Generator Step Up, Aux Transformers, and Aux Potential Transformer do not have any effect on the safety limits, limiting safety system settings, nor any limiting condition of operation. Therefore, the changes will not reduce the margin of safety as defined in any Technical Specification.
Qgg 59 '3-036 C E C DCP 90-9030C, E Unit1 C 0 C This proposed activity will make the necessary modifications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers located on selected piping systems.
I. No. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to be modified accordingly.
All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the Design Specifications and FSAR have been addressed in the reanalysis effort. As in the existing piping analysis, the applicable code design limits have been met to ensure piping integrity and system function; FSAR Section 15.0.3.1, 15.0.3.5.
No. The accident events/causes in FSAR Chapter 15.0 that are applicable to this proposed action involve equipment malfunctions/failures and pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations.
Therefore, the proposed change will not increase the probability of an accident and a new unanalyzed type of malfunction or accident will not be created by the removal or replacement of snubbers.
III. No. The proposed action reduces the overall snubber population for Unit 1. In case of snubber size change the same type of snubber as currently installed will be used where necessary - i.e., no new type of snubber will be introduced by these modifications. The bases, as defined in the Technical Specifications, deal with snubber test requirements and are not directly affected by this proposed action. Removing or replacing snubbers will not reduce the margin of safety as defined by the Technical Specification noted.
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)>ATE'Jg)93-037 C C: 90-9099, Unit Common ESC 0 C Reconfigure the Startup Transformer, T10, Tap Changer Alarm SUJHHlhBY'.
No. The proposed design eliminates a false alarm and enhances the performance of the startup tap changer alarm circuitry. There is a remote potential of tap changer failure due to malfunction of the new equipment and electrical cascading. However, the probability of such failure is very low and considered insignificant to the availability and performance of the tap changer.
II. No. The proposed action does not change the operation of the tap changer. Installation of this modification will be consistent with existing standards and engineering specification guidelines. This consistency with the existing standards and the unchanged function of the involved circuits, ensures that the proposed action will not create a possibility for an accident or malfunction of a different type than any evaluated in FSAR.
III. No. The proposed modification will not change the operation of the startup transformer, it simply enhances its performance by eliminating the tap changer nuisance alarm.
The Technical Specifications section 3/4.8 were reviewed. The Technical Specification Bases are not affected by the proposed modification. Therefore this modification will not reduce the plants margin of safety as defined in the Bases of the Technical specification.
SEPJJO: 93-038 E E C 90-9100, Unit Common DESC IP 0 C Reconfigure the Startup Transformer, T20, Tap Changer Alarm SSJEMhHY'.
No. The proposed design eliminates a false alarm and enhances the performance of the startup tap changer alarm circuitry. There is a remote potential of tap changer failure due to malfunction of the new equipment and electrical cascading. However, the probability of such failure is very low and considered insignificant to the availability and performance of the tap changer.
II. No. The proposed action does not change the operation of the tap changer.
Installation of this modification will be consistent with existing standards and engineering specification guidelines. This consistency with the existing standards and the unchanged function of the involved circuits, ensures that the proposed action will not create a possibility for an accident or malfunction of a different type than any evaluated in FSAR.
III. No. The proposed modification will not change the operation of the startup transformer, it simply enhances its performance by eliminating the tap changer nuisance alarm.
The Technical Specifications section 3/4.8 were reviewed. The Technical Specification Bases are not affected by the proposed modification. Therefore this modification will not reduce the plants margin of safety as defined in the Bases of the Technical SpeciTication.
SE~jO '3-039 C C: DCP 92-9088, Unit 1, 2,Common DESC Removal of ITT Hydramotor Actuator Auxiliary Switch. In addition, the modification will provide justification for the permanent removal of any spare unused auxiliary switch assembly and linkage from any series NH-90 ITT actuator.
I. No. The spare auxiliary switch has no wiring terminated directly to it and, as such, no connection to any circuitry results. The auxiliary switch linkage will be disconnected and removed from the main actuator rod assembly.
This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15.
II. No. Since seismic and environmental qualification of the actuators is maintained, this modification does not degrade their ability to perform their design basis function.
Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. ITT actuators are installed in the Standby Gas Treatment System and the Control Room Emergency Outside Air Supply System. These are Tech. Spec. related systems and are addressed in Tech. Specs. 3/4.6.5.3 and 3/4.7.2, respectively. The proposed modification does not adversely affect the logic, control, or operation of any Tech. Spec. related system. None of the parameters that are the bases for the Technical Specifications will be adversely impacted, therefore, no affect on any Tech.
Spec. margin of safety occurs.
QRB hK)93-040 C C: DCP 93-9012, Unit1 C 0 C E'odify Seat Retaining Ring On Valves on HV-141F032A8B in the Feedwater System.
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No. The design, and construction standards are not changed. The valve stroke time and leakage characteristics are unaffected by the modification. Therefore, the analysis described in FSAR Section 6.2.4 and Table 6.2-12 is still valid. The modification does not affect the design or performance of any equipment which is important to safety.
II. No. The modification creates the possibility for interferences that prevent the valves from opening or closing due to the increased disc outside diameter (change from recessed screws to hex head screws). However, adequate clearance for both installation and operation were confirmed by the vendors review of the shop fabrication drawings and will be reconfirmed by exercising the modified valves and observing internal clearances as part of the requirements of the DCP. Therefore, the risk of this type of failure is considered negligible.
No. Tech Spec basis 3/4.6.1.1 ensures that the release of radioactive materials from containment is restricted to the leak paths and rates assumed in the accident analyses.
Tech Spec basis 3/4.6.1.2 ensures that the total containment leakage value will not exceed the value assumed in the accident analyses at peak accident pressure. Lastly, Tech Spec basis 3/4.6.3 ensures that the containment atmosphere will be isolated from the outside environment in the event of radioactive release or containment pressurization.
S~5Q;.93-041 C E E C: NL-93-011, Unit1 C C G Leak seal repair and mechanical gagging of Main Steam to Steam Jet Air Ejector (SJAE) Control Valve HV1 07018 until U7-RIO or any outage of sufficient duration to remove.
MERCURY'.
No. As described in FSAR Section 10.4.2.5, HV10701B is interlocked with HV10721 to close upon initiation of the SJAE shutdown signal (i.e., closure of HV10721). With this Bypass installed, HV10701B will not close upon a system shutdown signal. The fact that HV10701B will not close upon closure of HV10721 does not impact positive isolation of the system.
In the highly improbable event that HV10702 (a motor operated valve) spuriously fails to the closed position the Gaseous Radwaste Treatment System will be automatically isolated. System operating procedures will be changed to reflect that HV10701B will remain open upon SJAE system shutdown.
Based upon the above discussion, the probability and/or consequence of an accident or malfunction important to safety is not increased.
II. No. FSAR Section 15.7.1 analyzes failures of, or leaks in the Gaseous Radwaste System.
The analysis therein looks at failure of the Air Ejector System.
The proposed action, leak sealing and gagging HV107018 in the open position, will in no way alter or affect the assumptions made in the FSAR analysis. This action will in all cases stay within the bounds of the FSAR analysis. No possibility exists for an accident or malfunction different from that described in the FSAR.
III. No. Technical Specifications 3.11.2.4 and 3.11.2.6 address the requirements for the Gaseous Radwaste Treatment System. T.S. 3.11.2.4 dictates the circumstances under which the offgas system must be operable, while T.S. 3.11.2.6 specifies Hydrogen and Oxygen limits for operation. After installing this Bypass Offgas System performance will be unaffected and in compliance with T.S. 3.11.2.4 and T.S. 3.11.2.6.
As defined in the Technical Specifications the margin of safety is not reduced by this action.
0 QRR,5g '3-042 C E E CE: NL-93-012, Unit2 E C Ee The presence of an oil sight glass on the gear case oil drain piping of Containment Radiation Monitor (CRM) Blower 2V22OB2 was identified. This does not conform with the design or qualified configuration and, as such, constitutes an unauthorized modification. This safety evaluation is an examination of that modified configuration.
I. No. The CRM system does not perform a safety related function, nor does it initiate any automatic function. Its purpose is indication and alarm. As such, the failure of this function does not affect the probability of occurrence of any incident, nor of any malfunction of any equipment important to safety caused directly.
II. No. The CRM System provides indication and alarm only, and does not perform any safety function. The physical failure of the sight glass could result in failure of the blower due to the lack of oil. The oil leakage is within the CRM panel and, as such, could affect CRM system equipment. As previously stated this equipment is not required and therefore, this does not constitute a malfunction that need to be analyzed in the FSAR.
III. No. The CRM system normal function is unaffected by this unauthorized modification and, as such, the unauthorized modification does not impact the design bases for Operational leakage as described in the Technical Specification Bases 3/4.4.3.2.
~E~: 93-043 C OSS E E C: DCP92-9020Rev.2, UnitCommon
- 1. Replace the 30 percent undervoltage relays (ITE Type 27D) on MCC 08565 bus
- 2. Add degraded voltage relays (ABB Type 27N) on MCC OB565 bus
- 3. Add undervoltage alarm circuit for MCC 08565 bus
- 4. Add a testability scheme to the MCC 08565 bus I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine the addition of a new undervoltage protection scheme to the Diesel Generator E MCC OB565 bus does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine the addition of the new undervoltage protection scheme to the Diesel Generator E MCC OBS65 bus does not create the possibility for an accident or malfunction of a different type.
III. No. Based upon a review of the Technical Specifications NUREG-0931 and NUREG-1042, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a new undervoltage protection scheme for the Diesel Generator E MCC OB565 bus, establishes the margin of safety for the scheme which is consistent with margin of safety for the Class IE 4.16
5KRJgg'3-044 C E C: DCP 93-9008, Unit1 0 C E'he scope of this modification is to replace the RHR flow recorder FR-EII-IR608.
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No. The replacement of the Bailey recorder in the Unit I Control Room will not have radiological consequences as a result of this modification. The replacement of this recorder will maintain its seismic mounting qualification and does not interfere with the logic, control or operation of any safety related plant system or components. This modification does not increase the probability of occurrence of an accident or the probability of occurrence of a malfunction of equipment as described in FSAR Chapters 6 and 15. II II. No. Replacing a recorder in the Control Room with a more efficient and reliable model will not result in any increased risk to public health and safety. No new accident scenarios or malfunctions of a different type can result. The new recorder will maintain separation and seismic mounting in the IC601 panel. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. Although the flow recorder is not specifically addressed in the Technical Specifications, the RHR system is a Tech. Spec. related item. The Tech. Specs. for RHR are 3/4.4 and 3/4.5. The proposed modification does not adversely affect the logic, control, or operation of any Tech. Spec. related system. None of the parameters that are the bases for the Technical Specification will be impacted. Therefore, no affect on any Tech. Spec. margin of safety occurs.
l S~Q.: 93-045 C C: DCP 92-3032/A/8/C - Unit 1 C
GE'he modification includes replacing portions of the existing second, third and fourth extraction steam piping between the main turbine and closed feedwater heaters.
SVIHHlhRY'.
No. FSAR Section 15.1.1 evaluates a decrease in core coolant temperature caused by loss of feedwater heating. Extraction piping provides steam to feedwater heaters for preheating reactor feedwater. Piping design calculations were reviewed to assess impact of proposed action on the original design basis. Planned piping replacements do not affect the extraction steam design basis.
Proposed modifications neither alter system function nor performance. Installation of a more erosion/corrosion resistant pipe material will ensure piping integrity and enhance plant reliability.
II. No. The second, third and fourth extractbn piping systems do not perform a safety related function nor do they interface with safety equipment or systems. The proposed actions ensure piping integrity and enhance plant reliability. Therefore, the planned extraction piping modifications do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The planned modifications will not reduce the margin of safety as defined in the basis for any Technical Specification. There is no Technical SpeciTication section applicable to the Extraction Steam piping. Piping replacements will not impact system operations.
Qgg.gg '3-046 C C: DCP 92-3017, Unit 2 0 C The purpose of this modification is to re-wire the control circuits of Appendix R Safe Shutdown Path 2 Motor Operated Valves (MOVs) to ensure their operability in the event of a control room fire.
S~UINlhBY'o.
The relocation of the limit and torque switches in the control circuit protects the Path 2 MOVs from an Appendix R fire in the Control Room. It ensures operability of the valves by enabling the action of limit and torque switches to control valve stroke under the postulated hot short condition. Therefore, this modification does not increase the probability of occurrence of an accident or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. The design criteria for this modification is to maintain each h/DV functional (i.e.,
automatic, manual, testability, and protective features) requirements. The accident analyses in Chapter 15 of the FSAR were reviewed and this modification does not create a possibility for an accident or malfunction of a different type.
III. No. The RHR system operability bases for Tech Specs are discussed in Sections 3/4.3.3, 3/4.4.9, 3/4.5.1, and 3/4-9-11. The RCIC system operability basis is discussed in Sections 3/4.3.5 and 3/4.7.3. The Reactor Recirculation System is discussed in Sections 3/4.3.4 and 3/4.4.1. The Reactor Water Cleanup system is integral to Reactor Coolant Chemistry and, therefore, Section 3/4.4.4 is applicable.
The relocation of the limit and torque switches in the control circuitry of the MOVs does not affect the operation or safety function of the valves.
Therefore, this modification does not reduce the margin of safety as defined in the basis for the Tech Spec sections identified.
5Q~O '3-047 C SS C: DCP 92-9034Z, Unit 1 Ee The proposed action is to add a permissive control logic into the automatic close circuitry of the non-Class IE 13.8 kV startup incoming feeder breakers IAIOIO4 and IA10204 which blocks closing of the breaker for a Unit I LOCA until the Unit 1 13.8 kV pump loads are shed by the Unit 1 13.8 kV bus load shed logic.
I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that addition of a permissive control logic into the automatic close circuitry of the 13.8 kV startup incoming feeder breakers 1A10104 and 1A10204 does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR.
Thus, the addition of a permissive control logic into the automatic close circuitry of the 13.8 kV startup incoming feeder breakers 1A10104 and 1A10204 does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed.
The termination of the internal panel wiring and cables within panels IC626, IC627, IC654 and switchgear 1A101 and 1A102 is in accordance with existing approved installation termination procedure.
Thus, the addition of a permissive control logic into the automatic close circuitry of the 13.8 kV startup incoming feeder breakers 1A10104 and 1A10204 does not create the possibility for an accident or malfunction of a different type.
III. No. Based upon a review of the Technical Specifications, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including Supplements 1 through 7, the addition of a permissive control logic into the automatic close circuitry of the 13.8 kV startup incoming feeder breakers 1A10104 and 1A10204 does not reduce the margin of safety as defined in the basis for any Technical Specification.
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SERE 93-O48 C 0 E E E CE: NL93-008, UnitCommon ESC 0 C Changes to utilize the solid radioactive waste processing services proposed by the Pacific Nuclear Drying (Dewatering), Solidification and Liquid Processing Systems.
I. No. Accidents involving the proposed equipment and containers could result in an airborne radioactivity release which passes directly to the environment via the plant vent. These type of accidents have been previously evaluated in FSAR 15.7.
The proposed action does not constitute an unreviewed safety question for accidents or malfunctions as previously evaluated in the FSAR.
II. No. The use of the proposed Pacific Nuclear Systems to be used as mobile radwaste processing systems as described FSAR 11.4.2.2 does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Solid Radioactive Waste (SRW), 3/4.11.3 provides the basis and limiting condition for SRW systems. Processing of waste streams using the proposed Pacific Nuclear system is performed in accordance with the Process Control Program. Proposed action does not affect technical specification or margin of safety provided in bases for technical specification.
t !RGB.hQ C OS D C 93-049 E ERE C:
C NL 92-033 (TP-069-040, TP-069-041), Unit Common N
The purpose of TP-069-040 is to reduce organic contaminants in Liquid Radwaste (LRW) collection/surge tanks.
The purpose of TP-069-041 is the decontamination of the LRW collection/surge tanks and associated piping.
No. TP-069-040, LRW Collection/Surge Tank Organic Reduction Using Hydrogen Peroxide, involves injection of test chemicals to the LRW collection/surge tanks to reduce organic contaminates followed by processing through the LRW filters and demineralizer to the sample tanks. This TP does not increase the radioactive inventory in the radwaste system and measured liquid activities will likely not be higher than normal since thorough tank cleaning per TP-069-041 will be performed as a prerequisite to this TP. TP-069-040 does not involve the evaporator processing steam including the evaporator concentrates waste tank and associated piping and support equipment.
For the above reasons, TP-069-040 and TP-069-041 do not increase the probability of occurrence or the consequences of the accident analysis in FSAR section 15.7.3 or 2.4.13.3.
No. The only failure that could be postulated as a result of this test procedure would be an integrity failure resulting in draining radioactive material and test chemicals contained in any of the affected components for TP-069-040 or TP-069-041. The test procedures limit the amount of catalase to be injected, therefore, limit the amount that could be spilled given an integrity failure.
The test procedures again limit the amount of peroxide to be injected, therefore, limit the amount that could be spilled given an integrity failure.
For these reasons as stated above, the proposed action does not create the possibilty for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. NO. TP-069-041, LRW Collection/Surge Tank Decontamination Using Hydrogen Peroxide, is intended to remove radioactivity from the liquid radwaste stream by using peroxide to free activity from LRW system tanks, piping, etc.
TP-069-040, LRW collection Surge Tank Organic Reduction Using Hydrogen peroxide, is intended to remove organic material from the liquid radwaste stream following thorough tank cleaning for activity removal per TP-069-041. Radwaste processing during testing will be in accordance with the Offsite Dose Calculation Manual in compliance with Technical Specification involving appropriate treatment and radioactive effiuents.
For these reasons, the test procedures do not significantly reduce the margin of safety.
55R59 '3-050 C C: DCP 93-3001, Unit 1 C OF C GE'eplace the present HY-CAL Remote Temperature Detectors (RTD) (TE-15751 thru TE-15770) during a Unit I RIO with new RTD assemblies that are not prone to moisture induced failure and will not utilize asbestos containing material.
No. The Suppression Pool Temperature Monitoring system (SPOTMOS) is a redundant safety related system with eight safety related RTDs providing input to each division (I and H). The hardware changeout to increase the reliability of the SPOTMOS RTDs will have no affect that could increase the probability of occurrence of any accident evaluated in the FSAR. Potential failure modes of the replacement RTDs (open, short, and drift) are the same as the existing sensors and would be detected by the electronics unit or by the plant computer system allowing the plant operator to remove the failed sensor from the temperature average calculation.
Therefore, this modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15, or in the Design Assessment Report (DAR).
II. No. Potential failure modes of the replacement RTDs (open, short, and drift) are the same as the existing RTDs and would be detected by the electronics unit or by the plant computer system allowing the plant operator to remove the failed sensor from the temperature average calculation.
This modification does not create the possibility of a malfunction of equipment of a different type since failure of a replacement RTD would have the same effect on the SPOTMOS system as failure of an old RTD had in the past.
III. No. Technical Specifications limiting conditions for operation 3.3.7.4, "Remote Shutdown Monitoring Instrumentation"; 3.3.7.5, "Accident Monitoring Instrumentation"; and 3.6.2.1, Depressurization Systems Suppression Chamber" ensure the availability of the SPOTMOS instrumentation and limit the bulk suppression pool temperature so that adequate suppression pool heat capacity is maintained and containment integrity is not challenged. The Technical Specifications do not specifically address the type of sensing instrument used as partofthe SPOTMOS system. The replacementof HY-CAL RTDs has no affect on any margins of safety defined in the basis for any Technical Specification.
QQE5Q '3-052 C 0 C: DCP 93-3041, Unit 1 E C ONO C This DCP completes the retirement of the Unit 1 Reactor Building and Turbine Building Sample Stations (RBSS and TBSS) that were abandoned under DCP 90-3026.
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I. No. None of the equipment or piping that is being removed from the plant by this DCP has an affect on equipment or systems that are important to safety. All of the equipment and piping that are being affected by this DCP were abandoned under previous DCP's. The affected piping systems are all nonsafety related.
Therefore, this DCP does not increase the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety.
II. No. This DCP does not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. This DCP removes equipment and piping from the plant that has been abandoned as a result of DCP 90-3026. DCP 90-3026 was the final modification in a series that installed new Turbine and Reactor Building sampling systems and abandoned the existing systems in place. This DCP does not create a possibility of a different type of accident or malfunction because the equipment is isolated and not functional. The removal of this equipment actually reduces the possibility for an accident or malfunction of any type because it eliminates unnecessary equipment and valves that could be inadvertently operated.
III. No. This modification does not reduce the margin of safety as defined in the basis for any Technical Specifications. The abandoned sample stations have been replaced by new sample stations and are not addressed in the Technical Specifications 3.4.b and 3.4.11.
S~E~O '3-051 C E C: DCP 93-9010, Unit 1 D C C E'CP 93-9010 relocates Area Radiation Monitor (ARM) RIT-13711 to outside of Room l-508 on EL 749'djacent to the new Reactor Building Sample Station (RBSS).
SLlEIElhHY'.
No. The area radiation monitoring system is described in FSAR Section 12.3.4-1. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15,, or in the Design Assessment Report (DAR). The hardware relocation provides monitoring in the vicinity of the relocated RBSS and returns the system to its original design basis.
Potential failure modes of the relocated ARM have not changed. This action has no affect that could increase the probability of occurrence of any accident evaluated in the FSAR.
II. No. The ARM equipment is supported seismically and conduit work or rework is performed in accordance with procedures which assure the relocation does not impact any safety related equipment in the Reactor Building. Thus the proposed relocation does not create the possibility of a accident of a different type than previously evaluated in the FSAR.
This modification does not create the possibility of a malfunction of a different type since all equipment is the same. Failure of the relocated equipment would have the same effect on the ARM system as failure of the equipment in its current location.
III. No. The events analyzed in Chapter 6 and 15 of the FSAR and the DAR specifically addresses the Instrument Line Break in FSAR section 15-6.2. This section notes that a general increase in area radiation monitoring reading is one type of indication for operator action to mitigate this event. The proposed action to relocate ARM RE-13711/RIT-13711 to the RBSS area restores the original plant design basis to provide continuous monitoring in the Reactor Building sampling station area, and serves as a early detection method of such leakage.
The Technical specifications do not govern operation of the RBSS area radiation monitor. The relocation returns the ARM system to original design intent. Thus this proposed action does not affect the margin of safety defined in the basis for any Technical Specification.
S~E~O'3-053 0 C: DCP 93-019, Unit Common C GE:
This modification:
- 1. Removes four eductors and the majority of the piping inside each of the six Liquid Radwaste (LRW) Collection and four LRW Surge Tanks.
- 2. Installs a 90'lbow rotated 45'n the recirc piping inside each of the ten tanks.
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I. No. This modification does not alter the design basis or operation as descrbed within FSAR Section 11-2.1, "Liquid Waste Management Systems". The possibility of a pipe rupture has been evaluated against the Accident Analyses in Chapter 15 of the FSAR, specifically Section 15.7.2 - Liquid Radwaste System Failure, and Section 15.7.3 - Postulated Radioactive Releases due to Liquid Radwaste Tank Failure and concluded there is no decrease to the design integrity of the system.
Therefore, the proposes action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. Implementation of this modification will not change the normal operation of the Liquid Radwaste Management Systems, specifically the Liquid Radwaste Processing Subsystem. The system will continue to operate in accordance with its original design bases, and the piping modifications comply with the original construction codes (ANSI B31.1) and regulations. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. None of these modifications affects the concentration of radioactive materials released in the liquid radwaste effluents (ref. 3/4.11.1.1), nor the doses of the radioactive material in those effluents (ref. 3/4.11.1.2), nor the operability/availability of the LRW processing Subsystem (ref. 3/4.11.1.3), and therefore, will not reduce the margin of safety as defined in the basis for any Technical Specification.
U II
QQ~() '3-054 C SS C: NL-93-002, Unit 2 D C Test of improved iron removal cation resin in condensate polisher.
WNIMhBX'o.
This test will have negligible adverse impacts on vessel chemistry and thus cannot increase the probability of occurrence or the consequences of an accident or malfunction of equipment important of safety. The test should reduce feedwater metals, thereby reducing the amount of crud in the primary system.
The consequences of this action were evaluated relative to the offsite dose resulting from the rupture of a Liquid Radwaste tank containing radioactive material. This action is bounded by the analysis for the rupture of the Concentrates Waste Tank, described in Section 15.7.3 of the FSAR.
II. No. The action will have minimal effects on plant operation and does not create a possibility for an accident or malfunction of a different type than previously evaluated.
This action could only result in one condensate demineralizer being removed from service. This would have negligible impact on the plant, since the design has one more demineralizer vessel than required for 100% operation.
III. No. The effects of this action on vessel chemistry impurity levels is small and will not challenge the margin of safety of vessel chemistry Technical Specifications (T.S.
3/4.4.4 8 314.5.5), nor will it have an adverse effect on cladding integrity.
S~E59 g3 055 C C: NL-93-003, Unit Common Installation of a Bypass which will replace a failed solid state trip unit on Breaker 52-210333 of the Reactor Bldg. Motor Generator (MG) set 1S246/1G.
SMHRlhBY'.
No. The Bypass provides the same degree of electrical circuit protection as described in FSAR Section 8.3.1.3.13 which describes Electric Circuit Protection Systems.
The Bypass does not affect the capability of maintaining the electrical isolation required between redundant power sources as described in FSAR Section 8.3.1.3.5, Class 1E Isolated Swing Bus.
The Bypass will not inhibit the swing bus and transfer switch from meeting their intended safety function, so that for a loss of off-site power and any single failure, the minimum required ECCS flow to meet 10CFR50 Appendix K criteria is always available.
II. No. The Preferred 480 VAC Supply for the Class IE Swing Bus MG Set IS246/IG202, is 480 Circuit Breaker 52-21033, at ESS Channel A Load Center IB210. ESS Channel A Load Center IB210 is supplied from 4kV Load Group A, via circuit breaker 52-20106. The FSAR clearly analyzes the loss of 4kV circuit breaker 52-20106 which is the source for ESS Channel A Load Center IB210 and subsequently the source for the Class 1E Swing Bus MG Set IS246/IG202. As such, the FSAR, in Table 8.3-9, states that the loss of all Load Group A 480VAC loads has no effect on safety function since the safety functions are provided by redundant equipment supplied by load groups B, C 8 D. From a bounding perspective, a postulated gross failure breaker 52-21033, Reactor Building MG Set IS246/IG202, as a result of this Bypass, cannot create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The installation of the subject bypass will not reduce the margin of safety as defined in the basis of Technical Specification 3.5, Emergency Core Cooling System, ECCS Operating 8 Shutdown The installation of the subject bypass will not reduce the margin of safety as defined in the basis of Technical Specification 3.8.3, On-Site Power Distribution Systems.
The installation of the Bypass as evaluated herein will restore Division I Low Pressure Control Injection (LPCI) to an operable status and thereby restore the plant to the original margin of safety associated with the Division I LPCI equipment.
.P QE~Q '3-056 C E C: DCP,92-3020A, D, F, H, I, Unit1 DE C 0 C E'his proposed activity will make the necessary modifications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers located on selected piping systems.
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No. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to be modified accordingly.
All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the Design Specifications and FSAR have been addressed in the reanalysis effort. As in the existing piping analysis, the applicable code design limits have been met to ensure piping integrity and system function; FSAR Section 15.0.3.1, 15.0.3.5.
II. No. The accident events/causes in FSAR Chapter 15.0 that are applicable to this proposed action involve equipment malfunctions/failures and pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations. Therefore, the proposed change will not increase the probability of an accident and a new unanalyzed type of malfunction or accident will not be created by the removal or replacement of snubbers.
III. No. The proposed action reduces the overall snubber population for Unit 1. In case of snubber size change the same type of snubber as currently installed will be used where necessary - i.e., no new type of snubber will be introduced by these modifications. The bases, as defined in the Technical Specifications, deal with snubber test requirements and are not directly affected by this proposed action.
Removing or replacing snubbers will not reduce the margin of safety as defined by the Technical Specification noted.
P' SE~) '3-057 C C: DCP 93-9014Z, Unit Common The proposed action for Diesel Generator C is to add a transient suppresser on the contactor coils of the Fuel Oil Transfer Pump OP514C, the Air Compressors OK507C1 and OK507C2 and a varistor on the 3FOT relay coil.
SDU55lhBY I. No. Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine if the proposed action has an effect on the spectrum of postulated initiating events for which transients or anticipated operational occurrences and accident conditions were analyzed.
The transient suppressor consists of very simple discrete components (resistor, capacitor and varistor) which are considered to be highly reliable. For the 3FOT relay, the addition of the varistor is, as indicated, a simple discrete component with high reliability.
Thus, the addition of a transient suppressor to the contacter coils of the Fuel Oil Transfer Pump OP514C, the Air Compressors OK507CI and OK507C2 and the addition of a varistor on the 3FOT relay coil does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine that the addition of a transient suppressor to the contacter coils of the Fuel Oil Transfer Pump OP514C, the Air Compressors OK507CI and OK507C2 and the addition of a varistor on the 3FOT relay coil does not create a possibility for an accident or malfunction of a different type.
III. No. Based upon a review of the Technical Specification, NUREG-0931 and NUREG-1042, the FSAR and NUREG-0776, Safety Evaluation Report for SSES, including supplements 1 through 7, the addition of a transient suppressor to the contacter coils of the Fuel Oil Transfer Pump OP514C, the Air Compressors OK507C1 and OK507C2 and the addition of a varistor on the 3FOT relay coil does not reduce the margin of safety as defined in the basis for any Technical Specification.
SREH093-056 C S E E C: DCP92-9005, Unit1 DESC C E'he modification will change the cleanup line isolation signal from Level 3 or high drywell pressure to Level 2 or high drywell pressure.
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No. The modification does not increase the probability of occurrence of the operational transients and accidents described in FSAR Sections 6.2, 6.3 and Chapter 15. This conclusion is based on the consideration that the isolation signal change from Level 3 to Level 2 does not degrade the operation of any equipment or system. Furthermore, the relays and level switches to be used for the Level 2 isolation signal are identical in design, material and construction to those currently used for the Level 3 isolation signal.
II. No. This modification does not create the possibility for an accident or malfunction of a different type than any analyzed in FSAR Sections 6.2, 6.3, Chapter 15 and the Design Assessment Report.
The modification uses spare terminals on existing relays to receive an input from a Level 2 switch. These relays are identical to those used to receive the Level 3 input signal. The actuation logic remains single failure proof.
Containment isolation signals are reviewed in Section 6.2.4 of the Safety Evaluation Report (NUREG-0776). This review was based on FSAR revisions up to and including number 33. Level 2 was found to be acceptable and therefore defines the basis for the margin of safety.
III. No. The reactor vessel low water level signal for isolation of the suppression pool cleanup line is contained in FSAR Table 6.2-12 and Technical SpecificationsTable3.6.3.-I.
The table in the FSAR currently indicates Level 3, while based on a review of historical revisions, it was determined to be Level 2 prior to Revision 36. The Technical Specifications for Unit ¹1 have always indicated Level 3. Therefore, it is concluded that since the proposed modification changes a containment isolation setpoint to a previously accepted value, a reduction in the margin of safety will not occur.
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SRB'J$ 2-'93-059 C: DCP 90-9032D, Unit 2 ESC 0 0 C G The proposed action will eliminate the alarms: AAH-23233, Unit 2 Circulating Water pH-High, AAL-23233, Unit 2 Circulating Water pH-Low, AALL-23233, Unit 2 Circulating Water pH-Low-Low, and CAHL-23236, Unit 2 Circulating Water Conductivity. The above alarms are local alarms on panel OC538 (Unit 2 side) - Circulating Water Acid Feed System Panel.
SVlmlhBE I. No. FSAR Section 10.4.5.1 states the design bases for the system. None of the design bases will be affected by this modification. The proposed modification does not adversely affect the ability of the Circulating Water System to maintain the circulating water at the required pH level.
The proposed modiTication deletes alarms annunciated from equipment which has experienced chronic problems and is no longer utilized for its original intended function. The function of pH monitoring and acid injection is performed manually by Chemistry and Operations personnel rather than by automatic injection via signals from monitors on panel OC538. The circulating water is maintained at the required pH levels by the manual control. Based on the above, the proposed modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety.
II ~ No. FSAR Section 10.4.5 - Circulating Water System was reviewed for the impact of the proposed change. The Circulating Water System performance requirements are met with the deletion of the subject alarms. Circulating water pH is maintained at the required level. Acid injection is performed manually based on chemistry recommendations from regular sampling. The proposed modification will not impact the operation of any plant systems that perform a safety function. Based on the above, the proposed modification will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.
III. No. The Unit 2 Technical Specifications have been reviewed for impact of the proposed modiTication, specifically Sections 3.0 and 4.0 - Limiting Conditions for Operation and Surveillance Requirements and Section B3/4 - Bases for Sections 3.0 and 4.0. The proposed modification will not affect the operation of any system as described in the Technical Specifications. The proposed modification has no impact on acceptance limits which form the basis for the Technical Specifications. Circulating Water - Acid Injection is not addressed in the Technical Specifications. Therefore, the margin of safety as defined in the bases of the Technical Specification, will not be reduced.
I S~E~O'3-060 C C DCP 92-9076, Unit 1 C
This modification will correct an improperly sized relief valve (PSV12645) on the pressure relief piping for Unit 1 Reactor Building Containment Instrument Gas (CIG) Accumulator 1T215.
I. No. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15, because the modification re-establishes the original design basis for the CIG System and allows it to be operated as described in the FSAR.
The affected portion of the CIG System serves no safety function and because the modification meets all the design requirements specified in the FSAR, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification replaces existing incorrectly sized pressure relief valve PSV12645 with a valve that satisfies ASME Section,VIII requirements. The new valve will not alter the function of the CIG piping nor will it degrade the operation of CIG Accumulator 1T215. This modification simply corrects a design error and allows the CIG piping in question to fulfill its original design basis. Since this modification eliminates a non-conforming condition and re-establishes the original design basis of the CIG System, the modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification does.not jeopardize or degrade the function or operation of any plant system governed by the Technical Specifications. None of the parameters that are bases for the Technical Specifications will be adversely impacted by this modification.
The replacement of pressure relief valve PSV12645 will ensure that the CIG System meets its original design basis as specified in the FSAR; consequently, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
SKKM93-Oe1 C S EFE ENCE DCP 93-9009, Unit 1 C C On the Unit 1 Feedwater System, the 12" thinned pipe will be cut at the weld centerline to the branch side of the 20" x 20" x 12" upstream feedwater tee and at a point (1 to 3 ft.) downstream.
This section of the 12" pipe will be replaced with in-kind material. Also, the weld overlay on the tee will be removed.
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I. No. Although no additional pipe breaks are postulated at the new 12" pipe weld, the existing rupture restraints (PR52, 251, 252, 605) locations, configurations, and load capacities are sufficient to restrain the Feedwater piping even if a break were to occur at the new weld location. Thus, this modification does not increase the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification does not create the possibility of an accident or malfunction of a different type than any evaluated in the FSAR since the feedwater 12" pipe and 20" x 20" x 12" tee are being restored to their original design condition using in-kind material.
III. No. Technical Specification 3/4.4.8 (Reactor Coolant System - Structural Integrity) is applicable to the feedwater pipe replacement and weld overlay removal. The removal of the weld overlay from the 20" x 20" x 12" tee and replacement of the thinned section of 12" piping and the 20" x 20" x 12" tee (if required) and 12" elbow (if required) restore the original design margins of safety since it replaces thinned piping components with those that provide full pressure integrity.
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~E]gg '3-062 C OS E C DCP 93-9016, Unit 1 C C E'his modification relocates the EAC on PR-211 and add one rigid shim to PR-220 for the required horizontal direction of pipe movement thus returning the restraints to their original design basis. Furthermore, the existing shim (or EAC) can be abandoned in place because they do not interfere with the normal pipe movement and would be inactive during a postulated pipe rupture of the main steam lines.
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No. FSAR Section 3.6, Protecting Against Dynamic Effects Associated With The Postulated Rupture Of Piping, and in particular, Subsection 3.6.1.2.1, Main Steam System, describe pipe whip restraints such as PR-211 and PR-220 as having only one function.
That function is to prevent damage to safety related components essential to reactor safe shutdown in the event of a main steam pipe rupture. This modification ensures proper gaps on pipe whip restraints PR-211 and PR-220 and will validate the original design basis. FSAR Chapters 6 and 15 have been reviewed, and based on the above, this modification will not increase the probability of occurrence of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification will enable pipe whip restraints PR-211 and PR-220 to perform their original design function of restraining the main steam piping during a postulated pipe rupture and preventing any damage to safety related components or equipment. Since this modification does not deviate from the original design basis, the modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification allows pipe whip restraints PR-211 and PR-220 to perform their original design function; to protect other safety related components from being damaged during a main steam line pipe rupture. Therefore, this modification does not reduce any margin of safety as defined in the bases for any Technical Specification.
QER5Q '3-063 C S C DCP 93-9023, Unit 1 The purpose of this modification is to eliminate the contamination of the Auxiliary Boiler System and potential release of radioactive gases to the environment by rerouting the 3/4" SP-HBD-3023 drain line from steam trap ST10702 back to the main condenser instead of to the Auxiliary Boiler Deaerator.
I. No. FSAR Chapters 6 and 15 have been reviewed with particular attention paid to FSAR Section 15.2.5 "Loss of Condenser Vacuum", for possible adverse consequences with respect to the installation of this modification. The new drain piping will be designed to meet all the requirements of the ASME/ANSI 831.1 Code: thereby satisfying the design requirements for the condenser air removal piping as specified in FSAR Table 3.2-1.
Neither system impacted by this modification performs any safety related functions and because the new drain piping from steam trap ST10702 will meet all design requirements specified in the FSAR, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification will allow the discharge from steam trap ST10702 to enter the HP, condenser instead of into the Auxiliary Boiler System where the non-condensable gases can be removed by air ejectors and delivered to the Gaseous Radioactive Waste System for processing. The new 3/4" discharge piping will be designed to meet all the requirements of ASME/ANSI B31.1 Code to ensure its structural integrity and guarantee that the HP condenser and the Condenser Air Removal System functions as intended per the FSAR. Following this modification, the Auxiliary Boiler System will be able to function per its design basis with a greatly reduced risk of contamination from the Main Steam System.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by the Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification.
Qgg.50 '3-064 C C DCP 92-9067, Unit 1 C
This modification will replace the existing Reactor Feed Pump Turbine (RFPT) Steam Admission Valves HV-12710A, B 8 C with new valves due to the extremely high frequency of repair on these valves.
I. No. The one for one replacement of the four inch steam admission valves (HV-12710A, B &
C) will have no affect on the Main Steam System's ability to operate the RFPT exactly as it has prior to this modification. The new valve will use the existing motor operator and have the same instrumentation and controls as before. This modification will meet all the design requirements of the ASME/ANSI B31.1 Code; thereby satisfying the design requirements for the Feed Pump Turbine Steam System as specified in FSAR Table 3.2-1.
II. No. The removal of the tubing on the Process Valve Stem Leakoff Collection System will provide a more open work area around the RFPT and a potential condenser leak will be eliminated with the removal of the leakoff isolation valves. The proposed modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Replacing the steam admission valves will have no adverse effects on the Main Steam System or the RFPT; it will, however, enhance the system's ability to operate with the addition of more reliable valves. The removal of the Process Valve Stem Leakoff Collection System from these valves will have no affect on the offsite iodine release rate limits given in Section 3/4.11.2 "Gaseous Effluents" of the Unit 1 Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification; therefore, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
QE~Q '3-065 C E ENCE SCP J93-1006, Unit 2 DESC 0 C The proposed action is to lower the setpoint of PDSL-68702 from 3.2" wg to 2.8" wg.
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- l. No. The FSAR does not discuss the isophase bus duct cooling system, as determined by reviewing sections 8.1, 8.2, 8.3 (Electrical Power Systems), 9.4 (HVAC and Auxiliary Systems), and 10.0 (Turbine Generator Equipment). The isophase bus duct and associated cooling system is a non-seismic, non-safety system. The setpoint change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. The proposed action is intended to improve system reliability and plant availability. The setpoint change will result in a system that will function as the original design intended.
No new accidents or malfunctions of equipment than those previously evaluated in the FSAR or FSAR are created.
III. No. The Technical Specifications do not address the isophase bus duct systems, nor do the proposed actions impact the bases for section 3/4.8.1 (AC Sources) or 3/4.8.3 (Onsite Power Systems).
S~~Q '3-066 C C NL-93-013, Unit 1,2 C GE:
Fuel Pool Cooling Chemical Decontamination tests - These tests will control the evolution of chemically decontaminating the Spent Fuel Pool Cooling system piping, heat exchangers, and pumps.
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I~ No. FSAR section 3.4 "Water Level Flood Design", discusses the use of floor drains to process fluid leakage or fire protection water. This function is provided by ensuring at least one open floor drain in each room. This draining function is not important to safety. FSAR 9.1.3.2 "Spent Fuel Pool Cooling and Cleanup System Description" discusses maintaining the Fuel Pool Heat Exchanger shell side at a minimum of 5 psi lower than the tube side service water pressure to minimize the possibility of radioactive contamination of the Service Water system. This differential pressure is important to prevent cross contamination but not important to safety. FSAR 17.2.9 "Control of Special Processes" describes that chemical cleaning be considered a Special Process. This Special Process determination is not important to safety, therefore the proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. The proposed action of chemically cleaning the Fuel Pool Cooling system will introduce, in a controlled manner, approved decon chemicals into the Fuel Pool Cooling systems. The systems will be isolated one at a time from their respective Fuel Pool. Both pools will be cross tied and cooled by the opposite operable Fuel Pool Cooling system. This pool cross tie is a normal plant evolution. The resultant isolated Fuel Pool Cooling system does not perform any function important to safety. It will be chemically cleaned in a procedurally controlled manner and returned to service. This proposed action does not create a possibility for an accident or malfunction of a different type than evaluated previously in the FSAR.
III No.
~ The Fuel Pool is in 3/4.9.9 of the Technical Specifications. It describes keeping 22 feet of water above the spent fuel. The proposed Fuel Pool Cooling Chemical Decontamination isolates the fuel pool from the Fuel Pool Cooling system, thus the Fuel Pool margin of safety is not compromised. The operability of the Fuel Pool Cooling Service Water Rad Monitors are in T.S. 3/4.7.10. These monitors will be out of service during each decon. The T.S. LCO action requires obtaining grab samples during that time, thus the Fuel Pool margin of safety is not compromised.
C NL-93-014, Unit 1 C
of TP-145-011, Feedwater Flow Measurements by Chemical Tracer, to GE'erformance determine the accuracy of the Unit 1 feedwater flow elements and the attached Feedwater flow instrumentation (FE-1N001A/8/C 8 FT-C32-1 N002A/B/C).
I~ No. The temporary injection skid and sampling stations will be attached hydraulically to the plant during the performance of TP-145-011. The temporary equipment will not affect any control logic or control room indication. Therefore, the only affect on the plant will occur if the tubing to the temporary equipment develops a leak. The temporary equipment will have isolation valves installed with the temporary tubing. Also, the root valves at the feedwater piping can be isolated. The effects of the leak will be the same as a leak on permanent feedwater piping and instrumentation and therefore is bounded by previously evaluated events. The proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
II. No. FSAR section 7.7.1.4.1.2 states that feedwater control is a power generation system and is classified as not related to safety. Section 7.7.2.4.2 states that feedwater control is not required for safe shutdown of the plant, nor is it required during or after accident conditions.
The Reactor Feed Pumps (RFP) will be operated within the bounds of all existing operating procedures throughout the duration of this procedure, therefore with one RFP in manual, accidents are bounded by existing analyses.
III. No. The proposed test will not change any Tech Spec water chemistry parameter except for conductivity. The Tech Spec basis for requiring conductivity measurements is to be able to detect abnormal conditions. From Bases 3/4.4.4 Chemistry - 'When the conductivity is within limits", (< 1pmho/cm), "the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits". By limiting the conductivity to< 0.3 @ho/cm during the test period, the ability to detect other impurities and pH above their limits will not be compromised.
SKE5093-Oe8 e C C DCP 93-3012A, Unit 1 ESC N C Replace valve HV-151F016A's worm sub-assembly and worm gear with a 60:1 ratio gearset to increase the actuator torque rating from 950 ft-Ibs to 1250 ft-lbs.
I. No. The modification has no impact on the dynamic qualification of the motor operator as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations.
The only equipment affected by this modification is HV-151F016A. Since the valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time is within the design basis, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR Section 6.2.4, Table 6.2-12 and 6.2.2.2 are still valid.
II. No. Since the proposed modification does not adversely impact the ability of HV-151F016A to perform its intended design function, does not increase the stroke time of the valve beyond that previously contained in the Design Basis for SSES (FSAR. Table 6.2-12), does not decrease the MOV's allowable seismic acceleration below the actual valve determined in the piping analysis, involves the replacement of components outside the valve pressure boundary, does not affect the ability to open or close the valve when required, and does not affect any other equipment, it is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR (Chapter 15).
III. No. The modification does not reduce the margin of safety defined in the Bases of one Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.6, 3/4.6.2 and 3/4.6.3 were reviewed for this evaluation.
Valve seat leakage and integrity are not affected. The valve's stroke time remains within the accident analysis bounds of 90 seconds. In fact, the decreased design stroke time improves system response. Therefore, this modification has no effect on the RHR System - Containment spray mode intended design function or the basis of the above Tech Specs.
~~igllgg '3-069 C C DCP 93-3012B, Unit 1 ESC C Replace valve HV-151F016B's worm sub-assembly and worm gear with a 60:1 ratio gearset to increase the actuator torque rating from 950 ft-Ibs to 1250 ft-lbs.
~SU gg~Q I. No. The modification has no impact on the dynamic qualification of the motor operator as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations.
The only equipment affected by this modiTication is HV-151F016B. Since the valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time is within the design basis, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR Section 6.2 4, Table 6.2-12 and 6.2.2.2 are still valid.
II. No. Since the proposed modification does not adversely impact the ability of HV-151F016B to perform its intended design function, does not increase the stroke time of the valve beyond that previously contained in the Design Basis for SSES (FSAR Table 6.2-12), does not decrease the MOV's allowable seismic acceleration below the actual valve determined in the piping analysis, involves the replacement of components outside the valve pressure boundary, does not affect the ability to open or close the valve when required, and does not affect any other equipment, it is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR (Chapter 15).
III. No. The modification does not reduce the margin of safety defined in the Bases of one Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.6, 3/4.6.2 and 3/4.6.3 were reviewed for this evaluation.
Valve seat leakage and integrity are not affected. The valve's stroke time remains within the accident analysis bounds of 90 seconds. In fact, the decreased design stroke time improves system response. Therefore, this modification has no effect on the RHR System - Containment spray mode intended design function or the basis of the above Tech Specs.
Qgg.@O '3-070 C E C DCP 90-9086, Unit 2 C 0 C The proposed action is to replace PCV 22643 and PCV 22648 with pressure regulating valves that have superior design features and to raise their setpoint pressure from 150 to 158 psig to reduce/eliminate problems.
I. No. This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Sections 6.3, 15.1.4, 15.2, 15.6.1. Per FSAR Section 9.3.1.5.3 the purpose of the safety related Containment Instrument Gas (CIG)
Nitrogen bottles and piping header is to provide an emergency backup supply to the ADS valves. The failure (open or closed) of the PCVs will only result in loss of one bank of nitrogen bottles which cannot be an initiating event for an accident.
II ~ No. Potential failure modes of the replacement PCVs are failure open or failure closed causing a loss of N2 bottles. Per FSAR Sections 9.3.1.5 and 5.2.2.4 and SER Section 9.3.1 the ADS valves have safety related accumulators with check valves as a redundant gas supply which eliminates the possibility of a single failure from causing a loss of ADS function. No new operating modes are created for the ADS or CIG System.
The modification does not create the possibility of an accident or malfunction of a different type.
III. No. The Technical Specifications discuss the CIG low pressure alarm system in Section 4.5.1, however, this alarm system is not affected by this modification. Likewise, the ability of the CIG System to perform its safety function at or above the alarm set point is not affected. The only other system that could potentially be affected is the ADS system. Technical Specification bases 3/4.5.1 and 3/4.5.2 describe the scenarios where the ADS system would be used to depressurize the vessel to enable the low pressure ECCS systems to provide cooling. These bases also discuss the number of ADS valves required by the safety analysis. The modification has no effect on these bases.
The modification does not reduce the margin of safety of the basis of any Technical Specification.
I Ii\8 E CE DCP 90-9095, Unit 1 The proposed action is to replace PCV 12643 and PCV 12648 with pressure regulating valves that have superior design features and to raise their setpoint pressure from 150 to 158 psig to reduce/eliminate problems.
I. No. This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Sections 6.3, 15.1.4, 15.2, 15.6.1. Per FSAR Section 9.3.1.5.3 the purpose of the safety related Containment Instrument Gas (CIG)
Nitrogen bottles and piping header is to provide an emergency backup supply to the ADS valves. The failure (open or closed) of the PCVs will only result in loss of one bank of nitrogen bottles which cannot be an initiating event for an accident.
II. No. Potential failure modes of the replacement PCVs are failure open or failure closed causing a loss of N2 bottles. Per FSAR Sections 9.3.1.5 and 5.2.2.4 and SER Section 9.3.1 the ADS valves have safety related accumulators with check valves as a redundant gas supply which eliminates the possibility of a single failure from causing a loss of ADS function. No new operating modes are created for the ADS or CIG System.
The modification does not create the possibility of an accident or malfunction of a different type.
III. No. The Technical Specifications discuss the CIG low pressure alarm system in Section 4.5.1, however, this alarm system is not affected by this modification. Likewise, the ability of the CIG System to perform its safety function at or above the alarm set point is not affected. The only other system that could potentially be affected is the ADS system. Technical Specification bases 3/4.5.1 and 3/4.5.2 describe the scenarios where the ADS system would be used to depressurize the vessel to enable the low pressure ECCS systems to provide cooling. These bases also discuss the number of ADS valves required by the safety analysis. The modification has no effect on these bases.
The modification does not reduce the margin of safety of the basis of any Technical Specification.
Qgg ~ 93-072 C OS E C SCP J931004, Unit Common The proposed action reduces the spray pond high temperature alarm setpoint from 85'F to 82'F.
I~ No. No FSAR analyzed accident or malfunction relies on the spray pond high temperature alarm for the initiation of mitigating actions. This alarm, or its failure cannot cause any FSAR accident or malfunction. Because mitigative actions for FSAR analyzed accidents/malfunctions do not rely on receiving or acting upon a spray pond high temperature alarm, changing this alarm setpoint cannot increase the consequences of these accidents. In addition, because the annunciation of a high spray pond temperature cannot cause any of the FSAR analyzed transients/accidents, the changing of this alarm setpoint cannot cause or increase the probability of occurrence of an FSAR analyzed accident or malfunction.
II. No. The reduction of the spray pond high temperature alarm setpoint from 85'F to 82' only affects the pond temperature at which this alarm is received. There are no automatic actions associated with the receipt of the alarm. Neither the receipt of a high temperature alarm nor its associated manual actions create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification 3.4.7.1 addresses spray pond parameters. Reducing the high temperature alarm setpoint from 85' to 82' enhances the ability to control spray pond temperature by providing information to the plant operator sooner to allow spray pond cooling to occur sooner. This will increase rather than reduce the margin of safety.
S~~ 93-073 C C NL-93-015, Unit 1 DESC IP 0 C The proposed bypass will isolate a leak from the plug on the valve body of 1 RV-LSL-10112A by closing two manual valves. The leak is a contributor to the Main Steam pipe tunnel elevated moisture.
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I. No. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not increase with the implementation of this bypass.
A review of FSAR Chapter 15.0 was performed to determine if the proposed action has any affect on the accident analysis already evaluated for SSES, and it was determined that the accident events/causes described in Chapter 15 are not affected by changes to the systems or components included in this bypass.
II. No. The applicable portion of the Main Steam system bypassed has no safety related function. Implementation of the bypass will not create a possibility for an accident or malfunction of a different type than any evaluated previously in Chapter 15 of the FSAR.
III. No. Based on the discussion in the above sections, this bypass has no affect on the basis of any Technical Specifications. Tech Specs 3/4.4.4.7 (MSIV's) and 3/4.6.1.4 (MSIV Leakage Control) were reviewed, and are not affected.
~E~ 93-074 C E C 'CP93-3002Z, Unit1 DE C ION 0 C The proposed action is to add an automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus.
No. The addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis or NUREG 0776. Thus, the proposed action does not increase the probability of occurrence of an accident. The proposed action does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel C control power in the Unit 1 Channel C 4.16KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction. The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action.
II ~ No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG 0776 were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events for which transients or anticipated operational occurrences and accident conditions were analyzed.
The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. Thus, the addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus does not create the possibility for an accident or malfunction of a different type.
III. No. The operability of the Unit 1 Channel C 4.16KV Bus is governed by Technical Specification 3/4.8.3. The bases for operability during operation is to ensure that sufficient power is available to supply the safety related equipment required for the safe shutdown and the mitigation and control of accident conditions. The proposed action provides an automatic transfer scheme which transfers the breaker trip circuitry of divisionalized loads supplied from the Unit 1 Channel C 4.16KV Bus from a normal control power source to an alternate control power source. The proposed action ensures operability of the Unit 1 Channel C 4.16KV Bus for a Unit 1 Channel A battery failure coincident with a LOCA/LOOP condition thereby maintaining the existing margin of safety for the Unit 1 Channel C 4.16KV Bus.
Thus the proposed action does not reduce the margin of safety.
C DCP 93-9025, Unit 1 This modification will provide sufficient 480VAC power to supply two Wyle snubber testers inside the Unit 1 Start-up and Test Room in the Turbine Building, Elev. 656', Area 1. This will involve installing scheduled cable from existing power panel 1PP123.
I. No. The proposed action is not a precursor to any offsite doses. The proposed change is already operational per a By-pass (¹1-92-004) to connect temporary power to the Wyle equipment in the Unit 1 Turbine Building. There could be no effect possible on the accidents that have radiological consequences as a result of this modification. Also, no new radiological consequences can occur. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
This modification does not interfere with the logic, control or operation of any safety-related plant system or components. Although this is not covered in the FSAR, a review of Chapters 6 and 15 were reviewed and this modification does not increase the probability of occurrence of an accident or malfunction of equipment important to safety.
II. No. The proposed modification is already operational per a By-pass (¹1-92-004) to connect temporary power to the Wyle equipment in the Unit 1 Turbine Building. This modification will install permanent power to this equipment.
Accident scenarios have previously been analyzed in FSAR Chapters 6 and 15.
Although this test equipment was not covered in these chapters, the engineering and installation criteria will be satisfied to ensure that no potential adverse effects occur to any existing plant equipment during the course of installation. The proposed modification does not involve changes in system operation nor add a more severe or different type of failure mode.
Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Snubber testing is required by Tech. Spec. 3/4.7.4, but does not specifically address the snubber test equipment nor its power source. The proposed modification will not challenge any safety systems. Therefore, the implementation of DCP 93-9025 will not reduce the margin of safety as defined in the bases of the Technical Specifications.
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~~O'3-076 C SS E E ENC DCP92-3023, Unit1 ESC I 0 C Replace the existing Electrical Protection Assembly (EPA) logic cards with the G.E. Group 7 cal d.
I~ No. The replacement of the EPA logic cards with the new G.E. Group 7 logic cards does not have any impact on the postulated initiating events identified in the engineered safety features and accident analysis of FSAR Chapters 6 and 15. The replacement of the existing EPA logic cards with new Group 7 logic cards will result in a more reliable protection system which will reduce spurious trips of the RPS from its power source circuits. Therefore, there is no increase in the probability of occurrence of an accident and will not increase the consequences of a design basis accident. The correction of these design deficiencies will result in increased reliability and improved operability of the EPAs. The prevention of chip lockup will eliminate blocking of the breaker trip when required and the reduction of the voltage band for the trip setpoint, will reduce premature trip signals. The replacement of the existing logic cards with Group 7 logic cards will not increase the probability of occurrence of a malfunction of equipment.
II. No. A review of FSAR Chapter 15 on accident analysis did not identify any postulated initiating events relative to the proposed action which would create the possibility for an accident of a different type. The Group 7 card is a plug-in replacement for the existing card with additional trip status/indication and circuit test features and does not modify the EPA protection circuit configuration.
The Group 7 card is an enhanced version of the existing card, redesigned to eliminate deficiencies. Therefore, the replacement of the existing EPA logic card with a G.E.
Group 7 logic card does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification Section 4.8.4.3 provides the surveillance requirements governing the operability of the RPS EPAs and establishes the setpoints for undervoltage, overvoltage and underfrequency relay actuation. The setpoints are applicable to the Group 7 logic cards which will replace the existing EPA cards.
Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
QQ~Q '3-077 C E CE NL-89-043, Unit 1 ESC 0 0 C In accordance with 10CFR50.59, if a nonradioactive system becomes contaminated and it is considered necessary to continue operation of the system as contaminated, an immediate safety evaluation of the operation of the system as a radioactive system must be performed. The Reactor Building Chilled Water and Closed Water systems are treated with sodium nitrate to control corrosion. Since both systems enter the drywell and are subjected to a neutron flux, some of the Na-23 will be activated in a reaction to sodium-24.
I~ No. Operating the Reactor Building Chilled Water and Reactor Building Closed Cooling Water systems with trace amounts of Na-24 will not affect the mechanical operation of the systems. The systems will operate the same as before since the low levels of contamination present will not degrade any component. Operating the system will not increase the probability of occurrence or the consequences of releasing activity to the environment above those evaluated in Section 15 of the FSAR.
II. No. The only probable consequence of operating the Reactor Building Closed Cooling and Chilled Water systems with the low level of Na-24 contamination is that it would get released to the environment unmonitored. This is no different than the potential unmonitored release evaluated in Section 15 of the FSAR. The consequences of this potential release is less than those analyzed previously.
III. No. The potential dose consequences from this action are well below the Tech Spec limit of 15 mrem/yr.
Qgg'Jg) '3-078 C OSS C NL-89-047, Unit 2 In accordance with 10CFR50.59, if a nonradioactive system becomes contaminated and it is considered necessary to continue operation of the system as contaminated, an immediate safety evaluation of the operation of the system as a radioactive system must be performed. The Reactor Building Chilled Water and Closed Water systems are treated with sodium nitrate to control corrosion. Since both systems enter the drywell and are subjected to a neutron flux, some of the Na-23 will be activated in a reaction to sodium-24.
I. No. Operating the Reactor Building Chilled Water and Reactor Building Closed Cooling Water systems with trace amounts of Na-24 will not affect the mechanical operation of the systems. The systems will operate the same as before since the low levels of contamination present will not degrade any component. Operating the system will not increase the probability of occurrence or the consequences of releasing activity to the environment above those evaluated in Section 15 of the FSAR.
II~ No. The only probable consequence of operating the Reactor Building Closed Cooling and Chilled Water systems with the low level of Na-24 contamination is that it would get released to the environment unmonitored. This is no different than the potential unmonitored release evaluated in Section 15 of the FSAR. The consequences of this potential release is less than those analyzed previously.
III No.
~ The potential dose consequences from this action are well below the Tech Spec limit of 15 mrem/yr.
S~E~O'3-079 C CE DCP 92-9050, Unit 1 C
GE'he proposed action will eliminate PDAHL-17431 and will install local pressure differential indication to monitor Turbine Building differential pressure.
SMMjMIRY'.
No. The Turbine Building HVAC System has no safety related function. The removal of a local panel alarm will not impact any functional requirements of the Turbine Building HVAC System, therefore, no adverse effects are possible. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15.
The elimination of PDAHL does not impact any safety related equipment since no connection to any safety related circuitry exits. This modification will not adversely affect any structure, system, or component in performing its safety function, and in fact will eliminate a nuisance alarm. There will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR.
This modification can have no effect on the accidents that have radiological consequences. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II ~ No. The proposed change does not conflict in any way with FSAR Section 9.4.4. The Turbine Building HVAC System, in part, is designed to maintain air flows from clean areas to potentially contaminated areas and from areas of potentially lower level contamination to areas of potentially higher level contamination through a filtered exhaust system. The proposed modification will not alter the above mentioned design basis. Therefore, this modificatio does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. There is no limiting condition for operation in the Technical Specification for the Turbine Building HVAC System. This modiTication will eliminate a nuisance alarm. The proposed changes will have no adverse affect on Turbine Building HVAC operation.
Therefore, the implementation of DCP 92-9050 will not reduce the margin of safety as defined in the bases of the Technical Specifications.
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~~) '3-080 C SS E C DCP 93-9010, Unit 1 D C DCP 93-9010 relocates Area Radiation Monitor (ARM) RIT-13711 to outside of Room I-508 on El. 749'djacent to the new Reactor Building Sample Station (RBSS). This change represents a direct one-for-one hardware move in that the existing radiation monitor is physically removed from its present location and reinstalled at the new Reactor Building Sample Station (1C210A, B) location.
I. No. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15, or in the Design Assessment Report. The hardware relocation provides monitoring in the vicinity of the relocated RBSS and returns the system to its original design basis. Potential failure modes of the relocated ARM have not changed. This action has no affect that could increase the probability of occurrence of any accident evaluated in the FSAR.
This modification does not increase the consequences of an accident since failure of the relocated ARM will have the same effect on the system as it would have before relocation and will in no way affect the offsite dose to the public. In fact it will provide quicker warning of instrument line break at RBSS with potential to reduce overall consequence of that event.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since a simple relocation of existing equipment is being made. Potential failure modes of the relocated ARM are the same as those at the existing location.
This modification does not create the possibility of a malfunction of a different type since all equipment is the same. Failure of the relocated equipment would have the same effect on the ARM system as failure of the equipment in its current location.
III. No. Plant Technical Specification limiting conditions for operation 3.3.7.1 "Radiation Monitoring Instrumentation" specifically addresses the main control room outside air intake radiation monitors and the area criticality monitors which include the new fuel storage vault and spent fuel storage pool area monitors. The Technical Specifications do not specifically address the RBSS ARM instrumentation.
The Technical Specifications do not govern operation of the RBSS area radiation monitor. The relocation returns the ARM system to original design intent. Thus this proposed action does not affect the margin of safety defined in the basis for any Technical Specification.
SRK5093-O81 C ER CE NL-92-015, Unit 1,2 D C I Bypass to install temporary HEPA Units to Unit 1 (2) Containment Purge Duct during Plant Outages: The purpose of this bypass is to install temporary HEPA filter units to the Containment Purge supply duct during plant outages to maintain a negative pressure inside primary containment, thereby ensuing airflow from areas of lesser potential contamination to areas of greater contamination.
Sggg~
I. No. FSAR Section 9.4.2 discusses the purposes of the Reactor Building HVAC System.
One function is to supply vent and purge air to primary containment. This function is a non-safety related function. Prohibiting its use during plant outages does not increase the probability or consequences of any accident or malfunction of equipment important to safety.
The Primary Containment Isolation function discussed in Section 6.2.4 is being affected by this Bypass in that isolation valve HV-15721 (25721) will be placed in its fail closed position and valves HV-15722, 23, 24, 25 (25722, 23, 24, 25) may be maintained in their open position. Placing these valves in the open position is acceptable and does not increase the probability or consequences of an accident or malfunction of safety related equipment since primary containment is not required in Conditions 1, 2, or 3.
II. No. A review of FSAR Sections 6.2.3, 6.2.4, 6.5.1, 6.5.3, 9.4.2, 15.0 could not identify any possible accidents or malfunctions of a different type than previously evaluated in the FSAR. This Bypass allows the airflow pattern to be from areas of lesser potential contamination to areas of greater potential contamination. Installation of this Bypass minimizes the possibility of an airborne incident spreading contamination outside of primary containment.
III. No. The only Tech. Spec. section affected by this Bypass is Section 3.6.3, Primary Containment Isolation. However, since the unit will not be in Conditions 1, 2, or 3, primary containment isolation is not required; therefore, there is no reduction in the margin of safety as defined in the basis section of Tech Specs.
V 4
FjgP gg '3-082 C EFE E C DCP93-3003Z, Unit1 C
GE'he proposed action is to add an automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus.
~SHElhBY'o.
The addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis or NUREG 0776. Thus, the proposed action does not increase the probability of occurrence of an accident. The proposed action does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel D control power in the Unit 1 Channel D 4.16KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction. The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG 0776 were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events for which transients or anticipated operational occurrences and accident conditions were analyzed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. Thus, the addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus does not create the possibility for an accident or malfunction of a different type.
III. No. The operability of the Unit 1 Channel D 4.16KV Bus is governed by Technical SpeciTication 3/4.8.3. The bases for operability during operation is to ensure that sufficient power is available to supply the safety related equipment required for the safe shutdown and the mitigation and control of accident conditions. The proposed action provides an automatic transfer scheme which transfers the breaker trip circuitry of divisionalized loads supplied from the Unit 1 Channel D 4.16KV Bus from a normal control power source to an alternate control power source. The proposed action ensures operability of the Unit 1 Channel D 4.16KV Bus for a Unit 1 Channel B battery failure coincident with a LOCNLOOP condition thereby maintaining the existing margin of safety for the Unit 1 Channel D 4.16KV Bus. Thus the proposed action does not reduce the margin of safety.
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Qgg ~O'3-083 c s C DCP 93-3010, Unit 1 0 C E This modification provides a new 24 ton trolley/hoist 1H205 with a lift capability sufficient to lower personnel hatch shield blocks from Elev. 719'o 670'ithout using the railroad bay overhead crane.
I. No. The new 24 ton trolley/hoist 1H205 and special rigging devices have been designed to address all safety impact concerns as well as meet all the NUREG-0612 requirements for liffing heavy loads in nuclear plants. FSAR chapters 6 and 15 have been reviewed and this modification - including inplant usage of the hoist and rigging - will have no affect on the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification complies with all the PP8L Licensing commitments to NUREG-0612 and NRC Generic Letter 85-11 for the lifting of heavy loads in nuclear plants and will be designed and tested so as to preclude any possibility of a dropped shield block that could damage the safety related components and prevent the safe shutdown of the plant. The actual use of this equipment will simulate rigging activities currently performed using the existing railroad bay overhead crane. To ensure proper application of this equipment, it is a requirement of NUREG-0612 and the Installation Instructions that a specific lifting procedure be prepared and reviewed prior to use. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III No. The Technical Specifications have been reviewed and the 24 ton trolley/hoist 1H205 is
~
not mentioned or addressed in any Technical Specification. Furthermore, since this modification has been designed to meet all the applicable requirements of NUREG-0612 and NRC Generic Letter 85-11 for the control of heavy loads at nuclear plants, seismically qualified for all safety impact concerns, the proposed actions taken for this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
GER~ 93-084 C C DCP 93-3011, Unit 2 C 0 C G This modification provides a new 24 ton trolley/hoist 2H205 with a lift capability sufficient to lower personnel hatch shield blocks from Elev. 719'o 670'ithout using the railroad bay overhead crane.
QQJY~INA +
I~ No. The new 24 ton,trolley/hoist 2H205 and special rigging devices have been designed to address all safety impact concerns as well as meet all the NUREG 0612 requirements for lifting heavy loads in nuclear plants. FSAR chapters 6 and 15 have been reviewed and this modification - including inplant usage of the hoist and rigging - will have no affect on the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification complies with all the PP8L Licensing commitments to NUREG-0612 and NRC Generic Letter 85-11 for the lifting of heavy loads in nuclear plants and will be designed and tested so as to preclude any possibility of a dropped shield block that could damage the safety related components and prevent the safe shutdown of the plant. The actual use of this equipment will simulate rigging activities currently performed using the existing railroad bay overhead crane. To ensure proper application of this equipment, it is a requirement of NUREG-0612 and the Installation Instructions that a specific lifting procedure be prepared and reviewed prior to use. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The Technical Specifications have been reviewed and the 24 ton trolley/hoist 2H205 is not mentioned or addressed in any Technical Specification. Furthermore, since this modification has been designed to meet all the applicable requirements of NUREG-0612 and NRC Generic Letter 85-11 for the control of heavy loads at nuclear plants, seismically qualified for all safety impact concerns, the proposed actions taken for this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
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E C DCP 92-3032 A/8/C, Unit 1 E'he modification includes replacing portions of the existing second, third and fourth extraction steam piping between the main turbine and closed feedwater heaters with Corronix clad piping and fittings.
SVRINlhEY'
~ No. FSAR Section 15.1.1 evaluates a decrease in core coolant temperature caused by loss of feedwater heating. Extraction piping provides steam to feedwater heaters for preheating reactor feedwater. Design calculations were reviewed to assess the impact of the piping replacements on the original design basis. The carbon steel portion of the Corronix clad piping, pressure boundary, is the same thickness as the originally installed extraction steam piping. Change in design parameters including weight, elastic modulus and thermal expansion coefficients were found to have insignificant effects on the pipe stress analysis, pipe supports, and supporting structures.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not increase with the implementation of the planned modifications. The accident events/causes described in FSAR Chapter 15 were reviewed in order to reach this determination.
II ~ No. The second, third and fourth extraction piping systems do not perform a safety related function nor do they interfere with safety equipment or systems. The installation of the Corronix clad material does not impact the original piping design stress analysis. Nor does this upgraded material create a chemical interaction that will impact operation of other plant systems or equipment. The proposed actions ensure piping integrity and enhance plant reliability. Therefore, the planned extraction piping modifications do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. There is no Technical Specification section applicable to the Extraction Steam piping.
Piping replacements will not impact system operations. The change in pipe material has no impact on the chemistry requirements for systems discussed in the Technical Specifications. The erosion/corrosion resistant piping material installed will provide a reduction in wear rates resulting in lower metal contributions in the feedwater system.
Therefore, the planned modifications will not reduce the margin of safety as defined in the basis for any Technical Specification.
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S~EgO '3-086 C C DCP 93-3020 A/B/C, Unit 2 C 0 C E'he modification includes replacing portions of the existing second, third and fourth extraction steam piping between the main turbine and closed feedwater heaters with Corronix clad piping and fittings.
S I. No. FSAR Section 15.1.1 evaluates a decrease in core coolant temperature caused by loss of feedwater heating. Extraction piping provides steam to feedwater heaters for preheating reactor feedwater. Design calculations were reviewed to assess the impact of the piping replacements on the original design basis. The carbon steel portion of the Corronix clad piping, pressure boundary, is the same thickness as the originally installed extraction steam piping. Change in design parameters including weight, elastic modulus and thermal expansion coefficients were found to have insignificant effects on the pipe stress analysis, pipe supports, and supporting structures.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not increase with the implementation of the planned modiTications. The accident events/causes described in FSAR Chapter 15 were reviewed in order to reach this determination.
II ~ No. The second, third and fourth extraction piping systems do not perform a safety related function nor do they interfere with safety equipment or systems. The installation of the Corronix clad material does not impact the original piping design stress analysis. Nor does this upgraded material create a chemical interaction that will impact operation of other plant systems or equipment. The proposed actions ensure piping integrity and enhance plant reliability. Therefore, the planned extraction piping modifications do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. There is no Technical Specification section applicable to the Extraction Steam piping.
Piping replacements will not impact system operations. The change in pipe material has no impact on the chemistry requirements for systems discussed in the Technical Specifications. The erosion/corrosion resistant piping material installed will provide a reduction in wear rates resulting in lower metal contributions in the feedwater system.
Therefore, the planned modifications will not reduce the margin of safety as defined in the basis for any Technical Specification.
C OS C DCP 92-9031, Unit Common SC The first action is to change the location of accelerometer VT-05701 from the existing location, 685'levation, to the lowest floor, 660'levation, of the ESSW pumphouse. The second action is to disable the audible alarm function on Seismic Monitoring System panel OC696.
I. No. The relocation meets the design basis requirements in Regulatory Guide 1.12 to monitor and record data on seismic ground motion from a Seismic Category I structure where the response is different from that of the Containment Structure. The relocation will eliminate the spurious seismic triggers of the system which occur at the present location due to pump and valve operating transients. With the elimination of the audible alarm on OC696 the commitment to Regulatory Guide 1.12 for control room annunciation is still met by the seismic system triggered alarm on panel OC653. These actions have no effect that could increase the probability of occurrence of any accident or malfunction evaluated in the FSAR.
The modification does not increase the consequences of an accident since any failure of the relocated accelerometer will have the same effect as it would have before relocation and will in no way affect the offsite dose to the public.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since a simple relocation of existing equipment is being made, and an unnecessary audible alarm is being removed.
This mo'dification does not create the possibility of a malfunction of a different type since all equipment is the same. Failure of the relocated equipment would have the same effect on the system as failure of the equipment in its current location.
III. No. The design of the Seismic Monitoring System is consistent with the requirements of Regulatory Guide 1.12 as, noted in the Technical Specification Bases, Section 3/4.3.7.2, and the Safety Evaluation Report NUREG-0776. The proposed action to relocate accelerometer VT-05701 to ESSW Pumphouse floor elevation 660'nd remove the audible alarm from panel OC696 maintains compliance with the requirements of Regulatory Guide 1.12 and does not affect the margin of safety defined in the basis for any Technical Specification.
S~E+Q '3-088 CE DCP 93-3024, Unit 1 D C C GE'odify HV-149F007 by replacing the actuator motor and installing current monitoring equipment in the valve limit switch compartment for VOTES testing.
SVlljICRE.
I. No. This modiTication has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations.
The only equipment affected by this modification is HV-149F007. The valve's pressure retaining capability is maintained, the leakage characteristics are unaffected.
Therefore, this modification will not increase the probability of occurrence or the consequences of an accident previously evaluated in Chapter 15 of the FSAR.
II No.
~ The proposed modification does not adversely impact the ability of HV-149F007 to perform its intended design function; does not change the stroke time of the valve; does not decrease the MOVs allowable seismic acceleration below the actual valve determined in the piping analysis; improves the ability to open or close the valve when required; does not affect valve leakage; does not affect the electrical distribution system; does not alter the valve limit switch compartment seismic or environmental qualification; and does not affect any other equipment in the RCIC system.
Based upon the above, it is concluded that the proposed modiTication does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR (Chapter 15).
III. No. The modification does not reduce the margin of safety defined in the Bases Technical Specification. Tech SPEC bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.7.3 and 3/4.3.5 were reviewed for the evaluation.
Valve integrity and seat leakage are not affected. The valve stroke time is within the accident analysis bounds of 20 seconds. Therefore, this modification has no effect on the RCIC system intended design function or the bases of the above Tech Spec.
Qgg ~ 93-089 C E E E CE. DCP93-3054, Unit1 DSC 0 0 C This modification installs a 480VAC, three-phase power panel (1PP102) and adjacent welding receptacles located near the alcove area on Elevation 699, Area 10, of the Turbine Building.
SMH5lhBY'.
No. The installation of power panel 1PP102 to supply power for welding equipment has no adverse affect on any plant systems or components.
FSAR Chapters 6 and 15 were reviewed and this installation would have no affect upon, and will therefore not increase the probability of occurrence or the consequences of any accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. FSAR 15.2.6 and 15A, Event 28, discuss accidents/transients related to the loss of AC power. Failure of the power panel or power feed cables could only result in the trip of the 1B130-032 breaker which is dedicated for this one non-Class 1E application.
Therefore, the installation of the power panel for the welding equipment and its tie to 480VAC load center 1B130 breaker 032 could not create a possibility for a different type accident or malfunction as discussed in these sections.
III. No. The Tech Spec Basis for Electric Power Systems, Section 3/4.8 requires sufficient power be available to supply safety related equipment required for safe shutdown and the mitigation and control of accident conditions. Also, power services to maintain the facility in the shutdown or refueling conditions for extended periods and sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The AC power supplies necessary to meet these requirements are listed in Sections 3/4.8.3.1 and 3/4.8.3.2. 480VAC load center 1B130 is not listed as necessary, nor will the power panel 1PP102 provide power required for operation or shutdown of the unit or for mitigation of the consequences of an accident. Therefore, the margin of safety is not reduced.
ling '3-090 C E E C NL-93-016, Unit2 0 C The proposed action is to continue system operation with possible leakage by the isolation valves for Unit 1 and Unit 2 Condensate Storage Tank (CST) berm water to the storm sewer.
The mode of operation will be an interim condition until a positive means can be installed to prevent any possible leakage.
I. No. The proposed action can have no affect on the accidents or the equipment important to safety as evaluated in the FSAR. The only effect is the possible release of water to the environment. Even if the CST would fail and discharge its entire contents, the offsite doses are a small fraction of those postulated for a liquid radwaste tank failure (FSAR section 15.7.3). This proposed action does not affect the probability or consequences of any radioactive liquid release accident evaluated in the FSAR.
Section 11.2 of the FSAR describes the liquid waste management system and notes that the liquid waste management system has no nuclear safety related function.
Section 11.2.1 describes the design features which prevent uncontrolled releases of outdoor storage tanks, such as tank level, alarms on overflow condition, and diked areas. This action does not affect any of those functions. Table 11.2-16 does identify that CST berm sumps can be routed to either liquid radwaste or a storm sewer.
II. No. This type of malfunction is evaluated in 15.7.2.1 Miscellaneous Small Release Outside Containment. The resulting offsite dose is negligible in comparison to the dose resulting from the postulated leakage from a liquid radwaste tank failure. Thus this action is evaluated in the FSAR and does not create the possibility for an accident or malfunction of a different type.
III. No. The potential dose consequences from this action are all many orders of magnitude below the Technical Specification limit of 15 mrem/year, thus the margin for safety as defined in the Technical Specification basis is not reduced.
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!jgP gg '3-091 C S E E E C DCP92-3023E, Unit1 C G The proposed action is to provide centrifugal fans (with built in filter) to blow ambient air into the Electrical Protection Assembly (EPA) panels. This enhancement should extend the operating life of the EPA components by lowering the assembly internal air temperature.
SMlHIQlkBY'.
No. This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Sections 7.2, 15.1, 15.2, 15.3, 15.4, and 15.6.
The proposed action improves the environmental operating conditions inside the EPA panels by removal of internal heat utilizing forced ventilation. This modification does not increase the probability of occurrence of a malfunction of equipment. The proposed forced ventilation modification is expected to extend the operating life of the electrical components and reduce the probability of occurrence of a malfunction of the EPAs.
This modification does not increase consequences of an accident as previously evaluated in the FSAR. Accident evaluations (FSAR 7.2.2, 15.1, 15.2, 15.3, 15.4, 15.6) take credit for Reactor Protection System (RPS) operation and result in conservative results in all cases.
II. No. This modification does not create the possibility of an accident of a different type.
FSAR Chapter 15 discusses events and accidents which activate those sensors which input the RPS and result in reactor trip. The proposed modification does not involve changes in system operation or contribute to a different type of failure mode.
This modification does not create the possibility of a malfunction of a different type.
Forced ventilation reduces the panel internal temperature below specified operating limits.
III. No. Technical Specification Section 2.2.1 defines RPS instrumentation trip setpoints and allowable values. This modification does not affect trip setpoints. Section 3.3.1 defines limiting conditions for operation of RPS instrumentation. This modification does not affect the minimum number of operable channels required for operation.
Since no changes are being made to the EPA protective circuitry, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
@gal5g '3-092 C E C DCP 93-9011, Unit 2 ESC C DCP 93-9011 relocates Area Radiation Monitor (ARM) RIT-23711 to Room li-508 on El. 749'djacent to the new Reactor Building Sample Station (RBSS). This change represents a direct one-for-one hardware move in that the existing radiation monitor is physically removed from its present location and reinstalled at the new Reactor Building Sample Station (2C210A,B) location.
~SU HHlhBY'.
No. The area radiation monitoring system is described in FSAR Section 12.3A.1. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15, or in the Design Assessment Report (DAR). The hardware relocation provides monitoring in the vicinity of the relocated RBSS and returns the system to its original design basis. Potential failure modes of the relocated ARM have not changed. This action has no effect that could increase the probability of occurrence of any accident evaluated in the FSAR.
This modification does not increase the probability of occurrence of a malfunction of equipment since it is relocating existing equipment to return the system to original design basis.
This modification does not increase the consequences of an accident since failure of the relocated ARM will have the same effect on the system as it would have before relocation and will in no way affect the offsite dose to the public. In fact it will provide quicker warning of instrument line break at RBSS with potential to reduce overall consequence of that event.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since a simple relocation of existing equipment is being made.
Potential failure modes of the relocated ARM are the same as those at the existing location. The ARM equipment is supported seismically and conduit work or rework is permitted in accordance with procedures which assure the relocation does not impact any safety related equipment in the Reactor Building. Thus the proposed relocation does not create the possibility of a accident of a different type than previously evaluated in the FSAR.
This modification does not create the possibility of a malfunction of a different type since all equipment is the same. Failure of the relocated equipment would have the same effect on the ARM system as failure of the equipment in its current location.
III. No. Plant Technical Specification limiting conditions for operation 3.3.7.1 "Radiation Monitoring Instrumentation" specifically addresses the main control room outside air intake radiation monitors and the area criticality monitors which include the new fuel storage vault and spent fuel storage pool area monitors. The Technical Specifications do not specifically address the RBSS ARM instrumentation.
The Technical specifications do not govern operation of the RBSS area radiation monitor. The relocation returns the ARM system to original design intent. Thus this proposed action does not affect the margin of safety defined in the basis for any Technical Specification.
C OSS C NL-92-020, Unit 182 C
evaluation of EO-100-113 which gives procedural guidance for mitigation of Anticipated E'afety Transient Without Scram (ATWS) events.
QQJgJQ~RY'.
No. Although Section 15.8 of the FSAR presents a brief discussion of ATWS, this accident is beyond the plant design basis, and therefore this section of the Safety Evaluation is not applicable.
II. No. The proposed actions are only executed after an ATWS event has occurred. Therefore, these actions are only executed when the plant is in a configuration which is beyond the design basis. Consequently, this section of the Safety Evaluation is not applicable.
III. No. The proposed actions do not affect any Technical Specification requirements.
Qgf~Q '3-094 C 0 E C NL-92-009, Unit Common Sulfur Hexafluoride (SFG) Tracer StudyNerification of STREAM: TP-099-012. This test simulates a release from the standby gas treatment system vent. However, the vent system is not actually involved in the test (i.e., no material is introduced into the vent system).
S~5lhBY'.
No. The maximum concentration of SF6 in the turbine building would be 42 ppm following a cylinder rupture. This concentration is insufficient to poison the offgas recombiner catalyst or adversely affect reactor water chemistry. This same concentration would depress the reading of the hydrogen analyzers by 5 ppm. This is a discrepancy of 0.05% of the setpoint value of the analyzers (1% hydrogen) and is therefore insignificant. A cylinder rupture in the control structure would result in a concentration of 314 ppm SF6 inside the control structure. This quantity poses no threat to control room habitability. Therefore, this test does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. All plausible accidents and malfunctions have been considered. This test consists only of releasing and measuring the concentration of SF6 and inert gas. This analysis assumes the worst case in that all SF6 is released into the plant. This worst case scenario results in no adverse safety impacts or effects on plant operation. The test does not create a possibility for an accident of malfunction of a different type than any evaluated previously in the FSAR.
III. No. The quantity of SF6 which could get into the offgas system is so small (42 ppm) that catalyst deactivation in the offgas recombiner will be insignificant. Therefore, the design of the offgas treatment system, as stated in Section 3.11.2.6 of the Unit 1 and Unit 2 Tech Specs, will not be affected. The sensitivity of the offgas hydrogen analyzers will be reduced by approximately 5 ppm in a worst-case scenario. This discrepancy is 0.05% of the analyzer's setpoint value and is therefore insignificant. Thus, the requirements of Section 3.3.7.11 of the Unit 1 and Unit 2 Tech Specs will not be compromised. Therefore, the requirements of Section 3.3.7.8 of the Unit 1 and Unit 2 Tech Specs will not be compromised. There are no other effects on Tech Spec requirements.
~ I4" Qgg gg '3-095 C E C NL-93-018, Unit 1'ESCR TIO C TP-162-025, Reactor Vessel Level Control During Feedwater Pipe Replacement.
TP-162-008, Isolation of Reactor Vessel Level Instrumentation.
The purpose of procedures TP-162-025 and TP-162-008 is to establish the plant conditions necessary to effect the repair and to return the plant to its as found condition.
I~ No. Based on the review of Chapter 15 of the FSAR, performance of TP-162-025 and TP-162-008 does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. Feedwater pipe replacement will be performed with Unit 1 defueled and the refueling gates installed. In addition, no fuel movement on Unit 1 or Unit 2 will be allowed. AII equipment removed from service will be in accordance with Technical Specifications. The reactor vessel barrier will maintain Zone 3 (and Zone 2) integrity. In-leakage rates to Zone 3 are well within the total in-leakage margin available and this will have a negligible effect on Standby Gas Treatment System (SGTS) drawdown requirements.
II. No. The plant conditions in effect during the performance of TP-162-025 and TP-162-008 (i.e., Unit 1 defueled, refueling gates installed, no fuel movements on Unit 1 or 2 allowed) will not create the possibility of an accident of a different type than evaluated previously in Chapter 15 of the FSAR. Secondary containment integrity (Zone 2 and 3) will not be adversely affected, and spent fuel pool water inventory will not be reduced.
III No. All equipment affected by performance of the proposed test procedures is either not
~
required to be operable, not affected, or the applicable LCO action requirements are met. The ability of secondary containment to minimize off-site releases of radioactivity is not compromised. Various leakage paths were evaluated that will exist between Zone 3 and Zone 1 during performance of TP-162-025. This additional leakage is a small percentage of the total in-leakage margin available. Thus, total secondary containment inleakage (including the additional leakage created by TP-162-025) will remain within the leakage assumed by Technical Specifications. In addition, the ability of SGTS to meet drawdown requirements is not compromised due to the negligible increase in in-leakage. Thus the margin of safety is not reduced. The applicable Technical Specification sections include 3.3.1, 3.3.2, 3.3.3, 3.3.4.1, 3.3.7, 3.5.2, 3.6.5.1, 3.6.5.3, 3.7.3.
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DCP 93-3045, Unit Common 0 C DCP 93-3045 will replace the Remote Data Analysis System (RDAS) PRIME 750 computers and their disk drives, located in the EOF computer room, with PRIME 9955-II computers, new disk drives, cartridge tape subsystems to automate system backups, and statistical multiplexers with additional channel capacity to provide terminal access capability for four additional terminals for Reactor Engineering personnel located in the S8A building.
I. No. RDAS is not safety-related (FSAR Section 7.5 c and Section 7.7.1.7.1.2). There are no safety-related systems in the EOF and the equipment does not interface with any safety-related equipment in the plant. The RDAS modifications do not change the POWERPLEX core monitoring software calculations of safety-related parameters which are monitored per Technical Specification surveillance requirements. RDAS is not essential for safe shutdown of the plant.
Therefore, the modification by DCP 93-3045 does not increase the probability of occurrence or consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. There are no safety-related systems in the EOF and the equipment does not interface with any safety-related equipment in the plant. The RDAS modifications do not change POWERPLEX core monitoring software calculations of safety-related parameters which are monitored per Technical Specification surveillance requirements. RDAS is not essential for safe shutdown of the plant. RDAS is not safety-related (FSAR Section 7.5 c and Section 7.7.1.7.1.2).
The failure modes after this modification remain the same, for the same systems, as that before the modification. The modification will in no way affect any safety-related systems. Therefore, the proposed modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. RDAS is not addressed in the Technical Specifications. RDAS does not perform any safety functions and does not interface with any equipment which performs a safety-related function.
The POWERPLEX Core Monitoring System software is used to monitor Technical Specification fuel thermal margin requirements, e.g. MCPR, LHGR (ref. Technical Specification Section 3/4.2 Power Distribution Limits).
The proposed hardware and software changes to RDAS do not affect or change the POWERPLEX fuel thermal margin calculation methodology (i.e. the calculation to determine the margin of safety) and will reduce the calculational time required to determine if fuel thermal limit violations exist, as will be verified by post installation parallel testing. Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
S~gg '3-097 C ENC SCP J-91-1045, Unit Common D C OFC G Setpoint Change Package SCP J-91-1045 increases the maximum allowable setpoint of the Reactor Building Ventilation System filter differential pressure switch to 1.4" WC. An instrument setpoint of 1.4" WC and a tolerance of + 0.1" WC would ensure that the annunciation occurs well below the maximum allowable differential pressure.
S I~ No. The setpoint changes will cause no deviation from the flow rate limits of conditioned air previously cited in the FSAR. The SCP requires no physical change to the plant design, function, accuracy, response time, reliability, actuation logic or to any other component or system. The SCP also does not require any physical modification to the switches or to any other component. The setpoint change is intended to allow a greater pressure differential across the filter and thereby reduce the costly maintenance requirements associated with frequent filter changes. The increased differential has no adverse effect on the filter or the fan system. A review of Section 9.4.3 of the FSAR indicates that the proposed change would neither change the performance of any equipment as described in the FSAR nor would it cause air flows to be outside those described in the FSAR. Therefore, the setpoint change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. A review of Section 9.4.3, Radwaste Building Ventilation System, of the FSAR indicates the setpoint change does not affect the design or function of the high differential pressure switch or of any other components. Also, the change does not cause any component or system to operate outside the performance boundaries described in the FSAR or Technical Specifications. Therefore, the setpoint change does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III No.
~ Neither the Radwaste Building Supply System nor the high differential pressure switch of this system is governed by the Technical Specification. Therefore, there is no reduction in the margin of safety as defined in any of the basis of the Technical Specification.
C DCP 90-3042, Unit 2 C
GE'he purpose of this change is to retire the portions of the Reactor and Turbine Sampling systems that were abandoned as a result of DCP 90-3025, which installed new Unit 2 Turbine and Reactor Building Sample Stations (TBSS and RBSS).
I~ No. None of the equipment or piping that is being removed from the plant by this DCP has an affect on equipment or systems that are important to safety. All of the equipment and piping that are being affected by this DCP was abandoned under previous DCP's.
This DCP does not increase the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety.
II. No. This DCP does not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. This DCP removes equipment and piping from the plant that has been abandoned as a result of DCP 90-3025. This DCP does not create the possibility of a different type of accident or malfunction because the equipment is isolated and not functional. The removal of this equipment actually reduces the possibility for an accident or malfunction of any type because it eliminates unnecessary equipment and valves.
III. No. Technical Specification 3/4.6.3 requires that the Reactor Building Closed Cooling Water containment isolation valves be operable. This DCP will have no effect on the operability of these valves since the work is being completed on a small branch line that will be isolated from the system while the work is being completed. Technical Specification 3/4.6.5 requires that secondary containment integrity be maintained. A register is being installed in the ductwork that is being disconnected from the RBSS to assume that proper airflow balance is maintained. Technical Specification 3/4.11.2.5 discusses that operability of the ventilation exhaust gas treatment systems.
Implementation of DCP 90-3042 does not require that the Zone II equipment compartment exhaust system be taken out of service. This DCP does not reduce the margin of safety as defined in the basis for any Technical Specification.
SEK59 83-O8 8 C SS E ERE CE 'SCP-J91-1015A8B, Unit Common Control Room and Control Structure HVAC:
The setpoint changes will increase the setpoints for the "Instrument Failure" function above the maximum flow range of the instrument, which is far above the maximum system flow rate, but yet sufficiently conservative to ensure that the instruments trip on an actual "Instrument Failure".
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I~ No. All safety-related Control room and Control Structure systems are designed to maintain functional integrity during a design basis accident. The systems are part of the Control Structure Habitability System (described in Section 6.4) which is specifically defined as an Engineering Safety Feature by the FSAR (Section 7.3.1). Implementation of this setpoint change has no affect on the functions of the equipment other than to improve the reliability of the low-flow switches. Therefore, this setpoint change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. This setpoint change does not affect the design or operation of the Control Room and Control Structure low-flow switches or HVAC fans. The setpoint change does not introduce a new, unanalyzed type of malfunction. As mentioned above, this change increases system reliability by eliminating the possibility of erroneous trips caused by faulty operation of the low-flow switches. This has the net affect of reducing the probability of an accident due to the low-flow switches. Improved reliability of a safety system does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The Control Room Emergency Outside Air System (CREOAS) is the only Control Structure Habitability System addressed by the Technical Specifications (Section 3/4.7.2). Implementation of this setpoint change will not adversely affect the operation of this system. The setpoint change enhances the reliability of the equipment, which under certain situations, receives its supply air from the CREOAS system. This change increases the reliability of the Control Structure HVAC System, and therefore, the Control Structure Chillers. It does not reduce the margin of safety as defined in the basis for any Technical Specification.
p E CE DCP 93-3040A, Unit Common D SC C E'his modification will cross-tie the domestic water header, which is supplied by well water, to the Clarified Water Storage Tank which will supply seal water to the circulating and service water pumps.
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FSAR Chapters 6 and 15 have been reviewed with particular attention paid to FSAR Section 15.2.3 and 15.2.5 for possible adverse consequences with respect to the installation of this modification. Loss of seal water to the Circulating Water Pumps could cause them to trip resulting in a turbine trip and loss of condenser vacuum. This modification will not result in an increase in the probability of occurrence of these events due to the storage capabilities of the Well Water and Clarified Water Storage Tanks. The probability that a component failure which would prevent the Potable Water or Raw Water Treatment from performing their design function is not increased by installing the cross-tie due to the reserve water which is stored and the ability to restart the clarifier. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The Potable Water and Raw Water Treatment Systems are non-radioactive and are not tied in any way to a radioactive system. The proposed action does not involve a precursor of or a significant contributor to any evaluated accidents involving offsite dose. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification will cross-tie the Potable and Raw Water Treatment Systems so that the clarifier can be shut down. The new 2" cross-tie pipe and back flow preventer will be designed to meet all requirements of ASME/ANSI B31.1 Code to ensure its structural integrity, and guarantee that the Potable and Raw Water Treatment Systems will be able to operate and function as speciTied in the FSAR. The proposed action will not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modiTication.
The modification will cross-tie the Potable to Raw Water Treatment Systems so that the clarifier can be shut down. The cross-tie pipe and back flow preventer will be designed to meet all applicable codes; consequently, the actions taken by this modiTication will not reduce any margin of safety as defined in the basis for any Technical Specification.
SEE.5Q '3-101 C E C DCP 92-9018Z, Unit 1 D C The proposed action is to add redundancy to each Unit 1 4.16kV bus degraded voltage timer reselect logic and to delete the Startup Transformer T10 and T20 voltage boost scheme.
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No. The proposed action does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of redundancy into each Unit 1 4.16kV bus degraded voltage timer reselect logic does represent an increase in the probability of occurrence of a malfunction of equipment. However, based on engineering judgment, the increase in probability is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations. The redundancy further assures that the reselect logic operates for a random single failure in either division of the Plant Auxiliary Load Shed Logic should the single failure occur coincident with the accident.
Thus, the addition of redundancy to each Unit 1 4.16kV bus degraded voltage timer reseiect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis and NUREG-0776 were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events for which transients or anticipated operational occurrences and accident conditions were analyzed.
The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. And, the addition of redundancy to each Unit 1 4.16kV bus degraded voltage timer reselect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The operability of the Degraded Voltage Timer Reselect Logicis governed by Technical Specification Section 3/4.3.3 with tables establishing the required number of operable channels, the setpoints and response times. The proposed change does not affect the required number of operable channels, the setpoints or the response times. The operability of the Startup Transformer T10 and T20 tapchanger is governed by Technical Specification Interpretations numbers 1-93-002 and 2-43-002. Deletion of the voltage boost scheme has no impact on the margin of safety as defined in the Technical Specifications. Thus, the addition of redundancy to each Unit 1 4.16kV bus degraded voltage timer reselect logic and the deletion of the Startup Transformer T10 and T20 voltage boost scheme does not reduce the margin of safety as defined in the basis for any Technical Specification.
A SEEKER~ '3-102 cos E E C DCP93-3047D, Unit1 C C DCP 93-3047D: Non-outage portion of DCP 93-3047 Reactor Vessel Water Level Reference Leg Backfill:
Includes the fabrication and installation of the new instrument racks and associated tubing connecting them to the Control Rod Drive (CRD) tap-offs installed per the "A" package, and to the Q/non-Q and ASME/ANSI boundary valve assemblies installed per the "B" and "C" packages.
I. No. The only adverse effect that could be postulated to result from this modification is an instrument rack or tubing becoming loose during a seismic event and colliding with any safety related equipment in the rooms where it is being routed. This could damage the safety related equipment and prevent it from performing its safety functions. However, this scenario is not considered credible. The tubing will be installed with considerations for safety impacts, which includes the use of Q supports. The racks have been designed to meet seismic requirements. Therefore, the modification addresses seismic concerns and does not increase the probability or consequences of an accident or malfunction.
II. No. Seismic concerns have been addressed in this modification to ensure that the installed instrument racks and tubing will not adversely affect any existing equipment. This ensures that no new accidents or malfunctions are created as part of this modification.
III No. The installed instrument racks and tubing will not connect to any existing systems or
~
components. Seismic concerns have been addressed to ensure that no adverse safety impacts are introduced to existing equipment. Therefore, the margin of safety as defined in the bases for any Technical Specifications is not affected by this modification.
F E E C DCP 93-3048D, Unit 2 D C DCP 93-3048D: Non-outage portion of DCP 93-3048 Reactor Vessel Water Level Reference Leg Backfill:
Includes the fabrication and installation of the new instrument racks and associated tubing connecting them to the Control Rod Drive (CRD) tap-offs installed per the "A" package, and to the Q/non-Q and ASME/ANSI boundary valve assemblies installed per the "B" and "C" packages.
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I. No. The only adverse effect that could be postulated to result from this modification is an instrument rack or tubing becoming loose during a seismic event and colliding with any safety related equipment in the rooms where it is being routed. This could damage the safety related equipment and prevent it from performing its safety functions. However, this scenario is not considered credible. The tubing will be installed with considerations for safety impacts, which includes the use of Q supports. The racks have been designed to meet seismic requirements. Therefore, the modiTication addresses seismic concerns'nd does not increase the probability or consequences of an accident or malfunction.
II. No. Seismic concerns have been addressed in this modification to ensure that the installed instrument racks and tubing will not adversely affect any existing equipment. This ensures that no new accidents or malfunctions are created as part of this modification.
III. No. The installed instrument racks and tubing will not connect to any existing systems or components. Seismic concerns have been addressed to ensure that no adverse safety impacts are introduced to existing equipment. Therefore, the margin of safety as defined in the bases for any Technical Specifications is not affected by this modification.
QP~Q '3-104 cos C . DCP93-3061, Unit1 C
a mass ring on the Unit E'nstall 1 generator coupling to the low pressure (LP) 'C'ain turbine rotor.
I~ No. FSAR Section 3.5.1.3 and responses to NRC questions 121.18 through 121.21 address the probability of a catastrophic wheel failure for a "shrunk on wheel" rotor.
These sections conclude that the probability of unacceptable damage to safety related components due to turbine missiles is 2.61E-10. (This FSAR section states that this probability is reduced with the monoblock rotors currently installed in the turbine).
FSAR Section 10.2 provides the design bases of the Turbine/Generator and discusses the protective functions of the steam valves, overspeed protection, and turbine disk integrity. The analysis performed in this section is valid for the addition of the mass ring.
FSAR Section 15.2.2 and 15.2.3 address generator load rejection and turbine trip. The modiTication does not impact the analysis performed in these sections. Therefore, the modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. The mass ring represents a potential new missile; however, GE documentation states that the analysis performed in FSAR 3.5.1.3 and 10.2 bound this modiTication. There are no other potential failure modes introduced by the modification. Therefore, the modification does not create a possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. Tech Spec section 3/4.3.8 ensures that the turbine overspeed protection system instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. The additional mass reduces the potential for overspeed. The proposed modification does not affect the basis of Tech Spec section 3/4.3.9 "Feedwater/Main Turbine Trip System Actuation Instrumentation." Tech Spec section 3/4.7.8 "Main Turbine Bypass System" is not affected by this modification.
Therefore, there is no reduction in the margin of safety as defined in the basis for any Tech Spec.
C OSS C DCP 93-3064, Unit 1 C C E Modify the ¹4, 6 and 8 main turbine bearings and bearing housings as follows: Drill and tap two 1/2" diameter by 1.469" deep holes, split equally apart from the vertical centerline of the bearing by 2.25", at twelve o'lock; Drill and tap the lower bearing housing to accept a sealing gland with 3/8" NPT threads.
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No. FSAR Section 3.5.1.3 and responses to NRC questions 121.18 through 121.21 address the probability of a catastrophic wheel failure for a "shrunk on wheel" rotor.
These sections conclude that the probability of unacceptable damage to safety related components due to turbine missiles is 2.61E-10. Based on the low probability the turbine missile hazard is not considered a design basis event. The modification does not impact the analysis performed in FSAR sections 10.2, 15.2.2, 15.2.3. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. A failure of the spot welds or fasteners on the torque collars or the bolting on thebox or brass ring antenna would result in these parts becoming loose inside the bearing housing. Based on the bearing/shaft clearances and the size of these components there is no possibility of this malfunction causing detrimental shaft or bearing surface damage. Bearing oil drains are sufficiently sized such that no one piece or combination of pieces could block oil flow from the affected bearing. The probability of the tack welds creating a stress riser in the bearing is negligible. There are no other potential failure modes introduced by the modification. Therefore, the modification does not create a possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. Tech Spec section 3/4.3.8 ensures that the turbine overspeed protection system instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. The proposed modification will not reduce the margin between anticipated wheel failure and overspeed protection system instrument settings. Tech Spec section 3/4.3.9 is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand. The proposed modification does not affect this Tech Spec. Tech Spec 3/4.7.8 "Main Turbine Bypass System" is not affected by this modification.
Therefore, there is no reduction in the margin of safety as defined in the basis for any Tech Spec.
C OS EFE E C DCP91-9071, Unit1 C 0 C E'his modification will replace the existing undervoltage and 'overvoltage relays with Class 1E C8 D relays, Model MBC-4650T. The existing overvoltage relays are Model MBC-4650T but are installed as non-Class 1E.
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No. FSAR Table 8.3-21 evaluates the individual components that make up the 125 VDC System. The undervoltage and overvoltage relays are not singled out or evaluated specifically in the failure mode analysis. The relays provide no system protective function. Upon evaluation of the application of these relays and proposed setpoints, no increase in the probability of an accident was found to occur.
The new C8D relay has equal or greater selectivity, repeatability and reset capability than the existing GE and C&D relays. The relay replacement coupled with the MDR series isolation relay and setpoint changes will align the system more to its design intent. It is therefore concluded that the replacement of the undervoltage and overvoltage relays and application of setpoints will not increase the probability of malfunction of equipment.
There are no other failures in the 125 VDC system which can be postulated as a result of the proposed changes. Therefore, it is concluded that there is no increase in the consequences of an accident.
II. No. The function of the new relays has not changed from that of the relays being replaced.
The setpoints were chosen to provide early indication of system trouble. No safety impacts exist for the proposed changes to the 125 VDC equipment. Therefore, there is no possibility of creating an accident of a different type.
An MDR Series relay, which is accepted as an approved electrical isolation device per SSES Specification E-1012 and FSAR Section 8.1.6.1.q.2 will provide the interface between the Class 1E control circuits and the non-Class 1E annunciator circuits. As such, there are no situations that could create the possibility of a malfunction of a different type.
III. No. The effect of the modifications on the current Division I and II 250 VDC System battery design margins was determined. It was determined that the modification has a minimal increase in the load of the 250 VDC System, but has no reduction in the margin of safety that is the basis for the load profiles presented in SSES Technical Specification Section 4.8.2.1.d.2(c).
Qgg ~ 93-107 C EFE E C NL-93-001, UnitCommon DE C The changes made to the FSAR in change notice 1880 updates Section 9.1.3 and Appendix 9A.
This update reflects changes in fuel design parameters and refueling sequence from those originally anticipated and utilized to design the fuel pool and cooling systems.
I. No. Appendix 9A is the only FSAR design base accident affected by the FSAR change 1880. The changes made therein are related to the consequences of a loss of fuel pool cooling event and not the causes of such an accident. Therefore, the probability of occurrence of a loss of fuel pool cooling event is not increased. The radiological consequences of the changes were evaluated. These values are well below the guideline values of 10CFR100 and the 1.5 rem thyroid guideline of regulatory guide 1.29. The changes are not considered to increase the consequences of the loss of fuel pool cooling event.
II. No. The changes made in FSAR change 1880 relate to the consequences of a loss of fuel pool cooling event. They do not relate to the causal factors of such an event. Thus, the changes made in FSAR change 1880 do not create a possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.
III No.
~ Emergency Service Water makeup has been shown to have sufficient capacity. The radiological impact has been shown to be small and below the regulatory limits established in NUREG-0776. Therefore, the margin of safety is not affected by the changes made to the FSAR in change 1880.
S Qgf~Q '3-108 CRO E E ENC DCP93-3023, Unit1 C C G Modify HV-144F004 by replacing the actuator, motor, yoke and yoke clamp. This change also involves the replacement of the circuit breaker, resetting of the breaker magnetic trip and the installation of a new overload heater coil in the associated motor control center cubicle of the motor operated valve (MOV). The modification also includes the deletion of the space heater circuit to the valve motor and limit switch compartment.
I. No. The modification has no impact on the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. The valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time is within the design basis. This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR Sections 6.2.4, Table 6.2-12 and 6.2.4.3.2.6 are still valid.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-144F004 to perform its intended design function; 2. the stroke time of the valve is within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability of the valve to open or close when required; 6. does not affect the valve leakage; 7. does not alter the DC control center seismic qualification; 8. does not affect the 250 VDC motor power supply system; 9. does not affect any other equipment in the Reactor Water Cleanup (RWCU)
System. The proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety defined in the bases for any Technical Specification. Tech Spec Bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, and 3/4.3.2 were reviewed for this evaluation. Valve seat leakage and integrity are not affected.
The valve's stroke time remains within the accident analysis bounds of 30 seconds.
The MOV's decreased design stroke time improves system response by decreasing the amount of potential leakage through the penetration. This modification has no effect on the RWCU intended design function or the basis of the above Tech Specs.
C . DCP93-3022, Unit1 D C C GE:
Modify HV-144F001 by replacing the actuator/motor, yoke and yoke clamp. The scope of this change also involves replacement of the circuit breaker, the resetting of the breaker magnetic trip, installation of a new overload heater coil in the associated motor control center cubicle and the installation of current monitoring equipment in the limit switch compartment for VOTES testing.
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No. The modification has no impact on the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. The valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time is within the design basis. This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR Sections 6.2.4, Table 6.2-12 and 6.2.4.3.2.6 are still valid.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-144F001 to perform its intended design function; 2. the stroke time of the valve is within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability of the valve to open or close when required; 6. does not affect the valve leakage; 7. does not alter the MCC or the valve limit switch compartment seismic and environmental qualification; 8. does not affect the motor power supply system; 9. does not affect any other equipment in the Reactor Water Cleanup (RWCU) System. The proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety defined in the bases for any Technical Specification. Tech Spec Bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, and 3/4.3.2 were reviewed for this evaluation. Valve seat leakage and integrity are not affected.
The valve's stroke time remains within the accident analysis bounds of 30 seconds.
The MOV's decreased design stroke time improves system response by decreasing the amount of potential leakage through the penetration. This modification has no effect on the RWCU intended design function or the basis of the above Tech Specs.
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DCP 90-3107A/B, Unit Common The proposed changes will remove the mapboards in the Security Control Center/Auxiliary Security Control Center (SCC/ASCC) and will restore the floor. The existing horns and the accompanying electronics in panel OC549A/B will be relocated to the Main Control Console (OC524A/B) in order to maintain the alarm function. New panel lights will be installed in OC524A/B to provide a visual indication to the console operator during any alarm situation. The proposed action eliminates incorrect/misleading representation of the security configuration of the plant while maintaining vital alarm functions.
I. No. The Security Data Management System has no safety related function. The removal of the Security Mimic Mapboards (OC549A/B) will not impact any functional requirements of any safety related components or equipment; therefore, no adverse effects are possible. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15.
I'he elimination of the mapboards does not impact any safety related equipment since no connection to any safety related circuitry exists. The relocation of the existing security horn and the addition of an indicator light will not interact with any safety related equipment. This modification will not adversely affect any structure, system, or component in performing its safety function. No common mode failures or adverse affects to any safety related equipment/circuits can result from these modifications. Therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR.
The proposed action does not involve a precursor of or a contributor to any evaluated accidents involving offsite dose. This modification can have no effect on the accidents that have radiological consequences. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. Removing a security mapboard and relocating/modifying an alarm function will not result in any risk to public health and safety. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible. The Security Plan has
,.been submitted as a separate document pursuant to 10CFR2.790(d) "Rules of Practice" (reference FSAR Section 13.6, Industrial Security). The proposed action in no way conflicts with security FSAR commitments. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. There is no limiting condition for operation in the Technical Specifications involving security operation. There are no surveillance requirements for the Security Data Management System.
There are no Technical Specification bases for any security equipment. The proposed changes will have no adverse affect on security operation. Therefore, the implementation of DCPs 90-3107A/B will not reduce the margin of safety as defined in the bases of the Technical Specifications.
C TP-134-038, Unit 1 C G The purpose of this procedure is to allow tie-in of a temporary outage chiller to the Reactor Building Chilled Water (RBCW) System during a Unit 1 Outage.
I. No. Sections 9.2.12.3 and 9.4.2 were reviewed. The FSAR does not specifically address the affected plant systems during outage conditions. However, installation of this temporary system will not impact operation of Primary Containment Isolation, the only safety function of the RBCW System. Therefore, this activity does not increase the probability or consequences of an accident previously evaluated in the FSAR.
II ~ No. The installation of this temporary system does not adversely impact the operation or safety function of any existing installed equipment. The only new failure possible would be failure of the temporary system itself. If the temporary chiller failed to provide cooling, there would be no new accident created since this failure is bounded by failure of the Reactor Building Chillers during plant operation. Since RBCW is a non safety related system, there is no impact on plant safety.
III. No. Tech Spec Section 3.6.3 addresses Primary Containment Isolation Valves and is not affected by the proposed action. Section 3.6.1.7 addresses Drywell Average Air Temperature during Conditions 1, 2, 3. Since the proposed activity will be performed during a unit outage, this section is not affected. Section 3.6.5.1 Secondary Containment Integrity is not affected since the installation and operation of the temporary chiller will be during periods when Secondary Containment (Zone 1) is not required or actions taken to ensure Secondary Containment operable. The margin of safety is not reduced for any Technical Specification basis.
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~Egg '3-112 CROSS C TP-234-036, Unit 2 DE C The purpose of this procedure is to allow tie-in of a temporary outage chiller to the Reactor Building Chilled Water (RBCW) System during a Unit 1 Outage.
I~ No. Sections 9.2.12.3 and 9.4.2 were reviewed. The FSAR does not specifically address the affected plant systems during outage conditions. However, installation of this temporary system will not impact operation of Primary Containment Isolation, the only safety function of the RBCW System. Therefore, this activity does not increase the probability or consequences of an accident previously evaluated in the FSAR.
II ~ No. The installation of this temporary system does not adversely impact the operation or safety function of any existing installed equipment. The only new failure possible would be failure of the temporary system itself. If the temporary chiller failed to provide cooling, there would be no new accident created since this failure is bounded by failure of the Reactor Building Chillers during plant operation. Since RBCW is a non safety related system, there is no impact on plant safety.
III No. Tech Spec Section 3.6.3 addresses Primary Containment Isolation Valves and is not
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affected by the proposed action. Section 3.6.1.7 addresses Drywell Average Air Temperature during Conditions 1, 2, 3. Since the proposed activity will be performed during a unit outage, this section is not affected. Section 3.6.5.1 Secondary Containment Integrity is not affected since the installation and operation of the temporary chiller will be during periods when Secondary Containment (Zone 1) is not required or actions taken to ensure Secondary Containment operable. The margin of safety is not reduced for any Technical Specification basis.
S~EgQ '3-113 C 'CP92-3031 Athru K, Unit2 D C 0 C GE'his proposed activity will make the necessary modiTications to existing pipe support/restraint configurations that are required in order to reduce the total number of mechanical snubbers located on selected piping systems. In addition, DCP 92-3031 A includes the addition of a permanent steam tunnel monorail. In addition, DCP 92-3031 F includes modifications to the reactor head insulation frame and associated pipe supports.
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~ No. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not increase with the implementation of the proposed action. The systems included in this proposed modification were reanalyzed to justify the removal of unnecessary snubbers and their pipe support configurations are to be modified accordingly. The reanalysis is in accordance with the original design basis and will be performed in accordance with accepted industry codes. All aspects of the existing qualification calculations have been addressed, all original interface design parameters such as equipment allowables have been considered and all system design requirements as specified in the Design Specifications and FSAR have been addressed in the reanalysis effort.
II. No. This proposed action will not create the possibility for an accident or malfunction of a different type than evaluated in the FSAR. The accident events/causes in FSAR Chapter 15.0 that are applicable to this proposed action involve equipment malfunction/failures and pipe breaks. The proposed action only reduces the number of seismic and hydrodynamic restraints (snubbers) on selected piping systems by reanalyzing each line using optimum restraint configurations.
III. No. The Technical Specification bases related to these proposed modifications are: 3/4 4.7; 3/4.5.1; 3/4.7.3; 3/4.5.2; 3/4.4.9; 3/4.6.2; 3/4.9.11; 3/4.3.7; 3/4.6.1; 3/4.1.5; 3/4.7.1; 3/4.4.8; 3/4.6.1; 3/4.6.3; 3/4.7.4. The proposed action will not affect the intended safety functions. The proposed action will not reduce the margin of safety as defined by the Technical Specifications noted.
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~SE gO 93-114 C E CE DCP 92-9035Z, Unit 2 DE C The proposed action is to add a permissive control logic into the automatic close circuitry of the non-Class 1E 13.8 KV startup incoming feeder breakers 2A10104 and 2A10204 which blocks closing of the breaker for a Unit 2 LOCA until the Unit 2 13.8 KV pump loads are shed by the Unit 2 13.8 KV bus load shed logic.
I~ No. The addition of a permissive control logic does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR, the design assessment report, the current Reload Analysis or NUREG 0776. Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of a permissive control logic does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay in the Unit 2 Division I and II Core Spray LOCA logic. Based on engineering judgment, the increase in probability due to additional relays is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of overall system malfunction. The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action The MDR isolation relays are qualified isolation devices.
II ~ No. Chapter 6 and 15 of the FSAR, the Design Assessment Report, the current Reload Analysis, and NUREG 0776 were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events for which transients or anticipated operational occurrences and accident conditions were analyzed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. The 13.8 KV startup incoming feeder breaker control schemes for breakers 2A10104 and 2A10204 do not comply with random single failure criteria. The proposed action does not create another malfunction which would contribute to creating a malfunction or coincident malfunctions of a different type.
III. No. The operability of the 13.8 KV startup incoming feeder breakers 2A1 0104 and 2A1 0204 and their control circuitry is not governed by Technical Specifications. The proposed action does not affect the operability requirements of the ECCS Instrumentation (Section 3/4.3.3) including Tables 3.3.3-1, 3.3.3-2 and 3.3.3-3. Thus, the addition of a permissive control logic does not reduce the margin of safety as defined in the basis for any Technical Specification.
Qgggg '3-115 C E C DCP 93-9030, Unit 2 C
GE'his modification will remove the Process Valve Stem Leakoff Collection System from all 14 of the Unit 2 bleeder trip valves (BTVs).
I. No. The removal of the Stem Leakoff Collection System from the BTVs and subsequent utilization of a single packing arrangement will serve to enhance the overall performance and reliability of the BTVs by eliminating additional frictional forces on the disc arm shaft assembly. The modification will satisfy all the design requirements of the ASME/ANSI B31.1 Code; thereby satisfying the design requirements as specified in FSAR Table 3.2-1. In addition, installation of single packing will improve the operation of the BTVs thereby lessening the probability of challenging equipment important to the safe shutdown of the plant. Consequently, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. The Process Valve Stem Leakoff Collection System is no longer necessary in order to meet off-site iodine release criteria of 10CFR50, Appendix I.
Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II ~ No. The existing BTVs have a history of technical problems related to their connection to the Process Valve Stem Leakoff Collection System. Deleting the stem leakoff connection to the BTVs will have no adverse affects on any system or system functions and will increase system performance and improve the environmental conditions of the area in which the BTVs are located. The proposed modification does not create a possibility for an accident or modification of a different type than any evaluated previously in the FSAR.
III. No. Removing the valve stem leakoff from the BTVs will have no adverse effects on the Extraction Steam, the Feedwater Pump Turbine and the Turbine Steam Seals and Drains Systems; to the contrary, the reliability of these systems and the BTVs will be greatly enhanced by this modification. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification; therefore, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
E E C DCP 93-9042, Unit Common C C This modification will replace 12" valves 022458 and 022459 in a 12" KBF line with a back-flow preventer and add a back-flow preventer at valve 022472 in a 1 1/2" KBF line per the requirements and directions of Part Vll of the Public Water Supply Manual.
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~ No. This modification merely installs effective back-flow prevention methods for cross-connections in accordance to DER requirements as outlined in Section Vll of the Public Water Supply Manual. FSAR Chapters 6 and 15 and Fire Protection Review Report Section 6.0 have been reviewed and based on this review, this modification will not increase the occurrence of an accident previously evaluated in the FSAR. The modification meets all of the design requirements in the FSAR, and will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. The proposed modification does not involve a precursor of or contributor to, any evaluated accidents involving offsite dose.
Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification adds acceptable back-flow prevention devices to the back-up Fire Protection System which meet all of the requirements of the National Fire Protection Code, the KBF Class piping of DWG M-199, and Section Vll of the Public Water Supply Manual. As such, this modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by the Unit 1 or 2 Technical Specifications, especially Section 3/4.7.6, Fire Suppression Water System. Installation of the back-flow preventers at the discharge of the Well Water Pumphouse Fire Pumps will impose minor flow restrictions, but will not reduce their capability for the back-up fire protection below design basis requirements. The modification to the back-up Fire Protection System meets the original design basis as specified in the FSAR, consequently, the actions taken by this modiTication will not reduce the margin of safety as defined in the basis for any Technical Specification.
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C DCP 93-3047, Unit 1 C
GE'o prevent significant non-condensible gas accumulation in the reference leg, a'backfill system will provide a continuous flow of reactor grade water into the reference legs connected to condensing chambers XY-14202 and XY-B21-1D002.
I~ No. The proposed action does not increase the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
The proposed modification will tie the Control Rod Drive (CRD) system discharge water header to instrument racks 1C224/1C225. This will provide a flowrate which will force non-condensible laden water back into the RPV. The only adverse effects that could be postulated to result from this modification is a reduction in reference leg integrity due to leakage, or adverse effects on the instrumentation from these legs. Both postulated effects have been evaluated. Design considerations ensure that the integrity of the reference leg is maintained during all modes of operation and that no adverse effects on plant equipment are introduced. The reliability of the level indication system is actually enhanced, and the backfill system will ensure that reference leg volume remains unaffected by non-condensible gases.
II. No. It may be postulated that a new equipment malfunction scenario could be introduced by the modification by leakage through all check valves coincident with loss of CRD pressure. Such a failure could disable both level channels, defeating safety systems and indications required during and after accident scenarios. The accident analysis in the FSAR assume the loss of one instrumentation channel only. In addition to this, the non-Q portions of the system could create safety impacts on existing equipment due to break in secondary containment. The common mode leakage scenario is not considered credible however, since the probability of check valve leakage has been determined to be acceptably low. The non-Q side of the backfill system will be installed with Q supports and seismic spans, although the tubing is non-Q. This will ensure that a break in the non-Q side will have no safety impact on existing equipment. There is reasonable assurance that reference leg integrity will not be compromised by this modification and that no safety impacts on existing equipment are created.
III No. The proposed action does not reduce the margin of safety as defined in the Technical
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Specifications. The modification will actually increase the margin of safety by preventing non-condensible gas accumulation in the reference legs.
ggs.gg '3-118 e C C DCP 93-3048, Unit 2 To prevent significant non-condensible gas accumulation in the reference leg, a backfill system will provide a continuous flow of reactor grade water into the reference legs connected to condensing chambers XY-24202 and XY-B21-2D002.
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No. The proposed action does not increase the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
The proposed modification will tie the Control Rod Drive (CRD) system discharge water header to instrument racks 2C224/2C225. This will provide a fiowrate which will force non-condensible laden water back into the RPV. The only adverse effects that could be postulated to result from this modification is a reduction in reference leg integrity due to leakage, or adverse effects on the instrumentation from these legs. Both postulated effects have been evaluated. Design considerations ensure that the integrity of the reference leg is maintained during all modes of operation and that no adverse effects on plant equipment are introduced. The reliability of the level indication system is actually enhanced, and the backfill system will ensure that reference leg volume remains unaffected by non-condensible gases.
II ~ No. It may be postulated that a new equipment malfunction scenario could be introduced by the modification by leakage through all check valves coincident with loss of CRD pressure. Such a failure could disable both level channels, defeating safety systems and indications required during and after accident scenarios. The accident analysis in the FSAR assume the loss of one instrumentation channel only. In addition to this, the non-Q portions of the system could create safety impacts on existing equipment due to break in secondary containment. The common mode leakage scenario is not considered credible however, since the probability of check valve leakage has been determined to be acceptably low. The non-Q side of the backfill system will be installed with Q supports and seismic spans, although the tubing is non-Q. This will ensure that a break in the non-Q side will have no safety impact on existing equipment. There is reasonable assurance that reference leg integrity will not be compromised by this modification and that no safety impacts on existing equipment are created.
III. No. The proposed action does not reduce the margin of safety as defined in the Technical Specifications. The modification will actually increase the margin of safety by preventing non-condensible gas accumulation in the reference legs.
Qgg'gg '3-119 C C NL-93-023, Unit 1 & 2 C
To maintain open the interior large double doors and roll-up door assemblies in the Turbine Buildings. Doors covered under this scope are:
Unit 1 and Common: Doors 3 and 4A on Elevation 656', Doors 33, 35, and 37 Elevation 676', Door 221 on Elevation 699'nit 2: Door 44 on Elevation 656', Doors 34, 36, and 38 on Elevation 676',
Door 244 on Elevation 699'o.
Because of the fusible link closure configuration, maintaining the large double doors and roll-up doors open does not impact their function as fire barriers (T.S. 3.7.7).
This safety evaluation primarily addresses the impact, if any, of maintaining open the large double doors and roll-up door assemblies has on the Turbine Building HVAC System, which performs no safety-related functions (FSAR 9.4.4.1). The proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Maintaining the large double doors and roll-up doors open does not impact their function as safety barriers. The safety evaluation addresses the impact on the Turbine Building HVAC System. Three conditions were identified and analyzed and are covered in Section I. The proposed action does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The SSES Technical Specifications does not directly or indirectly discuss the operation of the Turbine Building HVAC System, which performs no safety functions. The proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
C DCP 92-3034, Unit 2 C G This modification relocates the limit switches to the bracket on the shielded side of the Main Steam Isolation Valve (MSIV) operator. For MSIVs A8B, one limit switch is relocated to the north bracket. For MSIVs C8D, three limit switches are relocated to the south bracket. A reinforcing plate is added to the A8 D MSIV shielded bracket. The brackets on the unshielded side of each MSIV will be removed.
~SU gg~Q I. No. The proposed action does not increase the probability of occurrence nor consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR since this modification does not effect the safety function and satisfies all design criteria of the MSIVs and the limit switches as described in the FSAR (FSAR and NUREG-0776 Sections 6.2, 6.7, 7.1, and 7.3).
By eliminating the potential loss of the limit switches due to jet impingement forces this modification decreases the consequences of recirculation piping ruptures (FSAR Section 3.6).
II. No. There are no new failure modes of equipment important to safety associated with the activities of this modification. The function and design criteria of the MSIVs and limit switches remain the same. The relocation of the switches from one side of the MSIV to the other does not create the potential for a new type of unanalyzed accident or a new type of malfunction.
III. No. This modification does not effect the operation or closing time capability of the MSIVs and thus does not reduce the margin of safety defined in the basis of Technical Specification 3/4.4.7.
Technical Specification 3/4.6.1.4 requires that two independent MSIV leakage control system (LCS) subsystems shall be operable. Since relocating the MSIV-LCS limit switch from one side of the MSIV to the other and the addition of an electrical quick disconnect does not effect the operation of the limit switch or the MSIV-LCS, it does not reduce the margin of safety defined in the basis of Technical Specification 3/4.6.1.4.
This modification improves the availability of the inboard MSIV-LCS by relocating the LCS limit switch to avoid jet impingement effects.
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g~g '3-121 C SS E C DCP 91-9032Z, Unit1 C C GE'he existing Rosemount model 1151DP transmitters (FT-E11-1N007 ABB) are replaced with Rosemount Model 1153 transmitters. The new transmitters are installed at the same location as the existing transmitters in instrument racks 1C018 and 1C021.
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I. No. The FSAR was reviewed, in particular sections 7.5.1b.6 and 9.2.6, Chapter 15, and the responses to NRC questions. NUREG-0076 was reviewed.
The affected instruments are panel mounted indicators for control room operators. The indicators function identically as they did before the DCP.
The new installation replaces the existing transmitter with a similar model and relocates the square root function from the flow transmitters to the upper and lower relay rooms and the Remote Shutdown Panel. The new design does not introduce any new failure modes.
Therefore, overall system reliability is unchanged.
The consequences of a malfunction are not changed, because the replacement instrumentation functions identically to the existing instrumentation.
Therefore, this modification does not increase the probability or consequences of an accident or probability or consequences of a malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification installs new transmitters and square root extractors for the instrument loops.
The new installation separates the square root extraction function from the transmitters. This design change does not cause the loops to perform or function differently than the original loops.
Therefore, the proposed action does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification 3/4.3.7.4 addresses Remote Shutdown Monitoring Instrumentation and specifies operability requirements for RHR Service Water flow indication for remote shutdown.
The control room instrument loops are identified in FSAR section 7.5 as being part of the Safety Related Display Instrumentation, but are not identified in any Technical Specification. Neither the control room loops nor the Remote Shutdown Panel Loop contain any interlock function.
This modification has no effect on the operability of the instrument loops. Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
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S~E~ 93-122 C NL-93-017, Unit 1&2 DESC 0 C Reactor Water Cleanup/Residual Heat Removal (RWCU/RHR) Steam Leak Detection-Differential Temperature Monitoring Bypass for Temporary Monitoring Capability.
The purpose of this Bypass is to provide accurate monitoring, recording and trending of the RWCU Steam Leak Detection Differential temperature elements in the RWCU Penetration Room.
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~ No. The proposed action only affects the method by which the RHR Pump and RWCU Penetration room differential temperatures are monitored. The design basis of the steam leak detection system and the safety related function of the system isolation as described in FSAR Sections 7.3.1.1a.2 and 7.6.1a.4 are unaffected by this Bypass.
The Accident Analysis described in FSAR Section 15.6.4 is the bounding accident for steam leaks of the size the steam leak detection system is capable of detecting.
Installing a temporary recorder under this Bypass does not affect the probability of occurrence or the consequences of a steam leak accident outside containment.
II ~ No. This Bypass will only affect the method by which the Steam Leak Detection Differential Temperature is monitored in the RHR Pump and RWCU Penetration rooms. The system isolation function is unaffected by this Bypass. The same temperature element (TE) is utilized for both the recorder and the isolation logic, however the TE is a dual-head thermocouple. One head is used for the isolation logic and the other head is used by the recorder. The two heads are electrically independent, therefore the failure of the temporary recorder will not cause failure of the safety related isolation function. This Bypass does not create any accidents or malfunctions of a different type than previously evaluated in the FSAR.
III. No. The Steam Leak Detection Isolation function is defined in Technical Specifications Section 3.3.2, however the recorder/pre-isolation alarm function are not covered by any plant Technical Specification. Therefore, there is no reduction in the margin of safety as defined in the Technical Specifications.
S~~ 93-123 C SS E E CE'P-193-027, Unit 1 D C C E'nit 1 Main Turbine Torsional Vibration Test:
The turbine-generator rotor torsional natural frequencies will be measured from approximately 800 to 1900 rpm. This will be accomplished prior to and during initial synchronization of the unit. The primary objective of the test is to determine whether any torsional natural frequencies exist at or very near 120 Hz.
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FSAR section 3.5.1.3 describes the safety risk from the Turbine-Generator as a Turbine overspeed condition which causes a rotor wheel burst and subsequent missile generation.
GE has established that the probability of missile generation at design overspeed is statistically insignificant. There are two overspeed protection devices, at approximately 110% and at approximately 112%. Neither device will be impaired by this test and thus the probability of an overspeed condition and subsequent missile generation is not increased by this test. Section 10.2.1 of the FSAR provides the design bases of the Turbine-Generator.
The only function to be affected by this test is speed control, and the temporary GE speed control circuit will be able to control Turbine speed at the slower ramp rate of 7 rpm/min, as compared to the normal slow rate of 60 rpm/min. FSAR sections 15.2.2 and 15.2.3 address generator load rejection and turbine trip, respectively. These functions are in effect
> 30% Reactor power. Reactor power level is expected to be= 10 to 15% during the performance of this test, so these protective functions are not a concern. The performance of this test does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II~ No. The performance of this torsional test is within the bounds of Turbine Generator system design. Due to the fact that Turbine-Generator overspeed protective functions will not be defeated, this test does not create a possibility for an accident or malfunction of a different type than evaluated in the FSAR, as discussed in section I III No.
~ Tech Spec section 3/4.3.8 ensures that the turbine overspeed protection system instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. Performance of the torsional test will not reduce the margin between anticipated wheel failure and overspeed protection system instrument settings, nor will it affect the overspeed protection system. Tech Spec section 3/4.3.9 is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand. The main turbine bypass system is not affected by this test. The torsional test has no effect on the basis for the sections described above. Therefore, there is no reduction in the margin of safety as defined in the basis for any Tech Spec.
U SEED¹93-124 C DCP 92-9034F, Unit 1 DE C The proposed action is to remove the automatic restart of the Unit 1 Condensate Pump 1D after actuation of the Unit 1 13.8 KV Bus Load Shed. The proposed action involves wiring changes to remove the 62X1-10104 relay from the Unit 1 13.8 KV Bus Load Shed logic and to remove the 62X1-10104 relay contact from the 13.8 KV breaker 1A1 0209 close circuit.
MSHHliMK; I. No. The removal of the automatic restart of the Condensate Pump 1D after actuation of the Unit 1 13.8 KV Bus Load Shed does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus the proposed action does not increase the probability of occurrence of an accident. The removal of the automatic restart of the Condensate Pump 1D after actuation of Unit 1 13.8 KV Bus Load Shed does not increase the probability of occurrence of a malfunction of equipment. The removal of a component from the Condensate Pump 1D start circuit incrementally decreases the probability of occurrence of a malfunction in the non-Class 1E circuitry. The consequences of an accident is not changed by the proposed action.
The accident analysis does not take credit for the restart of the Unit 1 Condensate Pump 1D.
Preventing the automatic restart of the Unit 1 Condensate Pump 1D after actuation of the 13.8 KV Bus Load Shed scheme does not change the consequences of a malfunction of equipment.
Since an operating mode of the Condensate Pump 1D and it's impact on the ECCS equipment is eliminated, the consequences of a malfunction of equipment for that operating mode are eliminated.
II. No. Chapters 6 and 15 of the FSAR were reviewed to determine if the proposed action had the potential of creating a postulated initiating event which was not within the spectrum of events for which transient or anticipated operational occurrences and accident conditions were analyzed.
The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. The removal of the 62X1-10104 relay, the removal of internal panel wiring and the sparing of conductors in a cable is in accordance with existing approved procedures. With the proposed action the voltage on the Class 1E 4.16 KV buses recovers above the pickup value of the Bus Degraded Grid Undervoltage protective relays prior to operation of the Degraded Grid Timer. This assures that the voltage to the ECCS equipment is adequate for proper operation of the systems and a malfunction of a different type is not created.
III. No. The operability of the Unit 1 13.8 KV Bus Load Shed logic and the operability of the Unit 1 Condensate Pump 1D are not governed by Technical Specifications. The operability of the Class 1E 4.16 KV buses is governed by Technical Specification 3/4.8.3, Onsite Power Distribution Systems. The proposed action does not affect the bases for operability of the Onsite AC Power Distribution System during operation, shutdown or refueling conditions. Thus, the removal of the automatic restart of the Condensate Pump 1D after actuation of the 13.8 KV Bus Load Shed does not reduce the margin of safety as defined in the basis for any Technical Specification.
II 5gR59 '3>>125 C E C NL-93-019, Unit 1 D C OFC G A system functional flow test of the Reactor Water Cleanup (RWCU) System, will be performed to determine if a system flow of 297 GPM @60', which is equivalent to an RWCU flow of 396 GPM @532 F can be achieved and satisfactorily maintained by throttling open the demineralizer flow control valves FV-14566A 8 B.
I. No. The FSAR, chapters 3.6, 3.6A, 5.4.8, 6.2 and 15 were reviewed to determine if the proposed test has an effect on the spectrum of postulated accidents previously reviewed. The postulated accident for the RWCU system is a pipe break outside containment in the RWCU pump suction line downstream of the containment isolation valve HV-144F004. Therefore, the consequences of this accident are bounded by the Steam Line Break Accident.
The proposed test does not alter system configuration, change initial conditions or change the operating mode. The proposed test does not affect the postulated accident identiTied and does not increase the probability of occurrence of an accident.
The increased RWCU flow for the proposed test results in a negligible increase in the probability of an inadvertent RWCU system isolation. The proposed test results in no increase in the probability of a malfunction of equipment important to safety.
The consequences of a design basis accident inside or outside containment are not changed by the proposed test. The accident remains bounded by the Steam Line Break Accident, which is unaffected by the proposed test. There is no increase in dose to the public; therefore, the consequences of an accident are not increased. Since closure of the containment isolation valves would be in response to an out of specification test condition and not to an accident, the closure would be unintentional and undesirable. This is an operational concern and not a safety issue, therefore, plant safety is not reduced and radiation doses offsite are not increased.
II ~ No. The FSAR, chapters 3.6, 3.6A, 5.4.8, 6.2 and 15 were reviewed. All equipment has been evaluated by GE and proven to be adequate and within its design capability for the increased flow. The proposed test does not introduce any change to the RWCU system physically, or functionally, which preciudes the possibility of an accident of a different type. The logic or sequencing of the RWCU Engineered Safety Feature System for the proposed test is unchanged; therefore, no new malfunction of a different type exists.
III. No. Technical Specification 3/4.3.2 prescribes the limiting conditions of operation for isolation actuation instrumentation. The basis for the Technical Specifications is to ensure the effectiveness of instrumentation used to mitigate consequences of accidents by prescribing the operability trip setpoints for isolation of reactor systems. There are no other Tech Specs affected by the proposed test. Therefore, there is no reduction in the margin of safety of any Tech Spec.
Qgg.gg '3-126 C E E E CE DCP93-3040C, Unit Common DESC 0 C G Prior to installation of this mod, DCP 93-3040A will cross-tie well water to clarified water to supply seal water for the circulating and service water pumps. This proposed modification will allow a future modification to retire the clarifier. The proposed modification also installs a hose connection so that a backup source of domestic water can be made available.
Y I~ No. FSAR Chapters 6 and 15 have been reviewed with particular attention paid to FSAR Section 15.2.3 and 15.2.5 for possible adverse consequences with respect to the installation of this modification. Loss of seal water to the Circulating Water Pumps could cause them to trip resulting in a turbine trip and loss of condenser vacuum. This modification will not result in an increase in the probability of occurrence of these events due to the storage capabilities of the Well Water and Clarified Water Storage Tanks and due to the capability to make clearwater from a vendor truck within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of notification.
The Potable Water and Raw Water Treatment Systems and systems to which they connect, such as Service and Circulating Water, are not safety related. The probability that a component failure which would prevent the Potable Water or Raw Water Treatment from performing their design function is not increased by installing the hose connections due to the reserve water which is stored and the ability to restart the clarifier or vendor truck. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
II. No This modification will install hose connections allowing the use of a vendor clearwater truck as a backup to the well water system. These hose connection spool pieces will be designed to meet all the requirements of ASME/ANSI B31.1 Code to ensure its structural integrity, and guarantee that the Potable and Raw Water Treatment Systems will be able to operate and function as specified in the FSAR. The proposed action will not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No This modification does not jeopardize or degrade the function or operation of any plant system governed by Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification.
The hose connections will be designed to meet all applicable codes; 'consequently, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
I 0
! gg gg '3-127 C E C DCP 93-3002Z, Unit 1 DE C TIO C G The proposed action is to add an automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus. The proposed action involves the addition of a new Potter and Brumfield MDR relay across the normal DC control power to the trip circuitry of each breaker feeding a divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus.
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No. Chapters 6 and 15 of the FSAR were reviewed. The addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR.,Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of a automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel C control power in the Unit 1 Channel C 4.16 KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction. The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action. The MDR relays, the indicating lights, the fuses and the breakers are all qualified devices.
II. No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in either trip circuitry, the MDR relay or the light added to the trip circuitry of a divisionalized load supplied from the Unit 1 Channel C 4.16 KV Bus does not create a malfunction of a different type.
III. No. The operability of the Emergency Service Water System and the Residual Heat Removal Service Water System is governed by Technical Specification 3/4.7.1. The proposed action does not affect the operability requirements of the Service Water Systems. The proposed action does not affect the operability requirements of the Control Structure Heating and Ventilating System or the Control Structure Chilled Water subsystem. The operability of the Unit 1 Channel C 4.16 KV Bus is governed by Technical Specification 3/4.8.3. The proposed action ensures operability of the Unit 1 Channel C 4.16 KV Bus for a Unit 1 Channel A battery failure coincident with a LOCA/LOOP condition thereby maintaining the existing margin of safety for the Unit 1 Channel C 4.16 KV Bus.
tv SRR59 es-12S C CE DCP 93-3003Z, Unit 1 D C C GE'he proposed action is to add an automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus. The proposed action involves the addition of a new Potter and Brumfield MDR relay across the normal DC control power to the trip circuitry of each breaker feeding a divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus.
No. Chapters 6 and 15 of the FSAR were reviewed. The addition of an automatic transfer logic into the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of a automatic transfer logic to the breaker trip circuitry of each divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel D control power in the Unit 1 Channel D 4.16 KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction. The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action. The MDR relays, the indicating lights, the fuses and the breakers are all qualified devices.
II. No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in either trip circuitry, the MDR relay or the light added to the trip circuitry of a divisionalized load supplied from the Unit 1 Channel D 4.16 KV Bus does not create a malfunction of a different type.
III. No. The operability of the Emergency Service Water System and the Residual Heat Removal Service Water System is governed by Technical Specification 3/4.7.1. The proposed action does not affect the operability requirements of the Service Water Systems. The proposed action does not affect the operability requirements of the Control Structure Heating and Ventilating System or the Control Structure Chilled Water subsystem. The operability of the Unit 1 Channel D 4.16 KV Bus is governed by Technical Specification 3/4.8.3. The proposed action ensures operability of the Unit 1 Channel D 4.16 KV Bus for a Unit 1 Channel B battery failure coincident with a LOCA/LOOP condition thereby maintaining the existing margin of safety for the Unit 1 Channel D 4.16 KV Bus.
S~E~O 93-129 e C DCP 92-9035F, Unit 2 E CRI C G The proposed action is to remove the automatic restart of the Unit 2 Condensate Pump 2D after actuation of the Unit 2 13.8 KV Bus Load Shed. The proposed action involves wiring changes to remove the 62X2-10104 relay from the Unit 2 13.8 KV breaker 2A1 0209 close circuit.
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I. No. Chapter 6 and 15 of the FSAR were reviewed. The removal of the automatic restart of the Condensate Pump 2D after actuation of the Unit 2 13.8 KV Bus Load Shed does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus, the proposed action does not increase the probability of occurrence of an accident. The removal of the automatic restart of the Condensate Pump 2D after actuation of Unit 2 13.8 KV Bus Load Shed does not increase the probability of occurrence of a malfunction of equipment. The removal of a component from the Condensate Pump 2D start circuit incrementally decreases the probability of occurrence of a malfunction in the non-Class 1E circuitry. The consequences of an accident is not changed by the proposed action. The accident analysis does not take credit for the restart of the Unit 2 Condensate Pump 2D. Since an operating mode of the Condensate Pump 2D and it's impact on the ECCS equipment is eliminated, the consequences of a malfunction of equipment for that operating mode are eliminated.
II ~ No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. The removal of the 62X2-10104 relay, the removal of internal panel wiring and the sparing of conductors in a cable is in accordance with existing approved procedures. With the proposed action the voltage on the Class 1E 4.16 KV buses recovers above the pickup value of the Bus Degraded Grid Undervoltage protective relays prior to operation of the Degraded Grid Timer. This assures that the voltage to the ECCS equipment is adequate for proper operation of the systems and a malfunction of a different type is not created.
III. No. The operability of the Unit 2 13.8 KV Bus Load Shed logic, and the Unit 2 Condensate Pump 2D and its 13.8K Breaker 2A10209 control circuitry is not governed by Technical Specifications. The operability of the Class 1E 4.16 KV buses is governed by Technical Specification Section 3/4.8.3, Onsite Power Distribution Systems. The proposed action does not affect the bases for operability of the Onsite AC Power Distribution System during operation, shutdown or refueling conditions. Thus, the removal of the automatic restart of the Condensate Pump 2D after actuation of the Unit 2 13.8 KV Bus Load Shed does not reduce the margin of safety as defined in the basis for any Technical Specification.
STENO '3-130 C E E E CE DCP91-3023C,D, Unit182,Common This task installs a second measurement of cooling tower blowdown water flow discharge to the river.
This modification provides total water flow to the river to meet environmental requirements. In step one (DCP 91-3023D), all trenching for routing cable underground is completed, together with tasks required to support refilling the trench including pulling sensor cable. In step two (DCP 91-3023C), the rest of the modiTication is installed, tested and started-up.
I. No. Adding a new measurement of total water flow discharge to the river for use in theLiquid Radwaste (LRW) discharge valve logic does not increase the probability of occurrence of an accident as previously evaluated in the FSAR because the modification does not add or change any components whose failure would initiate an accident or which are used prior to an accident. This modification does not increase the consequences of an accident as previously evaluated in the FSAR with regard to radiological consequences or the health and safety of the public because the modification does not change the function or performance of the liquid radwaste system. This modification increases the reliability of the plant discharge flow measurement system, thus reducing the potential for a radioactive release to the public. Thus, this change does not increase the probability of occurrence of a malfunction of equipment as previously evaluated in the FSAR. This modiTication does not increase the consequences of malfunction of equipment as previously evaluated in the FSAR because the modification does not change the liquid radwaste system.
II. No. FSAR Sections 10.4.5, 11.2 and Chapters 6 and 15 were reviewed. The modification does not create the possibility of an accident of a different type than previously evaluated in the FSAR because the modiTication does not change the function or performance of the LRW system. The modification does not create a possibility for malfunction of a different type than any previously evaluated in the FSAR because the modification does not change the function or performance of the LRW.
III. No. Plant Technical Specification Sections 3.3.7.10 and 4.3.7.10, Tables 3.3.7.10-1 and 4.3.7.10-1 address radioactive liquid effluent monitoring instrumentation. In the tables, Item 3 addresses how many channels should be operable. This modification increases the number of channels, increasing reliability of the instrumentation. Technical Specification 3.11.1.3 specifies total off site radiation dose limits for a 31 day period. Although the Technical Specifications do not provide limits for waste discharge dilution water flow to the river, indirectly these doses imply a minimum water flow limit. The new instrumentation will measure the total water flow to the river, including both cooling tower flows and spray pond overflow, providing this information for the first time. It should also increase the implied margin of safety from the existing system, based on measurement of total flow, including spray pond overflow.
S~E~ 93-131 co C . DCP93-3057, Unit2 DEC NC Modify HV-249F008 by installing a new standard gear set consisting of motor pinion gear, worm shaft gear, worm gear and worm with an overall ratio of 82.17 to replace the existing non-standard gear set having an overall ratio of 83.35.
I. No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. Installing a new standard gear set for valve HV-249F008 will provide sufficient torque to ensure the valve functions during all design conditions. The ability to open or close the valve when required is not affected by the modification. Valve integrity and seat leakage are not affected. The new valve stroke time will be 14.25 seconds, which is within the FSAR limit of 20 seconds. Therefore, this modification will not increase the probability of occurrence or the consequences of an accident previously analyzed in Chapter 15 of the FSAR.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-249F008 to perform its intended design function; 2. does not increase the stroke time of the valve beyond that previously contained in the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. improves the ability to open or close the valve when required; 5. does not affect valve leakage; 6. involves the replacement of components outside the valve pressure boundary; 7. does not affect any other equipment in the RCIC system; 8.
does not affect motor horsepower rating, the power supply from the DC Motor Control Center (MCC), or the battery charger or battery bank associated with the DC MCC.
The proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III No.
~ Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.7.3 and 3/4.3.5 were reviewed for this evaluation. The modification does not reduce the margin of safety defined in the Bases Technical Specification.
QggJgg 93-132 COS E C ME-062-001, Unit 1 and 2 The Mechanical Stress Improvement Process (MSIP) is a stress-related Intergranular Stress Corrosion Cracking (IGSCC) mitigation measure. MSIP produces favorable stress patterns by removing residual tensile stresses and generating residual compressive stresses at the inside surface of piping weldments. ME-062-001 will be used to control the MSIP process to ensure a correct application.
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~ No. FSAR Section 5.2 discusses the material selection and fabrication/construction controls considered to avoid stress corrosion cracking of the reactor coolant pressure boundary and ECCS systems. MSIP will mitigate IGSCC by removing tensile residual stress and generating compressive residual stresses at the weldments. The application of MSIP as controlled by ME-062-001 will not cause pipe cracks or alter the stainless steel microstructure in such a way that would degrade piping integrity. MSIP can arrest the growth of existing shallow cracks in piping due to the generation of compressive stresses around the crack tip. The various accidents analyzed in Chapter 15 of the FSAR evaluate performance of the reactor vessel as a radioactive material barrier.
Since MSIP will not reduce piping or nozzle integrity but actually increase its resistance to IGSCC performance of the reactor coolant pressure boundary as a radioactive barrier will not be reduced. Therefore, MSIP does not increase the probability of occurrence or the consequences of accidents or malfunctions of equipment evaluated in the FSAR.
II. No. The proper application of MSIP as controlled by ME-062-001 will not cause pipe cracks or alter the stainless steel microstructure or ASME code stress levels. Since the integrity of the reactor vessel and RHR system is not degraded but actually enhanced due to decreasing probability of developing IGSCC, MSIP does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The bases for the reactor coolant system are described in Tech Spec B 3/4.4. MSIP makes the piping less susceptible to IGSCC by changing the residual stresses in the weld from tensile to compressive. MSIP does not compromise the integrity of the reactor coolant system pressure boundary and ASME code stress levels are not altered. Since the process only increases the probability of these systems being able to perform their safety function the margin of safety is not reduced.
II I
4
SREUK) '3-133 C OSS C . DCP93-3037, Unit2 SC C Supports on the Scram Discharge Volume (SDV) vent and drain lines require structural modifications. Changes include addition of pipe clamps, replacement of Unistrut channels and modifications to existing support components.
I~ No. The applicable design basis accidents are described in FSAR Sections 15.4, and 15A.6.5.3. The specific evaluations are described in FSAR Sections 4.6.2, 15.4.9, and 15A.6.5.3. Pipe support modifications required will increase design capacities of the previously installed structural components so that all supports are maintained within their appropriate design limits. Proposed modifications neither alter Control Rod Drive (CRD) system function nor performance. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR will not increase.
II. No. The planned modifications do not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR. Breaching of the SDV vent and drain piping is not required. All work as planned is external to the piping system and will result in increasing supporting capacities of the previously installed components. The proposed actions ensure piping and pipe support structural integrity and enhance Unit 2 reliability.
III. No. Technical Specification 3/4.1.3 is applicable to the SDV vent and drain piping required pipe support modifications. Planned modifications will insure applicable safety design margins by increasing pipe support hardware capacities in accordance with Code design requirements. The planned modifications will not reduce the margin of safety as defined in the basis for any Technical Specification.
S~E~ 93-134 C DCP 93-3042, Unit 2 E C E'CP 93-3042 will install condenser tube stakes in accessible areas of the condensers and top hat baffles over the air removal hoods of the upper bundles. The modification will limit tube vibrations in affected areas of the condensers and will decrease the occurrence of tube failures due to vibration and will result in extended tube life.
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No. This modification does not increase the probability of occurrence of an accident evaluated in the FSAR. FSAR Sections 10.4 and 15.2.7 were reviewed for previously evaluated accidents. The new self-locking stake design is an improvement to previous stake installations, resulting in reducing the probability of loosening during operation.
This modification does not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. The main condenser is not safety related or required for safe shutdown of the reactor or mitigation of a design basis accident. This modiTication does not increase the consequences of an accident as previously evaluated in the FSAR. There are no accidents evaluated in the FSAR which involve the condenser or require condenser operability. This modification does not increase the consequences of a malfunction of equipment important to safety. The worst case malfunction of equipment that can be postulated is failure of the condenser or a pump in the condensate/feedwater stream.
None of these items of equipment are safety related.
II~ No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR. The only accident that can be postulated as a result of this modification is loss of feedwater which has already been evaluated in the FSAR. This modification does not create the possibility of a malfunction of a different type than any evaluated previously in the FSAR. The modification is designed to reduce tube vibration and decrease the occurrence of tube failures due to vibration. No other different types of malfunction can be identified.
III. No. Technical Specifications 3.3.7.11, 3.4.4, 3.7.8, 3.11.2.6, and 3.11.2.7 were reviewed.
This modification does not change any instrumentation, setpoints, flow rates of noble gases, generation of hydrogen or oxygen and does not have any effect on radioactive gaseous effluents or control of radioactive leakage. This modificatio does not reduce the margin of safety as defined in the Technical Specification.
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~E~ 93-135 C E ERE CE DCP 93-9024, Unit Common SC To eliminate biannual bypasses and Radwaste HVAC system trips due to high ambient temperatures in the spring and fall, DCP 93-9024 changes the off-gas fans control system from high temperature to low flow control.
No. The proposed change does not increase the probability of occurrence of an accident as previously evaluated in the FSAR because the modification does not add or change any components whose failure would initiate an accident. The proposed change does not increase the consequences of an accident as previously evaluated in the FSAR with regard to radiological consequences or the health and safety of the public because the modification does not change the function or performance of the Radwaste Building main air heating, ventilation and air conditioning system. The proposed change does not increase the probability of occurrence of a malfunction of equipment as previously evaluated in the FSAR, because the modification does not change the Radwaste Building main air heating, ventilation and air conditioning system. The proposed change does not increase the consequences of malfunction of equipment as previously evaluated in the FSAR with regard to radiological consequences or the health and safety of the public because the modification does not change the function or performance of the Radwaste Building main air heating, ventilation and air conditioning system.
II~ No. FSAR Sections 6, 10.4.5, 11.2 and Chapter 6 and 15 were reviewed. The proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR because the modification does not change the function or performance of the Radwaste Building main air heating, ventilation and air conditioning system and malfunctions of this system are not evaluated in the FSAR.
III. No. There are no stated limiting conditions for operation in the Technical Specifications for the Radwaste Building HVAC System or stated margins of safety. The ambient temperature after installing the differential pressure switch, resetting the area temperature controllers and adding the off-gas high temperature alarm will still be maintained within the specified limits stated in the FSAR Design Bases Section 9.4.3.1.
Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specification.
S~E O.: 93-136 C DCP 93-3028, Unit 1 DE C Modify HV-155F003 by replacing the actuator, yoke, yoke clamp and disc. The scope of this DCP also involves the use of a size 2 reduced voltage reversing starter in lieu of the existing size 1 starter for valve power, resetting of the breaker magnetic trip, the installation of new overload heater coils in the DC control center cubicle and deletion of the space heater circuit to the valve operator.
DSU RlhBY I. No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. After implementation of the modification the MOV design stroke time will be well within the FSAR and Tech Spec limit of 50 seconds. This modification will not affect circuit signal logic. The valve leakage criteria is not changed by this modification.
Changing HV-155F003 motor actuator will provide sufficient thrust to ensure the valve functions during all design conditions. Therefore, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously analyzed in Chapter 15 of the FSAR.
II. No. The proposed modification: 1. enhances the ability of HV-155F003 to perform its intended design function, 2. does not increase the stroke time beyond the design basis;
- 3. does not decrease the MOV's allowable seismic accelerations below value determined in piping analysis, 4. does not alter the motor control center seismic qualification, 5. improves the ability to open or close the valve when required, 6. does not affect the valve leakage, 7. does not decrease the available voltage at the motor terminals below the required level, 8. does not affect the motor power supply system, 9.
and does not affect any other equipment in the HPCI system. It is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR (Chapter 15).
III. No. Tech Spec bases 3/4.5.1, 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.3 were reviewed for this evaluation. The modification does not reduce the margin of safety defined in the bases of any Technical Specifications. Valve integrity and seat leakage are not affected. The valve stroke time is within the accident analysis bounds of 50 seconds. There are no effects to any of the system/components by this modification. Therefore, this modification has no effect on the HPCI system intended design function or the basis on the above Tech Specs.
C DCP 93-3074, Unit 2 DEC 0C Modify the ¹4, 6 and 8 main turbine bearings and bearing housings as follows: Drill and tap two 1/2" diameter by 1 469" deep holes, split equally apart from the vertical centerline of the bearing by 2.25", at twelve o'lock; and drill and tap the lower bearing housing to accept a sealing gland with 3/8" NPT threads.
I. No. FSAR Section 3.5.1.3 concludes that the probability of unacceptable damage to safety related components due to turbine missiles is 2.61E-10. This FSAR section states that this probability is reduced with the monoblock rotors currently installed in the turbine.
Based on the low probability, the turbine missile hazard is not considered a design basis event. FSAR Section 10.2 provides the design bases of the Turbine/Generator.
The analysis performed in this section is not impacted by the modification. The modification does not impact the analyses performed in FSAR Sections 15.2.2 and 15.2.3. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. A failure of the spot welds or fasteners on the torque collars or the bolting on the box or brass ring antenna would result in these parts becoming loose inside the bearing housing. Based on the bearing/shaft clearances and the size of these components there is no possibility of this malfunction causing detrimental shaft or bearing surface damage. Bearing oil drains are sufficiently sized such that no one piece or combination of pieces could block oil flow from the affected bearing. The probability of the tack welds creating a stress riser in the bearing is negligible. Therefore, the modification does not create a possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III No. Tech Spec section 3/4.3.8 ensures that the turbine overspeed protection system
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instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. The proposed modification will not reduce the margin between anticipated wheel failure and overspeed protection system instrument settings, nor will it affect the overspeed protection system or the speed control valves in any other manner. The proposed modification does not affect the basis of Tech Spec section 3/4.3.9. The main turbine bypass system is not affected by the modification.
The modification has no effect on the bases for any Tech Spec. Therefore there is no reduction in the margin of safety as defined in the basis for any Tech Spec.
>r S~E@Q '3-138 CE DCP 93-3043A, Unit 2 ESC 0 C G DCP 93-3043A implements the following modifications to the Turbine/Generator Electrohydraulic Control (T/G EHC): Six indicators on the control room panel (2C651) will be replaced in-kind by indicators with expanded ranges so that uprated conditions will not cause them to read near the top of their scales. Diode function generator (DFG) cards must be recalibrated to assure that control valves CV-1, 2, 3 and 4 are operated in the most stable region under uprated conditions.
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No. The EHC Turbine Control System is described in FSAR Section 10.2.1 and 10.2.2.5.
The indicator replacement portion of this modification merely changes the scale range, all physical and electrical characteristics are identical. The remaining recalibration of EHC main turbine control equipment results in a small change in Turbine Control Valve positioning, but the sequencing of the TCV's remains the same. This DCP does not involve the addition of equipment or changes in the type of existing equipment.
Therefore, this modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15. This modification does not increase the probability of occurrence of a malfunction of equipment. The indicator replacement portion of this modiTication merely changes the scale range, all physical and electrical characteristics are identical. This modification does not increase the consequences of an accident since the EHC equipment has no function related to preventing uncontrolled release of radioactivity nor does it affect any equipment that does. This modification does not increase the consequences of a malfunction of equipment since the EHC equipment has no function related to preventing uncontrolled release of radioactivity nor does it affect any equipment that does.
II. No. The proposed action does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR. The indicator replacement portion of this modification merely changes the scale range, all physical and electrical characteristics are identical.. The recalibration of EHC equipment does not change the plausible failure modes of the equipment.
III. No. Functions performed by the equipment within the scope of this modification are not mentioned in the Technical Specifications, therefore this modification does not reduce the margin of safety for any Technical Specifications
C OSS E C DCP 93-3043B, Unit 2 C
E'CP 93-3043B adds a second Steamline Resonance Compensator (SLRC) card to both the primary and secondary pressure sensing loops. The new cards will be located in Electrohydraulic Control (EHC) cabinet 2C663 card rack A slots A52 and A69.
No. The EHC Turbine Control System is described in FSAR Sections 10.2.1 and 10.2.2.5.
This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Section 6 and 15. Although this DCP adds equipment to the EHC system pressure sensing loops, it merely adds a second card in series with the existing card to improve resonance damping. The function of the sensing loop in the EHC system and its interface with other equipment is unchanged.
This modification does not increase the probability of occurrence of a malfunction of equipment important to safety because the EHC system performs no safety function and its interface with other equipment is unchanged. This modification does not increase the consequences of an accident or malfunction of equipment since the EHC equipment has no function related to preventing uncontrolled release of radioactivity nor does it affect any equipment that does.
II~ No. This modification does not create the possibility of an accident of a different type than any evaluated in the FSAR because the plausible failure modes of the equipment and its interfaces with other equipment is unchanged. The second SLRC card merely provides additional EHC pressure signal damping. This modification does not create the possibility of a malfunction of equipment of a different type than any evaluated in the FSAR. The addition of a second SLRC card to each of the pressure sensing loops does increase the possibility of the loops failing because there are more components in the loop. However, this addition has no effect on safety because these loops have no safety function nor do they have any impact on equipment that does.
III. No. Review of the Technical Specifications shows that the EHC and functions performed by the equipment within the scope of this modification are not mentioned in the Technical Specifications. Further, the subject instrumentation does not have any impact on equipment mentioned in the Technical Specifications.
~SE ~O: 93-140 C DCP 92-3023, Unit 1 The purpose of the modification is to improve the overall reliability of the Reactor Protection System (RPS) power distribution system. This will be achieved by replacing the existing Electrical Protection Assembly (EPA) logic cards with the new G.E. logic cards. This modification will result in a reduction of spurious trip incidences of the RPS power system.
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The replacement of the EPA logic cards with the new G.E. logic cards does not have any impact on the postulated initiating events identified in the engineered safety features and accident analysis of FSAR Chapters 6 and 15. The replacement of the existing EPA logic cards with new logic cards will result in a more reliable protection system. Therefore there is no increase in the probability of occurrence of an accident.
The new G.E. logic cards have been redesigned to remedy most of the deficiencies of the existing EPA logic cards and therefore are better than the existing logic cards.
Therefore, the replacement of the existing logic cards with the new card will not increase the probability of occurrence of a malfunction of equipment. The new logic card mitigates several of the deficiencies of the existing logic card. This will result in a logic card with more reliable operation for the RPS buses. The replacement of the logic card within the EPA does not increase the consequences of a design basis accident.
The use of the new G.E. logic cards, which have the same function and setpoints as the existing EPA cards, will not affect the performance of equipment supplied by the RPS buses or increase the consequences of a malfunction of any equipment.
II~ No. A review of FSAR Chapter 15 on accident analysis did not identify any postulated initiating events relative to the proposed action which would create the possibility for an accident of a different type. The new card is an enhanced version of the existing card, redesigned to eliminate deficiencies. Therefore, does not create a possibility of a malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification Section 4.8.4.3 provides the surveillance requirements governing the operability of the RPS EPAs and establishes the setpoints for undervoltage, overvoltage and underfrequency relay actuation. The setpoints listed in Technical Specification Section 4.8.4.3.b are applicable to the new logic cards.
Therefore the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
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CE NL-93-026, Unit 1 and 2 The proposed action is to change the method of performing the monthly 4kV Bus Degraded Voltage Channel Functional Test. The proposed method accomplishes the test in a manner less likely to cause inadvertent actuation of bus trip and transfer relays and bus load shed relays.
I. No. The specific details of the Degraded Grid Relay Test logic are not described in the FSAR. The proposed action operates the existing relay logic in a more secure manner which reduces the chance of inadvertent actuation. This action actually reduces the probability of occurrence of an accident or malfunction of equipment important to safety by reducing the chance of major plant transients due to this test.
II. No. All required testing functions are completed by the proposed action thereby assuring the Operability of the Degraded Grid Relaying. Since the changes are designed to perform the existing logic in a sequence that improves the reliability of the existing test scheme, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Tech Spec 4.3.3.1 requires monthly functional tests of the degraded grid bus voltage relaying. The proposed action performs the required testing and in a more secure manner. Since all required testing is performed by the proposed action there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
SEE.59 '3-142 C E C DCP 93-3058F, Unit1 DESC I T 0 C G Eliminate the affected fibrous insulation on the Reactor Building Cooling Water (RBCW) system piping, within 7 pipe diameters of a postulated High Energy Line Break (HELB), in the drywell by replacing the existing stainless steel jacketed fiberglass insulation with stainless steel jacketed phenolic foam insulation.
I~ No. This modification does not increase the probability of an accident evaluated in the FSAR including FSAR Sections 6.3.2 and 15.6.5. Failure of insulation does not result in an accident evaluated in the FSAR. This reduction in material density results in a negligible weight change per foot on the pipe and therefore does not pose any pipe stress issues.
Therefore, the pipe stress analysis will not be affected and therefore will not increase the probability of a pipe break. This modification does not increase the probability of a malfunction previously evaluated in the FSAR. The RBCW piping insulation meets normal operating environmental parameters. By removing the stainless steel jacketed fibrous RBCW insulation inside the drywell and replacing it with a low specific gravity, stainless steel jacketed, phenolic foam insulation, the potential of blocking the ECCS suction strainers is reduced. This modification does not increase the consequences of an accident as previously evaluated in the FSAR. This modiTication does not increase the consequences of a malfunction of equipment important to safety. The proposed action does not affect the failure modes of any equipment important to safety.
II. No. The replacement phenolic foam insulation meets the requirements of FSAR Section 3.2.1, regarding safety impact item. The proposed action does not create a possibility nor increase the probability of an accident or malfunction of a different type than any evaluated previously in the FSAR. The small change in weight of the replacement insulation system is bounded by the existing piping analyses and as such the safety related primary containment isolation function of the RBCW system, according to FSAR Section 6.2.4.3.3.5 is not affected by this modification.
III. No. The modification has no effect on the bases of Technical Specification Section 3/4.5.1 and 3/4.5.3. As the replacement phenolic foam insulation has better thermal properties than the original design for the existing fiberglass insulation being replaced, it will have no effect on the margin of safety for this limit as long as proper installation practices are used and the vapor barrier is intact. Technical Specification Section 3/4.6.1.7 discusses the drywell average air temperature limiting conditions for operation of 135 degrees F. The modification has no effect on these bases. The modification does not reduce the margin of safety of the basis of any Technical Specification.
C 0 CE NL-93-025, Unit 1 C CH E'he proposed action involves a flush of radioactive residue existing within the Recirculation to Reactor Water Cleanup (RWCU) suction drain lines, the vessel bottom head drain line, and the Recirculation discharge valve drain lines.
I. No. The potential for large water inventory losses from the reactor pressure vessel is minimal. FSAR section 6.3.1.1.1 states that the ECCS makeup capability was sized to enable core flooding after the double-ended break of the recirculation line suction to reactor recirculation pumps. A conservative calculation was performed to determine the maximum flowrate that would be achieved if one of the two inch drain lines were left wide open, and the resultant flowrate would be approximately 800 GPM, which is well within the ECCS makeup capabilities. In addition, the procedure has a prerequisite step to ensure that all administrative controls are completed for Operation with Potential for Draining the Reactor Vessel in accordance with operational procedure.
II ~ No. This procedure involves operating equipment in its designated mode of operation.
While two of the flushes are technically classified as Operating Potentials To Drain Reactor Vessel, the necessary precautions are taken to prevent draining the vessel.
FSAR Section 6.3.1.1.1 encompasses vessel coolant losses of much greater magnitude than this procedure could cause.
III. No. This procedure complies with the following Technical Specification sections: 3.6.5.3, 3.6.5.1, 3.6.5.2 and 3.5.3, and does not reduce the margin of safety as defined in the basis for any Technical Specification
'kEE gg '3-144 E E ENC DCP93-9046, Unit1 C C G This modification will allow the addition of a bracket to various Masoneilan valves which will prevent the valve stem from rotating. The bracket will be installed on the valve by utilizing the existing upper limit switch bolting location on the valve yoke.
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No. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15, because the modification does not alter the original design basis of the valves nor any of their systems as described in the FSAR.
The modification meets all of the design requirements specified in the FSAR, and, therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. This modification ensures the remote indication of valve operator positioning for various Masoneilan valves in the plant resulting in protection of the integrity of the applicable systems. The proposed action does not involve a precursor of or a contributor to any evaluated accidents involving offsite dose. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. The proposed action has no adverse affect on valve operability, nor on the system dynamic qualification. The proposed modification also does not introduce any previously unanalyzed failure modes into the overall design of the valves or the systems in which they are contained. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than as previously evaluated in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by the Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification.
The actions taken by this modific'ation will not reduce any margin of safety as defined in the basis for any Technical Specification.
S~EJgg '3-145 C 0 C DCP 93-9047, Unit 2 D C C E'his modification will allow the addition of a bracket to various Masoneilan valves which will prevent the valve stem from rotating. The bracket will be installed on the valve by utilizing the existing upper limit switch bolting location on the valve yoke.
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No. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15, because the modification does not alter the original design basis of the valves nor any of their systems as described in the FSAR.
The modification meets all of the design requirements specified in the FSAR, and, therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. This modification ensures the remote indication of valve operator positioning for various Masoneilan valves in the plant resulting in protection of the integrity of the applicable systems. The proposed action does not involve a precursor of or a contributor to any evaluated accidents involving offsite dose. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. The proposed action has no adverse affect on valve operability, nor on the system dynamic qualification. The proposed modification also does not introduce any previously unanalyzed failure modes into the overall design of the valves or the systems in which they are contained. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than as previously evaluated in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by the Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification.
The actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
C S E E ENCE DCP93-9057, Unit1 0 C This modification provides for increasing the cable length of the signal cable for the Unit 1 Intermediate Range Monitor (IRM) 1G detector assembly through installation of a plug jack connection in the undervessel cable tray and addition of new cabling from the plug jack to the IRM 1G detector.
I. No. No potential safety impact items are created by this installation. No degradation in signal quality will occur. The cabling is electrically equivalent and the new cable connectors are completely compatible with the existing installation. The probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR Chapters 6 and 15, is not increased.
II ~ No. Since the proposed change does not adversely affect the operation of theIRM System, the on-site radiological consequences will not be impacted. No new radiological consequences can occur. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR. This modification will not result in any risk to public health and safety. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR. (Reference FSAR 7.6.1a.5.4)
III. No. Tech Spec Basis 3/4.3.1 specifies the minimum operability and surveillance requirements involving nuclear instrumentation. Similarly, Tech Spec 3/4.3.6 and Tech Spec 3/4.4.1 have been reviewed. The proposed modification has no impact on the above mentioned Tech Specs. The proposed modification does not adversely affect the logic, control, or operation of any safety related plant system or component.
Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specifications:
CE TP-159(259)-008, Unit 1 8 2 DE CR 0 C G The proposed action is use of the P.E. Main Steam Line (MSL) Plugs with fuel in or out of the Reactor Pressure Vessel (RPV) and performance of TP-159(259)-008 at a maximum MSL pressure of 45.5 psig. The action does not permit movement of fuel in the RPV while any plug is under a positive LE'hat could cause it to eject into a fuel bundle that is being moved in the RPV.
MRlhBY I. No. There are no accidents or malfunctions in the FSAR that are specifically related to the MSL Plug. However, there are accidents evaluated in the FSAR concerning load drops onto irradiated fuel. The specific event analyzed in the FSAR covering objects falling onto the core is the fuel handling accident described in Section 15.7.4. Since the P.E.
MSL plug is used for testing at the LLRT pressure, it is subject to ejection due to improper installation or component failure and, therefore, would be able to fall on the core. Since the MSL Plug is prevented from falling on the core when ejected, per the conditions stated in the proposed action, the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR, is not increased.
II. No. The MSL Plug has been demonstrated by analysis to remain intact during a seismic event. It is also designed to limit leakage from the reactor cavity through the MSL to less than 150 gpm, which is well within the makeup capability available with irradiated fuel in the RPV or in the spent fuel pool with the gates removed. With all of the fuel in the spent fuel pool and the gates installed the reactor cavity is isolated from the spent fuel pool. Therefore, a reactor cavity/reactor draindown event would not occur and an accident of a different type than previously evaluated in the FSAR would not be created.
III. No. The minimum water level specified in Tech Spec 3/4.9.8 is 22 feet above the reactor vessel flange. The normal water level during refueling operation is 24 feet above the RPV flange. The P.E. Modified/Upgraded Plugs are designed to limit leakage from an individual plug to less than 150 gpm if both the primary and secondary seals were to fail. This leakage value is well within the makeup capabilities of the system. With the gates installed, the spent fuel pool would be isolated from the RPV and a leak from a MSL Plug would not have an impact on Nuclear Safety.
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C SS E C DCP 93-9020, Unit 1 D C C GE'he action is to modify the control logic of valve F059 to automatically CLOSE the valve when the HPCI turbine is OFF and to OPEN the valve when the HPCI turbine is ON. This is done by adding a new annunciator isolation relay, and replace existing K12 trip relay with a relay with additional contacts.
No. Sections 3.6,6.3, and Chapter 15 of the FSAR were reviewed. Changing the control logic of cooling water valve F059 to OPEN when the HPCI turbine is ON and CLOSE when the HPCI turbine is OFF does not increase the probability of occurrence of an accident as previously evaluated in the FSAR because the modification does not add or change any components whose failure would initiate an accident or which are used prior to an accident.
This modification only changes the post accident operation of valve F059. This modification does not increase the consequences of an accident as previously evaluated in the FSAR because the modification does not change the accident mitigating design or performance of the HPCI system. This modification does not increase the consequences of a malfunction as previously evaluated in the FSAR because the modification does not change the HPCI system response to an accident. This modification results in a small probability that lube oil cooling will not be re-established on the second start due to F059 not opening. This potential for loss of lube oil cooling is offset by the elimination of the potential piping system damage on the second start due to the turbine flooding while it is tripped. It is concluded that this change to the control logic of cooling water valve F059 does not increase the overall probability of occurrence of a malfunction of HPCI.
II. No. FSAR Section 6.3 and Chapter 15 were reviewed. The HPCI turbine, pump, piping, support structures and initiation logic is not being changed by this modification. Changing valve F059 logic to parallel the HPCI turbine operation does not create the possibility of an accident of a different type than previously evaluated in the FSAR because the affected HPCI system components only operate in response to an accident and the modified components only effect post accident operations. The modification does not create a possibility for malfunction of a different type than any previously evaluated in the FSAR because the replacement trip relay has the same failure modes as the relay it replaces and failure of F059 is already addressed as a failure of the HPCI system III No.
~ Changing the control logic of cooling water valve F059 to OPEN when the HPCI turbine is ON and to CLOSE when the HPCI turbine is OFF does not reduce any margin of safety as defined in the basis for the Technical Specifications. Review of the Technical Specification Bases (Section 2.0) and Limiting Conditions for Operation and Surveillance Requirements for HPCI (Section 3/4.5.1; Page B3/4 5-1) show that there are no explicit margins of safety defined for this system. System initiation and flow delivery in response to an accident is not changed by this modification, hence existing system margins are not effected by this modification.
E E E C DCP92-9051, Unit2 E C 0 C The proposed action will eliminate PDAHL-27431 and will install local pressure differential indication to monitor Turbine Building differential pressure.
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The Turbine Building HVAC System has no safety related function. A review of the Susquehanna SES accident analysis regarding 10CFR part 100 commitments dictates that the only accident which could be affected by these alarms is the break of the main steam line outside containment. In the analysis of this leak, it is assumed that any releases from the main steam line go directly to the atmosphere and are not filtered through the Turbine Building ventilation system. Therefore, the Turbine Building differential pressure alarms would have no part in this analysis. The proposed action will not impact any functional requirements of the Turbine Building HVAC System, therefore, no adverse effects are possible. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15. The elimination of PDAHL does not impact any safety related equipment since no connection to any safety related circuitry exists. This modification will not adversely affect any structure, system, or component in performing its safety function, and in fact will eliminate a nuisance alarm. Therefore, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. Since the proposed change does not adversely affect the operation of the Turbine Building HVAC System, the on-site radiological consequences that are required to be evaluated for Control Room habitability will also not be impacted.
Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. The proposed action will not result in any risk to public health and safety. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible. The proposed change does not conflict in any way with FSAR Section 9.4.4, Turbine Building Ventilation. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III No.
~ There is no limiting condition for operation in the Technical Specification for the Turbine Building HVAC System. The proposed changes will have no adverse affect on Turbine Building HVAC operation. Therefore, the proposed action will not reduce the margin of safety as defined in the bases of the Technical Specifications.
'J S~E~O'3-150 C DCP 93-3047, Unit 1 C
GE'o prevent significant non-condensible gas accumulation (>150 ppmv) in the reference leg, a backfill system will provide a continuous flow of reactor grade water into the reference legs connected to condensing chambers XY-14202 and XY-B21-1D002 I. No. The proposed modificatio will tie the Control Rod Drive (CRD) system discharge water header to instrument racks 1C224/1C225. This will provide a flowrate which will force non-condensibie laden water back into the Reactor Pressure Vessel (RPV). The only adverse effects that could be postulated to result from this modification is a reduction in reference leg integrity due to leakage, or adverse effects on the instrumentation from these legs. Both postulated effects have been evaluated. It can be concluded that the backfill modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. Design considerations ensure that the integrity of the reference leg is maintained during all modes of operation and that no adverse effects on plant equipment are introduced. The reliability of the level indication system is actually enhanced, and the backfill system will ensure that reference leg volume remains unaffected by non-condensible gases.
II. No. It may be postulated that a new equipment malfunction scenario could be introduced by the modificatio by leakage through all check valves coincident with loss of CRD pressure. The common mode leakage scenario is not considered credible however, since the probability of check valve leakage has been determined to be acceptably low.
The valves will be determined to provide a leak tight barrier by testing. The backfill system tests also ensure that leakage in the reference legs will be promptly detected. If a leak were to develop or one already existed in the system, it is highly unlikely that leaks in separate reference legs would be of the same magnitude. If different leakage rates developed in separate reference legs, they would be detected via channel checks. Therefore, it can be concluded that the probability of check valve leakage leading to a common mode failure of the level instrumentation is not considered credible. There is reasonable assurance that reference leg integrity will be compromised by this modification and that no safety impacts on existing equipment are created.
III. No. The proposed action does not reduce the margin of safety as defined in the Technical Specifications. The modification will actually increase the margin of safety by preventing non-condensible gas accumulation in the reference legs. Currently, during RPV depressurizations level errors are possible due to non-condensible gases trapped in the reference legs coming out of solution. This modification removes the non-condensible gases in the reference legs, eliminating these potential level errors.
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!~EGO 93-1 51 C OSS C 'CP 93-3048, Unit 2 To prevent signiTicant non-condensible gas accumulation (>150 ppmv) in the reference leg, a backfill system will provide a continuous flow of reactor grade water into the reference legs connected to condensing chambers XY-24202 and XY-B21-2D002 I. No. The proposed modification will tie the Control Rod Drive (CRD) system discharge water header to instrument racks 2C224/2C225. This will provide a flowrate which will force non-condensible laden water back into the Reactor Pressure Vessel (RPV). The only adverse, effects that could be postulated to result from this modiTication is a reduction in reference leg integrity due to leakage, or adverse effects on the instrumentation from these legs. Both postulated effects have been evaluated. It can be concluded that the backfill modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. Design considerations ensure that the integrity of the reference leg is maintained during all modes of operation and that no adverse effects on plant equipment are introduced. The reliability of the level indication system is actually enhanced, and the backfill system will ensure that reference leg volume remains unaffected by non-condensible gases.
II. No. It may be postulated that a new equipment malfunction scenario could be introduced by the modification by leakage through all check valves coincident with loss of CRD pressure. The common mode leakage scenario is not considered credible however, since the probability of check valve leakage has been determined to be acceptably low.
The valves will be determined to provide a leak tight barrier by testing. The backfill system tests also ensure that leakage in the reference legs will be promptly detected. If a leak were to develop or one already existed in the system, it is highly unlikely that leaks in separate reference legs would be of the same magnitude. If different leakage rates developed in separate reference legs, they would be detected via channel checks. Therefore, it can be concluded that the probability of check valve leakage leading to a common mode failure of the level instrumentation is not considered credible. There is reasonable assurance that reference leg integrity will be compromised by this modification and that no safety impacts on existing equipment are created.
III No. The proposed action does not reduce the margin of safety as defined in the Technical
~
Specifications. The modification will actually increase the margin of safety by preventing non-condensible gas accumulation in the reference legs. Currently, during RPV depressurizations level errors are possible due to non-condensible gases trapped in the reference legs coming out of solution. This modification removes the non-condensible gases in the reference legs, eliminating these potential level errors.
S~E~ 93-152 C E E C DCP92-9040Z, Unit1 D C The proposed action is to: Replace the 93 percent Degraded Voltage Relays with similar degraded voltage relays having a narrower, adjustable dead band and a harmonic filter; Change the time delay setpoint of the 93 percent Degraded Voltage Relays from 1 second to 7 seconds; Change the time delay setpoint of the Degraded Voltage LOCA Timer Relays from 9 seconds to 3 seconds; Change the time delay setpoint of the Degraded Voltage Non-LOCA Timer from 5 minutes to 4 minutes and 53 seconds; Change the time delay setpoint of the Degraded Voltage Alarm Relays from 9 seconds to 3 seconds; Replace the 96.5 percent Incoming Feeder Undervoltage Relays with similar undervoltage relays having a harmonic filter.
I. No. The proposed action does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus the proposed action does not increase the probability of occurrence of an accident. The proposed actions represent no change in the probability of occurrence of a malfunction of equipment. The time delay setpoint change does not affect the reliability of the relays. The consequences of an accident are not changed by the proposed action. The proposed action does not affect the actuation of the degraded voltage protection logic or the bus transfer scheme. The replacement of the Incoming Feeder Undervoltage relays and Degraded Voltage relays do not cause relay setpoint shifts which could lead to misoperation of safety related equipment. The replacement of the Degraded Voltage Relays assures that the starting of the Condensate Pump 2D approximately 70 seconds after an accident does not create a delayed LOOP condition on any Class 1E 4.16 kV bus(es). The consequences of a malfunction of equipment is not affected by the proposed action.
II. No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in the Degraded Voltage Protection logic does not create a malfunction of a different type. A random single failure in the logic during normal operation has no impact since the circuitry is de-energized. A random single failure in the Incoming Feeder Undervoltage Protection logic does not create a malfunction of a different type. A random single failure in the logic during normal operation has the same affects as an inadvertent operation of the logic which has no impact.
III. No. The operability of the Incoming Feeder Undervoltage Relay is not governed by Technical Specifications. The operability of the Degraded Voltage Relays and Degraded Voltage Timer Relays is governed by Technical Specification 3/4.3.3, establishing the required number of operable channels, the setpoints and response times. The proposed action does not affect the required number of operable channels, the setpoints or the response times.
The proposed action does not reduce the margin of safety as defined in any Technical Specification.
QEE.59 '3-153 C C DCP 92-9041Z, Unit 2 DESC I 0 C ANGE'he proposed action is to: Replace the 93 percent Degraded Voltage Relays with similar degraded voltage relays having a narrower, adjustable dead band and a harmonic filter; Change the time delay setpoint of the 93 percent Degraded Voltage Relays from 1 second to 7 seconds; Change the time delay setpoint of the Degraded Voltage LOCA Timer Relays from 9 seconds to 3 seconds; Change the time delay setpoint of the Degraded Voltage Non-LOCA Timer from 5 minutes to 4 minutes and 53 seconds; Change the time delay setpoint of the Degraded Voltage Alarm Relays from 9 seconds to 3 seconds; Replace the 96.5 percent Incoming Feeder Undervoltage Relays with similar undervoltage relays having a harmonic filter.
I~ No. The proposed action does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus the proposed action does not increase the probability of occurrence of an accident. The proposed actions represent no change in the probability of occurrence of a malfunction of equipment. The time delay setpoint change does not affect the reliability of the relays. The consequences of an accident are not changed by the proposed action. The proposed action does not affect the actuation of the degraded voltage protection logic or the bus transfer scheme. The replacement of the Incoming Feeder Undervoltage relays and Degraded Voltage relays do not cause relay setpoint shifts which could lead to misoperation of safety related equipment. The replacement of the Degraded Voltage Relays assures that the starting of the Condensate Pump 2D approximately 70 seconds after an accident does not create a delayed LOOP condition on any Class 1E 4.16 kV bus(es). The consequences of a malfunction of equipment is not affected by the proposed action.
II. No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in the Degraded Voltage Protection logic does not create a malfunction of a different type. A random single failure in the logic during normal operation has no impact since the circuitiy is de-energized. A random single failure in the Incoming Feeder Undervoltage Protection logic does not create a malfunction of a different type. A random single failure in the logic during normal operation has the same affects as an inadvertent operation of the logic which has no impact.
III. No. The operability of the Incoming Feeder Undervoltage Relay is not governed by Technical Specifications. The operability of the Degraded Voltage Relays and Degraded Voltage Timer Relays is governed by Technical Specification 3/4.3.3, establishing the required number of operable channels, the setpoints and response times. The proposed action does not affect the required number of operable channels, the setpoints or the response times.
The proposed action does not reduce the margin of safety as defined in any Technical Specification.
3 Same '3-154 E E E C NL-93-024, Unit1 DEC 0 C Replace 232 irradiated SPC 9x9-2 fuel assemblies with 232 fresh SPC 9x9-2 assemblies. The remaining assemblies in the reactor core will be reshuffled to obtain a core configuration which will reliably provide the required energy for Cycle 8 operation and maintain the same margin of safety as in previous cycles under normal operating conditions, during anticipated operational occurrences, and for postulated accidents.
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I. No. The applicable sections of the FSAR which are affected by the core design change are Chapters 4, 5, 6, and 15. With the exception of core stability, for which changes in the core design itself can influence the probability of the occurrence of core instabilities or limit cycles, the core design change will not affect the failure mode of any plant system or component, nor will it affect the probability of occurrence of any transient or accident initiating event. PP8L has implemented NRC Interim Correction Actions as Technical Specifications, and the use of these Technical Specifications will assure the same low level of probability of the occurrence of core instabilities as with previous cores. The Unit 1 Cycle 8 core design change will not increase the probability of occurrence of an accident or malfunction of equipment important to safety, as previously evaluated.
II ~ No. The Unit 1 Cycle 8 core design change results in a core design which is mechanically, thermal-hydraulically, and neutronically very similar to previous core designs. It does not create any new or different initiating events, failures, or failure modes which have not been previously considered or evaluated, nor does it make any accidents or malfunctions previously considered incredible, any more credible. As a result of this change there are no new accidents or malfunctions which must be considered within the design basis of the plant.
III. No. The applicable Technical Specification sections include 2.0, 3.1, 3.2, and 3.4. The Unit 1 Cycle 8 core design change does not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The reload licensing analyses provide Unit 1 Cycle 8 operating limits that will maintain an equivalent margin of safety as currently defined in the basis of the applicable Technical Specification sections. Therefore, the implementation of the Unit 1 Cycle 8 core design change will not reduce the margin of safety as defined in the basis for any Technical Specification.
t C DCP 93-3052B, Unit 2 DE C C G DCP 93-3052B increases the Reactor Water Cleanup (RWCU) design process flow rate. To support the increased flow rate, the following instrumentation changes will be made as part of this modification: The setpoint for the non-regenerative heat exchanger discharge high temperature switch will be increased from 130 F to 135 F.; The setpoint for the non-regenerative heat exchanger discharge high-high temperature switch will be increased from 140 F to 145 F.; The RWCU filter demineralizer flow controllers will be adjusted to achieve up to a 15 GPM increase in flow through each filter/demineralizer.; The setpoint for the leak detection high flow instrumentation will be increased from 426 GPM to 462 GPM.; The flow indicator scales on Fl-G33-2R605A/B will be replaced to accommodate the higher flow rates.
I~ No. Review of FSAR Sections 5.2.5, 5.4.8.1.1, 7.3.1.1a.2, 7.6, 7.7.1.8, and 15.6 determined that the changes being proposed for this modification will not increase the probability of a postulated accident previously identified and evaluated in the FSAR. Since no new hardware or system configuration changes occur as a result of this modification, the RWCU System isolation response time, from time of break detection to isolation, is unchanged. No protection interface features of equipment important to safety are impacted by this proposed activity; therefore, the probability of occurrence of a malfunction of equipment is not increased by this modiTication. The consequences of any previously evaluated accident will not increase as a result of this modification since no radiological consequences are altered by the proposed changes to the RWCU system and monitoring instrumentation. The RWCU system operation and equipment is not changed as a result of this proposed modification.
This modification will not cause a malfunction of equipment, important to safety, previously evaluated in the FSAR.
II. No. This modification creates no new pathways to the environs. The possibility for an accident of a different type, than any previously evaluated in the FSAR has not been created since the functionality of the RWCU system and the RWCU leak detection system is unchanged.
The new temperature switch settings maintain sufficient temperature margin to protect the demineralizer media from unacceptable temperature damage and performance degradation and also does not increase the probability of inadvertent alarms/isolations. This modification does not create the possibility of a malfunction of a different type since the operational logic and functionality of the RWCU leak detection system is unchanged.
Ill. No. The new proposed Technical Specification limits will result in a negligible reduction in the margin between RWCU System isolation and the 4351 gpm flow. This instrumentation serves no safety function. Therefore, the revised setpoint will not reduce the margin of safety.
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~E~ 93-156 C OS C DCP 93-3013NB, Unit 1 Modify HV-151F028NB by replacing the actuator/motor, yoke and bolting. The scope of this DCP also involves the replacement of the circuit breaker, resetting of the breaker magnetic trip, the installation of a new overload heater coil in the motor control center cubicle of the motor-operated valve, the installation of current monitoring equipment in the valve limit switch compartment for VOTES testing and the deletion of the space heater circuit to the valve operator.
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I. No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. HV-151F028NB are required to close upon receipt of an isolation signal.
This modification will not affect circuit interlocks or signal logic. The valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time is within the design basis. Therefore, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR Section 6.2.4 and Table 6.2-12 are still valid.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-151F028NB to perform their intended design function; 2. the stroke time of the valve is within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability of the valve to open or close when required; 6. does not affect the valve leakage; 7. does not alter the MCC or the valve limit switch compartment seismic and environmental qualification; 8.
does not affect the motor power supply system; 9. does not affect any other equipment in the RHR - Suppression Pool Cooling/Spray Subsystem. The proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the Bases for any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.6.2, 3/4.6.1.6 and 3/4.3.2 were reviewed for this evaluation. Valve seat leakage and integrity are not affected. The valve's stroke time remains within the accident analysis bounds of 90 seconds. The MOV's decreased design stroke time improves system response. This modification has no effect on the Residual Heat Removal - Suppression Pool Cooling/Spray Subsystem intended design function or the basis of the above Tech Specs.
C DCP 93-3027, Unit 1 DE C IP C G Modify valve HV-155F002 by replacing the actuator, motor, yoke, yoke clamp and disc. The scope of this DCP also involves the installation of new overload heater coils in the motor control center cubicle of the Motor Operated Valve (MOV), the installation of current monitoring equipment in the valve limit switch compartment for VOTES testing and the deletion of the limit switch space heater circuit.
I. No. The modification has no impact on the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. The valve leakage criteria is not changed by this modification. LLRT requirements of FSAR Section 6.2.6.3 and table 6.2-22 will be maintained. The only equipment affected by this modification is HV-155F002. The valve's pressure retaining capability is maintained, the leakage characteristics are unaffected, and the stroke time remains within the design basis.
Therefore this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously analyzed in Chapter 15 of the FSAR. Also, the analysis described in FSAR section 6.2.4, Table 6.2-12 and 6.2.4.3.2.7 are still valid.
II. No. The proposed modification: 1. enhances the ability of HV-155F002 to perform its intended design function; 2. does not increase the stroke time of the valve beyond that contained in the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. does not affect the MCC and valve limit switch compartment seismic and environmental qualification; 5. improves the ability to open or close the valve when required; 6. does not affect valve leakage; 7. does not affect the motor power supply system; 8. does not affect any other equipment in the HPCI system. It is concluded that the proposed modiTication does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety defined in the bases of one Technical Specification. Tech Spec bases 3/4.5.1, 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.3 were reviewed for this evaluation. Valve seat leakage and integrity are not affected. The valve stroke time is within the accident analysis bounds of 50 seconds. There are no effects to ant other system/components by this modification. Therefore, this modification has no effect on the HPCI systems design function or the basis of the above Tech Specs.
t" 'r Qgg.~ 93-158 C E E CE DCP 93-3040B, Unit Common C G The proposed modification will fabricate a steel grating docking station that will allow access to the rear of the vendor demineralizer truck and clear the Circulating Water Pumphouse (CWPH) doorway. Two hose connections and related piping and valves will be installed to tie in the vendor demineralizer truck. These modifications will provide automatic isolation of the vendor truck on high silica or high conductivity, the ability to rinse in the truck to the neutralizer basin and the ability to isolate the plant piping and drain the vendor hoses during freezing conditions.
No. This modification will shift the inlet for the vendor makeup demineralizer from the Domestic Water Header to the Clarified Water 120 psig header. This modification will restore the clarified 120 psig header to its original purpose which is to supply water to the makeup demineralizers. FSAR Chapters 6 and 15 were reviewed with particular attention paid to FSAR Section 15.2.3 and 15.2.5. Loss of seal water to the Circulating Water Pumps could cause them to trip resulting in a turbine trip and loss of condenser vacuum. This modification will not result in an increase in the probability of occurrence of these events due to the storage capabilities of the Well Water and Clarified Water Storage Tanks and due to the capability to make clearwater from the vendor truck within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of notification. The Makeup Demineralized Water, Domestic Water and Raw Water Treatment Systems and the systems to which they connect, are not safety related. The probability that a component failure which would prevent the Domestic Water, Makeup Demineralized Water or Raw Water Treatment Systems from performing their design function is not increased by installing these hose connections or the cross-tie due to the reserved water which is stored and the ability to restart the clarifier or vendor truck. Also, the probability of contaminating the Demin Water Storage Tank is reduced by the installation of high quality silica and conductivity instruments.
Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
II. No. The Domestic Water, Makeup Demineralized Water and Raw Water Treatment Systems will be able to operate and function as specified in the FSAR. The proposed action will not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by Technical Specifications. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this'odification.
The modifications will be designed to meet all applicable codes; consequently, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
S~E5Q '3-159 C C DCP 93-9060A8 B, Unit 1 DEC I CE This modification will install new 6" manual gate valve to the 6" GBD-115 condensate piping upstream of the existing motor operated HV-10659A,B&C valves. These new gate valves will provide an alternate means of isolating the 6" fill lines leading to the drain coolers and feedwater heater strings A,B8C should the HV-10659A,B8C globe valves leak-by.
I. No. This modiTication will have no adverse affects on the operation or function of the Condensate System as defined in the FSAR. The modification will satisfy all the design requirements of the ANSI B31.1 Code; thereby, satisfying the design requirements as specified in FSAR Table 3.2-1. FSAR Sections 10.4.7, Chapter 6 and Chapter 15 have been evaluated. This modification will not increase the probability of occurrence or consequences of an accident previously evaluated in the FSAR. The modification will conform to all the design criteria as specified per the FSAR and the flow characteristics of the affected piping will be virtually unchanged. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification will meet all the design basis criteria as specified in the FSAR.
Therefore, the proposed actions do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III No. The piping and components affected by the actions taken in this
~ modification are not specifically addressed in the basis for any Unit 1 Technical Specification. The modification meets all applicable design criteria as speciTied in the FSAR. The proposed actions will not reduce any margin of safety as defined in the basis for any Technical SpeciTication.
SRRKQ 93-160 C S EFE E C DCP93-3050, Unit2 DESC C G DCP 93-3050 will revise the Safety Relief Valve (SRV) safety setpoints, revise the setpoints of the Main Steam Relief Valve (MSRV) relief pressure switches and make Safety Parameter Display System (SPDS), process computer and simulator changes required to incorporate the revised setpoints.
U No. This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Sections 5.2.2, 7.2.3 and 15.1, 15.2 and 15.6.
The revised setpoints and proposed groupings will not increase the probability of occurrence of this type of accident. This modification does not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. The margin between peak allowable pressure and the maximum safety setpoint is unchanged. There is no indication that SRV setpoint changes contribute to an increased probability of SRV malfunction. The Simmer Margin is slightly reduced but will be compensated for by more stringent leak test requirements during valve refurbishment. This modification does not increase the consequences of an accident as previously evaluated in the FSAR. The consequences of accidents described in the FSAR are not changed by this modification. This modification does not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the FSAR. Changing SRV setpoints does not make failure more likely or affect the consequences in any manner. The SPDS and Process computers are not safety-related and do not impact any FSAR analysis.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR. This modification does not involve any hardware changes or changes to system function. This modification does not create the possibility of a malfunction of a different type than any evaluated previously in the FSAR. There are no changes resulting from this modification other than pressure relief setpoints. The SPDS and Process computers are not safety-related and do not impact any FSAR analysis.
III. No. Technical Specification 3.4.2 describes the limiting conditions for operation of the SRVs. Calculations demonstrate that reactor pressure will be limited to within ASME Section III allowable values for the worst case upset transient. The margin of safety is inherent in the ASME III allowable pressure values. The margin of safety is not reduced by this modification.
Qgg.gg '3-161 C OS E E E CE DCP93-3058E, Unit1 Eliminate the affected fibrous NUKON insulation in the drywell, within 7 pipe diameters of a postulated High Energy Line Break (HELB), by replacing the existing NUKON insulation with metallic reflective insulation.
I~ No. This modification does not increase the probability of an accident evaluated in the FSAR including FSAR and SER Sections 6.2.2, 6.3.2, and 15.6.5. Failure of insulation does not result in an accident evaluated in the FSAR. This modification does not increase the probability of a malfunction previously evaluated in the FSAR. The malfunction of equipment located in the drywell due to environmental degradation will not increase as the metal reflective insulation will meet to the same heat transmission rate design criteria used for the existing NUKON insulation. This modification does not increase the consequences of an accident as previously evaluated in the FSAR. FSAR Sections 15.6.5, 6.1.1, and 6.2.2 evaluate accidents in which failure of the insulation can affect the consequences. By removing the NUKON insulation and replacing it with metal reflective insulation the potential of blocking the RHR and CS suction strainers is reduced. This modification does not increase the consequences of a malfunction of equipment important to safety. The proposed action does not affect the failure modes of any safety equipment.
II. No. The proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR. Failure of reflective metal insulation is evaluated in FSAR Section 6.2.2.3 and SER Section 6.2.2. The replacement reflective metal insulation meets the requirements of FSAR Section 3.2.1.
III. No. Technical Specification 3/4.5.1 and 3/4.5.3 concern ECCS operation and the Suppression Pool. The modification has no effect on these bases. Technical Specification bases 3/4.6.1.7 discusses the drywell average air temperature limit of 135 F. As the replacement reflective metal insulation has the same heat transmission rate design criteria as used for the existing NUKON insulation, it will have no effect on the margin of safety for this limit. This modification does not reduce the margin of safety of the bases of any Technical Specification.
S~E~O. 93-1 62 C E C DCP 93-3072, Unit 2 C AN Eliminate the affected fibrous NUKON insulation in the drywell, within 7 pipe diameters of a postulated High Energy Line Break (HELB), by replacing the existing NUKON insulation with metallic reflective insulation and eliminate the affected fibrous insulation on the Reactor Building Cooling Water (RBCW) system piping, within 7 pipe diameters of a postulated HEI B, in the drywell by replacing the existing, stainless steel jacketed, fiberglass insulation with, stainless steel jacketed, phenolic foam insulation.
SDUmlhHY No. This modiTication does not increase the probability of an accident evaluated in the FSAR including FSAR and SER Sections 6.2.2, 6.3.2, and 15.6.5. Failure of insulation does not result in an accident evaluated in the FSAR. This modification does not increase the probability of a malfunction previously evaluated in the FSAR. The malfunction of equipment located in the drywell due to environmental degradation will not increase as the replacement metal reflective insulation will meet to the same heat transmission rate design criteria used for the existing NUKON insulation. This modification does not increase the consequences of an accident as previously evaluated in the FSAR. FSAR Sections 15.6.5, 6.1.1, 6.2.2, 6.3.2.2.3.1, and 6.3.2.2.4.1 evaluate accidents in which failure of the insulation can affect the consequences. By removing the NUKON insulation and replacing it with metal reflective insulation the potential of blocking the RHR and CS suction strainers is reduced. This modiTication does not increase the consequences of a malfunction of equipment important to safety.
The proposed action does not affect the failure modes of any safety equipment.
II. No. The proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR. Failure of reflective metal insulation is evaluated in FSAR Section 6.2.2.3 and SER Section 6.2.2. The replacement reflective metal insulation meets the requirements of FSAR Section 3.2.1.
The RBCW system, according to FSAR Section 6.2.4.3.3.5, is not affected by this modification.
III. No. Technical Specification 3/4.5.1 and 3/4.5.3 concern ECCS operation and the Suppression Pool. The modification has no effect on these bases. Technical SpeciTication bases 3/4.6.1.7 discusses the drywell average air temperature limit of 135 As the replacement reflective metal insulation has the same heat transmission rate design criteria as used for the existing NUKON insulation, it will have no effect on the margin of safety for this limit. This modification does not reduce the margin of safety of the bases of any Technical Specification.
g)~gg '3-163 C C DCP 83-0151, Unit Common DE C I C N E'his modification enhances the operation of the Liquid Radwaste Filters. The modification replaces existing high and low level sensors with more reliable sensors, installs a redundant high level overfill control and installs a new analog level indication system which allows the operator to continuously monitor level during the fill process.
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I. No. The changes are not safety related and will be part of a non-safety related system. The proposed modification does not impact any station design features that are used in radioactive release analysis for postulated radwaste system failures. Failure analysis for the proposed changes would be enveloped by the Accident Analysis, "Postulated Radioactive Releases due to Liquid Radwaste Tank Failure," (FSAR Section 15.7.3).
The proposed control logic changes reduce the probability of misoperation of the liquid radwaste system and provide more reliable information to the operator. Therefore, the proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety as previously evaluated in the FSAR.
II. No. Worst case scenarios for any accident or malfunction of the radwaste system have been conservatively analyzed in the FSAR Section 15.7.3. The proposed changes do not involve a change in system operation or add a more severe type of failure mode which would have an effect on this evaluation. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification 3.11.1.3 relates only to liquid waste system operability. This modification will provide an upgrading of the existing filter level instrumentation. These changes will not degrade current system operability. The proposed change has no impact on the Technical Specification, and will not reduce the margin of safety as defined in the basis for any Technical Specification.
C NL-93-027, Unit 1 and 2 This safety evaluation justifies reclassification of the Fuel Grapple Assembly for the Seismic and Quality Assurance categories. The changes are in Seismic Category from I to NA, in Quality Assurance Requirement from Y to N, and in Comments to delete reference to General Note 23.
I. No. While retaining the same design and design features, the proposed action of changing FSAR Table 3.2-1 to reclassify the Fuel Grapple Assembly as non-seismic, non-Q does not create any new failure modes or effects that would cause the consequences of the existing Fuel Handling Accident to increase. Since this safety evaluation is for a non-Q version of the Fuel Grapple Assembly currently approved for use, with all the same design features and no design changes, and it is functionally tested before each use, there is no increase in the probability of occurrence of a Fuel Handling Accident.
FSAR Sections 9.4, 12.2, 13.1, 13.2, 14.2, 15, 18 were reviewed and no changes were deemed necessary for the proposed actions. Therefore the proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Because no design changes are being made, and the existing Fuel Handling Accident Analysis is still valid for the reclassified Fuel Grapple Assembly, the Fuel Grapple Assembly reclassification does not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.
III. No. The reclassification of the Fuel Grapple Assembly does not reduce the margin of safety in the basis of any Technical Specification. This is because of the Refueling Interlocks and operability surveillances are being retained as is and no new requirements need to be established.
E E CE DCP 93-9012, Unit 1 DEC 0 0 C Replace the eight, 48 seat retaining ring cap screws with 5/16" hex head cap screws and tack weld the screws into position on HV-141F032B. The screw material will be changed from commercial grade carbon steel to "Q1" austenitic stainless steel, and replace the hinge shaft on HV-141F032B and install a stainless steel spacer.
I. No. This modification does not increase the probability of an accident. The design, and construction standards are not changed. The new bolting material is in common use in the plant. The basic functions of any plant system are not changed. This modification does not increase the consequences of an accident. The valve stroke time and leakage characteristics are unaffected by the modification. Therefore, the analysis described in FSAR Section 6.2.4 and Table 6.2-12 is still valid. The replacement shaft and addition of the spacer will maintain the long term isolation capability of HV-141F032B. This modification reduces the probability of occurrence of a malfunction of equipment important to safety by reducing the likelihood of seat retaining ring screw failure. The modification will have no other affect on valve performance. This modification does not increase the consequences of a malfunction of equipment important to safety. The modiTication does not affect the design or performance of any equipment which is important to safety. The modification does not affect the containment isolation valve design bases or flow path integrity.
II. No. The modification creates the possibility for interferences that prevent the valves from opening or closing due to the increased disc outside diameter. However, adequate clearance for both installation and operation were confirmed by the vendors. Incorrect dimensions/tolerances of the shaft spacer could result in the disc binding or in the shaft slipping off of the bushings. However, these considerations have been incorporated into the design to preclude this type of malfunction. The modiTication does not create the possibility of accidents or malfunctions of a different type than evaluated in the FSAR. The use of appropriate design, material, and construction standards ensures the integrity of the valves.
III. No. The modification does not reduce the margin of safety defined in the Basis of any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, and 3/4.6.3 were reviewed for this evaluation. The stroke time and leakage characteristics of the valves will not be affected by the modification.
C E CE DCP 92-9056, Unit 1 ESC 0 C DCP 92-9056 will remove the temporary cross-tie between the evaporator concentrates and the distillate piping system and provide a permanent piping cross-tie between the Radwaste Evaporator Distillate Sample Tank, OT-321, and the Radwaste Evaporator Concentrates Storage Tank, OT-322.
I. No. Of the Design Basis Accidents described in FSAR Section 15, only Section 15.7 is affected by this modification. In addition, FSAR Section 2.4.13.3 describes the potential effect on groundwater quality of an accidental release of liquid radwaste due to rupture of the Concentrates Storage Tank. Design Activities ensure that the piping, valves and components utilized will meet the design requirements for system pressure, temperature, flow and materials. Therefore since the systems and components are designed for the required operating parameters, the modification does not increase the probability of occurrence of an accident or malfunction of equipment as previously evaluated in Section 15.7.3 or Section 2.4.13.3.
II. No. The only failure that could be postulated as a result of the permanent cross-tie and new processing mode would be an integrity failure resulting in draining the affected components. This failure has been previously evaluated in the FSAR. FSAR Section 15.7.3 analyzes the complete release of activity from the concentrates waste tank and bounds all liquid radwaste failures. The proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The following Technical Specifications are applicable to release of liquid effluents:
3/4.3.7.10, 3/4.11.1.1, 3/4.11.1.2, 3/4.11.1.3, and 3/4.11.4. The ODCM was changed previously as a result of Bypass 0-92-002 to detail the requirements necessary to meet TS 3/4.11.1.1, 11.1.2, 11.1.3, and 11.4. These changes are still valid for DCP 92-9056..
~Jg) '3-167 C OSS E C DCP 91-9003, Unit Common DESC 0 0 C each existing Laundry Drain Filter with a filter bank, one bank for "A" train and one bank for GE'eplace "B" train. The current Wye strainer will be replaced by a basket type strainer to provide easier and cleaner removal of strained debris.
I. No. DCP 91-9003 replaces the strainers and filters and ensures that the piping and components utilized meet the design requirements for system pressure, temperature and materials.
Therefore, the modification does not increase the probability of occurrence of an accident.
This modification does not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. A malfunction of the strainers or filters is not evaluated in the FSAR and the modification does not affect the performance or operability of the Laundry Waste Processing System. The components are classified as non-safety, non-seismic. This modification does not increase the consequences of an accident of equipment important to safety, as previously evaluated in the FSAR. Since the consequences of the Concentrates Storage Tank rupture are orders of magnitude greater than the consequences of a breach in the Laundry Drain Filters or Strainers, the consequences of an accident previously evaluated in the FSAR are not increased by this modification. This modification does not increase the consequences of a malfunction of equipment important to safety, as previously evaluated in the FSAR. The probability of occurrence of a malfunction of these components is not increased because they meet system design requirements for pressure and temperature and their function in the system has not changed.
II ~ No. This modification does not create the possibility for an accident of a different type than any evaluated previously in the FSAR. The pressure boundary integrity will be maintained by this modification. This modification does not create the possibility for a malfunction of a different type than any evaluated previously in the FSAR. A possible malfunction of the filters is cartridge plugging whereby the pressure differential across the cartridge could cause cartridge failure. The current filter cartridges have a maximum allowablehP of 70 psid, which is less than the maximum differential pressure that can be experienced across the filters, the Laundry Drain Pumps shutoff head. The replacement filter cartridges have a reduced differential pressure capability. Therefore, the potential for this type of malfunction currently exists and will not increase since the margin between alarm and filter differential pressure rating will be maintained.
III. No. The following Technical Specifications are applicable to radioactive liquids: 3/4.3.7.10, 3/4.11.1.1, 3/4.11.1.2 and 3/4.11.1.3. The proposed action does not affect the operability of the affected components nor the dose limits. Therefore, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
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5gR50 '3-'I68 CE DCP 93-3017NB, Unit 2 DEC I Modify HV-251F028NB by replacing the actuator/motor yoke, yoke bolting, actuator bolting and stem nut. The scope of this DCP also involves the replacement of the circuit breaker, resetting of the breaker magnetic trip, the installation of a new overload heater coil in the motor control center cubicle of the motor-operated valve, the installation of current monitoring equipment in the valve limit switch compartment for VOTES testing and the deletion of the space heater circuit to the valve operator.
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No. The modification has no impact on the dynamic qualification of theMotor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. MOV design stroke time will be decreased from 90 seconds to approximately 56 seconds. The new design stroke time is well within the FSAR limit. In addition, the decreased MOV design stroke time will improve system response. This modiTication will not affect circuit interlocks or signal logic. The change in the torque switch setting range and actuator torque rating will ensure sufficient thrust is available do that VOTES and LLRT testing requirements are met and the valves function during all design conditions. This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 on the FSAR.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-251F028NB to perform their intended design function; 2. maintains the valve stroke time within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability of the valve to open or close when required; 6. does not adversely affect the valve leakage; 7. does not alter the MCC or the valve limit switch compartment seismic and environmental qualification; 8. does not affect the motor power supply system; 9. does not affect any other equipment in the RHR-Suppression Pool Cooling/Spray Subsystem. It is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the Bases for any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.6.2, 3/4.6.1.6, and 3/4.3.2 were reviewed for this evaluation. Valve seat leakage and pressure boundary integrity are not affected. The valve's stroke time remains within the accident analysis bounds. This modification has no effect on the. Residual Heat Removal-Suppression Pool Cooling/Spray Subsystem intended design function or the basis of the above Tech Specs.
MES93-169 C S EFE E CE DCP93-3029, Unit2 C 0 C Modify HV-244F001 by replacing the actuator/motor, yoke and yoke clamp. The scope of this DCP also involves the replacement of two circuit breakers, the resetting of the breaker magnetic trip, the installation of new overload heater coils in the motor control center cubicle, the installation of current monitoring equipment in the limit switch compartment for VOTES testing and the deletion of the space heater circuit to the valve operator.
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No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. MOV design stroke time will be well within the FSAR limit. This modification will not affect circuit interlocks or signal logic. The change in the torque switch setting range and actuator torque rating will ensure sufficient thrust is available do that VOTES testing requirements are met and the valve functions during all design conditions. This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 on the FSAR.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-244F001 to perform their intended design function; 2. maintains the valve stroke time within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability of the valve to open or close when required; 6. does not adversely affect the valve leakage; 7. does not alter the MCC or the valve limit switch compartment seismic and environmental qualification; 8. does not affect the motor power supply system; 9. does not affect any other equipment in the Reactor Water Cleanup (RWCU) system. It is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the Bases for any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, and 3/4.3.2 were reviewed for this evaluation. Valve seat leakage and pressure boundary integrity are not affected. The valve's stroke time remains within the accident analysis bounds.
This modification has no effect on the RWCU intended design function or the basis of the above Tech Specs.
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HV-255F002 by replacing the actuator, motor, yoke, yoke clamp and disc. The scope of E'odify this DCP also involves the installation of new overload heater coils in the motor control center cubicle of the Motor Operated Valve (MOV), the installation of current monitoring equipment in the valve limit switch compartment for VOTES testing and the deletion of the space heater circuit to the valve operator.
I. No. The modification has no impact on the dynamic qualification of the MOV as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. MOV design stroke time will be well within the FSAR limit. This modification will not affect circuit signal logic.
Changing HV-255F002 motor actuator will provide sufficient thrust to ensure the valve functions during all design conditions. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 on the FSAR.
II. No. The proposed modification: 1. enhances the ability of HV-255F002 to perform their intended design function; 2. maintains the stroke time of the valve within the Design Basis; 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. does not alter the MCC or the valve limit switch compartment seismic and environmental qualification; 5. improves the ability to open or close the valve when required; 6. does not adversely affect the valve leakage;
- 7. does not affect the motor power supply system; 8. does not affect any other equipment in the HPCI system. It is concluded that the proposed modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the bases for any Technical Specification. Tech Spec bases 3/4.5.1, 3/4.6.1.1, 3/4.6.1.2, and 3/4.6.3 were reviewed for this evaluation. Valve seat leakage and pressure boundary integrity are not affected. The valve stroke time is within the accident analysis bounds. This modification has no effect on the HPCI systems design function or the basis of the above Tech Specs.
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93-171 C E C DCP 92-9057, Unit Common SC 0 C E This modification will: 1) add four 120V AC instrument power available lights in panel OC681.
This will minimize the time and the possibility for the secondary containment isolation system not actuating due to a blown fuse, 2) modify the secondary containment isolation logic with a time delay energize relay which would utilize N.O. contacts in series with the Zone 3 lockout relay, 3) determ and remove five spare relays.
S I. No. The spare relays that will be removed from panel OC876A-A, B-A and the addition of power available indicating lights in panel OC681 have no safety related function. The addition of a time delay relay to the Standby Gas Treatment System (SGTS) circuit will not interfere with the systems intended logic, control or operation, nor does it adversely affect the ability for SGTS to actuate either automatically or manually. This modification does not measurably increase the probability of occurrence or the consequences of the accidents described in FSAR Chapters 6 and 15. This modification significantly reduces the probability of unplanned Engineered Safety Function (ESF) actuations.
Therefore, this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR.
II. No. Since the proposed changes do not adversely affect the operation of the SGTS system, the on-site radiological consequences will not be impacted. There could be no effect on the accidents that have radiological consequences as a result of this modification. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. This modification will be implemented in the Secondary Containment Isolation System and the SGTS. These are addressed in Tech. Specs. 3/4.3.2 and 3.6.5.3. The proposed modification does not adversely affect the logic, control, or operation of any Tech. Spec. related system. None of the parameters that are the bases for the Technical Specifications will be impacted, therefore, no affect on any Tech. Spec.
margin of safety occurs. The removal of spare relays in panel OC876 will reduce the risk of losing secondary containment isolation due to a failed spare relay. The design basis of these systems will not be affected. Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specifications.
//PJ'S) '3-172 C SS C DCP 93-3051, Unit 2 D C E'ach of the Reactor Building Cooling Water System (RBCWS) chillers will be upgraded to increase chiller heat removal capacity. The upgrade consists of replacing shrouds and impellers with components of higher performance ratings.
I. No. FSAR Section 9.4.2 and 9.2.12.3 and Chapter 15 were reviewed and it is concluded that this modification does not increase the probability of occurrence of an accident evaluated in those sections. The upgrade capability of the chillers is necessitated by existing high heat loads during the summer and the expected increase in heat loads resulting from power uprate. The design basis chilled water supply temperature remains unchanged. This modification does not increase the probability of occurrence of a malfunction of equipment important to safety. The equipment involved in this modification is the RBCWS Chillers which do not perform any plant functions which are important to safety. This modification does not increase the consequences of an accident as previously evaluated in the FSAR.
FSAR Section 9.4.2 and 9.2.12.3 and Chapter 15 evaluate accidents or events which could affect the integrity of the fuel barrier and the Reactor Coolant Pressure Boundary, cause reactor pressure vessel pressure increases or raise fuel temperature. None of these accidents or events take credit for RBCWS or Service Water System (SWS) operation because they are non-safety systems. This modification does not increase the consequences of a malfunction of equipment important to safety. RBCWS chillers and the SWS are not safety related and the modiTication does not affect any other equipment.
II ~ No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR. The modification does not involve changes in system function or contribute to a different type of failure mode. Failure of the chillers does not initiate an accident of any type and does not challenge the safety function of other systems.
This modification does not create the possibility of a malfunction of a different type than any evaluated previously in the FSAR. The modification does not change the function or operating modes of the chillers - only the rating is changed.
III. No. This modification improves the ability of the system to achieve the design chilled water loop supply temperature of 50'F under all power uprate conditions. As a result, system performance using one chiller should provide more margin for maintaining the Technical Specification limit. There is no specific margin of safety defined in the basis of any Technical SpeciTication specifically for the chillers. Technical Specification 3.8.3.2 defines the limiting condition for operation of the power distribution system divisions for safe shutdown. This modification will not affect the safety function of the 4.16kV ES Switchgear and therefore will not reduce the margin of safety as defined in the basis for this Technical SpeciTication.
1' SREKO 93-173 C OS E E CE DCP93-9027Z, Unit2 D C 0 CH E'his design modification proposes installation of bypass switches in Control Room Panels 2C645 and 2C644 to provide the following: Capability to bypass Main Steam Isolation Valve (MSIV) Level 1 signal from the Control Room and capability to bypass Containment Instrument Gas (CIG) (Level 1 and high drywell pressure signal) from the Control Room to prevent MSIV closure.
No. The proposed action does not affect any of the postulated initiating events identified in chapters 6,7,15 of the FSAR. The proposed action does not increase the probability of occurrence of an accident regardless of the operating conditions including the Anticipated Transient Without Scram (ATWS) event. The modification does represent an increase in the probability of occurrence of a malfunction of equipment due to the addition of keylock selector switches, isolation relays, fuses and indication lights. It is engineering judgment that the increases in probability due to the additional switches, isolation relays, fuses and indicating lights is considered to be so small and insignificant that the change is within the error bounds associated with the original design calculation. This modification results in an insignificant increase in the likelihood of MSIV isolation failure on level 1 for accidents where this isolation is desired. This modification reduces the chance of core damage and containment failure during ATWS events. Therefore, this modification does not increase the probability of occurrence of a malfunction of equipment analyzed in the FSAR. The consequences of an accident are not changed by this proposed action. The addition of the bypass switches does not affect the safety functions of the existing components.
II. No. With the exception of FSAR Section 15.8,7.1.2a.1.30.1 andAppendix 15A which provides a brief discussion of ATWS, a review of FSAR chapters 6, 7 and 15 did not identify any postulated initiating event relative to the proposed action which would create the possibility for an accident of a different type. The components being added by this modification are dynamically qualified and analyzed for installation in their proposed locations. The failure of one logic channel for MSIVs and one division of CIG would not prevent the safety functions from being performed. Thus, the addition of the bypass function does not create the possibility for an accident or malfunction of a different type.
III. No. The proposed action does not affect any Technical Specification requirements. A modification has a minimal increase in the loads and a minor decrease in the design margins of the 125 VDC System, but has no reduction in the margin of safety that is the basis for the load profiles presented in Technical Specification Sections 4.8.2.1.d.2(b). The Technical Specification load profiles, which are used as the bases for the designs of the 125 VDC System, will not be affected by the proposed modification. Thus, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
E E CE DCP93-9018, Unit1 Part 1: To provide improved protection against overvoltage surges caused by lightning strikes, a transient voltage surge suppressor will be mounted inside of a new weatherproof enclosure on the back of the existing weatherproof enclosure for FT-11503 and connected across the incoming 120 Vac power feed.
Part 2: To vent air trapped in the 4" standpipe which extends up from the 24" blowdown pipe and supports the existing Marsh-McBirney flow element (FE-11503), a vent hole will be provided on the standpipe and above the cooling tower basin water.
I. No. None of the components being changed by this modification are important to safety or affect equipment important to safety as previously evaluated in the FSAR. The required functions of the systems to which these changed components are attached are unaffected. Overall system operation will remain unchanged, i.e., 5000 gpm flow will still be required as the minimum blowdown permissive for liquid radwaste discharge.
Based on the above, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification (a) provides better surge protection by replacing the existing metal oxide varistor with a new transient voltage surge suppressor and protects the power source with a surge suppressor diode and (b) vents the 4" standpipe. These two changes upgrade the performance of a non-safety related instrument loop. The new configuration will perform all the same functions as the existing configuration. No new functions or interfaces with other devices or systems will be introduced. The upgraded non-safety related instrument loop will continue to provide the 5000 gpm minimum blowdown flow interlock that is a permissive for radwaste discharge. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Radiological releases are discussed in sections 3.3.7.10, 4.3.7.10, and 3.11.1.1 of the Technical Specifications, which provides the bases and limiting conditions for operation for liquid effluent releases and associated instrumentation. The new configuration will not affect the existing instrument loop accuracy; therefore, the loop will continue to use 5000 gpm for the minimum blowdown flow permissive to ensure minimum dilution flow is available prior to allowing radwaste effluent discharge valves to be opened. Since the new configuration will not affect the minimum requirements for cooling tower blowdown dilution flow the proposed modification will not reduce the margin of safety as defined in the basis for any Technical Specification.
S~E~(+ 93-175 C DCP 93-9019, Unit 2 Part 1: To provide improved protection against overvoltage surges caused by lightning strikes, a transient voltage surge suppressor will be mounted inside of a new weatherproof enclosure on the back of the existing weatherproof enclosure for FT-21503 and connected across the incoming 120 Vac power feed.
Part 2: To vent air trapped in the 4" standpipe which extends up from the 24" blowdown pipe and supports the existing Marsh-McBirney flow element (FE-21503), a vent hole will be provided on the standpipe and above the cooling tower basin water.
I~ No. None of the components being changed by this modification are important to safety or affect equipment important to safety as previously evaluated in the FSAR. The required functions of the systems to which these changed components are attached are unaffected. Overall system operation will remain unchanged, i.e., 5000 gpm flow will still be required as the minimum blowdown permissive for liquid radwaste discharge.
Based on the above, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. This modification (a) provides better surge protection by replacing the existing metal oxide varistor with a new transient voltage surge suppressor and protects the power source with a surge suppressor diode and (b) vents the 4" standpipe. These two changes upgrade the performance of a non-safety related instrument loop. The new configuration will perform all the same functions as the existing configuration. No new functions or interfaces with other devices or systems will be introduced. The upgraded non-safety related instrument loop will continue to provide the 5000 gpm minimum blowdown flow interlock that is a permissive for radwaste discharge. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Radiological releases are discussed in sections 3.3.7.10, 4.3.7.10, and 3.11.1.1 of the Technical Specifications, which provides the bases and limiting conditions for operation for liquid effluent releases and associated instrumentation. The new configuration will not affect the existing instrument loop accuracy; therefore, the loop will continue to use 5000 gpm for the minimum blowdown flow permissive to ensure minimum dilution flow is available prior to allowing radwaste effluent discharge valves to be opened. Since the new configuration will not affect the minimum requirements for cooling tower blowdown dilution flow the proposed modification will not reduce the margin of safety as defined in the basis for any Technical Specification.
f S~E~ 93-176 C C DCP 93-3062ASB, Unit 2 D C DCP 93-3062A will re-span existing Unit 2 Reactor Recirculation system flow transmitters and corresponding electronics and will re-scale the associated indicators, recorders, and computer inputs.
The loop ranges all will change from 0-50,000 GPM to 0-60,000 GPM. Also, the scale of Jet Pump Flow Recorder RX-B21-2R613 will change from 0-125 percent to 0-125,000,000 pounds per hour.
DCP 93-3062B replaces PT-C32-2N008 Narrow Range Reactor Pressure Transmitter with a new transmitter having a range enough to support power uprate. This DCP also re-scales the pressure pen of PR-C32-2R609 and the following Narrow Range Reactor Pressure computer points: NFP01, NRDP, and NFP001Z.
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I. No. DCP 93-3062A: The Recirculation Flow Control System is described in FSAR Section 7.7.
This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 5 and 15. AII physical and electrical characteristics of the electronic loops are identical. The modification does not change the failure modes of the equipment.
DCP 93-3062B: The Feedwater Control System is described in FSAR Section 7.7. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR. The transmitter replacement portion of this modification replaces the transmitter with a new one which is physically and functionally identical except for the range capsule. The modification does not change the plausible failure modes of the equipment. This modification does not increase the probability of occurrence of a malfunction of equipment.
II. No. DCP 93-3062A: This modification does not create the possibility of an accident of a different type than any evaluated in the FSAR because the re-spanning and re-scaling of Recirculation Flow Control System flow instrumentation merely changes calibration and scale range, all physical and electrical characteristics are identical.
DCP 93-3062B: This modification does not create the possibility of an accident or malfunction of a different type than any evaluated in the FSAR because the re-scaling of the pressure recorder merely changes the scale range and the replacement of the pressure transmitter with the same type of pressure transmitter, but with a different range capsule, merely changes the transmitter range.
III. No. DCP 93-3062A: The scope of this modification is not mentioned in the Technical Specifications and are not related to safety margins.
DCP 93-3062B: The scope of this modification is not mentioned in the Technical Specifications. Feedwater System operation and flow in response to an accident is not changed by this modification, therefore existing system margins are not effected by this modification package.
SEED 93 177 C E CE DCP 93-3063, Unit 2 E C C This modification installs a 480VAC, three-phase power panel (2PP102) and adjacent welding receptacles located near the alcove area on Elevation 699, Area 22, of the Turbine Building. The power panel will be centrally located for any welding power requirements in the condenser bay, feedwater heater cells, or steam pipe tunnel areas.
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I. No. The installation of power panel 2PP102 to supply power for welding equipmenthas no adverse affect on any plant systems or components. FSAR Chapters 6 and 15 were reviewed and this installation would have no affect upon, and will therefore not increase the probability of occurrence or the consequences of any accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II ~ No. FSAR 15.2.6 and 15A, Event 28, discuss accidents/transients related to the loss ofAC power. Failure of the power panel or power feed cables could only result in the trip of the 2B130-033 breaker which is dedicated for this one non-Class 1E application. This modification does not create a possibility for a different type accident or malfunction as discussed in these sections, nor would this modiTication result in a different type accident/transient.
III. No. The AC power supplies necessary to meet these requirements are listed in Section 3/4.8.3.1 and 3/4.8.3.2 of Technical Specifications. 480VAC load center 2B130 is not listed as necessary, nor will the power panel 2PP102 provide power required for operation or shutdown of the unit or for mitigation of the consequences of an accident.
Therefore, the margin of safety is not reduced.
Qgg.~ 93-178 C 0 E E C DCP 93-9044, Unit Common SC T 0 C GE'his modification will provide the documentation to: a) allow the use of existing composite radwaste process shields in addition to existing steel process shields; b) assure the structural integrity of the composite process shields; and c) allow the use of more than two radwaste container storage compartments for processing during the mobile radwaste solidification process. The purpose of this modification is to revise/update all applicable design drawings and documents as necessary to document the allowable use of those shields and compartments as indicated above.
No. This modification provides the documentation for using additional process shields and Radwaste (RW) storage compartments during mobile radwaste solidification processing. This does not alter the design basis or operation of the Radwaste Building Structure as described within FSAR Section 3.2.3.4. This modification does not increase the probability of occurrence of the accidents described in FSAR Chapters 6 and 15, because the modification meets the original design basis and operation of the Radwaste Building as described in the FSAR. The Radwaste Building serves no safety function and because the modification meets all design requirements specified in the FSAR, there will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. Since the components affected by this modification will still comply with the original construction codes as specified in FSAR Table 3.2-1, there is no decrease to the design integrity of the Radwaste Building. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification will not alter the design basis or operation of the radwaste building and will comply with the original construction codes and regulatory requirements.
Therefore, this modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. This modification does not jeopardize or degrade the function or operation of any plant system governed by the Technical Specifications, specifically Section 3/4.11 and 6.15.
The Radwaste Building is not addressed in the Technical Specifications and will still meet the original design basis as specified in the FSAR; consequently, the actions taken by this modification will not reduce any margin of safety as defined in the basis for any Technical Specification.
E E CE. DCP93-9065, Unit1 SC C G This modification will take the following actions: 1. hanger SP-DCA-151-H2011 will be modified by replacing the carbon steel U-bolt with a stainless steel U-bolt and a portion of the inactive GE pipe whip restraint R4A will be removed, 2. hanger SP-DCA-150-H2011 will be modified by replacing the carbon steel U-bolt with a stainless steel U-bolt. This hanger will also be altered to prevent thermal contact with the GE pipe whip restraint, 3. hanger SP-DCA-151-H2010 will have the existing carbon steel U-bolt replaced with a new stainless steel U-bolt, 4. hanger SP-DCA-150-H2009 will have the existing undocumented stainless steel U-bolt replaced by this modification with an approved stainless steel U-bolt, 5. hanger SP-DCA-151-H2009 will also have its existing stainless steel U-bolt retorqued to an approved value, 6. hanger SP-DCA-151-H2008 will also have its existing stainless steel U-bolt retorqued to an approved value.
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I. No. FSAR Chapters 6 and 15 have been reviewed. This modification will insure the integrity of the Reactor Recirculation System piping by eliminating the hanger design flaw. The probability of occurrence of an accident previously evaluated in the FSAR will not be increased. This modification will improve upon the original hanger designs thereby making the recirculation piping less likely to fail. There will be no increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR. The proposed action does not involve a precursor of or contributor to any evaluated accidents involving offsite dose. There will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. This modification will not significantly change the original design arrangement of the small drain and vent piping hangers for the Reactor Recirculation System. Since the modification will ameliorate the support designs and be acceptable per all the requirements of the FSAR, this modification will not create a possibility for an accident or malfunction of a different type than any evaluated previously the FSAR.
III No. This modification does not jeopardize or degrade the function or operation of any plant
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system governed by the Technical Specification. None of the parameters that are the bases for the Technical Specifications will be adversely impacted by this modification.
This modification will allow the Reactor Recirculation piping system to meet all its original design basis as specified in the FSAR. The actions taken by this modiTication will not reduce any margin of safety as defined in the basis to any Technical Specification.
qW QKQJg) '3-180 C OS E E E CE DCP93-3030, Unit2 Modify HV-244F004 by replacing the actuator, motor, stem nut, yoke, and yoke clamp. The scope of this DCP also involves the replacement of the circuit breaker, resetting of the breaker magnetic trip, the installation of new overload heater coils in the DC control center cubicle, and deletion of the space heater circuit to the valve motor and limit switch compartment.
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No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. MOV design stroke time will be within the FSAR limit. This modification will not affect circuit interlocks or signal logic. Changing HV-244F004 motor/actuator provides the required range of torque switch settings, increases the torque rating to ensure sufficient thrust is available so that the valve functions during all design conditions are met. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 on the FSAR.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-244F004 to perform their intended design function; 2. the stroke time of the valve is within the Design Basis; 3. does not decrease the MOVs allowable seismic acceleration below the actual value determined in the piping analysis; 4. involves the replacement of components outside the valve pressure boundary; 5. improves the ability to open or close the valve when required; 6. does not adversely affect the valve leakage; 7. does not alter the DC control center seismic qualification; 8. does not affect any other equipment in the Reactor Water Cleanup (RWCU) system. It is concluded that the proposed modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the bases for any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.3.2 and 3/4.8.2 were reviewed for this evaluation. Valve seat leakage and pressure boundary integrity are not affected. The valve stroke time is within the accident analysis bounds.
This modification has no effect on the RWCU System intended design function or the basis of the above Tech Specs.
'LEE 5g '3-181 C S E E E C DCP93-3031, Unit2 Modify HV-249F007 by replacing the actuator motor, and installing current monitoring equipment in the valve limit switch compartment for VOTES testing.
I. No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. This modification will not affect circuit signal logic. Changing HV-249F007 motor will provide sufficient thrust to ensure the valve functions during all design conditions. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in Chapter 15 on the FSAR.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-249F007 to perform its intended design function; 2. does not change the stroke time of the valve; 3.
does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis; 4. improves the ability to open or close the valve when required; 5. does not adversely affect the valve leakage; 6. involves the replacement of components outside the valve pressure boundary only; 7. does not affect the electrical distribution system; 8. does not alter the valve limit switch compartment seismic and environmental qualification; 9. does not affect any other equipment in the RCIC system. It is concluded that the proposed modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety as defined in the Bases Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.3, 3/4.7.3 and 3/4.3.5 were reviewed for this evaluation. Valve seat leakage and pressure boundary integrity are not affected. The valve stroke time is not changed and remains within the accident analysis bounds. This modification has no effect on the RCIC systems design function or the basis of the above Tech Specs.
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S~E~Q '3-182 C C DCP 92-3024E, Unit 2 D C 0 C The proposed action is to provide centrifugal fans (with built-in filter) to blow ambient air into the Electrical Protection Assembly (EPA) panels. Forced ventilation with ambient air will prevent the buildup of heat within the EPA panels. This enhancement should extend the operating life of the EPA components by lowering the assembly internal air temperature.
5L8RlhBY I. No. The RPS is described in FSAR Section 7.2. System analysis is addressedin Section 7.2.2.
This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Sections 7.2, 15.1, 15.2, 15.3, 15.4, and 15.6. The proposed action improves the environmental operating conditions inside the EPA panels by enhancing internal heat removal. This modification does not increase the probability of occurrence of a malfunction of equipment. The proposed forced ventilation modification is expected to extend the operating life of the electrical components and reduce the probability of occurrence of a malfunction of the EPA's. The modification does not affect Reactor Protection System (RPS) trip sensors or their settings. This modification does not increase the consequences of an accident since thermal related failures of logic cards and other circuitry is no different than an accident previously evaluated. This modification does not increase the consequences of a malfunction of equipment important to safety. Malfunctions of the EPA panels could result in reactor trip. This trip is the fail safe action, therefore, the consequences of a malfunction are not increased.
II. No. This modification does not create the possibility of an accident of a different type. The proposed modification does not involve changes in system operation or contribute to a different type of failure mode. Loss of both RPS power supplies results in a reactor SCRAM, which ensures a safe condition for the plant and initiates isolations to ensure containment integrity. This modification does not create the possibility of a malfunction of a different type.
The modified EPAs are dynamically qualified. Forced ventilation reduces the panel internal temperature below specified operating limits. Reliability and useful life of the EPA is expected to improve as a result of this modification.
III. No. Technical Specification Section 2.2.1 defines RPS instrumentation trip setpoints and allowable values. This modification does not affect trip setpoints. Section 3.3.1 defines limiting conditions for operation. This modification does not affect the minimum number of operable channels required for operation. Section 4.8.4.3 provides the surveillance requirements governing the operability of the RPS EPA's and establishes the setpoints for undervoltage, overvoltage and underfrequency relay actuation. This modification does not affect the setpoints listed in this section. Since no changes are being made to the EPA protection circuitry, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
4 S~E~Q '3-183 C C DCP 93-5004, Unit Common SC 0 0 C DCP 93-5004 will add twelve position keypads to entrance doors and turnstiles in the North and South Gatehouses.
SQRl5hBY I. No. The Security Data Management System (SDMS) is not a safety-related system (FSAR Section 13.6). The addition of cipher function at doors and turnstiles will not impact any functional requirements of safety-related equipment or components; therefore, the possibility of adverse effects do not exist. This modification does not increase the probability of occurrence of an accident or malfunction of equipment described in FSAR Chapters 6 and 15.
II. No. The addition of cipher keypads at entrance doors and turnstiles will not jeopardize public health and safety. No new scenarios or malfunctions of a different type can occur because no harmful safety system effects are possible as a result or consequences of the proposed action. The proposed addition does not conflict with security FSAR commitments. Therefore, DCP 93-5004 does not create the possibility for an accident or malfunction of a different type as previously evaluated in the FSAR.
III. No. There is no limiting condition for operation in the Technical Specifications related to security operation. The SDMS has no surveillance requirements declared in the Technical Specification. The proposed addition will only enhance the administrative access controls of the security operation. Therefore, this DCP will not reduce the margin of safety as defined in the Technical Specification basis.
E E CE NL-93-024, Unit 1 DE C As a result of the revised core loading, this change replaces 224 irradiated SPC 9x9-2 fuel assemblies with 224 fresh SPC 9x9-2 assemblies. The remaining assemblies in the reactor core will be reshuffled to obtain a core configuration which will reliably provide the required energy for Cycle 8 operation and maintain the same margin of safety as in previous cycles under normal operating conditions, during anticipated operational occurrences, and for postulated accidents.
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The applicable sections of the FSAR which are affected by the core design change are Chapters 4, 5, 6, and 15. With the exception of core stability, for which changes in the core design itself can influence the probability of the occurrence of core instabilities or limit cycles, the core design change will not affect the failure mode of any plant system or component, nor will it affect the probability of occurrence of any transient or accident initiating event. The Unit 1 Cycle 8 core design change will not increase the probability of occurrence of an accident or malfunction of equipment important to safety, as previously evaluated. Infrequent events are expected to occur with a much lower frequency and limiting faults are not expected to occur at all during the life of the plant.
As long as the plant continues to operate within the same operating restrictions as previous cycles and in accordance with Unit 1 Cycle 8 Core Operating Limits Report, the radiological consequences of postulated accidents will be within the limits established and previously approved by the NRC. The revised Unit 1 Cycle 8 core loading will not increase the consequences of an accident or malfunction of equipment important to safety, as previously evaluated.
II. No. The Unit 1 Cycle 8 core design change results in a core design which is mechanically, thermal-hydraulically, and neutronically very similar to previous core designs. It does not create any new or different initiating events, failures, or failure modes which have not been previously considered or evaluated, nor does it make any accidents or malfunctions previously considered incredible, any more credible. There are no new accidents or malfunctions which must be considered within the design basis of the plant.
III. No. The applicable Technical Specification sections include 2.0, 3.1, 3.2, and 3.4. The Unit 1 Cycle 8 core design change does not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The reload licensing analyses provide Unit 1 Cycle 8 operating limits that will maintain an equivalent margin of safety as currently defined in the basis of the applicable Technical Specification sections referenced above. Therefore, the implementation of the revised Unit 1 Cycle 8 core loading will not reduce the margin of safety as defined in the basis for any Technical Specification.
SKB~i 93-185 C S E C DCP 87-3001 (Rev 2), Unit 1 During the U1-7RIO replace the existing hold down beams on all twenty jet pumps with new beams which have undergone an improved heat treatment known as high temperature annealing and aging (HTA). This treatment significantly improves their resistance to Intergranular Stress Corrosion Cracking (IGSCC).
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I. No. The new beams being installed have undergone an improved heat treatment which improves the resistance to IGSCC. This reduces the probability of failure of these components. FSAR Section 5.4.1.4 and 15.3 have been reviewed. Change out of the hold down beams does not affect the ability to reflood the core to the top of the jet pump. This change does not have an impact on the reactor recirculation system malfunctions that pose threats of damage to the fuel barrier.
II. No. Since the replacement hold down beams with the improved treatment are functionally identical in every respect to the original hold down beams, they do not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. Technical Specification 3.4.1.2 addresses jet pump operation. All jet pumps shall be operable at all times. Since this modification will not affect the operability of any jet pump, the margin of safety is not reduced. Replacement of the jet pump hold down beams with the beams with improved heat treatment reduces their susceptibility to cracking due to IGSCC and would therefore not reduce any margin of safety.
Qgg.~ 93-1 86 C SS C DCP 91-3022, Unit 1 ME-1RF-003 Rev 1 C
E'his procedure controls the proposed activity of installing a jet pump sensing line clamp on one jet pump instrument sensing line to complete DCP 91-3022 and changeout of 20 jet pump holddown beams per DCP 87-3001.
I. No. The proposed activity does not increase the probability of occurrence or the consequences of accidents or malfunctions of equipment important to safety as evaluated in the FSAR. All equipment is being used within its design basis except that 19 jet pumps are being disassembled and a rod block is being jumpered out. FSAR Section 5.4.1.4 specifies that the Jet Pump assembly provides a flooded volume of 2/3 core height, however, this work is being performed with the core offloaded and therefore, the core height is zero. The FSAR Chapter 6 and 15 events are evaluating situations when fuel is in the reactor. Since no fuel will be in the vessel when this activity will be performed there are no negative effects on safety.
II ~ No. The proposed activity does not create the possibility of an accident or malfunction of a different type than evaluated previously in the FSAR. This is because the core will be defueled during the activity, no new failure modes are created, and the plant will be in an improved condition after the modification is installed.
III. No. The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification. The jet pump operability Tech Spec 3/4.4.1.2 is only required during Condition 1 and 2. Since all this work is being completed with the core defueled no decrease in any margin of safety will occur.
SERMON93-187 C E E CE TP-214-006 RO, Unit 2 C G The Unit 2 Reactor Building Closed Cooling Water (RBCCW) System has developed a leak in the Unit 2 drywell. Two possibilities are a stem packing leak from either HV-21345 or HV-21346.
The purpose of this test procedure is to place HV-21345 and HV-21346 on their backseat one at a time and determine if leakage from RBCCW decreases.
I. No. Per FSAR 9.2.2.1 and 9.2.2.3 RBCCW has no safety related function. This piping is designed to withstand the Safe Shutdown Earthquake such that its failure or loss of function will not impair safety related systems inside containment, will not affect containment isolation on RBCCW, or prevent a safe shutdown of the plant. This test procedure will maintain one operable isolation valve in each RBCCW penetrations at all times and will restore HV-21345 and HV-21346 to operable status at the conclusion of the procedure. Backseating of HV-21313 and HV-21314 will not be performed while HV-21345 and HV-21346 are on the backseat. Therefore, this proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
II ~ No. Loss of RBCCW to Reactor Recirculation pump seal and motor bearings is a possibility and loss of a primary containment isolation valve is another possibility. Per FSAR 9.2.2.3, loss of RBCCW will not prevent safe shutdown of the plant. Primary containment integrity will be maintained because the outboard containment isolation valves, HV-21313 and MV-21314 are available at all times during the test and can be closed to prevent RBCCW drain down. Therefore, the proposed action does not create a possibility for an accident or malfunction of a different type than evaluated previously.
III. No. Tech Spec 3.6.1.1 requires that primary containment integrity be maintained. Tech Spec 3.6.1.2 addresses the leakage requirements from containment isolation valves.
Placing HV-21345 and HV-21346 on the backseat will not change the leakage characteristics of the valve when it is closed. If these valves would break during backseating, the current ILRT and LLRT values will not significantly change. Tech Spec 3.6.3 addresses the operability of containment isolation valves. If either HV-21345 or HV-21346 fails during backseating, this Tech Spec requires isolating the penetration which will stop RBCCW flow to both Reactor Recirculation Pumps.
Therefore, the proposed action does not reduce the margin of safety as defined in Technical specifications.
E C DCP 93-3049, Unit 1 C GE:
DCP 93-3049 will revise the Safety Relief Valve (SRV) safety setpoints, revise the setpoints of the Main Steam Relief Valve (MSRV) relief pressure switches and make the Safety Parameter Display System (SPDS), process computer and simulator changes required to incorporate the revised setpoints.
This modification does not increase the probability of occurrence of an accident evaluated in the FSAR including FSAR Section 5.2.2, 7.2.3 and 15.1, 15.2 and 15.6.
Accidents described in the referenced FSAR sections rely on the SRVs to provide a relief or safety function. In none of these cases are the SRVs listed as the cause or initiator of an accident, therefore, modifications to the setpoints cannot increase the probability of these accidents occurring. This modification does not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. The margin between peak allowable pressure and maximum pressure and the maximum safety setpoint is unchanged. Design stresses remain unchanged. There is no indication that SRV setpoint changed contribute to an increased probability of SRV malfunction. This modification does not increase the consequences of an accident as previously evaluated in the FSAR. The consequences of accidents described in the FSAR are not changed by this modification. This modification does not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the FSAR. The SRV failure mode would be failure to open and relieve pressure. Changing SRV setpoints does not make failure more likely or affect the consequences in any manner. The SPDS and Process computers are not safety-related and do not impact any FSAR analysis.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR. This modification does not involve any hardware changes or changes to system function. This modification does not create the possibility of a malfunction of a different type than any evaluated previously in the FSAR. There are no changes resulting from this modification other than pressure relief setpoints. Increased pressures are bounded by original design conditions, therefore, structural integrity is unaffected. The SPDS and Process Computers are not safety-related and do not impact any FSAR analysis.
Technical Specification 3.4.2 describes the limiting conditions for operation of the SRVs. The margin of safety is inherent in the ASME III allowable pressure values.
Future fuel cycle analyses will account for the revised SRV setpoints and operability requirements. The margin of safety is not reduced by this modification.
l S~bQ '3-189 C ERENC . DCP 93-3018A, Unit 2 DESC 0 C Replace valve HV-251F016A's worm sub-assembly and worm gear with a 60:1 ratio gear set.
This will change the overall actuator ratio from 212.5:1 to 150:1. The result will increase the actuator torque rating from 950 ft-Ibs to 1250 ft-lbs. This modification will provide a range of acceptable torque switch settings for VOTES testing to meet Generic Letter 89-10.
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No. The modification has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. The new design stroke time is below the FSAR and Tech Spec limit. The decreased MOV design stroke time will improve system response. This modification will not affect circuit interlocks or signal logic. The ability to open or close the valve when required will not be affected by this modification.
II. No. The proposed modification: 1. does not adversely impact the ability of HV-251F016A to perform its intended design function, 2. does not increase the stroke time of the valve beyond that previously contained in the Design Basis, 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis, 4. involves the replacement of components outside the valve pressure boundary, 5. does not affect the ability to open or close the valve when required, 6.
does not affect any other equipment. It is concluded that the proposed modification does not create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety defined in the Bases of any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.6, 3/4.6.2 and 3/4.6.3 were reviewed. The above Tech Spec bases are dependent upon the integrity of the primary containment. The Containment Spray Cooling system is relied upon to preclude containment over pressurization and maintain primary containment integrity.
With the result of the proposed modification being decreased valve stroke time and increased available thrust, the above Tech Spec bases and FSAR analysis will be better supported by the implementation of this modification. Valve seat leakage and integrity are not affected. The MOV's decreased design stroke time improves system response. Therefore, this modification has no effect on the RHR System - Containment spray mode intended design function or the basis of the above Tech Specs.
~E@O '3-190 C E C 'CP 93-3018B, Unit2 Replace valve HV-251F016B's worm sub-assembly and worm gear with a 60:1 ratio gear set.
This will change the overall actuator ratio from 212.5:1 to 150:1. The result will increase the actuator torque rating from 950 ft-Ibs to 1250 ft-lbs. This modification will provide a range of acceptable torque switch settings for VOTES testing to meet Generic Letter 89-10.
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I. No. The modiTication has no impact on the dynamic qualification of the Motor Operated Valve (MOV) as discussed in Section 3.9.3.2b.2 of the FSAR since the actual accelerations at the valve during a dynamic event are less than the allowable accelerations. The new design stroke time is below the FSAR and Tech Spec limit. The decreased MOV design stroke time will improve system response. This modification will not affect circuit interlocks or signal logic. The ability to open or close the valve when required will not be affected by this modiTication.
II ~ No. The proposed modification: 1. does not adversely impact the ability of HV-251F016B to perform its intended design function, 2. does not increase the stroke time of the valve beyond that previously contained in the Design Basis, 3. does not decrease the MOV's allowable seismic acceleration below the actual value determined in the piping analysis, 4. involves the replacement of components outside the valve pressure boundary, 5. does not affect the ability to open or close the valve when required, 6.
does not affect any other equipment. It is concluded that the proposed modification does not, create the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III. No. The modification does not reduce the margin of safety defined in the Bases of any Technical Specification. Tech Spec bases 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.6, 3/4.6.2 and 3/4.6.3 were reviewed. The above Tech Spec bases are dependent upon the integrity of the primary containment. The Containment Spray Cooling system is relied upon to preclude containment over pressurization and maintain primary containment integrity.
With the result of the proposed modification being decreased valve stroke time and increased available thrust, the above Tech Spec bases and FSAR analysis will be better supported by the implementation of this modification. Valve seat leakage and integrity are not affected. The MOV's decreased design stroke time improves system response. Therefore, this modification has no effect on the RHR System - Containment spray mode intended design function or the basis of the above Tech Specs.
C DCP 93-5007, Unit 1, 2, Common C ON C GE'he primary Emergency Offsite Dose Projection model currently in use at SSES, called STREAM, is being replaced by a new model, the "Meteorological Information and Dose Assessment System" (MIDAS). The STREAM dose projection model software and associated RAMTEK terminals will be removed.
I. No. The proposed action does not affect safety related equipment and has no effect on probability of occurrence or consequences of an accident. The dose projection model is used as one of the response tools for assessing possible doses to members of the public. The new model will provide equivalent assessment capabilities for projecting doses to members of the public. It, therefore, does not affect the consequences of an accident.
II ~ No. The dose projection system has no connection to or effect on any plant system or component. It has no use during normal operations and, therefore, cannot cause a new type of accident.
III. No. The dose projection model has no effect on any basis for any Technical Specification.
S~~O'3-192 C 'CP 93-3075, Unit1 C OF C GE'he modification installs a non-safety related level probe on a dynamically designed and installed support bracket which is welded to the Spent Fuel Pool liner above the normal water level.
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No. The Spent Fuel Pool is dis'cussed in FSAR Section 9.1.2 and Electrical Class 1E requirements are contained in FSAR Section 8.1.6.1.q.5. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15. The non-safety related remote indication of Spent Fuel Pool level and temperature provides no interlocks with existing plant equipment and is independent of all installed circuits that do provide interlocks with existing plant equipment. This modification does not increase the probability of occurrence of a malfunction of equipment important to safety since: no system or component interlocks are provided; new component design features provide for seismic/dynamic installation for protection of the fuel pool, the 1E power source, and the safety related components located in 1C644; appropriate electrical isolation has been provided for protection of the 1E power source in the event of a fault in the non-1E portion of the loop. This modification does not increase the consequences of an accident since failure of any component installed by this DCP will in no way initiate or change the outcome of any accident.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since failure of the new components installed by this DCP would result only in loss of control room indication of Spent Fuel Pool level and temperature. This modification does not create the possibility of a malfunction of equipment of a different type since all potential failure modes have been considered, and appropriate design features have been implemented.
III. No. This modification is an enhancement to the plant to provide control room indication for monitoring Spent Fuel Pool level and temperature. The instruments provide a monitoring function only and make no change to normal spent fuel pool level or to the minimum level required (22 feet) by Technical Specification 3/4.9.9. There is no change made to any margin of safety by the components added in this DCP.
r.J g)~gg: 93-193 CR E C DCP 93-9041, Unit 1 This modification will remove the Kaowool from raceway that does not require wrapping. It will also provide the inspection of the installation method of the Kaowool to determine whether or not the installation was performed to the requirements. The modification will specify the inspection and documentation requirements. This modification also decreases the scope of the visual inspections of fire rated assemblies required by Tech. Spec. 3/4.7.7.
I. No. The removal of the Kaowool fire barrier material does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety. The fire barrier material was installed for compliance with Appendix R. Based on the method of compliance with Appendix R the raceway identified does not require wrapping to ensure safe shutdown in the event of an Appendix R fire.
II. No. The only physical impact that this modification has on any equipment or components is the removal of fire barrier material on electrical conduits, trays, and junction boxes.
This modification does not interfere with the logic, control or operation of any plant systems or components. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.
III. No. Technical Specification 3/4.7.7 requires the operability of fire barriers to ensure that fire damage is limited. This design feature minimizes the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The removal of the fire barriers on the raceway does not reduce the margin of safety required by 3/4.7.7.
SER59 93-194 C OS E ERE CE DCP93-9056, Unit2 This modification relocates relay 52YX20105 to eliminate interference with the cubicle compartment door.
I. No. The existing relay mounting configuration makes it incumbent upon the electrician, for his own personal safety, to apply extraordinary measures involving safeguard devices when working in the switchgear cubicle. The proposed action relocates the auxiliary relay and restores design basis personnel safety features. The relay relocation in 2A20105 has no adverse affect on any plant systems or components. No scheme logic changes occur. FSAR Chapters 6 and 15 were reviewed and the installation would have no affect upon, and will therefore not increase the probability of occurrence or the consequences of any accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. Relocating the 52YX20105 relay in switchgear cubicle 2A20105 will not result in any risk to public health and safety. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible.
The relocation of an auxiliary relay to enhance personnel safety and to maintain control circuit requirements for the Core Spray Pump 2A does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR.
III. No. Tech Spec 3/4.8 requires sufficient power be available to supply safety related equipment required for safe shutdown and the mitigation and control of accident conditions. The AC power supplies necessary to meet these requirements are listed in Sections 3/4.8.3.1 and 3/4.8.3.2. The implementation of DCP 93-9056 has no adverse impact on the operation of the 2A201 bus. Therefore, this modification does not reduce the margins of safety as defined in the basis for any Technical Specifications.
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S~E~) 93-1 95 C DCP 93-9040, Unit 2 DESC T 0 C This modification provides for increasing the cable length for the Unit 2 Intermediate Radiation Monitor (IRM) 2F detector assembly through installation of a plug jack connection in the undervessel cable tray and addition of new cabling from the plug jack to the IRM 2F detector.
The proposed action shall maintain the operability of the IRM System.
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No. No potential safety impact items are created by this installation. No degradation in signal quality will occur. The cabling is electrically equivalent and the new cable connectors are completely compatible with the existing installation, the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR Chapters 6 and 15, is not increased.
II. No. Since the proposed change does not adversely affect the operation of the IRM System, the on-site radiological consequences will not be impacted. Also, no new radiological consequences can occur. Therefore, there will be no increase in the consequences of an accident of malfunction of equipment important to safety as previously evaluated in the FSAR. No new accident scenarios or malfunctions of a different type can result because no adverse equipment or circuit effects are possible. Therefore, this modification does not create the possibility for an accident or malfunction of a different type than any evaluated in the FSAR. (Reference FSAR 7.6.1a.5.4.)
III. No. Tech. Spec. Basis 3/4.3.1 specifies the minimum operability and surveillance requirements involving nuclear instrumentation. Similarly, Tech. Spec. 3/4.3.6 and Tech. Spec. 3/4.4.1 have been reviewed. The proposed modification has no impact on the above mentioned Tech. Specs. The proposed modification does not adversely affect the logic, control, or operation of any safety related plant system or component.
Therefore, this modification does not reduce the margin of safety as defined in the basis for any Technical Specifications, and shall not adversely impact the Reactor Protection System response time.
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E C DCP 93-3046A,B,C, Unit 2
~D~~ - installs signal/power cables for the vibration probes from the outboard side of containment penetration 2W102B to the control structure penetrations and installs power feeder cable from lighting panel OLP2D and Uninterruptible Power Source (UPS) distribution panel 2Y688A to the Vibration Data Acquisition System location, but does not terminate cables. ~D~~ - performs all in-containment work including new transducer installation, cable/conduit routing between the inboard side of the containment penetration and the vibration transducers and termination on the inboard side of 2W102B. QQP~-
installs Vibration Data Acquisition System, completes all cable routing and all termination work outside containment.
I. No. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR. The installation of this equipment or its failure would not be an initiating cause for the events analyzed in FSAR Sections 6 and 15. The new vibration transducers do not intrude on the reactor coolant pressure boundary and are seismically installed in a manner which precludes increasing the failure probability of the pump. These actions have no effect that could increase the possibility of occurrence of any accident evaluated in the FSAR. This modification does not increase the probability of occurrences of a malfunction of equipment important to safety. This modification enhances diagnostic capabilities relating to pump dynamic behavior and can provide early indication of potential pump problems, but is not important to safety. The modification does not increase the consequences of an accident since the modification does not affect any safety functions of the system. The modification enhances pump monitoring capabilities and will in no way affect the offsite dose to the public. The modification does not increase the consequences of malfunction of equipment since this equipment's failure to produce valid data is no different than the present configuration without the additional vibration probes and Vibration Data Acquisition System.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since non-safety equipment is being added to monitor pump vibration and failure of this equipment will not result in an accident. FSAR Sections 15.2.2, 15.2.3 and 15.2.6 discuss generator load rejection, turbine trip and loss of a.c. power events.
The new circuits are independent of and separated from the Recirculation Pump Trip logic. Thus the proposed modification does not create the possibility of an accident of a different type than previously evaluated in the FSAR. The modification does not create the possibility of a malfunction of a different type. Failure of the proposed equipment does not involve a change in system operation or add a different failure mode which would affect this evaluation.
III. No. Recirculation Pump vibration instrumentation is not discussed in the Technical Specification.
This modification has no affect on system operation. The proposed action does not affect the margin of safety defined in the basis for any Technical Specifications.
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S~E~O 93-197 C DCP 93-3076, Unit 2 DEC I The modification installs a non-safety related level probe on a dynamically designed and installed support bracket which is welded to the Spent Fuel Pool liner above the normal water level.
RVJRlYlhBY No. The Spent Fuel Pool is discussed in FSAR Section 9.1.2 and Electrical Class 1E requirements are contained in FSAR Section 8.1.6.1.q.5. This modification does not increase the probability of occurrence of any accident evaluated in the FSAR, including FSAR Sections 6 and 15. The non-safety related remote indication of Spent Fuel Pool level and temperature provides no interlocks with existing plant equipment and is independent of all installed circuits that do provide interlocks with existing plant equipment. This modification does not increase the probability of occurrence of a malfunction of equipment important to safety since: no system or component interlocks are provided; new component design features provide for seismic/dynamic installation for protection of the fuel pool, the 1E power source, and the safety related components located in 2C644; appropriate electrical isolation has been provided for protection of the 1E power source in the event of a fault in the non-1E portion of the loop. This modification does not increase the consequences of an accident since failure of any component installed by this DCP will in no way initiate or change the outcome of any accident.
II. No. This modification does not create the possibility of an accident of a different type than any evaluated previously in the FSAR since failure of the new components installed by this DCP would result only in loss of control room indication of Spent Fuel Pool level and temperature. This modification does not create the possibility of a malfunction of equipment of a different type since all potential failure modes have been considered, and appropriate design features have been implemented.
III No. This modification is an enhancement to the plant to provide control room indication for
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monitoring Spent Fuel Pool level and temperature. The instruments provide a monitoring function only and make no change to normal spent fuel pool level or to the minimum level required (22 feet) by Technical Specification 3/4.9.9. There is no change made to any margin of safety by the components added in this DCP.
~ 4 C DCP 93-9068A8 B, Unit 1 0 C This modification will install a new actuator stem with flats at the coupling ends and a new anti-rotation device used in conjunction with the existing stem connector for each RHR Motor Operated Valve (MOV) HV-151F017A8 B.
SIJJgg~R No. This modification will ensure the valves function properly to support Low Pressure Coolant Injection and Shutdown Cooling modes of RHR. This modification will eliminate the possibility of misalignment of the limit switches and subsequent motor'ailure that existed with the original stem coupling design. FSAR Sections 5.4.7, 5.4.7.1.1.2, 6.2.1.1.3.3.1.3, 6.2.1.1.3.3.1.6, 6.3 and 15 have been evaluated. The actions taken in this modification will not increase the probability of occurrence of an accident previously evaluated in the FSAR. This modification will enhance the overall reliability of the MOVs, while allowing them to fulfill their design functions. There will be no increase in the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the FSAR. This modification does not adversely affect or change the operation or function of MOVs HV-151F017A8B or the RHR System.
The proposed actions do not involve a precursor of or contributor to any evaluated accidents involving offsite dose. Therefore, there will be no increase in the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
II. No. The MOVs will function as originally intended following the modification. This modification will eliminate the possibility of motor operator failure as a result of misaligned limit switches following repeated closures of the valves over the life of the plant. The proposed actions taken by this modification do not create a possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
III. No. The use of the new anti-rotation device will have no adverse affects on the RHR system performance. The opening and closing time of valves HV-151F017A8B will remain unchanged by this modification. The proposed actions will not reduce any margin of safety as defined in the basis for any technical specification
S~ER O.: 93-199 C OSS C DCP 93-9067, Unit 2 C GE The pipe crack in the Reactor Building Closed Cooling Water (RBCCW) return line is to be permanently repaired using a weld overlay technique to provide sufficient wall thickness across the crack location to restore the pressure integrity and mechanical strength of the joint. A new pipe support is to be added to the piping to minimize the loads that will be imposed upon the repaired joint. The weld repair and pipe support addition are for the purposes of restoring structural integrity and have no impact upon RBCCW system functions.
S A I. No. The planned modifications do not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR. FSAR Section 15.3 analyzes the effect of recirculation pump trips/seizures/and shaft breaks. The proposed actions restore the required structural integrity of the piping system and redistribute the applied loads to minimize the chance for another failure in the same region. This action does not change the original design functions of the affected systems.
II. No. The planned modifications do not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR. The proposed actions restore the structural integrity of the piping system and redistribute the applied loads away from the relatively sensitive equipment nozzle region.
III. No. This modification restores the structural adequacy the RBCCW piping system so that it will be able to perform its original design function. The changes made will have no impact upon the function or operation of any plant system governed by the Technical Specifications. Therefore, the actions proposed by this modification do not reduce the margin of safety as defined in the basis for any Technical Specification.
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QgQJg)93-200 C SS C NL-93-029, Unit Common C C GE ECO 93-6050 is being issued to replace the existing River Water Makeup (RWMU) discharge valves, XV-02804A,B,C,D with high performance butterfly valves using a hard seat design. The new valves offer more reliable sealing and reduced need for maintenance.
I. No. RWMU is not adequate to account for the spray pond losses due to evaporation.
RWMU normally feeds both cooling towers providing make-up to replace the water that is evaporated and blown down from the towers. With a volume of 7 million gallons in each tower, several hours exist from the point that make-up is lost until there is not enough water to maintain condenser back pressure even if 100% power is maintained.
When make-up to the tower is lost, power is reduced to 60%. Loss of RWMU feed to the clarifier results in the loss make-up to the clarified water tank. Loss of the clarifier or clarified water has no safety impact. No condition will exist that increases the probability of occurrence or consequence of an accident or malfunction of equipment important to safety as a result of the work described.
II. No. The FSAR does not describe the RWMU system, it only references it as a water source. The RWMU system provides no safety function and affects no equipment important to safety. The proposed action will not create a possibility for an accident or malfunction of a different type than evaluated in the FSAR.
III No.
~ No Technical Specification addresses the RWMU system. No Technical Specification uses the RWMU system as part of its basis. Therefore the margin of safety has not been diminished.
S I
C DCP 93-3074, Unit 2 0 C Modify the ¹4, 6 and 8 main turbine bearings and bearing housings as follows: 1. drill and tap two 1/2" diameter by 1.469" deep holes, split equally apart from the vertical centerline of the bearing by 2.25", at twelve o'lock; 2. drill and tap the lower bearing housing to accept a sealing gland with 3/8" NPT threads. The hole will be located approximately 8" outboard of the bearing face axially, and approximately 8" below the horizontal centerline.
S I~ No. FSAR Section 3.5.1.3 addresses the probability of a catastrophic wheel failure for a "shrunk on wheel" rotor. These sections conclude that the probability of unacceptable damage to safety related components due to turbine missiles is 2.61E-10. The turbine missile hazard is not considered a design basis event. This analysis bounds the modification. No new permanent potential missiles are created by the modification.
FSAR Section 10.2 provides the design bases of the Turbine/Generator and discusses the protective functions of the steam valves, overspeed protection, and turbine disk integrity. The analysis performed in this section is not impacted by the modification.
The modification does not impact the analyses performed in FSAR Sections 15.2.2 and 15.2.3. The modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. Based on the bearing/shaft clearances and the size of these components there is no possibility of this malfunction causing detrimental shaft or bearing surface damage.
Since strain gages are attached by very low energy capacitor discharge welding, there will be negligible metallurgical impact, and no possibility of inducing stress risers in the shaft. The probability of the tack welds creating a stress riser in the bearing is negligible. There are no other potential failure modes introduced by the modification.
Therefore, the modification does not create a possibility for an accident or malfunction of a different type than previously evaluated in the FSAR.
III No. Tech Spec section 3/4.3.8 ensures
~ that the turbine overspeed protection system instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. The proposed modification will not reduce the margin between anticipated wheel failure and overspeed protection system instrumentation settings. The proposed modification does not affect the basis of Tech Spec section 3/4.3.9. The main turbine bypass system is not affected by the modification. The modification has no effect on the bases for the sections described above, nor are there any other Tech Specs impacted. Therefore, there is no reduction in the margin of safety as defined in the basis for any Tech Spec.
- Et
'P-293-027, Unit 2 The purpose of this test is to determine the precise location of the Unit 2 turbine-generator rotor system torsional resonant frequencies. This test is recommended to determine if a modification is needed to install additional mass to the main turbine rotor. The turbine-generator rotor torsional natural frequencies will be measured from approximately 800 to 1900 rpm.
No. FSAR section 3.5.1.3 describes the safety risk from the Turbine-Generator as a Turbine overspeed condition which causes a rotor wheel burst and subsequent missile generation. Neither device will be impaired by this test and thus the probability of an overspeed condition and subsequent missile generation is not increased by this test.
Section 10.21 of the FSAR provides the design bases of the Turbine-Generator, including a list of Electrohydraulic Control (EHC) system functions and alternator excitation ratings. The only EHC function to be affected by this test is the speed control. The temporary GE speed control circuit will be able to control Turbine speed at the slower ramp rate of = 7 rpm/min. This is conservative from a speed control standpoint. FSAR sections 15.2.2 and 15.2.3 address generator load rejection and turbine trip. Reactor power level is expected to be 10 to 15% during the performance of this test, so these protective functions are not a concern. The performance of this test does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the FSAR.
II. No. The performance of this torsional test is within the bounds of Turbine Generator system design. Due to the fact that Turbine-Generator overspeed protective functions will not be defeated, this test does not create a possibility for an accident or malfunction of a different type than evaluated in the FSAR.
III. No. Tech Spec section 3/4.3.8 ensures that the turbine overspeed protection system instrumentation and the turbine speed control valves are operable and will protect the turbine from excessive overspeed. Performance of the torsional test will not reduce the margin between anticipated wheel failure and overspeed protection system instrument settings, nor will it affect the overspeed protection system. Tech Spec section 3/4.3.9 is provided to initiate action of the feedwater system/main turbine trip system in the event of failure of feedwater controller under maximum demand. The torsional test does not affect the basis of this Tech Spec. Tech Spec section 3/4.7.8 states that the required operability of the main turbine bypass system is consistent with the assumptions of the feedwater controller failure analysis in the cycle specific transient analysis. The main turbine bypass system is not affected by this test. Therefore, there is not reduction in the margin of safety as defined in the basis for any Tech Spec.
1~
~ENO '3-203 C OS E E CE DCP93-3004Z, Unit2 DE C C G The proposed action is to add an automatic transfer logic to the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel C 4.16 KV Bus. The proposed action involves the addition of a new MDR relay across the normal DC control power to the trip circuitry of each breaker feeding a affiliated load supplied from the Unit 2 Channel C 4.16 KV Bus.
I~ No. Chapters 6 and 15 of the FSAR were reviewed. The addition of an automatic transfer logic into the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel C 4.16 KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of a automatic transfer logic to the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel C 4.16 KV Bus does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel C control power in the Unit 2 Channel C 4.16 KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction.
The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action. The MDR relays, the indicating lights, the disconnect switches, the fuses and the breakers are all qualified devices.
II ~ No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in either trip circuitry, the MDR relay or the light added to the trip circuitry of a affiliated load supplied from the Unit 2 Channel C 4.16 KV Bus does not create a malfunction of a different type.
III No.
~ The operability of the Emergency Service Water System and the Residual Heat Removal Service Water System is governed by Technical Specification 3/4.7.1. The proposed action does not affect the operability requirements of the Service Water Systems. The proposed action does not affect the operability requirements of the Control Structure Heating and Ventilating System or the Control Structure Chilled Water subsystem. The operability of the Unit 2 Channel C 4.16 KV Bus is governed by Technical Specification 3/4.8.3. The proposed action ensures operability of the Unit 2 Channel C 4.16 KV Bus for a Unit 2 Channel A battery failure coincident with a LOCA/LOOP condition thereby maintaining the existing margin of safety for the Unit 2 Channel C 4.16 KV Bus.
S~~Q '3-204 C S C . DCP93-3005Z, Unit2 D C The proposed action is to add an automatic transfer logic to the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel D 4.16 KV Bus. The proposed action involves the addition of a new MDR relay across the normal DC control power to the trip circuitry of each breaker feeding a affiliated load supplied from the Unit 2 Channel D 4.16 KV Bus.
SMHElhEY'o.
Chapters 6 and 15 of the FSAR were reviewed. The addition of an automatic transfer logic into the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel D 4.16 KV Bus does not affect any of the postulated initiating events identified in Chapters 6 and 15 of the FSAR. Thus, the proposed action does not increase the probability of occurrence of an accident. The addition of a automatic transfer logic to the breaker trip circuitry of each affiliated load supplied from the Unit 2 Channel D 4.16 KV Bus does represent an increase in the probability of occurrence of a malfunction of equipment due to the additional relay and indicating light in the breaker control circuitry and the connection of an isolation system to the Channel D control power in the Unit 2 Channel D 4.16 KV switchgear. Since the failure rate of the equipment being controlled is so much larger than the failure rate of these components, it is engineering judgment that the increase in probability due to the additional relay, indicating light and isolation system is considered to be so small or insignificant that the change is within the error bounds associated with the original design calculations and does not constitute a significant increase in probability of the overall system malfunction.
The consequences of an accident is not changed by the proposed action. The consequences of a malfunction of equipment is not affected by the proposed action. The MDR relays, the indicating lights, the disconnect switches, the fuses and the breakers are all qualiTied devices.
II. No. Chapter 6 and 15 of the FSAR were reviewed. The review did not identify a postulated initiating event which would create the possibility for an accident of a different type. A random single failure in either trip circuitry, the MDR relay or the light added to the trip circuitry of a affiliated load supplied from the Unit 2 Channel D 4.16 KV Bus does not create a malfunction of a different type.
III. No. The operability of the Emergency Service Water System and the Residual Heat Removal Service Water System is governed by Technical Specification 3/4.7.1. The proposed action does not affect the operability requirements of the Service Water Systems. The proposed action does not affect the operability requirements of the Control Structure Heating and Ventilating System or the Control Structure Chilled Water subsystem. The operability of the Unit 2 Channel D 4.16 KV Bus is governed by Technical Specification 3/4.8.3. The proposed action ensures operability of the Unit 2 Channel D 4.16 KV Bus for a Unit 2 Channel B battery failure coincident with a LOCA/LOOP condition thereby maintaining the existing margin of safety for the Unit 2 Channel D 4.16 KV Bus.
~gg) '3<<205 C 0 CE DCP 93-9026Z, Unit 1 DESC 0 C This design modiTication proposes installation of bypass switches in Control Room Panels 1C645 and 1C644 to provide the following: Capability to bypass Main Steam Isolation Valve (MSIV) Level 1 signal from the Control Room and capability to bypass Containment Instrument Gas (CIG) (Level 1 and high diywell pressure signal) from the Control Room to prevent MSIV closure.
SUIMI~+
No. The proposed action does not affect any of the postulated initiating events identified in Chapters 6,7,15 of FSAR and NUREG-0776. The proposed action does not increase the probability of occurrence of an accident regardless of the operating conditions including the Anticipated Transient Without Scram (ATWS) event. This modification results in an insignificant increase in the likelihood of MSIV isolation failure on level 1 for accidents where this isolation is desired. The contacts on both MSIV bypass switches, or both the CIG bypass switches, must fail by changing state to the closed position and a true level 1 signal must be present. The probability of this combination of events is more than a thousands times less likely than occurrence of core damage and containment failure. This modification does not increase the probability of occurrence of a malfunction of equipment analyzed in the FSAR. Bypass switches are operated only during an ATWS plant event which is beyond the design basis. In fact the consequences of an ATWS would be lessened because primary containment would not be challenged. This modification does not increase the consequences of an accident previously evaluated in the FSAR. The bypass switches on channel basis for MSIV, Div. 1 and on division basis for CIG are being added to eliminate the need for manual installation of bypass jumpers during an ATWS plant event. A single failure within a single channel of MSIVs or CIG would not affect the safety functions of these systems.
II. No. With the exception of FSAR Section 15.8,7.1.2a.1.30.1 and Appendix 15A which provides a brief discussion of ATWS, a review of FSAR chapters 6, 7 and 15 and NUREG-0776 did not identify any postulated initiating event relative to the proposed action which would create the possibility for an accident of a different type. Consequently there is no possibility of creating an accident of a different type than any previously evaluated in the FSAR. The components being added by this modification are dynamically qualified and analyzed for installation in their proposed locations. The subject components are rated for use in their proposed Class 1E applications. Thus, the addition of the bypass function does not create the possibility for an accident or malfunction of a different type.
III. No. The proposed action does not affect any Technical Specification requirements. The Technical Specification load profiles, which are used as the bases for the designs of the 125 VDC system, will not be affected by the proposed modification. Thus, the proposed action does not reduce the margin of safety as defined in the basis for any Technical Specification.
QQMQ '3-206 C S C NL-93-030, Unit Common C E Drawing E118-85 Sht. 2 required breaker 1B227-072 to be installed with a magnetic trip unit that has a range of 750-1500 amps. Contrary to this requirement, a breaker with a range of 875-1750 amps was installed. Since the trip setpoint for this breaker is 1200 amps, either breaker can be calibrated to trip at the design overcurrent condition. Both the installed breaker and the breaker specified by the design drawings are Quality Class Q2 and are suitable for use in Load Center 1B227. Either breaker is technically acceptable.
I. No. The Reactor Building Recirculation System is described and discussed in FSAR Sections 6.5.3 and 9.4.2. A review of these sections indicates that the nonconforming condition described in NCR ¹ 93-0293 will not prevent the Reactor Building Recirculation Fan OV201B from performing as originally designed. Since there is no change in the ability of the Recirculation Fan OV201B to perform its safety function, the accidents described in Chapter 15 remain unaffected by this nonconforming condition.
Therefore, there is no increase in the probability of occurrence or consequences of an accident as previously specified in the FSAR.
II. No. FSAR Sections 6.5.3 and 9.4.2 were reviewed. Since the installed breaker was tested in service an can provide motive power to Recirculation Fan OV201B and the breaker was calibrated and tested to trip in the event of a 1200 amp overcurrent condition, there is no reason to believe that a different accident or equipment malfunction that any previously evaluated in the FSAR would occur.
III. No. The Reactor Building Recirculation Fan OV201B is not specifically covered by either Unit 1 or Unit 2 Technical Specifications. However, Technical Specification Interpretation Nos. 1-87-010 and 2-87-010 cover this equipment. This interpretation indicates that the Recirculation Fan OV201B is governed by Technical Specification 3.6.5.3. A review indicates that the basis for Technical Specification 3.6.5.3 is
¹ unaffected by the nonconforming condition described in NCR 93-0293 and therefore there is no reduction in the margin of safety.
C E E C SCP i=932121, Unit 1 ESC C G i )
A Type L3.10A overload heater is installed in 1D274022 for HV-155F066 while the design output documents require a Type L9.10A Overload heater for this valve. SCP E932121 formalizes the acceptability of the Type L3.10A overload heater for use in the HV-155F066 valve circuit.
SUUEIHhBY'.
No. Using a Type L3.10A 0/L does not affect any of the postulated initiating events identified in FSAR Chapters 6 and 15 since the 0/L trip function is bypassed during normal and emergency operation. Thus, the proposed action does not increase the probability of occurrence of an accident. Using Type L3.10A 0/L does not increase the probability of occurrence of a malfunction of equipment because this 0/L is bypassed during normal and emergency operation. Using Type L3.10A 0/L does not change the consequences of an accident since HV-155F066 does not change position for the accident and HPCI is still available to mitigate the effects of the accident. Using Type L3.10A 0/L does not change the consequences of a malfunction of equipment because HV-155F066 does not change positions for an accident and the 0/L is bypassed if this valve is manually operated from the control room after the accident.
II. No. Using a Type L3.10A 0/L for HV-155F066 does not create the possibility of an initiating event which creates the possibility of an accident of a different type than identified in FSAR Chapter 6 and 15 because the 0/L trip function is bypassed during normal and emergency operation. Using a Type L3.10A 0/L in HV-155F066 does not create the possibility of a malfunction of a different type because the 0/L trip function is bypassed during normal and emergency operation which is the same as the existing operation.
III. No. The operability of the HPCI is governed by Technical specification Section 3/4.5.1. The bases for operability of the HPCI System is to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. HV-155F066 does not change position for the accident and HPCI is still available to mitigate the effects of the accident. Therefore, the proposed action does not reduce the margin of safety as defined in the bases for any Technical Specification.
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