ML041900394
| ML041900394 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/30/2004 |
| From: | Rosalyn Jones Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML041900394 (31) | |
Text
Duke R. A.JONES rPowers Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax June 30, 2004 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
Subject:
Oconee Nuclear Site Docket Nos. 50-269, 50-270, 50-287 10 CFR 50.59 Annual Report Attached are descriptions of Oconee facility changes, tests, and experiments which were completed subject to the provisions of 10 CFR 50.59 between January 1, 2003, and December 31, 2003. This report is submitted pursuant to the requirement of 10 CFR 50.59 (d)(2).
If there are any questions, please contact, Graham Davenport at (864) 885-3044.
Very tr y yours, R. . ones, Site Vice President Oco egNuclear Site Attachment 5LV7 www. duke-energy. corn
U. S. Nuclear Regulatory Commission June 30, 2004 Page 2 xc: Mr. W. D. Travers Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 Mr. M. C. Shannon Senior NRC Resident Inspector Oconee Nuclear Site Mr. L. N. Olshan Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Attachment 1 Oconee Facility Changes - 2003 Description Page I. Nuclear Station Modifications 1- 6 II. Minor Modifications 7 - 13 Hi. Procedures 14 - 16 IV. Selected Licensee Commitments 17 - 19 V. Temporary Modifications 20 - 22 VI. Updated Final Safety Analysis Report 23 - 25 VII. Calculations 26 - 27 VIII. Miscellaneous 28
I. NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Steam Generators (OTSG)
The modification NSM ON-13086 Part 000 Phase 1 provides the BWC Report No.: 006K-LR-01, Input Document for Replacement replacement Once Through Steam Generators (ROTSGs.) The original Once Through Steam Generator Safety Evaluation and updated UFSAR steam generators installed in Oconee Unit I were Once Through Steam Chapter 6, 10, and 15 analyses describe ROTSG design and fabrication Generators (OTSGs) manufactured by Babcock & Wilcox. The and operation under normal, transient, and accident conditions. The replacement steam generators (ROTSGs) are manufactured by Babcock contents of these reports support the conclusion that the ROTSGs will
& Wilcox Canada. The ROTSG occupies the same physical envelope support normal and transient plant operation with no adverse impact as the OTSG. Differences between the OTSG and ROTSG are and that the existing licensing basis is maintained with the ROTSGs.
intended to improve the operation, maintainability, and accident Utilizing these reports and other supporting information, the 10 CFR performance. 50.59 Evaluation performed for Modification ON-13086 Part 000, Phase I concluded that no 10 CFR 50.59(cX2) criteria exist that would require a License Amendment Request. This modification did not involve any Unreviewed Safety Questions or safety concerns.
I
NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Steam Generators (OTSG)
The modification NSM ON-13086 Part 000, Phase 2 replaced Once BWC Report No.: 006K-LR-01, Input Document for Replacement Through Steam Generators (ROTSGs). The original steam generators Once Through Steam Generator Safety Evaluation and updated UFSAR installed in Oconee Unit 1 were Once Through Steam Generators Chapter 6, 10, and 15 analyses describe ROTSG design and fabrication (OTSGs) manufactured by Babcock & Wilcox. The replacement steam and operation under normal, transient, and accident conditions. There generators (ROTSGs) are manufactured by Babcock & Wilcox Canada. are no changes to the physical interfaces with the reactor coolant, main The ROTSG occupies the same physical envelope as the OTSG. steam, feedwater or other connected systems. Normal operating Differences between the OTSG and ROTSG are intended to improve conditions and plant transients have been reviewed or reanalyzed to the operation, maintainability, and accident performance. There are no include the ROTSG design and are documented in References 4 and 8, changes to the physical interfaces with the reactor coolant, main steam, respectively. The contents of these reports support the conclusion that feedwater or other connected systems. Normal operating conditions the ROTSGs will support normal and transient plant operation with no and plant transients have been reviewed or reanalyzed to include the adverse impact and that the existing licensing basis is maintained with ROTSG design and are documented in BWC Report 006K-LR-Oland the ROTSGs. Utilizing these reports and other supporting information, ROTSG Accident Analysis Summary report. the 10 CFR 50.59 Evaluation performed for Modification ON-13086 Part 000, Phase 2 concluded that no 10 CFR 50.59(cX2) criteria exist that would require a License Amendment Request. This modification did not involve any unreviewed safety questions or safety concerns.
(UFSAR 03-24) 2
NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Building Electrical Penetrations The modification NSM ON-13071 replaced four Reactor Building A paragraph in UFSAR Section 3.8.1.5.4 was revised to account for electrical penetration assemblies (EPAs) in ONS Unit 1.The four EPAs differences in the construction of the Conax penetrations and existing replaced by NSM ON-13071ALI were manufactured by Viking D. G. O'Brien/Vildng penetrations. The UFSAR revision does not Industries, Inc. and by the D. G. O'Brien Company. Viking is no change or affect the design basis for electrical penetrations. The longer in business and D. G. O'Brien no longer supplies EPAs under an revision only documents the differences in penetrations due to the use Appendix B program. These manufacturers have ceased production of of penetrations supplied by different manufacturers. There are no EPAs and all related replacement components changes to the Technical Specifications or Selected Licensee Commitments due to this modification. This modification did not involve any Unresolved Safety Questions or safety concerns.
(UFSAR changes 03-32 and 03-42) 3
NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Condensate and Feedwater System The modification NSM ON-23100 improved overall Condensate and As a result of this modification, UFSAR Section 7.6.1.2.2.1 was revised Feedwater System reliability by adding a "Manual/Fail Open" switch on to add the Condensate/Feedwater low suction pressure as a runback to the Hotwell Pump (HWP) Discharge Header Valve, 2C-10; revising the the CTPD subsystem. The CTPD subsystem initiates load limiting and Condensate Booster Pump/Main Feedwater Pump protection trip runback fiactions to restrict operation within prescribed limits. The circuits (low suction pressure circuits) and adding an automatic CTPD is restrained by a maximum load limiter, a minimum load Integrated Control System (ICS) runback upon low suction pressure of limiter, a rate limiter and a runback limiter (UFSAR Section the Condensate Booster Pumps or Main Feedwater Pumps. 7.6.1.2.2.1). This UFSAR Section currently contains a listing of those conditions under which the limiter acts to runback and/or limit the CTPD. The new Condensate/Feedwater low suction pressure runback is appropriately added to that list. There are no changes to the Technical Specifications or Selected Licensee Commitments due to this modification. This modification did not involve any Unresolved Safety Questions or safety concerns. (UFSAR change 03-32 and 03-42) 4
NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Building Electrical Penetrations This 10 CFR 50.59 evaluation determined that none of the 10 CFR The modification NSM ON-23071 replaced four Reactor Building 50.59 criteria were met for Part A or Part B of this NSM. UFSAR electrical penetration assemblies (EPAs) in ONS Unit 1. The four EPAs Section 3.8.1.5.4 was revised to include specific codes and code replaced by NSM ON-23071 Part A were manufactured by Viking sections for which the converted mechanical penetrations conform.
Industries, Inc. and by the D. G. O'Brien Company. These UFSAR Figure 3-20 was revised to include information on the new manufacturers have ceased production of EPAs and all related spare penetrations. UFSAR Sections 3.7.3.9, 6.2.2.2.7, 6.2.3.2, replacement components. Part B of the NSM will isolate the LPSW 9.2.2.2.3, and 9.4.6.2 were revised to reflect changes due to Part B of piping associated with the RBACs from the LPSW supply to the "B" this NSM. UFSAR Figures 6-3, 6-4, 6-9, 8-4, and 9-12 were revised.
Reactor Building Cooling Unit (RBCU). For UFSAR Figures 6-3, 6-4, and 9-12, the original figure were revised to indicate it is for Units I and 3 and a second page for this figure was added to reflect the Unit 2 design. UFSAR Table 6-7 and 7-3 were revised. SLC Section 16.9.12 was revised. No technical specification changes are required. This modification did not involve any Unresolved Safety Questions or safety concerns. (UFSAR change 03-32 and 03-42) 5
NUCLEAR STATION MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: SSF Pressurizer Heaters The modification NSM ON-53110 installed additional pressurizer The 10 CFR 50.59 evaluation determined that a Licensee Amendment heater control from the Standby Shutdown Facility (SSF) to assist in Request (LAR) is not required. UFSAR sections 5.4.6.2, 9.6.2, and making up heat lost by the pressurizer. This NSM also allowed these 9.6.3.2. and UFSAR Figure 9-40 was changed accordingly. Changes to additional heaters to be powered from the SSF. the bases to Technical Specification 3.4.9 and 3.10.1 are required. No SLC changes or other SLC bases changes are required. This modification did not involve any Unresolved Safety Questions or safety concerns. (UFSAR change 03-26, 03-27, 03-28) 6
II. MINOR MODIFICATIONS (ONOEs)
EVALUATION
SUMMARY
DESCRIPTION SYSTEM: Reactor Building Spray (BS)
The net effect of the change will be to slow down the actuators so that Minor modification OE-12466 modified the actuators on lBS-I and the valve stroke time will increase to approximately 16 seconds from its lBS-2 (BS header isolation valves) by installing new gears. The current value of approximately 12 seconds. No fission product barriers overall gear ratio was changed from 42.5:1 to 56.64:1. The stroke time are affected. Neither plant operating procedures nor test procedures are of the valves was changed from 12 seconds to a new stroke time of 16 affected. There is no impact upon safety analyses or dose calculations.
seconds These changes enhance the expected performance of the Reactor Building Spray system by increasing stem thrust on these valves, thus increasing their margin with respect to meeting the requirements of GL 89-10. This modification did not involve any Unresolved Safety Questions or safety concerns. No changes to UFSAR or Technical Specifications are required.
7
MINOR MODIFICATION (ONOEs)
EVALUATION
SUMMARY
DESCRIPTION SYSTEM: Emergency Feedwater (EFW)
The activity will not create any condition which will cause a LOCA, Minor Modification OE-17428 isolates Low Pressure Service Water LOOP, or any other accident analyzed in the FSAR. The accident (LPSW) and High Pressure Service Water (HPSW) to the Turbine mitigation functions of the LPSW and EFW Systems are not adversely Driven Emergency Feedwater Pump (TDEFWP) Jacket Cooler. OE- affected. The TDEFWP is not adversely affected if cooling water is 17428 also downgrades the applicable service water piping to the jacket isolated to the pump bearings. No new failure modes are postulated.
cooler from Duke Class F (i.e. QA-1 and seismic) to Duke Class G No malfunctions of SSCs important to safety will occur. Fission product (non-QA and non-seismic). barriers are unaffected. This modification does not involve any Unresolved Safety Questions or safety concerns. UFSAR Sections 9.2.2.1, 9.2.2.2.2, 10.4.7.2.2, SLC 16.9.8a and Technical Specification Basis B3.7.7 will need to be changed.
8
MINOR MODIFICATION (ONOEs)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Keowee Fire Suppression Minor Modification ONOE-17547 removed the ceramic frangible discs The removal of the frangible discs has no consequential effects on the in the Keowee Carbon Dioxide system delayed nozzles. These discs ability of the Keowee C02 system from performing its fire mitigation fracture when the delayed discharge piping becomes charged to allow function because the nozzles will not have to rely on C02 system C02 to discharge. The primary discharge nozzles do not have these pressure buildup to fracture before discharging C02 to a Keowee discs. Normally exposed to the outside environment these discs are generator fire. The removal of these ceramic discs from the delayed installed to prevent moisture, dirt, and insects from internally discharge nozzles will have no effect to the operational characteristics.
contaminating the discharge piping. In the Keowee application the The Keowee generator housings are sealed to ensure proper air cooling discharge nozzles are located inside the sealed generator housing; and to minimize gas leakage during a C02 actuation. Sealing of the therefore there is no requirement for their installation. The generators' generator housing also minimizes insect and other environmental casing sealing is maintained to ensure that the carbon dioxide which is conditions from potentially clogging the nozzles. No changes to discharged remains within the generator casing for a length of time UFSAR or Technical Specifications are required. This modification sufficient to suppress fires. does not involve any Unresolved Safety Questions or safety concerns.
9
MINOR MODIFICATION (ONOEs)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: High Pressure Service Water (HPSW)
The HPSW System is fully capable of performiing its functions. The Minor modification ONOE-17770 changed the setpoint for automatic proposed changes do not degrade the effectiveness of actions required stop of the A and B High Pressure Service Water (HPSW) pumps and to mitigate any design basis events. The proposed changes do not the Elevated Water Storage Tank (EWST) "full" indication. ONOE- introduce the possibility for a malfunction of an SSC with a different 17770 also changed the setpoints such that the pumps will start at result, because the proposed changes do not introduce a new 60,000 gallons and 50,000 gallons for the Base and Standby pumps, malfunction. There are no new malfunctions that need to be considered respectively.
based on the proposed changes. The proposed changes do not involve any changes to a method of evaluation described in the UFSAR. No changes are required to Technical Specifications. This modification does not involve any Unresolved Safety Questions or safety concerns. A change was made to UFSAR Section 9.5.1.2 and SLC 16.9.7. (UFSAR change 03-31) 10
MINOR MODIFICATIONS (ONOEs)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Coolant System (RCS)
Minor Modification ONOE-17823 issues ONEI-3 10-001 (Ref. 2) which The operating limits provided by this activity do not impact the safety contains the operating constraints in support of the "RCS Tave analyses or dose consequences and present no safety issues relative to the reduction coast down operation" of Unit 1 at EOC-21. UFSAR transient and accident analyses (Reference 8). The failure modes for the Reactor Coolant and Main Steam Systems and other related SSCs remain the same. This activity provides conservative limits for operation for the Main Steam, Reactor Coolant and Main Feedwater Systems that are within those system's design limits and that will not result in an accident initiator. No Technical Specification or other SAR document changes are required by this activity. This modification does not involve any Unresolved Safety Questions or safety concerns.
11
MINOR MODIFICATIONS (ONOEs)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: High Pressure Service Water (HPSW)
Minor Modification ONOE-18169 revised design documents to require This modification will not adversely affect how UFSAR described that HPSW-20 (A HP Line Supply to Aux Bldg) and HPSW-21 (B Hdr functions are performed. With HPSW-20 and lHPSW-21 normally closed, Isol Post Indicator Valve) be normally closed to reduce the the Auxiliary Building headers will be supplied water via the 4 inch consequences of a postulated Auxiliary Building flood. diameter line containing 3HPSW-453. There is no adverse effect on containment integrity and no new release paths are created. This modification does not require any design basis limits for a fission product barrier. The proposed changes do not involve any changes to a method of evaluation described in the UFSAR. This modification does not involve any Unresolved Safety Questions or safety concerns. The changes do not require a change to Technical Specifications. UFSAR Section 9.5.1.5.2 was changed. (UFSAR change 03-33) 12
MINOR MODIFICATION (ONOEs)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Equipment Database (EDB)
The changes are administrative or editorial in nature. The changes have Minor modification ONOE-10732 made administrative and editorial no effect on the radiological consequences of an accident The changes changes to several Oconee flow diagrams (OFDs) and the Equipment will not prevent or degrade the effectiveness of actions required to Database (EDB). The changes corrected minor drawing errors, clarified mitigate any accident. The changes do not require a change to Technical design parameter flags, and changed unit designators for several valves, to Specifications. This modification does not involve any Unresolved be consistent with guidance in other design documents. Safety Questions or safety concerns. UFSAR Figure 9-10 was revised due to unit designator changes for certain valves. (UFSAR change 03-09) 13
Z. PROCEDURE
S DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Standby Shutdown Facility (SSF)
The actions taken involve a procedure to safely shut the unit down in The procedure AP/O/A/1700/25 change provides specific guidance for a the case of a fire in the plant. It does not involve a method of confirmed fire in the Control Room, Cable Room, Equipment Room or evaluation. Therefore, this activity does not result in a departure from a Turbine Building. For a fire in these areas, a series of decision steps method of evaluation described in the UFSAR. The changes do not will direct the operator to electrically transfer the SSF 600VAC MCC constitute a change to the SAR. Furthermore these changes do not affect to be powered and controlled from the SSF. This action will result in Tech Specs. Based on this information, the procedural changes will not the removal of the Engineered Safeguards (ES) signal to the change the design function of the system described in the UFSAR. No containment isolation valves HP-3, HP-4, HP-20, RC-5 and RC-6. changes to UFSAR or Technical Specifications are required. This 50.59 does not involve any Unresolved Safety Questions or safety concerns.
14
PROCEDURES DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Low Pressure Service Water (LPSW)
AP/1,2,3/A/1700/011 was revised to provide guidance to align LPSW Providing LPSW to the CCSWPs during periods of low lake level to the CCSWPs once lake level reaches 791 feet to ensure LPSW is increases the reliability of the Chillers. With LPSW aligned to the aligned prior to de-entrainment concerns. CCSWPs, LPSW flow to safety related loads are not adversely affected during ES events. The CCW Pumps are operated consistent with normal operating procedures. No new failure modes are postulated. No adverse interactions were determined to exist. No malfunctions of SSCs important to safety will occur. The accident mitigation fumctions of the LPSW and WC Systems are not adversely affected. The CCW Punps are operated consistent with normal operating procedures.
Fission product barriers are unaffected. Evaluation methodologies, as described in the UFSAR, are unaffected. No changes to UFSAR or Technical Specifications are required. This 50.59 does not involve any Unresolved Safety Questions or safety concerns.
15
PROCEDURES EVALUATION
SUMMARY
DESCRIPTION SYSTEM: High Pressure Injection (HPI), Low Pressure Injection (LPI) and Reactor Building Spray (RBS)
The temperature limits for the HPI, LPI and RBS pump rooms will not TT/1&2/A/0170/017 (HPI and LPI / RBS Pump Room Heatup) will be be exceeded. Thus the HPI, LPI and RBS pumps will not be adversely performed to obtain room heatup rates and steady state temperatures for affected. The HPL LPI and RBS systems will be capable of performing the HPI Pump Room and LPI / RBS Pump Rooms. No Technical their accident mitigation functions during and following the test. The Specifications (TS) or Selected Licensee Commitments (SLC) action manipulation of AHUs and exhaust fans can not create an accident of a statements are entered during the test. different type. No new failure modes are introduced. Since temperature limits of the HPI, LPI and RBS pump rooms are maintained within limits, the possibility of a malfunction of an SSC is not created. The test procedure does not alter plant safety limits, set points, or design basis limits for a fission product barrier. This activity does not involve a change in a n evaluation methodology as described in the UFSAR. This activity does not require a Technical Specification or other licensing change. This 50.59 does not involve any Unresolved Safety Questions or safety concerns.
16
IV. SELECTED LICENSEE COMMITMENTS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: High Pressure Service Water (HPSW)
Selected Licensee Commitments (SLC) Manual, Section 16.9.8a was The changes have no effect on the radiological consequences of a revised to establish a minimum level of 70,000 gallons for the Elevated LOOP. By maintaining lake level within the limits plus measurement Water Storage Tank. Also, add a surveillance requirement to verify the error, the ECCW siphon header will be fully capable of performing its minimum level periodically. Changes to the Bases section are made to function of mitigating a LOOP. The proposed changes will not prevent explain the basis for the minimum leveL or degrade the effectiveness of actions required to mitigate a LOOP.
The proposed changes do not involve any changes to a method of evaluation described in the UFSAR. These changes do not require a change to Technical Specifications. No changes to the UFSAR are required. This SLC did not involve any Unresolved Safety Questions or safety concerns.
17
SELECTED LICENSEE COMMIMENTS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: High Pressure Service Water (HPSW)
This activity revised Selected Licensee Commitments (SLC) Manual, This change to SLC 16.9.8a does not change its original purpose nor its Section 16.9.8a to add a commitment for HPSW to be available to original requirements. It merely adds new scope to the commitment to provide backup cooling water to the TDEFWP bearing oil cooler. include the HPSW supply for the TDEFWP bearing oil coolers. The Minor changes to the Bases section are also made. proposed changes to SLC 16.9.8a are consistent with existing requirements in the UFSAR and the Technical Specifications.
The proposed change does not affect any margins of safety defined in the basis for any technical specification. The proposed change does not affect any safety limits or limiting safety system settings. No plant safety limits, setpoints, or design parameters are adversely affected.
There is no impact to the nuclear fuel, cladding, Reactor Coolant System (RCS), or containment integrity. This change does not require a change to Technical Specifications or other licensing documents and did not involve any Unresolved Safety Questions or safety concerns.
18
SELECTED LICENSEE COMMITMENTS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Low Pressure Service Water (LPSW)
No new components are being added to the facility. The safety-related This activity clarifies that both lLPSW-139 and 2LPSW-139 are functions of the LPSW Systems are maintained. The activity does not.
required to be operable if either unit is in the mode of applicability. adversely affect LPSW flow used for normal or accident operation.
Minor changes to the Bases section are also made. Therefore, the activity will not create any condition which will cause a LOCA, LOOP, or any other accident analyzed in the FSAR. The LPSW System remains single failure proof. The accident mitigation functions of the LPSW Systems are not adversely affected. No new failure modes are postulated. No new equipment is being added and no new adverse interactions were determined to exist No malfunctions of SSCs important to safety will occur. Fission product barriers are unaffected. Evaluation methodologies, as describe in the UFSAR, are unaffected. This change does not require a change to Technical Specifications or other licensing documents and did not involve any Unresolved Safety Questions or safety concerns.
19
V. TEMPORARY MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Building Cooling Units (RBCU)
Auxiliary Coolers are used only for normal RB cooling and are isolated Temporary Modification ONTM-2145 installed blind inserts in flanges for 1C3 RB auxiliary cooler coil to isolate a non- repairable leak in the on ES actuation. All required emergency RB cooling is accomplished coil. This is the fourth of the 16 auxiliary cooler coils to be isolated in by the RBCUs and RB Spray systems. This Aux cooler coil is NOT.
this manner. Auxiliary cooler coils are not required for RB cooling to required by licensing documents. Isolation of this coil is required for-meet design basis, however they are used to maintain building containment integrity to the active LPSW penetrations. Reactor temperatures below limits for normal operation per SLC 16.6.13. building temperatures during normal operation will be affected to some NSM-13031 is planned for IEOC 22 to replace all sixteen of these extent. (This is the fourth of the sixteen aux cooler coils to be isolated due to leaks.) Removal of this coil from service is not expected to coils.
cause the upper limits for RB average temperature (to maintain LPI system operability) to be reached during the peak temperature conditions for the upcoming operating cycle. However, if that operating limit is reached, actions to address the resulting conditions are bounded by requirements of SLC 16.6.3. This impact of this activity is bounded by conditions previously evaluated in the Oconee operating license. This temporary modification does not involve any Unresolved Safety Questions or safety concerns. No changes to UFSAR or Technical Specifications are required.
20
TEMPORARY MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Coolant (RCS)
Temporary Modification ONTM-2163 was implemented due to failure of Use of pressurizer pressure compensated pressurizer level in lieu of Pressurizer Temperature RTD 3RC RD0043B. ONTM-2163 does two temperature compensated level is conservative. No new failure modes distinct things: 1)replaces Temperature Compensated Pressurizer Level are introduced. Technical Specification 3.4.12, LTOP System, Bases
- 3 with Pressure Compensated Pressurizer Level #3 for control board was revised because the bases specifically credits temperature indications and for ICS pressurizer level control; 2) revises OAC alarm compensated pressurizer level to meet the analytical results. A review fimctions to employ pressure compensated pressurizer level in lieu of of the bases shows that it is not necessary to depend solely on temperature compensated pressurizer level. temperature compensated level, and that use of pressure compensated pressurizer level is conservative. SLC 16.7.11(1) requires surveillance of pressurizer temperature, but does not require any action if the indication fails. This temporary modification does not involve any Unresolved Safety Questions or safety concerns. No changes to UFSAR or Technical Specifications are required.
21
TEMPORARY MODIFICATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Building Cooling Units (RBCU)
Temporary Modification ONTM-2146 installed blind inserts in flanges Auxiliary Coolers are used only for normal RB cooling and are isolated for IB RB auxiliary cooler coil to isolate a non- repairable leak in the on ES actuation. All required emergency RB cooling is accomplished coil. This is the fifth of the 16 auxiliary cooler coils to be isolated in by the RBCUs and RB Spray systems. This Aux cooler coil is NOT this manner. Auxiliary cooler coils are not required for RB cooling to required by licensing documents. Isolation of this coil is required for meet design basis, however they are used to maintain building containment integrity to the active LPSW penetrations. Reactor temperatures below limits for normal operation per SLC 16.6.13. building temperatures during normal operation will be affected to some NSM-13031 is planned for IEOC 22 to replace all sixteen of these extent (This is the fourth of the sixteen aux cooler coils to be isolated coils. due to leaks.) Removal of this coil from service is not expected to cause the upper limits for RB average temperature (to maintain LPI system operability) to be reached during the peak temperature conditions for the upcoming operating cycle. This impact of this activity is bounded by conditions previously evaluated in the Oconee operating license. This temporary modification does not involve any Unresolved Safety Questions or safety concerns. No changes to UFSAR or Technical Specifications are required.
22
VI. UFSAR CHANGES (Pkg. 03-08)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Component Cooling (CC)
There has been an industry trend of changing system corrosion This 50.59 Evaluation is being performed to assess using sodium inhibitors from chromate-based to molybdate-based. This evaluation molybdate in the Component Cooling System as a corrosion inhibitor consisted of comparing the two corrosion inhibitors to ensure their instead of potassium chromate. Because the UFSAR specifically effects on components in the CC System were essentially equivalent.
mentions "chromate-phosphate" water treatment in the Component As part of this evaluation, a detailed materials evaluation was-Cooling System, this activity will require a change to UFSAR Section performed and a detailed industry practice survey was conducted by 18.32. GO Chemistry (Reference #3). This materials evaluation encompassed all materials of construction of the CC System piping and wetted components. No Technical Specification changes are required and did not involve any Unresolved Safety Questions or safety concerns. The UFSAR Section 18.312 was updated accordingly.
23
UFSAR CHANGE (Pkg. 03-33)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: High Pressure Service Water (HPSW)
Design documents are revised to require that HPSW-20 (A HP Line There has been an industry trend of changing system corrosion Supply to Aux Bldg) and HPSW-21 (B Hdr Isol Post Indicator Valve) be inhibitors from chromate-based to molybdate-based. This evaluation normally closed to reduce the consequences of a postulated Auxiliary consisted of comparing the two corrosion inhibitors to ensure their Building flood. effects on components in the CC System were essentially equivalent The change did not require a change to Technical Specifications and did not involve an Unresolved Safety Questions or safety concerns.
UFSAR Section 9.5.1.5.2 was revised accordingly.
24
UFSAR CHANGE (Pkg. 03- 12 )
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Steam Generator (OTSG)
The offsite doses reported in UFSAR Chapter 15 resulting from The activity involves revising offsite dose analyses for the SGTR and postulated Steam Generator Tube Rupture and Main Steam Line Break MSLB accidents using an NRC accepted computer code, LOCADOSE.
accidents were recalculated using the LOCADOSE comnputer code, which While LOCADOSE was already used in portions of the most recent has been previously approved for use in this application by the NRC. calculations of SGTR and MSLB dose, the current analysis approach applies the LOCADOSE code modeling over the full course of the accident The thermal hydraulic input used in the re-baselined dose analyses is identical to the input currently reported in the UFSAR, including the total amount of steam released to the environment. The fission product source term is also identical to that currently reported in the UFSAR. This 50.59 did not involve any Unresolved Safety Questions or safety concerns. The UFSAR Chapter 6 was revised accordingly with the new re-baseline dose results, and excessive detail from the Chapter 15 description.
25
VII. CALCULATIONS DESCRI PlON EVALUATION
SUMMARY
SYSTEM: Reactor Coolant (RCS)
The O1C22 Reload Design Safety Analysis Review (REDSAR),
A 10 CFR 50.59 evaluation has been performed for the Oconee Nuclear Station Unit 1, Cycle 22 (O1C22) core reload and is attached to performed in accordance with Nuclear Engineering Division workplace calculation OSC-8471. The impact of any other plant changes made procedure NE-102, "Workplace Procedure for Nuclear Fuel Management", serves as the overall justification for operation of the concurrent with the refueling outage is not addressed in the O1C22 10 CFR 50.59 evaluation. Oconee Unit 1 Cycle 22 core reload. The SAPP and MA sections of the REDSAR checklist documented evaluations of the OIC22 physics parameters. The reload safety evaluation documented in OSC-8471 confirm the updated final safety analysis report (UFSAR) Chapter 15 accident analyses remain bounding with respect to the O1C22 safety analysis reactor physics parameters. The safety analysis reactor physics parameters method is described in topical report DPC-NE-3005-PA.
This 50.59 did not involve any Unresolved Safety Questions or safety concerns. No changes to UFSAR or Technical Specifications are required.
26
CALCULATIONS DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Coolant (RCS)
Computer code SIMULATE-3 version 4.02 (Duke designation The two new code versions were determined to generate essentially the sim3dk06) is being replaced with SIMULATE-3 version 6.04.03 (Duke same results as the older code versions currently in use. Therefore, designation sim3dkl0) for all reload design and safety analysis prior NRC approval is not required to begin using the new code calculations pertaining to the Oconee Nuclear Station. The new code versions. The activity will not create any condition which will cause a version offers an improved steam table model, improved iteration LOCA, LOOP, or any other accident analyzed in the FSAR. No new convergence criteria, and much more input error checking than the failure modes are postulated. No new equipment is being added and no older version. Computer code TABLES-3, whose sole finction is to new adverse interactions were determined to exist. No malfimctions of generate a nuclear cross-section library for SIMULATE-3, is also being SSCs important to safety will occur. Fission product barriers are upgraded from version 4 to version 5. unaffected. Evaluation methodologies, as describe in the UFSAR, are unaffected. No changes to UFSAR or Technical Specifications are required and did not involve any Unresolved Safety Questions or safety concerns.
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VIm. MISCELLANEOUS PROBLEM INVESTIGATION PROCESS (PIP)
DESCRIPTION EVALUATION
SUMMARY
SYSTEM: Reactor Building Spray (RBS)
Low Pressure Injection (LPI)
Due to test results which indicate that the LPI/RBS pumps may not be Use of WC in lieu of non-essential LPSW for maintaining the operable when subjected to the anticipated post-accident environment environment in the LPI/RBS pump rooms does not create a different in the LPIIRBS pump rooms, a temporary compensatory action will use type of accident or a malfunction of an SSC important to safety with a Non-QA WC rather than the non-essential LPSW header as a cooling different result than any previously evaluated in the UFSAR. WC and medium for the recirculating fan-coil AHUs in the LPI/RBS pump LPSW are both cooling water systems and both are currently required rooms. by ONS Technical Specifications. No significantly different malfunction has been identified as a result of this compensatory action.
This change of cooling medium does not impact the design basis limit for a fission product barrier or involve a method of evaluation described in the UFSAR. No changes to UFSAR or Technical Specifications are required and did not involve any Unresolved Safety Questions or safety concerns.
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