ML15028A381
| ML15028A381 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 01/28/2015 |
| From: | Navin P D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML15028A381 (24) | |
Text
Exelon Generation 10CFR 50.59, 10CFR 72.48, NEI 99-04 (SECY 00-0045)January 28, 2015U. S. Nuclear Regulatory Commission Attn.: Document Control DeskWashington, DC 20555-0001 Peach Bottom Atomic Power Station (PBAPS),
Units 1, 2 and 3 andPBAPS Independent Spent Fuel Storage Installation (ISFSI)Facility Operating License No. DPR-12Renewed Facility Operating License Nos. DPR-44 and DPR-56NRC Docket Nos. 50-171, 50-277, 50-278, and 72-29 (ISFSI)
Subject:
Biennial 1OCFR 50.59 and 10CFR 72.48 Reports for the Period 1/1/2013 through12/31/2014 and Annual Commitment Revision Report for the Period 1/1/14 through12/31/14Enclosed are the 2013-2014 Biennial 10CFR 50.59 and 10CFR 72.48 Reports and the 2014Annual Commitment Revision Report as required by 10CFR 50.59(d)(2),
10CFR 72.48, andSECY-00-0045 (NEI 99-04).There are no new regulatory commitments contained in this transmittal.
If you have any questions or require additional information, please contact D. J. Foss at 717-456-4311.Sincerely, Patrick D. NavinPlant ManagerPeach Bottom Atomic Power Stationcc: Senior Resident Inspector, USNRC, PBAPSCommonwealth of Pennsylvania Document Control Desk, USNRC, Washington DCCCN: 14-105Attachments_..
2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportExelon NuclearPeach Bottom Atomic Power StationDocket Nos. 50-17150-27750-27872-292013-2014 Biennial 10CFR 50.59 and 10CFR 72.48 Reports and the 2014 Commitment Revision ReportThese reports are issued pursuant to reporting requirements for Peach Bottom Atomic PowerStation Units 1, 2 and 3. These reports address tests and changes to the facility and procedures as they are described in the Peach Bottom Final Safety Analysis Report and Independent FuelStorage Safety Analysis Report for the TN-68 Spent Fuel Cask. These reports consist of thosetests and changes that were implemented between January 1, 2013 and December 31, 2014.Also, this report identifies commitments that were revised during 2014 and require reporting inaccordance with the guidelines of NEI 99-04, Managing Regulatory Commitments Made byPower Reactor Licensees to the NRC Staff endorsed by SECY-00-0045.
TABLE OF CONTENTS10CFR 50.59 Report 210CFR 72.48 Report 14Commitment Revision Report 161 2013-2014 Biennial IOCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportExelon NuclearPeach Bottom Atomic Power StationUnits 1, 2 and 3Docket Nos. 50-171, 50-277, and 50-278BIENNIAL 10CFR 50.59 REPORTJANUARY 1, 2013 THROUGH DECEMBER 31, 2014EVALUATION SUMMARIES Title: Structural Analysis Changes for Reactor Building Floor Loading forIndependent Spent Fuel Storage Installation (ISFSI) Cask (ECR 12-00326)
Units Affected:
2/3Year Implemented:
2013Brief
Description:
This activity was to revise calculations associated with postulated ISFSI cask tipping and slidingevents in the Reactor Building and to approve modifications to the plant required to comply withthe revised calculations.
The seismic analyses of the reactor buildings with a TN-68 cask atspecific locations were found to have non-conservative assumptions.
Applying conservative assumptions resulted in reactor building floor stresses exceeding code allowable limits, but stilloperable.
The calculation changes and modifications for this activity assure that code allowable stress limits will be met for postulated design conditions.
Summary of Evaluation:
The modification relocates the placement of the ISFSI cask on elevation 234' to a morestructurally supported area of the floor. The only physical change to the plant is leveling thefloor at the new cask locations on elevation 234'. Additionally, the engineering change requiresthe use of low friction plates to be placed under the casks at both 135' and 234' elevations.
These plates are required to prevent cask tip-up and return to floor loads during a seismicevent. The ECR also issues the calculations for ISFSI cask loads in the spent fuel pool.The activity also processes a calculation for sliding due to tornado winds at the 234' elevation.
The low friction plates used to prevent tip up during a seismic event could result in cask slidingfrom tornado wind loads. The sliding distance is small and has been found to be acceptable.
There are no physical modifications to the plant involved with the calculation.
The newermethodology used in the analyses has been accepted by the NRC and is referenced in theStandard Review Plan. Therefore, the change in methodology was acceptable.
The floor slabremains fully qualified for the seismic design loads, so there is no increased risk to nuclearsafety. It was determined that the activity did not result in a departure from a method of2 2013-2014 Biennial 10CFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportevaluation described in the UFSAR used in establishing the design bases or in the safetyanalyses.
Title: Replacement of the Automatic Voltage Regulators for Main Generators (ECRs 11-00381 and 12-00256)
Units Affected:
2/3Year Implemented:
Unit 3 -2013, Unit 2 -2014Brief
Description:
This activity involved the replacement of the Main Generator Automatic Voltage Regulator (AVR) for both Units 2 and 3. The AVR is designed to maximize electrical stability of the maingenerator during all design basis transients.
The new AVR is also designed to support the maingenerator throughout its entire operating range. Utilizing updated technology, the new AVR willincrease electrical stability during transients.
The new AVR does not have any negative impacton the design bases or any UFSAR design functions and performance parameters.
However,this change was considered as a change to a UFSAR described design function due to theexistence of potential common cause failures that were not previously evaluated.
Summary of Evaluation:
As part of the AVR implementation, a power system stability study was performed in order todetermine AVR settings and implement the Power System Stabilizer (PSS) function.
The studyincludes the descriptive information,
- analyses, and referenced documents for the turbine,generator and exciter as well as the offsite power system and the stability studies for theelectrical transmission grid at current and future EPU power levels. The study verified stability for the postulated loss of the nuclear unit, the largest operating unit on the grid, or the mostcritical transmission line. The study assured the probability of a loss of offsite power (LOOP) tothe plant as a result of implementation of the AVR and PSS is unaffected.
The 50.59 Evaluation determined that this change did not more than minimally increase thefrequency or consequences of a previously evaluated accident or create the possibility of a newaccident since no accident initiators are involved.
It does not increase the likelihood ofoccurrence of a previously evaluated malfunction of an SSC important to safety because theaffected equipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
3 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportTitle: Seismic Response Spectra Methodology Change to Support Main Steamand HPCI piping Modifications in Primary Containment (ECRs 11-00369and 12-00178)
Units Affected:
2/3Year Implemented:
Unit 3 -2013, Unit 2 -2014Brief
Description:
This activity involved an engineering analysis of Main Steam (MS) and High Pressure CoolantInjection (HPCI) System Turbine steam supply piping stresses and installation of associated supports inside Containment between the reactor pressure vessel (RPV) nozzles and the mainsteam anchors at the Containment for operation at the Extended Power Uprate (EPU)conditions.
As a result, new snubber type pipe supports were installed on various steam lines.The pipe stress analyses were performed using new seismic response spectra developed specifically for the MS lines inside the Containment.
This required an update to the UFSAR,Appendix C for both Units 2 and 3. The development of new spectra for Maximum CredibleEarthquake (MCE) is based on the same methodology that was used for the existing seismicresponse spectra for Design Earthquake (DE).Summary of Evaluation:
The steam flow rate through the main steam (MS) piping increases for operation at theExtended Power Uprate (EPU) condition resulting in increased fluid transient loads. Therefore, new stress analyses for MS and HPCI steam piping have been performed in the designanalyses.
This has resulted in a change to the loading on the supports of the MS and HPCITurbine steam supply piping. This activity addressed the impact on MS and HPCI steam supplypiping and associated supports inside Containment and supporting Reactor Building (Drywell) structure.
As a result of the new stress analyses, pipe supports inside containment are affected.
Theloading on the existing pipe supports has changed to the extent requiring modifications of someof the supports on MS and HPCI piping and installation of new snubbers on MS lines. Thepiping stress analyses used new seismic response spectra which have been refined to reducethe built-in conservatism.
The refinement includes using the higher damping values consistent with values listed in the UFSAR Table C.2.1 for generating response spectra for both seismicDesign Earthquake (DE) and Maximum Credible Earthquake (MCE) independently instead ofconservatively using a multiplier of 2.4 x DE Widened Response Spectra (WRS) for MCEresponse spectra.
It was determined that the activity did not result in a departure from amethod of evaluation described in the UFSAR used in establishing the design bases or in thesafety analyses.
4 2013-2014 Biennial IOCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportTitle: Removal of Radiographic Testing Requirements for High Pressure ServiceWater (HPSW) and Emergency Service Water (ESW) (ECRs 11-00379and 13-00426)
Units Affected:
2/3Year Implemented:
2013Brief
Description:
The two activities removed mandatory Radiographic Examination Testing (RT) requirements associated with the Class 3 High Pressure Service Water (HPSW) and Emergency ServiceWater (ESW) systems' piping butt welds. RT is not required by the original Construction Code(ANSI B31.1-1967) for this Class 3 piping. The proposed change will allow RT, MagneticParticle Testing (MT), or Dye Penetrant Testing (PT) to be performed for final acceptance ofHPSW piping butt welds.Summary of Evaluation:
Maintenance and modifications of the HPSW and ESW Systems (Class 3 piping) are governedby Section XI. Section Xl specifies hydrostatic or system leakage testing but refers to theoriginal construction requirements for most of the other requirements for repair/replacement activities.
The original Construction Code for the HPSW System piping is ANSI B31.1-1967.
RT of Class 3 piping is not required by the original Construction Code and is not the only NDEmethod allowed by the latest Edition and Addenda of the ASME Section III Code currently endorsed by the NRC. The latest NRC endorsed Section III NDE requirements for butt weldsgreater than two inches nominal pipe size is RT, MT, or PT.This activity is not a physical change to the HPSW system. This activity ensures finalinspection of welds in accordance with NRC endorsed
- methods, and does not change HPSWsystem operating modes or design functions.
Applying NRC endorsed NDE methods to finalexamination of piping butt welds assures the welds are of acceptable code quality so there is nochange in the probability that an inferior quality weld will be introduced into service.
In additionthe proposed change does not change the function of any other any other safety relatedsystem. This activity does not introduce the possibility of an accident because an adverseeffect on the HPSW system would not be an initiator of any accident and no new failure modesare being introduced.
The 50.59 Evaluation determined that this change did not increase thefrequency or consequences of a previously evaluated accident or create the possibility of a newaccident since no accident initiators are involved.
It does not increase the likelihood ofoccurrence of a previously evaluated malfunction of an SSC important to safety because theaffected equipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
5 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportTitle: Reactor Pressure Vessel and Drywell Strongback Qualification Tests (ECR13-00378)
Units Affected:
2/3Year Implemented:
2013Brief
Description:
This activity added detail to the UFSAR regarding exceptions to the requirements of ANSIN14.6-1978, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials",
for the reactor head strongback
/carousel and the drywell head strongback.
The ANSI standard includes a requirement that"materials for load bearing members shall be subjected to a drop weight test in accordance withASTM E 208 or a Charpy impact test in accordance with ASTM A 370..." This activity justified the omission of this testing for several components of the strongbacks.
It also justified the useof the drywell head strongback during refueling outage P3R19.Summary of Evaluation:
The activity justified continued use of the lifting devices without performance of the materialtesting.
With regard to the load testing deficiency, use of the drywell head strongback isjustified during refueling outage P3R19 only, with the requirement that the full 300% load testbe performed prior to use beyond P3R19. These exceptions to the requirements of ANSIN14.6 are related to initial testing requirements only. The devices maintain the design safetyfactors required by ANSI N14.6, and must continue to satisfy all pre-use NDE requirements.
The methods, procedures and steps for the use of these lifting devices are not affected by theallowed exceptions to the initial testing requirements.
The Peach Bottom commitment to performing these lifts in a single failure proof configuration isas stated in the UFSAR: "the criteria of NUREG-0612, Phase II are met, except for alternatives which may be approved on a case-by-case basis in accordance with station procedures."
Detailed consideration of the exceptions to the required testing provided in the Evaluation concluded that the omitted testing does not affect the safety of the lifting devices.
Thesedevices maintain their design safety factors and undergo NDE testing to ensure their continued integrity.
The designer of these devices has concurred that there is no concern for theircontinued ability to function as intended based on their low stress usage and lack ofsusceptibility to brittle failure.The 50.59 Evaluation determined that this change did not increase the frequency orconsequences of a previously evaluated accident or create the possibility of a new accidentsince no accident initiators are involved.
It does not increase the likelihood of occurrence of apreviously evaluated malfunction of an SSC important to safety because the affectedequipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety because6 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportthere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
Title: Condensate Storage Tank (CST) Standpipe Addition (ECR 12-00227)
Units Affected:
2Year Implemented:
2014Brief
Description:
This activity modified the CST by adding a standpipe in the tank. The standpipe will preventdraining of CST to the condenser hotwell in the event of spurious opening of the hotwellmakeup valves. Under Extended Power Uprate (EPU) conditions, the CST inventory dedicated to High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) suction iscredited for Station Blackout (SBO), Anticipated Transient without Scram (ATWS) andAppendix R events and therefore, this modification preserves the availability of these systems.This activity does not impact plant operations at nominal CST levels; however, a newaction of opening existing manually operated isolation valves is required to allow the CSTs tocontinue to perform their design function of providing a backup water supply to the CRD pumpswhen CST inventory is below the height of the installed standpipe.
Summary of Evaluation:
The installation of the standpipe in the CST does not adversely impact design bases or safetyanalyses as described in the UFSAR. This activity does not adversely impact plant operations, design bases or safety analyses as described in the UFSAR.The 50.59 Evaluation determined that this change did not increase the frequency orconsequences of a previously evaluated accident or create the possibility of a new accidentsince no accident initiators are involved.
It does not increase the likelihood of occurrence of apreviously evaluated malfunction of an SSC important to safety because the affectedequipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
7 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportTitle: Reactor Feedpump Turbine (RFPT) Replacement
-Electrical
/Instrumentation (ECR 13-00265)
Units Affected:
2Year Implemented:
2014Brief
Description:
In conjunction with the Reactor Feed Pump Turbine (RFPT) replacement, the mechanical overspeed device and trip mechanism was replaced with an electrical overspeed device and tripmodule. This change from functionally diverse to functionally equivalent overspeed protection fundamentally altered the means of performing this function and affects a design function of theTurbine Driven Feedwater Pump Control as described in UFSAR Section 7.10.4.Summary of Evaluation:
As a result of this activity, the primary overspeed function is performed by a new protection device that is functionally equivalent to the previous device. While both devices aremicroprocessor based, they are electrically diverse and not subject to a common mode failure.Using guidance provided in NUREG/CR-6303 (Method for Performing Diversity and Defense-in-Depth Analyses of Reactor Protection Systems),
the new overspeed protection device is designdiverse (different architecture),
equipment diverse (same manufacturer of fundamentally different designs),
signal diverse (same parameter sensed by different sensors) and softwarediverse (different program architecture) as compared to the previous system. As such, thesame defense-in-depth will be provided by the electrically diverse redundant overspeed devicesas with the existing functionally diverse overspeed devices.The proposed facility change will not alter the manner in which the RFPT, RFPT Speed Control,Feedwater
- Control, Lube Oil or 125 VDC systems are controlled or operated.
The samemonitoring and protective functions will be performed by the modified system as currently performed by the existing RFPT instrumentation and controls.
This facility change does notaffect any Nuclear Safety Related components.
With the same defense-in-depth, the facility change does not increase the likelihood of amalfunction of equipment important to safety or the frequency of accidents evaluated in theUFSAR. Although the new digital equipment has different modes of failure, the effect of thesefailures is the same and does not create the possibility of a different accident or the malfunction of equipment important to safety with a different result.No new system interfaces are created and no physical changes are made to a steam path orbarrier that could alter or affect the consequences of an accident.
The radiological consequences of the malfunctions and accidents currently evaluated are not affected and arebounding for this facility change.8 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportTitle: Allowance of Synthetic Roundslings for NUREG-0612 Heavy Load LiftsUnits Affected:
2/3Year Implemented:
2014Brief
Description:
The activity revised UFSAR Section 10.4.11.1.5 to allow use of Twin-Path Extra TPXCSynthetic Roundslings constructed with K-Spec fiber used in combination with engineered softeners and abrasion protection
- devices, in addition to the currently referenced ANSIStandard B30.9-1971 slings. The activity was limited to slings used for NUREG-0612 HeavyLoad Lifts. The reason for the change is to allow for use of an additional type/style of sling thathas been developed since the issuance of the 1971 Standard.
Specifically, the proposedactivity will allow for the use of a particular "synthetic roundsling" for single failure proof heavyload lifts.Summary of Evaluation:
This type of synthetic roundsling was developed after the issuance of the current UFSARapproved ANSI B30.9-1971 standard.
The synthetic roundsling is included in ASME B30.9-2010, "Slings".
Synthetic roundslings are fabricated from core yarns wound together withmultiple turns and enclosed in protective cover(s).
Synthetic roundslings offer similar capacities as the other type of slings, but with greater flexibility and lighter weight. As a result synthetic roundslings have become the preferred sling for rigging activities.
It is acceptable to allow use of Twin-Path Extra TPXC Synthetic Roundslings constructed withK-Spec fiber used in combination with engineered softeners and abrasion protection
- devices, inaddition to the currently referenced ANSI Standard B30.9-1971 slings. The proposed UFSARchange limits the subset of synthetic roundslings to be used for NUREG 0612 heavy load lifts to"Twin-Path Extra TPXC Synthetic Roundslings constructed with K-Spec fiber. This style ofsynthetic roundsling provides required rated load capacities, superior fiber on fiber abrasionresistance, tell-tail overload and damage inspection
- features, and when combined with"Engineered Softeners" cut resistance protection.
Based on the improved material properties, sling construction, and the improved ability toinspect the roundsling, the Twin-Path Extra TPXC Synthetic Roundslings constructed with K-Spec fiber, along with engineered softeners and abrasion protection devices meet the intent ofthe NUREG 0612 heavy load handling requirements.
The 50.59 Evaluation determined that this change did not increase the frequency orconsequences of a previously evaluated accident or create the possibility of a new accidentsince no accident initiators are involved.
It does not increase the likelihood of occurrence of apreviously evaluated malfunction of an SSC important to safety because the affectedequipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated 9
2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportin the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
Title: Surveillance Interval Change of 4 kV Undervoltage RelaysUnits Affected:
2/3Year Implemented:
2014Brief
Description:
The activity involved a change to the frequency of the performance of the 4kV Undervoltage Relays and LOCA LOOP Functional Tests from 24 months (1R) to 48 months (2R). Althoughthe Technical Specification (TS) Surveillance Requirements (SRs) are controlled in TS 5.5.14,Surveillance Frequency Control Program (SFCP), some of the testing affected also changedUFSAR requirements.
Technical Requirements Manual (TRM) Appendix A (incorporated intothe UFSAR be reference) requires testing of HGA and SV relays on an every refueling frequency that were formerly in the PBAPS Custom TS, but were relocated into the UFSAR aspart of the transition to Improved Technical Specifications (ITS). These relays control thetripping of loaded breakers, fast transfer permissives, dead bus start of the diesel generator and sequential loading of vital loads. The test frequency also changes the licensing basiscommitment to the test frequencies at a frequency contained in Regulatory Guide 1.9 Revision.
3 as identified in "Improved Technical Specification (ITS), 3.8.1".Summary of Evaluation:
The frequency change will not prevent any of the associated SSCs included within the test fromperforming their design function as described in the UFSAR. Peach Bottom is committed toRegulatory Guide 1.9, Revision 3, Selection, Design, Qualification and Testing of Emergency Diesel Generator Units used as Class 1E Onsite Electric Power Systems at Nuclear PowerPlants. The testing interval specified in this regulatory guide is once every refueling outage.The commitment to the Reg. Guide contents is documented in Technical Specification BasesTS B3.8.1.The evaluation concluded that the change to the testing frequency would not have a significant adverse impact on the reliability of 4kV Undervoltage relays and LOCA LOOP logic. Many ofthe components being tested by the subject tests are also subject to other tests on a morefrequent basis. Some of the components are normally operating or rotated/cycled in and out ofservice while the plant is operating.
The 50.59 Evaluation determined that this change did not increase the frequency orconsequences of a previously evaluated accident or create the possibility of a new accidentsince no accident initiators are involved.
It does not more than minimally increase the likelihood of occurrence of a previously evaluated malfunction of an SSC important to safety because the10 2013-2014 Biennial IOCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportaffected equipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
Title: Time Increase for Suppression Pool Cooling (SPC) Operation forExtended Power Uprate (ECR 10-00478)
Units Affected:
2/3Year Implemented:
2014Brief
Description:
This activity involved changing of the time requirement to implement suppression pool cooling(SPC) during a station blackout (SBO) event. The Extended Power Uprate (EPU) analysisincluded in the License Amendment Request (LAR) and NRC Safety Evaluation Report (SER)assumed that alternate AC power was available in one hour following the initiation of the SBOand suppression pool cooling (SPC) was also initiated at the same time. It was identified thatoperators would require an addition 30 minutes, following the availability of AC power, to initiateSPC.Summary of Evaluation:
The SBO event was revised to incorporate a change to the initiation time of RHR in SPC mode.The original analysis was evaluated for a 60 minute initiation time. The new initiation time isevaluated for 90 minutes.
This longer period of time before initiation, will increase the peaksuppression pool temperature and peak drywell pressure, thus reducing the NPSHA and NPSHmargin for the RHR pumps. The margin is reduced from 5.3 ft to 4.75 ft. However, sincepositive margin is available for the pumps to function adequately, the mitigation of the SBOevent is still acceptable.
Also, the minor increases in peak suppression pool temperature andpeak drywell pressure are acceptable since there is sufficient design margin for theseparameters.
The 50.59 Evaluation determined that this change did not increase the frequency orconsequences of a previously evaluated accident or create the possibility of a new accidentsince no accident initiators are involved.
It does not increase the likelihood of occurrence of apreviously evaluated malfunction of an SSC important to safety because the affectedequipment does not interfere with any previously evaluated.
It does not increase theconsequences of a previously evaluated malfunction of equipment important to safety becausethere are no consequences associated with the activity.
It does not create the possibility for amalfunction of an SSC important to safety with a different result than any previously evaluated 11 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportin the UFSAR because no new failure modes are introduced.
It does not result in a designbasis limit for a fission product barrier as described in the UFSAR being exceeded or alteredbecause no system parameters will change as a result of this activity.
Title: Control Room Habitability Program Changes due to Construction of aNearby Power PlantUnits Affected:
2/3Year Implemented:
2014Brief
Description:
The scope of this activity was to issue a new hazardous chemical analysis due to theconstruction of an off site power plant. The station is committed to Regulatory Guides 1.78 Rev.0 and 1.95 Rev. 0 which give several "levels" of requirements, depending on proximity tohazardous chemicals and station ventilation design. Since initial licensing of the facility, a co-generation power plant was constructed within 5 miles of the Main Control Room (MCR) HVACintake. This co-generation plant contains hazardous chemicals of sufficient quantity such thatcrediting the low probability of a hazardous chemical event occurring cannot be the only methodto ensure control room habitability.
This activity performs control room habitability evaluations in accordance with the Regulatory Guides to demonstrate that the function of the CRE toprotect occupants will be maintained.
This activity utilized the HABIT code in order to evaluatethe dispersion of the hazardous chemicals.
HABIT is an NRC approved code for use in thisapplication.
Since the previous method for evaluating chemicals was based on low probability, this was considered a change in methodology for PBAPS.Summary of Evaluation:
The 50.59 evaluation determined that this activity can be completed without a licenseamendment.
The activity performed control room habitability evaluations in accordance withRegulatory Guide 1.78 and Regulatory Guide 1.95 to demonstrate that the MCR / MCREVsystems will perform their design function during a hazardous chemical event. The currentmethodology for demonstrating that the MCR / MCREV systems will perform their designfunction is to credit the low probability of an event occurring.
The activity used a method forassessing a specific chemical's effect on control room habitability in the event of a release forthose chemicals which have a greater than negligible probability of occurrence.
The method ofperforming a detailed evaluation for chemical events is described in Regulatory Guides (RG)1.78 and 1.95. RG 1.78 and RG 1.95 contain NRC approved methodologies for performing detailed evaluations and assessing the impact on control room habitability.
Since the activityutilized an NRC approved methodology intended for the specific application, it was notconsidered a departure from a method of evaluation described in the UFSAR used inestablishing the design bases or in the safety analyses.
12 2013-2014 Biennial IOCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportThere were no 10CFR 50.59 Evaluation Reports performed
/ implemented for Unit 1during this reporting period.End of 1OCFR 50.59 Report13 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportExelon NuclearPeach Bottom Atomic Power StationIndependent Spent Fuel Storage Installation (ISFSI)Docket No. 72-29BIENNIAL 10CFR 72.48 REPORTJANUARY 1, 2013 THROUGH DECEMBER 31, 2014EVALUATION SUMMARIES Title: TN-68 Cask Lid Poison Plate Conductivity ChangeUnits Affected:
ISFSI TN-68 Casks -Certificate No. 1027, Amendment 1Year Implemented:
2014Brief
Description:
The cask vendor discovered that suppliers could not obtain the thermal conductivity for theType D poison plates for a new order of ISFSI casks for PBAPS. This change is addressed in avendor calculation which analyzed the effect on the thermal, structural and confinement designfunctions of the TN-68 cask. The poison plates provide the necessary criticality control andprovide the heat conduction path from the fuel assemblies to the cask cavity wall. Theproposed change does not affect the criticality function since the required minimum arealdensity of Boron-10 remains unchanged.
Reducing the thermal conductivity of the poison plateincreases the maximum temperature of basket components.
The increased temperatures affect the structural and confinement design functions.
The increased temperatures may alsoaffect the clearances between the cask components and the maximum internal pressure.
Summary of Evaluation:
There are no departures from methods of evaluation described in the TN-68 SAR to evaluatethermal, structural and confinement functions in the calculation.
The maximum fuel claddingtemperature for normal, off-normal, vacuum drying, and hypothetical fire accident caseconditions increased by at most 11 OF, but remain well below the allowable limits specified in theapplicable NRC guidance document.
The time at which the maximum fuel claddingtemperature reaches the allowable temperature limit of 1058°F for buried accident case withlower poison plate conductivity is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> shorter than the design basis model reported in theTN-68 SAR. The effect of the temperature increase on the internal fuel rod pressure and stressare discussed and found to be within allowable limits. The basket plate temperatures increaseby, at most, 11 OF for all considered conditions due to reduction of the poison plate conductivity.
The calculation demonstrates that the cask internal pressure remain well below the designpressure for all considered conditions, the structural evaluation of the basket components asdescribed in the UFSAR remain bounding and adequate clearances exist between various14 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportcomponents such that TN-68 cask with the proposed change continues performing its structural and confinement functions as designed.
Based on the above discussion, the thermal, structural and confinement functions of TN-68 cask affected by reducing the poison plate conductivity remain within the appropriate limits and continue to satisfy their respective design requirements.
End of 10CFR 72.48 Report15 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportExelon NuclearPeach Bottom Atomic Power StationUnits 1, 2 and 3Docket Nos. 50-171, 50-277, and 50-278COMMITMENT REVISION REPORTJANUARY 1, 2014 THROUGH DECEMBER 31, 2014CHANGE SUMMARIES Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 2/3/89, Response to NRC Inspection Report85-42T00306The Radiation Materials Shipping Coordinator will perform asupervisory sign-off to verify inclusions and proper placement ofrestraints.
Quality Control will do performance-based monitoring to verify conformance with requirements.
Summary of Justification:
Upgrades in the radwaste shipping program and procedures have resulted in substantial improvements in ensuring appropriate actions are performed involving shipments of radioactive material.
Upgraded procedure quality and site operating practices justify the allowance for nottracking this commitment any longer. This commitment is considered to be historical in nature.The corrective actions taken were effective and the station is in compliance with requirements.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 6/30/89, Progress Report for Implementing Control Room Enhancements T00315Revise T-200 Emergency Procedure Nomenclature Summary of Justification:
Upgrades in the procedure program have resulted in substantial improvements in procedure quality.
Standard nomenclature for procedures is in place. The corrective actions taken wereeffective and the station is in compliance with requirements.
This commitment is considered ashistorical and may be deleted from future commitment programmatic tracking since upgradedindustry
/ PBAPS standards have eliminated the need for detailed tracking of this commitment.
16 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportUpgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 1/31/91, Response to Limerick NRCInspection Report 90-80T00999Minor revision to an emergency operating procedure and bases toalert the operators to the effective level range of the suppression pool temperature monitoring instruments Summary of Justification:
Upgrades in the procedure program have resulted in substantial improvements in procedure quality.
The corrective actions taken were effective and the station is in compliance withrequirements.
This commitment is considered as historical and may be deleted from futurecommitment programmatic tracking since upgraded industry
/ PBAPS standards haveeliminated the need for detailed tracking of this commitment.
Upgraded procedure quality andsite operating practices justify the allowance for deleting this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
NRC Inspection Report 91-31 dated 2/7/92 (Cover Page)TO 1730Monthly testing of the Emergency Service Water (ESW) systemSummary of Justification:
Based upon satisfactory, consistent trending of ESW flow testing over the past 10+ years, thereis adequate assurance that decreasing the frequency of testing is not risk significant.
Currentmeasured ESW flow rates through the emergency diesel generators as well as the plant ESWring headers that support emergency equipment reveal that there is significant ESW flowmargin. Engineering has determined that it is acceptable to measure ESW flow rates on a 12week frequency.
This change in frequency will not cause any adverse impact to systemperformance.
Letter Source:Exelon Tracking No.:Nature of Commitment:
NRC Inspection Report 91-21 dated 8/26/91 (Attachment 2)TO 1749Addition of Net Positive Suction Head (NPSH) information into the17 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change ReportStandby Liquid Control system lesson plansSummary of Justification:
Upgrades in the training program have resulted in substantial improvements in ensuringappropriate training is administered to personnel, including the Standby Liquid Control systemand NPSH. This commitment is considered to be historical in nature. The corrective actionstaken were effective and the station is in compliance with requirements.
There is no longer aneed to track this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 9/9/91, Response to NRC Inspection Report91-16TO 1874Aspects of the Operating Experience Assessment Program will beenhanced to ensure that information capture and trainingconcerns are adequately addressed Summary of Justification:
Upgrades in the operating experience program have resulted in substantial improvements in theassessment of operating experience.
The corrective actions taken were effective and thestation is in compliance with expectations.
This commitment is considered as historical andmay be deleted from future commitment programmatic tracking since upgraded industry
/PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 12/3/76, Response to NRC Inspection Report76-35/25T03020Upgrade to surveillance test for analysis of release rates tofacilitate supervisory review for compliance to limitsSummary of Justification:
The accounting for particulates and iodine is appropriately included in procedures.
Theserequirements were subsequently moved from the Technical Specifications to the Offsite DoseCalculation Manual (ODCM). The corrective actions taken were effective and the station is incompliance with requirements.
This commitment is considered as historical and may bedeleted from future commitment programmatic tracking since upgraded industry
/ PBAPS18 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportstandards have eliminated the need for detailed tracking of this commitment.
Upgradedprocedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 1/6/78, Response to NRC Inspection Report77-37T03071Control of chemistry instrumentation background and sourcechecksSummary of Justification:
The control of chemistry instrumentation background and source checks have beensubstantially improved since this commitment was made. The corrective actions taken wereeffective and the station is in compliance with requirements.
This commitment is considered ashistorical and may be deleted from future commitment programmatic tracking since upgradedindustry
/ PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 1/6/78, Response to NRC Inspection Report77-37T03072Control of chemistry laboratory reagents from being used in theperformance of analyses of reactor coolantSummary of Justification:
The control of chemistry reagents have been substantially improved since this commitment wasmade. The corrective actions taken were effective and the station is in compliance withrequirements.
This commitment is considered as historical and may be deleted from futurecommitment programmatic tracking since upgraded industry
/ PBAPS standards haveeliminated the need for detailed tracking of this commitment.
Upgraded procedure quality andsite operating practices justify the allowance for deleting this commitment.
Letter Source:Letter to NRC dated 2/23/94, Response to NRC Inspection Report93-2519 2013-2014 Biennial IOCFR 50.59 and IOCFR 72.48 Reports and 2014 Annual Commitment Change ReportExelon Tracking No.:Nature of Commitment:
T03256Revise maintenance procedure to address the use of stroke timesand stroke lengths as acceptance criteria for motor-operated valve (MOV) actuator performance Summary of Justification:
Upgrades in the MOV program and procedures have resulted in substantial improvements.
Thecorrective actions taken were effective and the station is in compliance with expectations.
Thiscommitment is considered as historical and may be deleted from future commitment programmatic tracking since upgraded industry
/ PBAPS standards have eliminated the needfor detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 10/30/85, Response to NRC Inspection Report 85-31/28T03326Develop a procedure to perform a final comparison of the linerserial number on the proposed shipping papers with that recordedon the applicable fuel floor operating procedure Summary of Justification:
Upgrades in the radwaste shipping program have resulted in substantial improvements in thecontrol of radwaste since this commitment was made. The corrective actions taken wereeffective and the station is in compliance with requirements.
This commitment is considered ashistorical and may be deleted from future commitment programmatic tracking since upgradedindustry
/ PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 12/31/86, Response to NRC Inspection Report 86-21/22T03339Generate procedures to require company approval of radwastecomputer programs prior to their use20 2013-2014 Biennial 1OCFR 50.59 and IOCFR 72.48 Reports and 2014 Annual Commitment Change ReportSummary of Justification:
Upgrades in the radwaste shipping program have resulted in substantial improvements in thecontrol of radwaste since this commitment was made. The corrective actions taken wereeffective and the station is in compliance with requirements.
This commitment is considered ashistorical and may be deleted from future commitment programmatic tracking since upgradedindustry
/ PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 6/13/78, Response to NRC Inspection Report78-09-12T03342Revise primary containment vacuum breaker surveillances toeither perform a bypass test or evaluate differential pressure toensure that the vacuum breakers are closedSummary of Justification:
Upgrades in surveillances have resulted in substantial improvements in the conduct ofsurveillance tests. The corrective actions taken were effective and the station is in compliance with requirements.
This commitment is considered as historical and may be deleted from futurecommitment programmatic tracking since upgraded industry
/ PBAPS standards haveeliminated the need for detailed tracking of this commitment.
Upgraded procedure quality andsite operating practices justify the allowance for deleting this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 3/16/95, Response to NRC Inspection Report95-01T03909Improve control of shielding installations with engineering changedocumentation, 10 CFR 50.59 reviews and health physicsprocedures Summary of Justification:
Upgrades in the control of shielding have resulted in substantial improvements.
The corrective actions taken were effective and the station is in compliance with requirements.
Thiscommitment is considered as historical and may be deleted from future commitment programmatic tracking since upgraded industry
/ PBAPS standards have eliminated the needfor detailed tracking of this commitment.
Upgraded procedure quality and site operating 21 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportpractices justify the allowance for deleting this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 7/7/97, Response to NRC Inspection Report97-02T04024Establish training requirements for personnel performing scaffoldinstallation Summary of Justification:
Upgrades in the training program have resulted in substantial improvements in ensuringappropriate training is administered to personnel involved with scaffolds.
A standard trainingprogram is in place in accordance with improved industry
/ PBAPS standards.
Therefore, thiscommitment is considered to be historical in nature. The corrective actions taken were effective and the station is in compliance with requirements.
There is no longer a need to track thiscommitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 1/29/98, Response to NRC inspection Report97-07T04047Upgrade vendor training to emphasize the need for open dialoguewith supervision Summary of Justification:
Upgrades in the training program have resulted in substantial improvements in ensuringappropriate training is administered to vendor personnel including open communications.
Astandard training program is in place in accordance with improved industry
/ PBAPS standards.
Therefore, this commitment is considered to be historical in nature. The corrective actionstaken were effective and the station is in compliance with requirements.
There is no longer aneed to track this commitment.
Letter Source:Exelon Tracking No.:Nature of Commitment:
Letter to NRC dated 6/3/98, Response to NRC Inspection Report98-01T04143Revise operations manual to reflect expectation that when the22 2013-2014 Biennial 1OCFR 50.59 and 1OCFR 72.48 Reports and 2014 Annual Commitment Change Reportcontrol room supervisor moves to other areas of the control roomthat another senior licensed operator is in the controls areaSummary of Justification:
Upgrades in the operations administrative procedures have resulted in substantial improvements to operations conduct.
The corrective actions taken were effective and thestation is in compliance with requirements.
This commitment is considered as historical andmay be deleted from future commitment programmatic tracking since upgraded industry
/PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
Letter Source: Letter to NRC dated 7/10/98, Response to NRC Inspection Report98-05Exelon Tracking No.: T04422Nature of Commitment:
Improve performance monitoring of plant equipment by systemmanagersSummary of Justification:
Upgrades in the conduct of plant engineering have resulted in substantial improvements in theperformance monitoring of plant equipment.
The corrective actions taken were effective andthe station is in compliance with requirements.
This commitment is considered as historical andmay be deleted from future commitment programmatic tracking since upgraded industry
/PBAPS standards have eliminated the need for detailed tracking of this commitment.
Upgraded procedure quality and site operating practices justify the allowance for deleting thiscommitment.
End of Commitment Revision Report23