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#REDIRECT [[1CAN081403, License Amendment Request Update the Reactor Coolant System Pressure and Temperature and the Low Temperature Overpressure Protection System Limits]]
| number = ML14241A240
| issue date = 08/27/2014
| title = Arkansas Nuclear One, Unit 1 - License Amendment Request Update the Reactor Coolant System Pressure and Temperature and the Low Temperature Overpressure Protection System Limits
| author name = Browning J G
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000313
| license number = DPR-051
| contact person =
| case reference number = 1CAN081403
| package number = ML14241A319
| document type = Letter
| page count = 53
}}
 
=Text=
{{#Wiki_filter:EntergyEntergy Operations, Inc.1448 S.R. 333Russellville, AR 72802Tel 479-858-3110 Jeremy G. BrowningSite Vice President Arkansas Nuclear One1 CAN081403 August 27, 2014U.S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, DC 20555
 
==SUBJECT:==
REFERENCES License Amendment RequestUpdate the Reactor Coolant System Pressure and Temperature and theLow Temperature Overpressure Protection System LimitsArkansas Nuclear One, Unit 1Docket No. 50-313License No. DPR-511 Entergy Letter to NRC, "Request for Exemption from Certain10 CFR 50.61 and 10 CFR 50, Appendix G Requirements,"
datedMarch 20, 2014 (1 CAN031403)
(ML14083A640)
: 2. NRC Letter to Exelon Nuclear, "Three Mile Island Nuclear Station, Unit 1-Exemption from Certain Requirements of 10 CFR Part 50, Appendix Gand 10 CFR 50.61, For Initial RTNDT Values for Linde 80 Welds(TAC No. MF0425),"
dated December 13, 2013 (ML13324A086)
: 3. NRC Letter to Exelon Nuclear, "Three Mile Island Nuclear Station, Unit 1-Issuance of Amendment RE: Revision to the Pressure andTemperature Limit Curves and the Low Temperature Overpressure Protection Limits (TAC No. MF0424),"
dated December 13, 2013(ML13325A023)
: 4. NRC Letter to Duke Energy Carolinas, LLC, "Oconee Nuclear Station,Units 1, 2, and 3; Issuance of Amendments Regarding Pressure
-Temperature Limits (TAC NOS. MF0763, MF0764, and MF0765),"
datedFebruary 27, 2014 (ML14041A093)
 
==Dear Sir or Madam:==
In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR)Section 50.90, Entergy Operations, Inc. (Entergy) is submitting a request for an amendment tothe Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) to revise the ReactorCoolant System Pressure and Temperature (P/T) Limits (TS 3.4.3); Pressurizer (TS 3.4.9);Pressurizer Safety Valves (TS 3.4.10);
and Low Temperature Overpressure Protection (LTOP)System (TS 3.4.11).
The current limits are applicable to 31 Effective Full Power Years(EFPYs).
The proposed limits are applicable to the end of the current period of extendedoperation (54 EFPY).
1 CAN081403 Page 2 of 3The P/T limits for the ANO-1 reactor pressure vessel were developed in accordance with therequirements of 10 CFR 50, Appendix G, using the analytical methods and flaw acceptance criteria of American Society of Mechanical Engineers (ASME) Code Section XI, Appendix G,and NRC approved AREVA Topical Report BAW-10046A, Revision 2.The projected fluence values at 54 EFPY are based on the NRC approved methodology presented in BAW-2241 P-A, Revision 2.The initial reference temperature for nil-ductility transition (RTNDT) values of the reactor vesselbeltline welds (Linde 80 welds) were determined using methods provided in Topical ReportBAW-2308, Revisions 1-A and Revision 2-A, rather than the methodology described withinTopical Report BAW-10046A, Revision
: 2. The methodology in BAW-10046A, Revision 2, wasused to evaluate the other beltline components (non-Linde 80 materials).
Entergy requested an exemption from the requirements of 10 CFR 50.61 to allow use of the alternate initial RTNDTvalues provided in BAW-2308, Revisions 1-A and 2-A (Reference 1). The subsequent analyses assumed this exemption request was approved.
The exemption request is currently being reviewed by the NRC.Attachment 1 provides a description and assessment of the proposed TS changes.Attachment 2 provides markup pages of existing TS and TS Bases to show the proposedchanges.
Attachment 3 provides revised (clean) TS pages. Attachment 4 provides a copy ofAREVA Topical Report ANP-3300, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY," June 2014. This report provides the technical basis for theproposed changes.
The values associated with the Pressurized Thermal Shock assessment are provided in Attachment 5.The current 31 EFPY P/T limits are estimated to be reached in April 2015. Entergy requestsapproval of the proposed license amendment by March 1, 2015, effective immediately with theamendment being implemented within 30 days of approval.
In accordance with 10 CFR 50.91(a)(1),
"Notice for public comment,"
the analysis regarding the issue of no significant hazards consideration (NSHC) using the standards in 10 CFR 50.92is being provided to the Commission in accordance with the distribution requirements in10 CFR 50.4.This amendment and the separate exemption request are similar to those approved for ThreeMile Island Nuclear Station, Unit 1 (References 2 and 3) and Oconee Nuclear Station,Units 1, 2, and 3 (Reference 4). Three Mile Island and the three units at Oconee all haveBabcock & Wilcox reactor vessels with Linde 80 welds similar to the ANO-1 reactor vessel.In accordance with 10 CFR 50.91 (b)(1), a copy of this application and the reasoned analysisabout NSHC is being provided to the designated Arkansas state official.
If you have any questions or require additional information, please contact Stephenie Pyle at479-858-4704.
1 CAN081403 Page 3 of 3I declare under penalty of perjury that the foregoing is true and correct.Executed on August 27, 2014.Sincerely, JG /rwcAttachments:
: 1. Description and Assessment of the Proposed Changes2. Proposed Technical Specification and Bases Changes (mark-up)
: 3. Revised (clean) Technical Specification Pages4. ANP-3300, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at54 EFPY" June 20145. Pressurized Thermal Shock Assessment cc: Mr. Marc L. DapasRegional Administrator U. S. Nuclear Regulatory Commission, Region IV1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear OneP.O. Box 310London, AR 72847U. S. Nuclear Regulatory Commission Attn: Ms. Andrea E. GeorgeMS O-8B1One White Flint North11555 Rockville PikeRockville, MD 20852Mr. Bernard R. BevillArkansas Department of HealthRadiation Control Section4815 West Markham StreetSlot #30Little Rock, AR 72205 Attachment I to1 CAN081403 Description and Assessment of the Proposed Changes Attachment 1 to1CAN081403 Page 1 of 13DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGES1.0 DESCRIPTION In accordance with 10 CFR 50.90, Entergy Operations, Inc., (Entergy) requests an amendment to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit 1 (ANO-1).The purpose of this License Amendment Request is to revise the pressure
/ temperature (PIT)limits in ANO-1 Technical Specification (TSs). The proposed amendment will revise the reactorcoolant system heatup, cooldown, and inservice leak hydrostatic test limitations for the ReactorCoolant System (RCS) to a maximum of 54 Effective Full Power Years (EFPY) in accordance with 10 CFR 50, Appendix G, "Fracture Toughness Requirements."
The 54 EFPY time period will bound the operation of ANO-1 until the end of the current periodof extended operation (i.e., 60 calendar years). Entergy has assumed that the unit wouldoperate with an average capacity factor of 90% over this time period.Further, the proposed amendment also revises other ANO-1 TSs requirements to reflect therevised PIT limits of the reactor vessel. These changes rely on NRC approved methodologies for determining allowable P/T limits.The proposed change includes the following TS revisions:
" TS Section 3.4.3 ("RCS Pressure and Temperature (P/T) Limits")
is being revised toincorporate updated figures for PIT curves and reactor coolant pump restrictions.
Thesefigures have been recalculated to account for 54 EFPYs of plant operation.
* TS 3.4.9,"Pressurizer";
TS 3.4.10, "Pressurizer Safety Valves";
and TS 3.4.11, "LowTemperature Overpressure Protection (LTOP) System" are being revised to account for54 EFPYs of operation.
The changes include the electromatic relief valve lift setpointbeing changed from 460 pounds per square inch -gauge (psig) to 563.8 psig. Theenable temperature for this valve is being revised from 262 OF to 248 OF. Thesechanges are as a result of the revised LTOP analyses and are consistent with the newP/T limits. Instrument uncertainty has not been included in these values.TS Bases changes have been provided for information only.2.0 TECHNICAL ANALYSISTo address plant operation through the period of extended operation (54 EFPY), neutronfluence projections were updated; reactor vessel embrittlement analyses performed, andupdated P/T and LTOP limits were developed.
The P/T limits for the ANO-1 reactor vessel weredeveloped in accordance with the requirements of 10 CFR 50, Appendix G, utilizing theanalytical methods and flaw acceptance criteria of ASME Code Section XI, Appendix G andTopical Report BAW-10046A.
Attachment 1 to1CAN081403 Page 2 of 13Beltline Region Determination Of particular interest in this analysis is the reactor vessel beltline, which is defined in 10 CFR 50,Appendix G, as the region of the reactor vessel that "directly surrounds the effective height ofthe active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limitingmaterial with regard to radiation damage."
The beltline region experiences increased embrittlement over the operating period of the reactor vessel as a result of accumulated fastneutron radiation from the core.10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements,"
provides the requirements to monitor changes in the fracture toughness properties of ferriticmaterials in the reactor vessel beltline resulting from the exposure to neutron irradiation and thethermal environment.
Appendix H to 10 CFR 50 states that no material surveillance program isrequired for reactor vessels for which it can be conservatively demonstrated by analytical
: methods, that the peak neutron fluence at the end of operating period will not exceedI E+17 neutrons/square centimeter (n/cm2) with energy greater than one million electron volts(E > 1 MeV). Appendix G to 10 CFR states, "To demonstrate compliance with the fracturetoughness requirements of Section IV of this appendix, ferritic materials must be tested inaccordance with the ASME Code and, for the beltline materials, the test requirements ofAppendix H of this part." Therefore, the fracture toughness requirements of 10 CFR 50,Appendix G, for the reactor vessel beltline are applicable to the reactor vessel materials withprojected neutron fluence values greater than 1 x 1017 n/cm2 (E > 1 MeV) at the end of theoperating period.During operation, the physical region of the reactor vessel with fluence that exceeds this levelcan expand as a result of several factors, including power uprates, increased operating periodsdue to license renewal, and modified fuel design. The result is that changes in fracturetoughness properties resulting from neutron embrittlement may occur in materials where theeffects of radiation damage may not have been considered previously when developing the P/Tlimits for the vessel. In particular, this may be true for reactor vessel nozzle materials when thenozzles are positioned immediately above or below the active core height.Fluence Determination Based on the considerations above, the fluence analysis performed for the latest cavitydosimetry exchange was expanded to determine if previously unevaluated regions of the reactorvessel had crossed the fluence threshold.
The inside wetted surface neutron fluence valueswere determined following the method from BAW-2241 P-A, Revision
: 2. BAW-2241 P-A hasbeen reviewed and accepted by the NRC, and is in compliance with NRC Regulatory Guide(RG) 1.190,"Calculational and Dosimetry Methods for Determining Pressure Vessel NeutronFluence,"
March 2001. All the 1/4 vessel wall thickness (T) and 3/4T fluence values weregenerated using the inside wetted surface fluence values and the methodology from RG 1.99,Revision
: 2. The fluence values are provided in Tables 3-1 and 3-2 of the AREVA report inAttachment 4 of this submittal.
How ANO-1 meets the NRC requirements and conditions for the use of methodologies such asBAW-2241 is discussed in a later portion of this attachment.
Attachment 1 toI CAN081403 Page 3 of 13P/T Limits Determination The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring thesafety of the primary system in light water-cooled reactors.
A method for guarding against brittlefracture in reactor pressure vessels is described in Appendix G to the ASME Boiler andPressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power PlantComponents."
This method utilizes fracture mechanics concepts and the reference temperature for nil-ductility transition, RTNDT. The RTNDT is defined as the greater of the drop weight nil-ductility transition temperature (per American Society for Testing and Materials (ASTM) E208)or the temperature at which the material exhibits 50 foot-pounds absorbed energy and 35 milslateral expansion minus 60 'F. The RTNDT of a given material is used to index that material to areference stress intensity factor curve (K,,). The K1, curve appears in Appendix G of ASMECode Section XI. When a given material is indexed to the K1, curve and applied thermal stressintensity factors and unit pressure stress intensity factors determined, then the allowable pressures can be obtained for this material as a function of temperature.
Operating P/T limitscan then be determined for a given heatup or cooldown temperature
-time histories.
The RTNDT of the reactor vessel materials must be adjusted to account for the effects ofirradiation.
Neutron embrittlement and the resultant changes in mechanical properties of agiven pressure vessel steel are monitored by a surveillance program.
The increase in theCharpy V-notch temperature is added to the unirradiated RTNDT to adjust it for neutronembrittlement.
This adjusted RTNDT (ART) is used to index the material to the K1, curve, whichin turn, is used to set new operating limits. These new limits take into account the effects ofirradiation on the vessel materials.
The ART is defined as the sum of the initial RTNDT, the mean value of the adjustment inreference temperature caused by irradiation (ARTNDT),
and a margin (o) term. The ARTNDT is aproduct of a chemistry factor (CF) and a fluence factor (FF). The CF is dependent upon theamount of copper and nickel in the material and may be determined from tables in RG 1.99,Revision 2, or from surveillance data. The FF is dependent upon the neutron fluence at themaximum postulated flaw depth. The margin term is dependent upon whether the initial RTNOTis a plant-specific or a generic value and whether the CF was determined using the tables inRG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNIT, the copper and nickel contents, the neutron fluence, and thecalculational procedures.
RG 1.99, Revision 2, describes the methodology to be used incalculating the margin term.For the Linde 80 welds, alternate initial RTNDT values were used per the NRC-approved TopicalReport BAW-2308, Revision 1-A and Revision 2-A. In order to utilize these alternate initialRTNDT values, an exemption request in accordance with 10 CFR 50.12 has been previously submitted (Reference 1). Using Linde 80 weld metal initial RTNDTS from BAW-2308 requires aminimum CF of 167.0 and a margin term of 28.0 'F.During the development of the new limits, AREVA informed Entergy that the generic RTNOT usedin reactor vessel integrity calculations is non-conservative.
The generic initial RTNDT and its standard deviation are determined in BAW-1 0046, Revision 2A,for a population of SA-508 Class 2 forgings ordered subsequent to 1971. These values areused to determine the ART for other SA-508 Class 2 forgings when sufficient material test datais not available to determine a heat specific initial RTNOT, in accordance with RG 1.99, Attachment 1 to1CAN081403 Page 4 of 13Revision
: 2. The forgings in Babcock & Wilcox (B&W) designed reactor pressure vessels aregenerally ordered prior to 1971. When RTNDTS from forgings manufactured prior to 1971 areincluded in the SA-508 Class 2 population, the newly calculated generic mean and standarddeviation increase calculated ARTs.AREVA now believes that the currently used dataset is not the most representative of the vesselforgings at the operating plants. AREVA further finds that the most representative datasets arethose grouped by the manufacturer that performed the forging process.
Using the moreapplicable
: dataset, the resultant ART can be higher, thus indicating that the current genericvalue may be less conservative (and potentially non-conservative).
AREVA has determined the appropriate new RTNDT and its uncertainty for ASTM SA-508Class 2 forgings based upon these findings, and has confirmed that the ART values will notincrease by greater than 4.5 OF.A review of the ART calculations that support the development of the current P/T curvesindicates that the limiting material is not affected at ANO-1. Only the non-limiting materials areaffected.
Therefore, the resultant P/T curves are unaffected by this finding.The ANO-1 RV contains both axially and circumferentially oriented welds. Therefore, the P/Tlimits are based on the postulation of both axial and circumferential flaws in the most limitingaxial and circumferential welds and the postulation of an axial flaw in the most limiting forgingmaterial of the reactor vessel.The limits are generated for normal operation heatup, normal operation
: cooldown, inservice leakand hydrostatic (ISLH) test conditions, and reactor core operations.
These limits are expressed in the form of curves of allowable pressure versus temperature.
The uncorrected P/T limitswere determined for 54 EFPY. Pressure correction factors were determined between thepressure sensor locations and the various regions of the reactor vessel.Attachment 4 provides a summary of the technical basis leading to the development of the newP/T limits. Instrument uncertainty was not included in the limits listed in this attachment.
Thesewill be applied in the appropriate operating procedures.
Low Temperature Overpressurization Protection LimitsLow Temperature Overpressurization Protection (LTOP) limits were based on the ASME Code,Section XI, Article G-2215. This article requires that the LTOP system ensures that themaximum pressure from the limiting P/T curve is not exceeded when the 1/4T temperature isless than the ART+ 50 OF. During a cooldown, the coolant temperature is always less than (orequal to) the 1/4T temperature; therefore, it is conservative to use the coolant temperature asthe LTOP enable set-point.
: However, during a heatup, the 1/4T temperature is always less thanthe corresponding coolant temperature.
To support the development of the LTOP system limits,the temperature differences between the reactor coolant in the downcomer region and the 1/4Twall locations are determined for the maximum heatup rate transient.
The current LTOP enabling temperature and electromatic relief valve (ERV) maximum liftsetpoint are 262 OF and 460 psig, respectively.
The proposed values are 248 OF and563.8 psig, based on the criteria specified in Appendix G of the ASME Code, Section XI. Theselimits do not include instrument uncertainties.
These will be applied in the operating procedures.
Attachment 1 to1 CAN081403 Page 5 of 13Pressurized Thermal ShockA Pressurized Thermal Shock (PTS) assessment for the ANO-1 reactor vessel beltline materials with fluences greater than 1 E+17 n/cm2 was performed in accordance with 10 CFR 50.61. ThePTS screening criterion is 270 OF for plates, forgings, and axial weld materials and 300 OF forcircumferential weld materials.
The controlling material are the Lower Shell axial welds, WF-18, with a predicted RTPTS value of191.1 OF. The remaining results are provided in Attachment 5 of this submittal.
UDoer Shelf Enerav and Eauivalent Marains AnalysisNeutron fluence is part of the basis for Upper Shelf Energy (USE or CXUSE) and Equivalent Margins Analysis (EMA). The current analysis supported the license renewal update whichdemonstrated compliance with 10 CFR 50, Appendix G, IV.A.1. Reference 8 is the SE for theANO-1 renewed license.The current analysis remains bounding for the projected end of life fluence, except for the lowershell plate and upper shell plate axial (longitudinal) welds, WF-18. The USE and EMAcalculations also remain bounding for close to 54 EFPY as the fluence calculated per BAW-2241 P-A methodology following Cycles 21, 22, and 23 is lower, or only marginally higher, thanthe conservative fluence used in BAW-2251A.
The copper content has also decreased.
Comparing the quarter thickness fluence with BAW-2251A and the most recent ART yields48 EFPY 54 EFPY Estimated Location Fluence Fluence 48 EFPYNumber (n/cm2) (nlcm2) USE1Nozzle belt forging (NBF) AYN-131 / 528360 7.11E+18 8.48E+172  98Upper Shell (US) Plate C5120-2 I C5120-2 8.06E+18 7.90E+18 65US Plate C5114-2 / C5114-2 8.06E+18 7.90E+18 82Lower Shell (LS) Plate C5120-1 / C5120-1 7.73E+18 7.78E+18 61LS Plate C5114-1 / C5114-1 7.73E+18 7.78E+18 74(71)NB to US Circumferential WF-182-1
/ 821T44 7.11E+18 7.14E+18 46(55)(Circ.) Weld (100%)US Longitudinal Weld WF-18 / 8T1762 5.82E+18 6.32E+18 49(Both 100%)US to LS Circ. Weld (100%) WF-1 12 / 406L44 7.73E+18 7.60E+18 41 (44)LS Longitudinal Weld WF-18 / 8T1762 5.71E+18 6.79E+18 49(Both 100%)12RG 1.99 Revision 2 Position 1 (RG 1.99, Revision 2 Position 2)Start at 8.4 inches portion of (lower) NBFAll reactor vessel locations not listed above have inside surface fluences below 1E+17 n/cm2.
Attachment 1 to1CAN081403 Page 6 of 13In Reference 8, the NRC made the following determination:
Although not discussed by the applicant, Appendix G to 10 CFR Part 50 requires that reactorvessel beltline materials have Charpy USE levels in the transverse direction for the basemetal and along the weld for the weld material according to the ASME Code, of no less than75 ft. lbs. (102 J) initially, and must maintain Charpy USE levels throughout the life of thevessel of no less than 50 ft. lbs. (68 J). However, Charpy USE levels below these criteria maybe acceptable if it is demonstrated in a manner approved by the Director, Office of NuclearReactor Regulation, that the lower values of Charpy upper-shelf energy will provide marginsof safety against fracture equivalent to those required by Appendix G of Section XI of theASME Code.The 48 EFPY CG USE values determined for the ANO-1 reactor beltline materials are given inBAW-2251A, Table 4-4. The T/4 fluence values in this table were calculated in accordance with the ratio of the clad-to base metal interface fluence to T/4 fluence values (i.e., neutronfluence lead factors at T/4) determined in the last reactor vessel surveillance program report.Table 4-4 shows that the CvUSE is maintained above 50 ft-lbs for all base materials (platesand forgings),
but weld materials nearly always fall below the 50 ft-lb limit at 48 EFPY.Appendix G of 10 CFR Part 50 provides for this situation by allowing lower values of CvUSE ifit is demonstrated that the lower CvUSE will provide margins of safety against fractureequivalent to those required by Appendix G to Section XI of the ASME Boiler and PressureVessel Code. An equivalent margins analysis was performed for 48 EFPY, and the resultsreported in Appendix A to BAW-2251A for service levels A, B, C, and D. For service levels Aand B, the results demonstrate that there is sufficient margin beyond that required by theacceptance criteria of Appendix K to Section XI of the ASME Code (1995 Edition).
Forservice levels C and D, the most limiting transient was evaluated.
Again, the results showedthat there is a sufficient margin beyond that required by the acceptance criteria of Appendix Kto Section XI of the ASME Code.As mentioned earlier in this evaluation, the applicant submitted a response to an RAI for ANO-1 regarding Supplement 1 to GL-92-01, Revision
: 1. This response was BAW-2325, Revision1. The "best estimate" chemistry composition (copper and nickel) was reported in BAW-2325, Revision
: 1. Best estimate chemistry compositions were also reported in BAW-2251A, andwere summarized in Table A-1 of Appendix A to BAW-2251A for the various reactor vesselmaterials.
The copper composition reported in BAW-2251A is equivalent to, or exceeds, thecopper content reported in BAW-2325, Revision
: 1. In addition, the 48 EFPY fluenceestimates were recalculated using the methodology described in Appendix B of BAW-2251A.
It was shown that the fluence estimates listed in BAW-2251A remain conservative.
Therefore the CvUSE values, given in Table 4-4 of BAW-2251A, remain conservative.
The Appendix K analysis, from Section Xl of the ASME Boiler and Pressure Vessel Codeinvolves a quantitative assessment of the impact of low CvUSE on reactor vessel integrity.
InAppendix K analysis, cracks are postulated at the inner reactor vessel wall. Since theneutron fluence decreases with depth into the vessel, the Appendix K analysis methodassumes the fracture toughness at the crack tip will be greater than that at the inner wall ofthe vessel. The applicant's analysis was carried out using conservative stress assumptions for service levels A, B, C, and D for 48 EFPY. The analysis, given in Appendix B ofBAW-2251A, shows that for service levels A and B, there is sufficient margin beyond thatrequired by the acceptance criteria of Appendix K to Section XI of the ASME Code (1995Edition).
For service levels C and D, the most limiting transient was evaluated, and again theanalytical results demonstrated that there is a sufficient margin beyond that required by Attachment 1 to1 CAN081403 Page 7 of 13Appendix K to Section XI of the ASME Code. The applicant concludes that evaluations for allfour service levels show the adequacy of safety against fracture for the ANO-1 vessel for48 EFPY.The staff found the B&WOG evaluation of the Charpy USE acceptable for all ANO-1materials for the period of extended operation because the 48 EFPY analysis reported inAppendix B of BAW-2251A, and referenced in this application, meets the provisions of10 CFR 54.21 (c)(1)(ii) and applies to ANO-1.3.0 REGULATORY ANALYSIS3.1 No Significant Hazards Consideration Determination Entergy Operations, Inc. (Entergy) proposes a change to the Arkansas Nuclear One, Unit 1(ANO-1) Technical Specifications (TSs) to revise the pressure
/ temperature (P/T) limits for thereactor coolant system.Entergy has evaluated the proposed changes to the TS using the criteria in 10 CFR 50.92 andhas determined that the proposed changes do not involve a significant hazards consideration.
Basis for no significant hazards consideration determination:
As required by 10 CFR 50.91(a),
Entergy analysis of the issue of no significant hazards consideration is presented below:1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
NoThe proposed change will revise the heatup, cooldown, and inservice leak hydrostatic testlimitations for the Reactor Coolant System (RCS) to a maximum of 54 Effective Full PowerYears (EFPY) in accordance with 10 CFR 50, Appendix G. This is the end of the period ofextended operation.
: Further, the proposed amendment revises the enable temperature and the lift setpoint for Low Temperature Overpressurization Protection (LTOP)requirements to reflect the revised P/T limits of the reactor vessel. The P/T limits weredeveloped in accordance with the requirements of 10 CFR 50, Appendix G, utilizing theanalytical methods and flaw acceptance criteria of Topical Report BAW-10046A, Revision 2, and American Society of Mechanical Engineers (ASME) Code, Section XI,Appendix G. These methods and criteria are the previously NRC approved standards forthe preparation of P/T limits. Updating the P/T limits for additional EFPYs maintains thelevel of assurance that reactor coolant pressure boundary integrity will be maintained, asspecified in 10 CFR 50, Appendix G.The proposed changes do not adversely affect accident initiators or precursors, and do notalter the design assumptions, conditions, or configuration of the plant or the manner inwhich the plant is operated and maintained.
The ability of structures,
: systems, andcomponents to perform their intended safety functions is not altered or prevented by theproposed
: changes, and the assumptions used in determining the radiological consequences of previously evaluated accidents are not affected.
Therefore, this change does not involve a significant increase in the probability orconsequences of an accident previously evaluated.
Attachment 1 to1CAN081403 Page 8 of 132. Does the proposed change create the possibility of a new or different kind of accidentfrom any accident previously evaluated?
Response:
NoThe proposed changes incorporate methodologies that either have been approved oraccepted for use by the NRC (provided that any conditions
/ limitations are satisfied).
The PIT limits and LTOP limits will provide the same level of protection to the reactorcoolant pressure boundary as was previously evaluated.
Reactor coolant pressureboundary integrity will continue to be maintained in accordance with 10 CFR 50,Appendix G, and the assumed accident performance of plant structures, systems andcomponents will not be affected.
These changes do not involve any physical alteration of the plant (i.e., no new or different type of equipment will be installed),
and installed equipment is not being operated in a new or different manner. Thus, no new failuremodes are introduced.
Therefore, this change does not create the possibility of a new or different kind ofaccident from an accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?Response:
NoThe proposed changes do not affect the function of the reactor coolant pressure boundary orits response during plant transients.
By calculating the P/T limits and associated LTOP limitsusing NRC-approved methodology, adequate margins of safety relating to reactor coolantpressure boundary integrity are maintained.
The proposed changes do not alter the mannerin which safety limits, limiting safety system settings, or limiting conditions for operation aredetermined.
These changes will ensure that protective actions are initiated and theoperability requirements for equipment assumed to operate for accident mitigation are notaffected.
Therefore, this change does not involve a significant reduction in a margin of safety.Based upon the reasoning presented above, Entergy concludes that the requested changeinvolves no significant hazards consideration, as set forth in 10 CFR 50.92(c),
"Issuance ofAmendment."
3.2 Applicable Regulatory Requirements/Criteria The NRC has established requirements in Title 10 of the Code of Federal Regulations (10 CFR)Part 50, to protect the integrity of the reactor coolant pressure boundary in nuclear power plants.The NRC staff evaluates the P/T limits based on the following regulations and guidance:
Appendix G to 10 CFR 50 requires that P/T limits be at least as conservative as those obtainedby applying the methodology of Appendix G to Section XI of the American Society forMechanical Engineering (ASME), Boiler and Pressure Vessel Code. Appendix G to 10 CFR 50also provides minimum temperature requirements that must be considered in the development of the P/T limit curves. Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement ofReactor Vessel Materials and Its Impact on Plant Operations,"
advised licensees that the NRC Attachment 1 to1 CAN081403 Page 9 of 13staff would use Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor VesselMaterial,"
Revision 2, to review P/T limits. RG 1.99, Revision 2 contains methodologies fordetermining the increase in transition temperature and the decrease in upper-shelf energy(USE) resulting from neutron radiation.
The GL 92-01, "Reactor Vessel Structural Integrity, "Revision 1, requested that licensees submittheir reactor pressure vessel (RPV) materials property data for their plants to the NRC staff forreview. GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess datafrom other licensees that could affect their RV integrity evaluations.
Standard Review Plan (STP), Branch Technical Position (BTP) 5-3, Revision 3, of NUREG-0800, provides an acceptable method of determining the P/T limit curves for ferritic materials inthe beltline of the RV based on the linear elastic fracture mechanics methodology of Appendix Gto Section XI of the ASME Code. The basic parameter of this methodology is the stressintensity factor, Kic, which is a function of the stress state and flaw configuration.
ASME Code,Section XI, Appendix G, requires a safety factor of 2.0 on stress intensities resulting frompressure during normal and transient operating conditions, and a safety factor of 1.5 on thesestress intensities for hydrostatic testing curves.The flaw postulated in the ASME Code, Section XI, Appendix G, has a depth that is equal to 1/4of the RPV beltline thickness and a length equal to 1.5 times the RV beltline thickness.
Thecritical locations in the RPV beltline region for calculating heatup and cooldown P/T limits arethe 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximumdepth of the postulated inside surface and outside surface defects, respectively.
Themethodology found in Appendix G to Section XI of the ASME Code requires that the adjustedreference temperature (ART or adjusted RT NOT) be determined by evaluating material propertychanges due to neutron irradiation.
The ART is defined as the sum of the initial RTNDT, themean value of the adjustment in reference temperature caused by irradiation (ARTNDT),
and amargin (a) term. The ARTNDT is a product of a chemistry factor (CF) and a fluence factor (FF).The CF is dependent upon the amount of copper and nickel in the material and may bedetermined from tables in RG 1.99, Revision 2, or from surveillance data. The FF is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF wasdetermined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term isused to account for uncertainties in the values of the initial RTNDT, the copper and nickelcontents, the neutron fluence and the calculational procedures.
RG 1.99, Revision 2, describes the methodology to be used in calculating the margin term.AREVA Topical Report BAW-2308, Revision 1-A and Revision 2-A provide NRC-approved alternate initial RT NDT and associated ai values for various heats of Linde 80 beltline weldmaterials for RPV integrity evaluation applications.
Section 50.60 of 10 CFR imposes fracture toughness and material surveillance programrequirements, which are set forth in 10 CFR 50, Appendices G and H. In the "Definitions" section of Appendix G, paragraph G.II.D(ii) states, "For the reactor vessel beltline materials, ARTNDT must account for the effects of neutron radiation."
In the "Fracture Toughness Requirements"
: section, paragraph G.IV.A states in part, " ... the values of RTNDT and Charpyupper-shelf energy must account for the effects of neutron radiation, including the results of thesurveillance program of Appendix H of this part." The effects of neutron radiation aredetermined, in part, by estimating the neutron fluence on the reactor vessel.
Attachment 1 to1CAN081403 Page 10 of 13RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining thepressure vessel neutron fluence with respect to the General Design Criteria (GDC) contained inAppendix A of 10 CFR 50. In consideration of the guidance set forth in RG 1.190, GDC 14, 30,and 31 are applicable.
GDC 14 requires the design, fabrication,
: erection, and testing of thereactor coolant pressure boundary so as to have an extremely low probability of abnormalleakage, of rapidly propagating
: failure, and of gross rupture.
GDC 30 requires among otherthings, that components comprising the reactor coolant pressure boundary be designed, fabricated,
: erected, and tested to the highest quality standards practical.
GDC 31 pertains tothe design of the reactor coolant pressure
: boundary, stating:The reactor coolant pressure boundary shall be designed with sufficient margin to assure thatwhen stressed under operating, maintenance,
: testing, and postulated accident conditions, (1) the boundary behaves in a non-brittle manner and (2) the probability of rapidlypropagating fracture is minimized.
The design shall reflect consideration of servicetemperatures and other conditions of the boundary material under operating maintenance,
: testing, and postulated accident conditions and the uncertainties in determining (1) materialproperties, (2) the effects of irradiation on material properties, The construction permit for ANO-1 was issued by the Atomic Energy Commission (AEC) onDecember 6, 1968, and an operating license was issued on May 21, 1974. The ANO-1operating license was issued based on compliance with the proposed GDC published by theAEC in Reference 2 (hereinafter referred to as "draft GDC"). The AEC published the final rulethat added Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants,"in Reference 3 (hereinafter referred to as "final GDC" or "GDC"). In accordance withReference 4, the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes ANO-1.ANO-1 Safety Analysis Report (SAR) section 1.4.10 incorporates the current GDC 14. SARSection 1.4.26 discusses GDC 30, and GDC 31 is discussed in SAR Section 1.4.27.3.4 Precedence This amendment and the separate exemption request are similar to the ones approved forThree Mile Island Nuclear Station, Unit 1 (References 5 and 6) and Oconee Nuclear Station,Units 1, 2, and 3 (Reference 7). Three Mile Island and the three units at Oconee all haveBabcock & Wilcox reactor vessels with Linde 80 welds similar to the ANO-1 reactor vessel.3.5 Topical Report Conditions The methodologies described in three separate topical reports were used in the development ofthis submittal.
These topical reports are:BAW-10046A, Revision 6, "Methods of Compliance with Fracture Toughness andOperational Requirements of 10 CFR 50, Appendix G"BAW-2241 PA, Revision 2, "Fluence and Uncertainty Methodologies" BAW-2308, Revisions 1-A and 2-A, "Initial RTNDT of Linde 80 Weld Materials" Each of these topical reports have been reviewed and approved by the NRC.
Attachment 1 to1CAN081403 Page 11 of 13BAW- 10046The Safety Evaluation (SE) associated with BAW-1 0046 states that the NRC staff hasdetermined that the methods are acceptable for application to the generation of P/T limit curvesfor pressurized water reactor (PWR) applications.
The Staff found the report to be acceptable for referencing in licensing applications to the extent specified and under the limitations delineated in the report and in the associated SE.A review of the SEs for all the revisions of this topical report did not identify any additional limitations in the use of this topical.
It should be noted that the current 31 EFPY P/T limits weredeveloped using this methodology.
BAW-2241A review of the SEs associated with BAW-2241 P-A, Revision 2 demonstrates that this revisionis an extension of the PWR calculational methodology for application to boiling water reactors.
This revision is not applicable to ANO-1.The SE for Revision 1 of the topical report concluded that the proposed methodology isacceptable for referencing in licensing applications for determining the pressure vessel fluenceof Westinghouse, Combustion Engineering and B&W designed reactors.
In addition, there arethree limitations imposed in the SE for Revision 1 of the topical.
These limitations involvedanalysis of reactor designs not included in BAW-2241 P-A database (e.g., partial length fluenceassembly designs),
changes in cross sections from those reviewed by the Staff, and any otherchanges in methodology.
The design of the ANO-1 has been included in the BAW-2241P-A database.
There are no changes in the cross sections or other changes to the methodology inthe current application.
The NRC found that methodology presented in Revision 0 of BAW-2241 was acceptable fordetermining the pressure vessel fluence of B&W designed reactors and to be referenced inB&W designed reactor licensing actions.
Three limitations were listed in the SE for this revision.
These include that the methodology is applicable only to B&W designed reactors; changes incross sections from those reviewed by the Staff, and provide the staff with a record of futuremodifications of the methodology, ANO-1 is a B&W designed reactor.
As noted above, thereare no changes in the cross sections from that previously reviewed and subsequent
: changes, ifany, have been presented to the NRC and the Staff has reviewed those changes.
See thediscussions above related to Revisions 1 and 2 of the topical.BAW-2308The SE for BAW-2308, Revision 2-A provides an NRC-approved alternate initial RTNDT andassociated a values for the Linde 80 weld material present in the beltline region of the reactorpressure vessels at Oconee Units 1, 2 and 3.
Attachment 1 to1 CAN081403 Page 12 of 13The following Conditions and Limitations are stated in the SE for BAW-2308, Revision 1-AAny license who wants to utilize the methodology of TR BAW-2308, Revision 1 as outlined initems (1) through (3) above, must request an exemption, per 10 CFR 50.12, from therequirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so.Condition and Limitation (2) requires that a minimum chemistry factor of 167.0 OF be appliedwhen the methodology of Regulatory Guide 1.99, Revision 2, and 10 CFR 50.61 is used toassess the shift in nil-ductility transition temperature due to irradiation.
Condition and Limitation (3) requires that a value of 0a = 28.0 OF be used to determine themargin term, as defined in Topical Report BAW-2308, Revision 1 and Regulatory Guide 1.99,Revision 2.This exemption request was submitted to the Staff via Reference
: 1. The analyses performed tosupport the exemption request included the values listed in Condition and Limitations 2 and 3.As of the date of this submittal, the exemption request is being reviewed by the NRC Staff.As demonstrated above, the limitations and conditions imposed on the three topical reports thatwere utilized in the development of the ANO-1 P/T limits have been satisfied and the reports areapplicable to ANO-1.4.0 ENVIRONMENTAL CONSIDERATION The proposed change would change a requirement with respect to installation or use of a facilitycomponent located within the restricted area, as defined in 10 CFR Part 20, and would changean inspection or surveillance requirement.
: However, the proposed change does not involve(i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released
: offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposedchange meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
 
==5.0 REFERENCES==
 
1 Entergy Letter to NRC, "Request for Exemption from Certain 10 CFR 50.61 and10 CFR 50, Appendix G Requirements,"
dated March 20, 2014 (1CAN031403)
(ML14083A640)
: 2. Federal Register (32 FR 10213) on July 11, 19673. Federal Register (36 FR 3255) on February 20, 19714. NRC Staff Requirements Memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223
-Resolution of Deviations Identified During the Systematic Evaluation Program,"
datedSeptember 18, 1992 (ML003763736)
Attachment 1 to1CAN081403 Page 13 of 135. NRC Letter to Exelon Nuclear, "Three Mile Island Nuclear Station, Unit 1 -Exemption from Certain Requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, For InitialRTNDT Values for Linde 80 Welds (TAC No. MF0425),"
dated December 13, 2013(ML13324A086)
: 6. NRC Letter to Exelon Nuclear, "Three Mile Island Nuclear Station, Unit 1 -Issuance ofAmendment RE: Revision to the Pressure and Temperature Limit Curves and the LowTemperature Overpressure Protection Limits (TAC No. MF0424),"
dated December 13,2013 (ML13325A023)
: 7. NRC Letter to Duke Energy Carolinas, LLC, "Oconee Nuclear Station, Units 1, 2, and 3;Issuance of Amendments Regarding Pressure
-Temperature Limits (TAC NOS.MF0763, MF0764, and MF0765),"
dated February 27, 2014 (ML14041A093)
: 8. NRC Cover Letter to Entergy, "Arkansas Nuclear One, Units 1, License Renewal SafetyEvaluation Report,"
dated April 12, 2001 (ML011030091)
(SER ML011020554)
Attachment 2 to1CAN081403 Proposed Technical Specification and Bases Changes (mark-up)
RCS P/T Limits3.4.3FIGURE 3.4.3-1RCS Heatup Limitations to 5431- EFPYC/)(L0~CDf-JCLU)2400220020001800160014001200100080060040020000 50 100 150 200 250300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
('F)Notes:1. These-Ccurves are not adjusted for instrument error and shall not be used for operation.
: 2. When DHR is in operation with no RCPs operating, the DHR system return temperature shall be used.3. RCP Operating Restrictions:
RCS TEMPT >< 30250 OF302500FF > T > -25100 OF2252?F&#xfd; > T>:&#xfd; 84 ?FT < 10084 OFRCP RESTRICTIONS None532No RCPs operating
: 4. Allowable Heatup Rates:RCS TEMP60 OF < T < 84 OFT > 84 OFH/U RATE< 15 &deg;F/hrH_4As allowed by applicable curveIANO-13.4.3-5Amendment No. 215 RCS P/T Limits3.4.3FIGURE 3.4.3-2RCS Cooldown Limits to 5434 EFPY0~CI-(9LU-JofCLU)0~CC)2400220020001800160014001200100080060040020000 50 100 150 200 250 300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
(-F)Notes:1. This curve is not adjusted for instrument error and shall not be used for operation.
: 2. A maximum step temperature change of 25 OF is allowable when securing all RCPs with theDHR system in operation.
This change is defined as the RCS temperature prior to securingall the RCPs minus the DHR return temperature after the RCPs are secured.
When DHR isin operation with no RCPs operating, the DHR system return temperature shall be used.3. RCP Operating Restrictions:
RCS TEMPT >> 30250 'F*30250&deg;F
&#xfd;&#xfd;> T> 2t26100 OF2250F >T .R1 FT < 10084 OFRCP RESTRICTIONS None<3No RCPs operating
: 4. Allowable Cooldown Rates:RCS TEMPT > 280 OF280&deg;F > T > 150 OFC/D RATE100 &deg;F/hrHR50 &deg;F/hrHR (Note 5)STEP CHANGE5 50 OF in any 1/2 hrHR< 25 OF in any 1/2 hrHRAmendment No. 215ANO-13.4.3-6 RCS P/T Limits3.4.3T < 150 OF 25 &deg;F/hrHR < 25 OF in any 1 hrHiR*eI44 ee...te...3
-. 4 2.4.h..,hG
'P,reduc~ed to 30 '..F in 15 hours..ANO-13.4.3-7Amendment No. 215 RCS P/T Limits3.4.3FIGURE 3.4.3-3RCS Inservice Hydrostatic Test H/U & C/D Limits to 54341 EFPYLu--J0LI-U)(nu-iL)c-)2400220020001800160014001200100080060040020000 50 100 150 200 250 300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
(-F)Notes:1. This curve is not adjusted for instrument error and shall not be used for operation.
: 2. All Notes on Figure 3.4.3-1 are applicable for heatups.
This curve is based on a heatuprate of < 90 &deg;F/HR.3. All Notes on Figure 3.4.3-2 are applicable for cooldowns.
ANO-13.4.3-8Amendment No. 215 Pressurizer 3.4.93.4 REACTOR COOLANT SYSTEM (RCS)3.4.9 Pressurizer LCO 3.4.9The pressurizer shall be OPERABLE with:a. Pressurizer water level < 320 inches; andb. A minimum of 126 kW of Engineered Safeguards (ES) bus poweredpressurizer heaters OPERABLE.
---------------------------
NOTE -------------------------------------------
OPERABILITY requirements on pressurizer heaters do not apply inMODE 4.APPLICABILITY:
MODES 1, 2, and 3,MODE 4 with RCS temperature
> 262248&deg;F.
ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Pressurizer water level not A.1 Restore level to within 1 hourwithin limits, limits.B. Required Action and B.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition A not ANDmet.B.2 Be in MODE 4 with RCS 24 hourstemperature
_< 2622480F.C. Capacity of ES bus C.1 Restore pressurizer heater 72 hourspowered pressurizer capacity.
heaters less than limit.D. Required Action and D.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition C not ANDmet.D.2 Be in MODE 4. 12 hoursANO-13.4.9-1Amendment No. 215, 24.
Pressurizer Safety Valves3.4.103.4 REACTOR COOLANT SYSTEM (RCS)3.4.10 Pressurizer Safety ValvesLCO 3.4.10Two pressurizer safety valves shall be OPERABLE.
-I -ILJr-.N -----------------------------------------------
: 1. Only one pressurizer safety valve is required to be OPERABLE inMODE 3, and in MODE 4 with RCS temperature
> 262.248 OF.2. The lift settings are not required to be within limits for entry into MODE 3or the applicable portions of MODE 4 for the purpose of setting thepressurizer safety valves under ambient (hot) conditions.
This exception is allowed for 36 hours following entry into MODE 3 provided apreliminary cold setting was made prior to heatup.3. Not applicable in MODE 3, and in MODE 4 with RCS temperature
> 26-2248 OF during hydrostatic tests in accordance with ASME Boilerand Pressure Vessel Code, Section III.4. The provisions of LCO 3.0.3 are not applicable in MODE 3, and inMODE 4 with RCS temperature
> 2-62248 OF.APPLICABILITY:
MODES 1, 2, and 3,MODE 4 with RCS temperature
> 262248 OF.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One pressurizer safety A.1 Restore valve to OPERABLE 15 minutesvalve inoperable in status.MODES 1 or 2.B. Required Action and B.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition A notmet.ORTwo pressurizer safetyvalves inoperable inMODES 1 or 2.ANO-13.4.10-1Amendment No. 215 Pressurizer Safety Valves3.4.10CONDITION REQUIRED ACTION COMPLETION TIMEC. Required pressurizer C.1 Be in MODE 4 with RCS 18 hourssafety valve inoperable in temperature
< 242248 OF.MODE 3 or MODE 4 withRCS temperature
> 262248 OF.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each required pressurizer safety valve is In accordance withOPERABLE in accordance with the Inservice Testing the Inservice Program.
Following
: testing, as-left lift settings shall Testing Programbe within +/- 1%.ANO-13.4.10-2Amendment No. 215 LTOP System3.4.113.4 REACTOR COOLANT SYSTEM (RCS)3.4.11 Low Temperature Overpressure Protection (LTOP) SystemLCO 3.4.11An LTOP System shall be OPERABLE with high pressure injection (HPI)deactivated and the core flood tanks (CFTs) isolated and:----NOTES-1. HPI deactivation and CFT isolation not applicable during ASME SectionXI testing.2. HPI deactivation not applicable during fill and vent of the RCS.3.4.HPI deactivation not applicable during emergency RCS makeup.HPI deactivation not applicable during valve maintenance.
: 5. CFT isolation is only required when CFT pressure is greater than orequal to the maximum RCS pressure for the existing RCS temperature allowed by the pressure and temperature curves provided in LCO 3.4.3,"RCS Pressure and Temperature (P/T) Limits."a. Pressurizer level such that the unit is not in a water solid condition andan OPERABLE electromatic relief valve (ERV) with a setpoint of< 46r0563.8 psig; or------------------
--- NOTES ---------------------
: 1. Pressurizer level not applicable as allowed by Emergency Operating Procedures.
: 2. Pressurizer level not applicable during system hydrotest.
: b. The RCS depressurized and the RCS open.APPLICABILITY:
MODE 4 with RCS temperature
< 26-2248 'F,MODE 5,MODE 6 when the reactor vessel head is on.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Pressurizer level not A.1 Restore pressurizer level to 1 hourwithin required limits, within required limits.ANO-13.4.11-1Amendment No. 215 LTOP System3.4.11B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.3 RCS Pressure and Temperature (P/T) LimitsBASESBACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due tosystem pressure and temperature changes.
These loads are introduced by startup(heatup) and shutdown (cooldown) operations, and unit transients.
This LCO limitsthe pressure and temperature changes during RCS heatup and cooldown, within thedesign assumptions and the stress limits for cyclic operation.
Figures 3.4.3-1, 3.4.3-2, and 3.4.3-3 contain PIT limit curves for heatup, cooldown, inservice hydrostatic
: testing, and physics testing at RCS temperatures
< 525 OF, andthe maximum rate of change of reactor coolant temperature.
The methods andcriteria employed to establish operating pressure and temperature limits aredescribed in BAW-10046A (Ref. 1). These limit curves are applicable through T40y-onefifty-four effective full power years (EFPY) of operation.
The pressure limit isadjusted for the pressure differential between the point of system pressuremeasurement and the limiting component for the various operating reactor coolantpump combinations.
Each P/T curve defines an acceptable region for normal operation below and to theright of the limit curve. The curves are used to develop operational guidance for useduring heatup or cooldown maneuvering.
The LCO establishes operating limits that provide a margin to brittle failure of thereactor vessel. The vessel is the component most subject to brittle failure due to thefast neutron embrittlement it experiences during power operation, and the LCO limitsapply mainly to the vessel. The limits do not apply to the pressurizer, which hasdifferent design characteristics and operating functions.
10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits for materialfracture toughness requirements of the reactor coolant pressure boundary (RCPB)materials.
Reference 2 requires an adequate margin to brittle failure during normaloperation, abnormalities, and system hydrostatic tests. It mandates the use of theAmerican Society of Mechanical Engineers (ASME), Boiler and Pressure VesselCode, Section III, Appendix G (Ref. 3).Linear elastic fracture mechanics (LEFM) methodology is used to determine thestresses and material toughness at locations within the RCPB. The LEFMmethodology follows the guidance given by 10 CFR 50, Appendix G; ASME Code,Section III, Appendix G; and Regulatory Guide 1.99 (Ref. 4). For the LInde 80 weldmaterials present in the ANO-1 reactor vessel beltline an alternative approach fordetermining the adiusted reference nil-ductility temperature as described in TopicalReport BAW-2308, Revisions 1-A and 2-A (Ref. 12). The Master Curve methodology is accepted with exemption from the requirements of 10 CFR 50.61 (ref. 13) and 10CFR 50, Appendix G (Ref.2)ANO-1 B 3.4.3-1 Amendment Ne. 215Rev. 32 LTOP System3.4.11BACKGROUND (continued)
The major components of the reactor coolant pressure boundary have beenanalyzed in accordance with Appendix G to 10CFR50.
Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressureboundary, are given in FT4 DoumentR-77-42:8569O-4CALC-14-E-0100-08 (Ref. 5).The. service period was.
year. from-hat-assumed performed by the NRC staff. The iing wld mati. b dthe-B&WV-OwRers.
Group-integratedReaetor-Vessel*.-Material.-Su~-vei.anee-.
P.Fofamd-e-ea~4hes--4e4ttiffwe4 mat -r.~i&41ieus"d nthe latestr *- ntaBW-re~e
-BAW-1 {-4Thechemical composition of the limiting weld material is reported in the B&W report,BAW-2-t-1-P2317 (Rev. 7). The effect of neutron irradiation on the nil ductility reference temperature (RTNDT) of the limiting weld material is reported in F-T4Calculations a2-.124591-.7--09..and32--12*77-1.6-roCALC-14-E-0100-02 (Rev. 8) andCALC-14-E-0100-09 (Ref. 14).The actual shift in the RTNDT of the vessel beltline region material will be established periodically by removing and evaluatihg the irradiated reactor vessel materialsurveillance specimens, in accordance with Appendix H of 10 CFR 50 (Ref. 9). Thesespecimens are installed near the inside wall ef this-or--in other similar reactor vesselsin the core region. The operating PIT limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 3.Prior to reaching thir4efift-foour effective full power years of operation, Figures3.4.3-1, 3.4.3-2, and 3.4.3-3 must be updated for the next service period inaccordance with 10 CFR 50, Appendix G. The service period must be of sufficient duration to permit the scheduled evaluation of a portion of the surveillance datascheduled in accordance with the latest revision of Topical Report BAW-1 543 (Ref.6) and Topical Report BAW-2308 (Ref. 12). The highest predicted adjustedreference temperature of all the beltline region materials is used to determine theadjusted reference temperature at the end of the service period. The basis for thisprediction is submitted for NRC staff review at least 90 days prior to the end of theservice period.The PIT limit curves are composite curves established by superimposing limitsderived from stress analyses of those portions of the reactor vessel and head thatare the most restrictive.
At any specific
: pressure, temperature, and temperature rateof change, one location within the reactor vessel will dictate the most restrictive limit.Across the span of the P/T limit curves, different locations are more restrictive, and,thus, the curves are composites of the most restrictive regions.The heatup curve represents a different set of restrictions than the cooldown curvebecause the directions of the thermal gradients through the vessel wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between theouter and inner walls.The calculation to generate the inservice hydrostatic testing curve uses different safety factors (per Ref. 3) than the heatup and cooldown curves. The testing curvealso extends to the RCS design pressure of 2500 psia.ANO-1 B 3.4.3-2 No. 215Rev. 32 LTOP System3.4.11LCO (continued)
The limits for the rate of change of temperature control the thermal gradient throughthe vessel wall and are used as inputs for calculating the P/T limit curves. Thus, theLCO for the rate of change of temperature restricts stresses caused by thermalgradients and also ensures the validity of the PIT limit curves.The limit curves include the limiting pressure differential between the point of systempressure measurement and the pressure on the reactor vessel region controlling thelimit curve. However, the limit curves are not adjusted for possible instrument errorand should not be used for operation.
The heatup and cooldown rates stated are intended as the maximum changes intemperature in one direction in the stated time periods.
The actual temperature linear ramp rate may exceed the stated limits for a shorter time period provided thatthe maximum total temperature difference does not exceed the limit and that atemperature hold is observed to prevent the total temperature difference fromexceeding the limit for the stated time period.The acceptable PIT combinations are below and to the right of the limit curves whichare applicable for the first 34-ifty-four EFPY. The limit curves include the limitingpressure differential between the point of system pressure measurement and thepressure on the reactor vessel region controlling the limit curve. However, the limitcurves are not adjusted for possible instrument error and should not be used foroperation.
Violating the LCO limits places the reactor vessel outside of the bounds of the stressanalyses and can increase stresses in other RCPB components.
The consequences depend on several factors, as follows:a. The magnitude of the departure from the allowable operating P/T regime or themagnitude of the rate of change of temperature;
: b. The length of time the limits were violated (longer violations allow thetemperature gradient in the thick vessel walls to become more pronounced);
andc. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The RCS P/T limits Specification provides a definition of acceptable operation forprevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2).Although the PIT limits were developed to provide guidance for operation duringheatup or cooldown (MODES 3, 4, and 5) or inservice hydrostatic
: testing, theirapplicability is at all times in keeping with the concern for nonductile failure.
Thelimits do not apply to the pressurizer.
ANO-1 B 3.4.3-3 Amendment No. 215Rev. 32 LTOP System3.4.11ACTIONS (continued)
D.1 and D.2 (continued)
ASME Code, Section XI, Appendix E (Ref. 10), may also be used to support theevaluation.
: However, its use is restricted to evaluation of the vessel beltline.
Condition D is modified by a Note requiring Required Action D.2 to be completed whenever the Condition is entered.
The Note emphasizes the need to perform theevaluation of the effects of the excursion outside the allowable limits. Restoration alone, per Required Action D.1, is insufficient because higher than analyzed stressesmay have occurred and may have affected RCPB integrity.
SURVEILLANCE REQUIREMENTS SR 3.4.3.1, SR 3.4.3.2, SR 3.4.3.3, and SR 3.4.3.4Verification that operation is within the limits of the appropriate figure is requiredevery 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits arespecified in hourly increments, 30 minutes permits assessment and correction forminor deviations within a reasonable time.Surveillance for heatup, cooldown, or inservice hydrostatic testing may bediscontinued when the definition given in the relevant unit procedure for ending theactivity is satisfied.
The acceptable P/T combinations are below and to the right of the limit curves whiG4... .ppli.able fGr the f"FSt 31 EFPYs. The limit curves include the limiting pressuredifferential between the point of system pressure measurement and the pressure onthe reactor vessel region controlling the limit curve. However, the limit curves are notadjusted for possible instrument error and should not be used for operation (asidentified in Note 1 on each applicable Figure).SR 3.4.3.1 is modified by a Note that requires this SR to be performed only duringsystem heatup operations with fuel in the reactor vessel. This SR refers toFigure 3.4.3-1 which provides applicable heatup limitations, including reactor coolantpump (RCP) operating restrictions and allowable heatup rates. Figure 3.4.3-1 Note 2identifies that when the decay heat removal system is operating with no RCPsoperating, the indicated DHR system return temperature to the reactor vessel is theappropriate temperature indicator.
ANO-1 B 3.4.3-4 Amendment No. 215Rev. 32 RCS P/T LimitsB 3.
 
==4.3REFERENCES==
: 1. BAW-10046A, "Methods of Compliance with Fracture Toughness andOperational Requirements of 10CFR50, Appendix G", Rev. 2, June 1986.2. 10 CFR 50, Appendix G, Fracture Toughness Requirements.
: 3. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.4. Regulatory Guide 1.99, Revision 2, May 1988.5. FTI DOcTument 77-1258-569-01-CALC-14-E-0100-08.
ANO-1 Corrected P-T Limitsfor 60 Years (54 EFPY).6. BAW-1543, Master Integrated Reactor Vessel Matepa4Surveillance Program(latest revision).
: 7. BAW--421r-PkFadiation Induc +R-arpyWpe-Se fr-gy--fReactor Vessel We!dsBAW-2313, Revision 6, B&W Fabricated Reactor VesselMaterials and Surveillance Data Information.
: 8. FT-! Galculations 32 1245917 00 @nd 32 12-577!&
CALC-14-E-0100-02, ANO-1 ART (Adjusted Reference Temp) Values at 54 EFPY.9. 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance ProgramRequirements.
: 10. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.11. 10 CFR 50.36, Technical specifications.
: 12. BAW-2308, Revisions 1-A and 2-A, Initial RTNDT of Linde 80 Weld Materials
: 13. 10 CFR 50.61, Fracture Toughness Requirements for Protection AgainstPressurized Thermal Shock Events14. CALC-14-E-0100-09, ANO-1 Fluence Analysis Report, Cycles 21, 22, and 23 forRV Beltline.
ANO-1 B 3.4.3-9 Amendmont No. 215Rev. 32 Pressurizer B 3.4.9LCOThe LCO requirement for the pressurizer to be OPERABLE with a water levels 320 inches ensures that a steam bubble exists prior to criticality.
Limiting themaximum operating water level preserves the steam space for pressure control.
TheLCO has been established to ensure the capability to establish and maintain pressurecontrol for steady state operation and to minimize the consequences of potential overpressure transients.
Requiring the presence of a steam bubble is also consistent with analytical assumptions.
The LCO requires a minimum of 126 kW (nominal) of pressurizer heaters OPERABLE.
To be considered
: OPERABLE, the required heaters must be powered from an ES bus.NUREG-0578 (Ref. 1) specifies that the minimum required pressurizer heaters arecapable of being powered from redundant, emergency diesel generator backed sources.This provides assurance that sufficient heater capacity is available to provide RCSpressure control during a loss of off-site power. The amount needed to maintainpressure is dependent on the insulation losses, which can vary due to tightness of fit andcondition.
APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatesteffect on RCS temperature, resulting in the greatest effect on pressurizer level and RCSpressure control.
Thus Applicability has been designated for MODES 1 and 2. TheApplicability is also provided for MODE 3 and, for pressurizer water level, for MODE 4with RCS temperature
> 2-62248 OF. The purpose is to prevent water solid RCSoperation during heatup and cooldown to avoid rapid pressure rises caused by normaloperational perturbations, such as reactor coolant pump startup.
The temperature of2-62248 OF has been designated as the cutoff for applicability because LCO 3.4.11, "LowTemperature Overpressure Protection (LTOP),"
provides a requirement for pressurizer level at or below 262248 OF. The LCO does not apply to MODE 5 with loops filledbecause LCO 3.4.11 applies and provides adequate overpressure protection.
Thisparameter value does not contain allowances for instrument uncertainty.
Additional allowances for instrument uncertainty are contained in the implementing procedures.
The LCO does not apply to MODES 5 and 6 with partial loop operation.
In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply. In the event of aloss of offsite power, the initial conditions of these MODES give the greatest demand formaintaining the RCS in a hot pressurized condition with loop subcooling for an extendedperiod. The Applicability is modified by a Note stating that the OPERABILITY requirements on pressurizer heaters do not apply in MODE 4. For MODE 4, 5, or 6, theneed to control pressure (by heaters) to ensure loop subcooling for heat transfer issignificantly reduced when the Decay Heat Removal System is in service, and therefore the LCO is not applicable.
ANO-1 B 3.4.9-3 Amondment No. 215Rev. 2-7-,39,35 Pressurizer B 3.4.9ACTIONSWith pressurizer water level outside the limit, action must be taken to restore pressurizer operation to within the bounds assumed in the analysis.
This is done by restoring thepressurizer water level to within the limit. The 1 hour Completion Time is considered tobe a reasonable time for adjusting pressurizer level.B.1 and B.2If the water level cannot be restored, reducing core power constrains heat input effectsthat drive pressurizer insurge that could result from an anticipated transient.
By shuttingdown the reactor and reducing reactor coolant temperature to at least MODE 3, thepotential thermal energy of the reactor coolant mass for mass and energy releases isreduced.Six hours is a reasonable time based upon operating experience to reach MODE 3 fromfull power in an orderly manner and without challenging unit systems.
Further pressureand temperature reduction to MODE 4 with RCS temperature
< 62.248 OF places theunit into a MODE where the LCO is not applicable.
The 24 hour Completion Time toreach the non-applicable MODE is reasonable based upon operating experience.
C. 1If the required pressurizer heaters are inoperable, restoration is required in 72 hours.The Completion Time of 72 hours is reasonable considering the anticipation that ademand caused by loss of offsite power will not occur in this period. Pressure controlmay be maintained during this time using non-ES bus powered heaters.D.1 and D.2If the Required Action and associated Completion Time are not met, the unit must bebrought to a MODE in which the LCO does not apply. To achieve this status, the unitmust be brought to MODE 3 within 6 hours and to MODE 4 within the following 6 hours.The Completion Time of 6 hours is reasonable, based on operating experience, to reachMODE 3 from full power conditions in an orderly manner and without challenging unitsystems.
Similarly, the Completion Time of 12 hours to reach MODE 4 is reasonable based on operating experience to achieve power reduction from full power conditions inan orderly manner and without challenging unit systems.ANO-1 B 3.4.9-4 Amendment No. 215Rev. 37 Pressurizer B 3.4.9SURVEILLANCE REQUIREMENTS SR 3.4.9.1This SR requires that pressurizer water level be maintained below the upper limit toprovide a minimum space for a steam bubble. The value specified for pressurizer leveldoes not contain an allowance for instrument error. Therefore, additional allowances forinstrument uncertainties must be provided in the implementing procedures.
The 12 hourinterval has been shown by operating practice to be sufficient to regularly assess the levelfor any deviation and verify that operation is within safety analyses assumptions.
Alarmsare also available for early detection of abnormal level.SR 3.4.9.2The SR requires sufficient pressurizer heaters which are connected to an ES bus verifiedto be capable of providing the required capacity.
(This may be done by testing the powersupply output and by performing an electrical check on heater element continuity andresistance.)
The Frequency of 18 months is considered adequate to detect heaterdegradation and has been shown by operating experience to be acceptable.
REFERENCES
: 1. NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations,"
July 1979.2. 10 CFR 50.36 Technical specifications.
ANO-1B 3.4.9-5AmendmnRt No. 215Rev. 37 Pressurizer Safety ValvesB 3.4.10B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.10 Pressurizer Safety ValvesBASESBACKGROUND The purpose of the two spring loaded pressurizer safety valves is to provide RCSoverpressure protection (Ref. 1). Operating in conjunction with the Reactor Protection System (RPS), two valves are used to ensure that the Safety Limit (SL) of 2750 psig isnot exceeded for analyzed transients during operation in MODES 1 and 2. One safetyvalve is required for MODE 3 and portions of MODE 4. For the remainder of MODE 4,MODE 5, and MODE 6 with the reactor head on, overpressure protection is provided byoperating procedures and LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP)."The self actuated pressurizer safety valves are designed in accordance with therequirements set forth in the ASME Boiler and Pressure Vessel Code, Section III(Ref. 2). The required lift pressure is 2500 psig + 1%, -3%. The safety valvesdischarge steam from the pressurizer to a quench tank located in the reactor building.
The discharge flow is indicated by acoustic flow monitoring
: devices, by an increase intemperature downstream of the safety valves, and by an increase in the quench tanktemperature,
: pressure, and level.The upper and lower as-left pressure limits are based on the +/- 1% tolerance requirement for lifting pressures above 1000 psig. The lift setting is for the ambientconditions associated with MODES 1, 2, and 3. This requires either that the valves beset hot or that a correlation between hot and cold settings be established.
The pressurizer safety valves are part of the primary success path and mitigate theeffects of postulated accidents.
OPERABILITY of the safety valves ensures that theRCS pressure will be limited to 110% of design pressure.
The consequences ofexceeding the ASME pressure limit could include damage to RCS components, increased
: leakage, or a requirement to perform additional stress analyses prior toresumption of reactor operation.
APPLICABLE SAFETY ANALYSESThe overpressure protection analysis (Ref. 3) is based on operation of both safetyvalves and assumes that the valves open at the high range of the setting (2500 psigsystem design pressure plus 1%). One pressurizer code safety valve is capable ofpreventing overpressurization in MODE 3 and in MODE 4 with RCS temperature
> 262248 'F since its relieving capacity is greater than that required by the sum of theavailable heat sources, i.e., pump energy, pressurizer
: heaters, and reactor decay heat(Ref. 1 and 4). These valves must accommodate pressurizer insurges thatANO-1 B 3.4.10-1 Amendment No. 215Rev. 37 Pressurizer Safety ValvesB 3.4.10APPLICABLE SAFETY ANALYSES (continued) could occur during a startup, rod withdrawal, or ejected rod event. The startup accidentestablishes the minimum safety valve capacity.
The startup accident is assumed tooccur at low power. Single failure of a safety valve is neither assumed in the accidentanalysis nor required to be addressed by the ASME Code. Compliance with thisSpecification is required to ensure that the accident analysis and design basiscalculations remain valid.In MODES 1 and 2, pressurizer safety valves satisfy Criterion 3 of the10 CFR 50.36 (Ref. 5). In MODE 3 and MODE 4 above the LTOP enable temperature, the pressurizer safety valves satisfy Criterion 4 of 10 CFR 50.36.LCOThe two pressurizer safety valves are set to open at the RCS design pressure(2500 psig) and within the specified tolerance to avoid exceeding the maximum RCSdesign pressure SL, to maintain accident analysis assumptions and to comply withASME Code requirements.
The upper and lower as-left pressure tolerance limits arebased on the +/- 1% tolerance requirements (Ref. 2) for lifting pressures above 1000 psig.The limit protected by this Specification is the reactor coolant pressure boundary(RCPB) SL of 110% of design pressure.
Inoperability of one or both valves could resultin exceeding the SL if a transient were to occur.The consequences of exceeding the ASME pressure limit could include damage to oneor more RCS components, increased
: leakage, or additional stress analysis beingrequired prior to resumption of reactor operation.
The LCO is modified by four Notes. Note 1 states that in MODE 3 and MODE 4 withRCS temperature above 262248 OF, only one pressurizer safety valve is required to beOPERABLE.
In this condition, one pressurizer safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greaterthan the sum of the available heat sources.Note 2 allows entry into MODE 3, and into MODE 4 with RCS temperature
> 2-622480F,with the lift settings potentially outside the limits. This permits testing of the safety valvesat high pressure and temperature near their normal operating range, but only after thevalves have had a preliminary cold setting.
The cold setting gives assurance that thevalves are OPERABLE near their design condition.
Only one valve at a time will beremoved from service for testing.
The 36 hour exception is based on an 18 hour outagetime for each of the two valves. The 18 hour period is derived from operating experience that hot testing can be performed in this timeframe.
Note 3 states that the LCO is not applicable in MODE 3, and in MODE 4 with RCStemperature
> 2622488F during hydrostatic tests in accordance with ASME Boiler andPressure Vessel Code, Section Ill. During hydrostatic tests, the code safeties must begagged to prevent them from relieving at the target test pressure.
RCS pressure iscarefully observed and compensatory measures are in place to provide assurance thatthe pressure is appropriately controlled during the performance of hydrostatic tests.ANO-1 B 3.4.10-2 Amondment No. 215Rev. 37 Pressurizer Safety ValvesB 3.4.10LCO (continued)
Note 4 states that the provisions of LCO 3.0.3 are not applicable in MODE 3, and inMODE 4 with RCS temperature
> 262248 OF. In the event no code safety valve isOPERABLE in this MODE, the Required Actions ensure that the RCS is placed in acondition in which the ERV is capable of relieving any potential LTOP pressuretransient.
The parameter value (262248 OF) does not contain allowances for instrument uncertainty.
Additional allowances for instrument uncertainty are contained in theimplementing procedures.
APPLICABILITY In MODES 1, 2, and 3, and portions of MODE 4 above the LTOP enable temperature, OPERABILITY of pressurizer safety valve(s) is required to ensure adequate relieving capacity is available to keep reactor coolant pressure below 110% of its design valueduring certain accidents.
The LCO is not applicable in MODE 4 with RCS temperature
< 2-62248 &deg;F, in MODE 5,nor in MODE 6 when the reactor vessel head is on because LTOP protection isprovided.
Overpressure protection is not required in MODE 6 with the reactor vesselhead removed.The parameter value (262248 OF) does not contain allowances for instrument uncertainty.
Additional allowances for instrument uncertainty are contained in theimplementing procedures.
ACTIONSA._1With one pressurizer safety valve inoperable in MODES 1 and 2, restoration must takeplace within 15 minutes.
The Completion Time of 15 minutes reflects the importance ofmaintaining the RCS overpressure protection system. An inoperable safety valvecoincident with an RCS overpressure event could challenge the integrity of the RCPB.B.1If the Required Action and associated Completion Time of Condition A are not met, or ifboth pressurizer safety valves are inoperable in MODES 1 and 2, the unit must bebrought to a MODE in which the requirement does not apply. To achieve this status, theunit must be brought to at least MODE 3 within 6 hours. The 6 hours allowed isreasonable, based on operating experience, to reach MODE 3 from full powerconditions in an orderly manner and without challenging unit systems.
The change fromMODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure),
lowers the potential for large pressurizer
: insurges, and thereby removes the need foroverpressure protection by two pressurizer safety valves.ANO-1 B 3.4.10-3 Amendment No. 215Rev. 37 Pressurizer Safety ValvesB 3.4.10ACTIONS (continued)
C.1With the required pressurizer code safety valve inoperable, the RCS overpressure protection capability is significantly reduced and an overpressure event could challenge the integrity of the RCPB. Therefore, the unit must be placed in a condition in which therequirement does not apply. To achieve this status, the unit must be brought to at leastMODE 4 with RCS temperature at or below the LTOP enable temperature within 18hours. The 18 hours allowed is reasonable, based on operating experience, to reach alow temperature within MODE 4 without challenging unit systems.
With RCStemperature at or below 262248 OF, overpressure protection is provided by LTOP.SURVEILLANCE REQUIREMENTS SR 3.4.10.1SRs are specified in the Inservice Testing Program.
Pressurizer safety valves are to betested in accordance with the requirements of the ASME OM Code (Ref. 6), whichprovides the activities and the Frequency necessary to satisfy the SRs. No additional requirements are specified.
The pressurizer safety valve setpoint is + 1%, -3% for OPERABILITY (Ref. 7); however,the valves are reset to +/- 1% during the Surveillance to allow for drift.REFERENCES
: 1. SAR, Section 4.2.4.2. ASME, Boiler and Pressure Vessel Code, Section III, Article 9, Summer 1968.3. SAR, Section 4.3.8.4. SAR, Section 4.3.11.4.
: 5. 10 CFR 50.36.6. ASME, Boiler and Pressure Vessel Code, Section XI.7. ASME OM Code -2001.ANO-1B 3.4.10-4 t No. 215Rev. 37 LTOP SystemB 3.4.11BACKGROUND (continued) accommodate a coolant insurge and prevent a rapid pressure
: increase, allowing theoperator time to stop the increase.
The ERV, with reduced lift setting, or the RCSvent is the overpressure protection device that acts as backup to the operator interminating an increasing pressure event.With HPI deactivated, the ability to provide RCS coolant addition is restricted.
Toallow for coolant addition, the LCO does not require the makeup function to bedeactivated.
Due to the lower pressures associated with the LTOP MODES and theexpected decay heat levels, the makeup function can provide flow through themakeup control valve.ERV Requirements As designed for the LTOP, the ERV is signaled to open if the RCS pressure reachesa limit set in the LTOP actuation circuit.
The LTOP actuation circuit monitors RCSpressure and determines when an overpressure condition is approached.
When themonitored pressure meets or exceeds the setting, the ERV is signaled to open.Maintaining the lowered setpoint ensures the Reference 1 limits will be met in anyevent analyzed for LTOP.RCS Vent Requirements Once the RCS is depressurized, adequate pressure relief capability may be providedby a vent path to the reactor building atmosphere which is capable of relieving theflow of the limiting LTOP transient and maintaining pressure below PIT limits. Therequired vent capacity may be provided by one or more vent paths. Acceptable RCSvent paths include any of the following:
removing a pressurizer safety valve, lockingthe ERV in the open position and disabling its block valve in the open position, orsimilarly establishing a vent by removing a steam generator (SG) primary manway,removing a SG primary hand hole cover, removing all control rod drive top closureassemblies (excluding reactor vessel level probe), or removing a pressurizer manway. The vent path(s) must be above the level of reactor coolant, so as not todrain the RCS when open.APPLICABLE SAFETY ANALYSESSafety analyses (Refs. 4, 5, 6, and 7) demonstrate that the reactor vessel can beadequately protected against overpressurization transients during shutdown.
Thepressure and temperature limits are derived from fracture mechanics analyses.
Transients are then evaluated to determine a required ERV setpoint and other unitconditions that will ensure that the P/T limits are not exceeded.
Fracture mechanics analyses (Lusing the safety rnargins of Reieence
: 8) established the temperature of LTOP Applicability at 2-2248 OF. Above this temperature, thepressurizer safety valves provide the reactor vessel overpressure protection.
Theactual temperature at which the allowable pressure falls below the pressurizer.
ANO-1B 3.4.11-2Amendment No. 215 LTOP SystemB 3.4.11APPLICABLE SAFETY ANALYSES (continued) safety valve setpoint increases as vessel material ductility decreases due to neutronembrittlement.
P/T limits are periodically determined using neutron fluenceprojections and the results of examinations of the reactor vessel material irradiation surveillance specimens.
The Bases for LCO 3.4.3 discuss these examinations.
Forthe current limits, vessel materials are assumed to have a neutron irradiation accumulation equivalent to 3454 effective full power years (EFPYs) of operation.
Each time the P/T limit curves are revised, the LTOP is re-evaluated to ensure thatits functional requirements can still be met. The ERV setpoint is revised ifnecessary.
Transients that are capable of overpressurizing the RCS at low temperature result ineither excessive mass input or excessive heat input. Such transients include:
HPIactuation, CFT discharge, energization of the pressurizer
: heaters, failing the makeupcontrol valve open, loss of decay heat removal, starting a reactor coolant pump(RCP) with a large temperature mismatch between the primary and secondary coolant systems, and addition of nitrogen to the pressurizer.
Without controls, HPIactuation and CFT discharge would be transients that result in exceeding P/T limitswithin the 10 minute period in which time no operator action can be assumed to takeplace. For the remaining events, operator action after that time precludes overpressurization.
This specification prevents exceeding the P/T limits by: 1) limiting the capability forrapid mass input to the RCS; and 2) ensuring that adequate vent capability exists toaccommodate inadvertent mass or energy addition to the RCS. Pressurizer level isalso limited to ensure that increasing pressure during a transient will be slow enoughto preclude exceeding pressure limits within the 10 minutes assumed to be requiredfor operator action to mitigate the transient.
Mass input into the system is limited bydisabling HPI (with specific exceptions) and by deactivating pressurized CFTdischarge isolation valves in the closed position with their power breakers open (withspecific exceptions).
The analyses demonstrate that HPI transients involving oneHPI pump can be accommodated by the ERV without exceeding the maximumallowable pressure.
The ERV setpoint is determined by modeling LTOP performance assuming the mostlimiting LTOP transient of a makeup control valve failing open. Pressure overshoot beyond the setpoint resulting from signal processing and valve stroke times isconsidered.
The resulting ERV setpoint ensures the fReference 1 limits will not beexceeded.
Vent capability is required to ensure that the maximum allowable pressure is notexceeded in the event of full opening of the makeup control valve while one makeuppump is running.
Acceptable vent paths have adequate capacity at a systempressure of 100 psig which is less than the maximum RCS pressure on the P/T limitcurve in LCO 3.4.3.ANO-1B 3.4.11-3Amendment No. 215 LTOP SystemB 3.4.11APPLICABLE SAFETY ANALYSES (continued)
The ERV is an active component.
Therefore, its failure represents the worst casesingle active failure of LTOP features.
The other vent paths are passive and notsubject to active failure.The LTOP satisfies Criterion 2 of 10 CFR 50.36 (Ref. 9).LCOThe LCO requires an LTOP system OPERABLE with a limited coolant inputcapability and a pressure relief capability.
To limit coolant input, the LCO requiresthe HPI deactivated, and the CFT discharge isolation valves closed and deactivated.
For pressure relief, the LCO requires the pressurizer coolant level to be below a levelwhich represents a water solid condition, and the ERV OPERABLE with a lowered liftsetting or the RCS depressurized and a vent established.
HPI deactivation requires that the HPI system be incapable of causing a significant increase in RCS pressure (motor operated valves de-activated closed, HPI pumpbreakers racked down, or other configurations that prevent inadvertent HPIactuation).
CFT isolation requires the CFT discharge valves to be closed and thecircuit breakers for the motor operators to be opened.The HPI deactivation and CFT isolation requirements are modified by five Notes.Note 1 indicates that the requirements are not applicable during ASME Section XItesting.
This exception provides for required testing during these shutdownconditions rather than at power when the HPI and CFTs are required to beOPERABLE for the ECCS function.
Note 2 indicates that the requirements are notapplicable for the HPI deactivation during fill and vent of the RCS. The HPI pumpsare used for this normal makeup function and must be available.
Specific procedural controls are provided to prevent overpressurization during this activity.
Note 3indicates that the requirements are not applicable for the HPI deactivation duringemergency RCS makeup. This exception is necessary to enhance the responsecapability to a loss of decay heat removal event without violating the TS (Ref. 10).Note 4 indicates that the requirements are not applicable for the HPI deactivation during valve maintenance.
This exception allows maintenance to be performed during these shutdown conditions rather than at power when the HPI is required tobe OPERABLE for the ECCS function.
Note 5 states that CFT isolation is onlyrequired when CFT pressure is more than or equal to the maximum RCS pressurefor the existing RCS temperature, as allowed in LCO 3.4.3. This is acceptable sincethe CFT can not be the source of an overpressurization event when its pressure isless than the allowable RCS pressure.
The pressurizer is considered to represent a water solid condition when coolant levelis > 445180 inches, -hen RCS
,.+.4.50 inches, when RGpressure is 100 psig. Although a vapor space still exists with pressurizer levelabove these values, from an analytical point of view, the unit is considered to bewater solid. TheseThis parameter valu~esdoes not contain allowances for instrument error.ANO-1 B 3.4.11-4 Amendment No. 215Rev. 2 LTOP SystemB 3.4.11LCO (continued)
The pressurizer level requirements are modified by two Notes. Note 1 indicates thatthe requirements are not applicable during operation allowed by the Emergency Operating Procedures (EOPs). This exception provides for use of the "feed andbleed" process when necessary as determined by the EOPs. Note 2 indicates that therequirements are not applicable during RCS hydrotesting.
Specific procedural controls are provided to prevent overpressurization during this activity.
OPERABLE pressure relief capability may be provided by an OPERABLE ERV, or bydepressurizing the RCS and providing an alternate RCS vent path. For the ERV to beconsidered
: OPERABLE, its block valve must be open, its lift setpoint must be set at _<460563.8 psig, testing must have proven its ability to open at that setpoint, andmotive power must be available to the ERV and its control circuits.
With the RCSdepressurized, acceptable alternate vent paths include removing a pressurizer safetyvalve, locking the ERV in the open position and disabling its block valve in the openposition, removing a SG primary manway, removing a SG primary hand hole cover,removing all control rod drive top closure assemblies (excluding reactor vessel levelprobe), or removing a pressurizer manway.APPLICABILITY This LCO is applicable in MODE 4 with RCS temperature
< 2-62248 OF, in MODE 5,and in MODE 6 when the reactor vessel head is on. The Applicability temperature of26.2248 OF is established by fracture mechanics analyses.
The pressurizer safetyvalves provide overpressure protection to meet LCO 3.4.3 P/T limits above 2-62248 OF.With the vessel head off, overpressurization is not possible.
LCO 3.4.3 provides the operational P/T limits for all MODES. LCO 3.4.10,"Pressurizer Safety Valves,"
requires the pressurizer safety valves OPERABLE toprovide overpressure protection during MODES 1, 2, and 3, and MODE 4 above2-92248 OF.The parameter value (2-2-248 OF) does not contain allowances for instrument uncertainty.
Additional allowances for instrument uncertainty are contained in theimplementing procedures.
ACTIONSA.1, B.1, and B.2With the pressurizer level not within its required limits, the time for operator action in apressure increasing event is reduced.
The postulated event most affected in theLTOP MODES is failure of the makeup control valve, which fills the pressurizer relatively rapidly.
Restoration is required within 1 hour.ANO-1 B 3.4.11-5 Ameondment No. 215Rev. 37 LTOP SystemB 3.4.11SURVEILLANCE REQUIREMENTS SR 3.4.11.1Verification of the pressurizer level at <1 105180 inches wherr RCS pressure is-.0psig, by observing controlroom or other indications ensures that the unit is not in a water solid condition and thata cushion of sufficient size is available to reduce the rate of pressure increase frompotential transients (Ref. 311). This parameter does not contain allowances forinstrument error.The 30 minute Surveillance Frequency during heatup and cooldown must beperformed for the LCO Applicability period when temperature changes can causepressurizer level variations.
This Frequency may be discontinued when theseevolutions are complete, as defined in unit procedures.
Thereafter, the Surveillance isrequired at 12 hour intervals.
These Frequencies are shown by operating practice sufficient to regularly assessindications of potential degradation and verify operation within the safety analysis.
SR 3.4.11.2 and SR 3.4.11.3Verifications must be performed that the HPI is deactivated, and each pressurized CFT is isolated.
These Surveillances ensure the minimum coolant input capability willnot create an RCS overpressure condition to challenge the LTOP. The Surveillances are required at 12 hour intervals.
The 12 hour intervals are shown by operating practice to be sufficient to assesscoolant input capability and verify operation within the safety analysis.
SR 3.4.11.4OPERABLE pressure relief capability must be provided to prevent overpressurization due to inadvertent full makeup system operation.
Such a vent keeps the pressurefrom full makeup flow within the LCO limit. OPERABLE pressure relief capability maybe provided by an OPERABLE ERV, or by depressurizing the RCS and providing analternate RCS vent path.For the ERV to be considered
: OPERABLE, its block valve must be open, its liftsetpoint must be set at < 460563.8 psig, testing must have proven its ability to open atthat setpoint, and motive power must be available to the two valves and their controlcircuits.
The parameter value of 464(563.8 psig does not contain allowances forinstrument uncertainty.
Additional allowances for instrument uncertainty are contained in the implementing procedures.
ANO-1 B 3.4.11-6 Amendmcnt Ne. 215Rev. 37 LTOP SystemB 3.4.11SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.11.4 (continued)
With the RCS depressurized, acceptable alternate vent paths include:
a) removing apressurizer safety valve; b) locking the ERV in the open position and disabling itsblock valve in the open position; c) removing a SG primary manway; c) removing a SGprimary hand hole cover; d) removing all control rod drive top closure assemblies (excluding reactor vessel level probe); and e) removing a pressurizer manway.For a vent path not locked open, the Frequency is every 12 hours. For a locked openvent path, the required Frequency is every 31 days.The Frequency intervals are considered adequate based on operating practice todetermine adequacy of pressure relief capability and verify operation within the safetyanalysis.
SR 3.4.11.5The performance of a CHANNEL CALIBRATION is required every 18 months. TheCHANNEL CALIBRATION for the LTOP ERV opening logic, including the ERVsetpoint, ensures that the ERV will be actuated at the appropriate RCS pressure byverifying the accuracy of the instrument string. The calibration can only be performed in shutdown.
The 18 month Frequency considers a typical refueling cycle and industry acceptedpractice.
REFERENCES
: 1. 10 CFR 50, Appendix G, Fracture Touqhness Requirements.
: 2. Generic Letter 88-11, Pressurizer Surqe Line Thermal Stratification.
: 3. ANO-1 LTOP Safety Evaluation Report (1CNA058302) dated May 5, 1983.4. Response to NRC Request for Additional Information (1CANl 17608) datedNovember 15, 1976.5. Response to NRC Request for Additional Information (lCAN127602) datedDecember 3, 1976.6. Response to NRC Request for Additional Information (1CAN037716) dated March24, 1977.ANO-1 B 3.4.11-7 Amcndmcnt No. 215Rev. 37 LTOP SystemB 3.4.11REFERENCES (continued) 7.J A Aand-Ope-ating
: Licinse, GnereA-h3-INAG39703 d-cated Marfoh 14,1-9W-Deleted.
: 8. ANO-l.. Request fo.r-ExemptIOn.-44.GAN1 I960q),-
dated -NevembeF-2-6,.
1-996j--and
-x 4r-en-Re u f--, F R 50,940,-1GNA 3W 4-9-97Deleted.
: 9. 10 CFR 50.36, Technical specifications.
: 10. ANO-1 License Amendment Request (lCAN059008),
dated May 22, 1990, andOperating License Amendment 138, (1 CNA1 19002) dated November 1, 1990.11. CALC-14-E-0100-13, ANO-1 Pressurizer Model for LTOP Design BasesTransient (54 EFPY).ANO-1B 3.4.11-8Amcndmont No. 215Rev. 37 RCS P/T Limits3.4.3Attachment 3 toI CAN081403 Revised (clean) Technical Specification Pages RCS P/T Limits3.4.3FIGURE 3.4.3-1RCS Heatup Limitations to 54 EFPYC/)0-I-00C/)C/)CLC/)024002200200018001600140012001000800600400200050 100 150200 250 300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
('F)Notes:1. Curves are not adjusted for instrument error and shall not be used for operation.
: 2. When DHR is in operation with no RCPs operating, the DHR system return temperature shall be used.3. RCP Operating Restrictions:
RCS TEMPT > 250 OF250&deg;F > T > 100 OFT < 100 OFRCS TEMP60 OF < T < 84 OFT > 84 OFRCP RESTRICTIONS None<3No RCPs operating
: 4. Allowable Heatup Rates:H/U RATES 15 &deg;F/hrAs allowed by applicable curveANO-13.4.3-5Amendment No. 215 RCS P/T Limits3.4.3FIGURE 3.4.3-2RCS Cooldown Limits to 54 EFPY6'(n0~C/)(n-IX-01U)2400220020001800160014001200100080060040020000 50 100 150 200 250 300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
('F)Notes:1. This curve is not adjusted for instrument error and shall not be used for operation.
: 2. A maximum step temperature change of 25 OF is allowable when securing all RCPs with theDHR system in operation.
This change is defined as the RCS temperature prior to securingall the RCPs minus the DHR return temperature after the RCPs are secured.
When DHR isin operation with no RCPs operating, the DHR system return temperature shall be used.3. RCP Operating Restrictions:
RCS TEMPT > 250 OF250_&deg;F > T > 100 OFT < 100 OF4. Allowable Cooldown Rates:RCS TEMPT > 280 OF280 0F > T > 150 OFT < 150 OFRCP RESTRICTIONS None<3No RCPs operating STEP CHANGE< 50 &deg;F in any 1/2 hr< 25 &deg;F in any 1/2 hr< 25 OF in any 1 hrAmendment No. 215C/D RATE100 &deg;F/hr50 &deg;F/hr25 &deg;F/hrANO-13.4.3-6 RCS P/T Limits3.4.3FIGURE 3.4.3-3RCS Inservice Hydrostatic Test H/U & C/D Limits to 54 EFPY0-0~i-UJ24002200200018001600140012001000800600400200000 50 100 150 200 250 300 350 400 450 500 550 600RCS COLD LEG TEMPERATURE
(-F)Notes:1. This curve is not adjusted for instrument error and shall not be used for operation.
: 2. All Notes on Figure 3.4.3-1 are applicable for heatups.
This curve is based on a heatuprate of < 90&deg;F/HR.3. All Notes on Figure 3.4.3-2 are applicable for cooldowns.
ANO-13.4.3-7Amendment No. 215 Pressurizer 3.4.93.4 REACTOR COOLANT SYSTEM (RCS)3.4.9 Pressurizer LCO 3.4.9The pressurizer shall be OPERABLE with:a. Pressurizer water level < 320 inches; andb. A minimum of 126 kW of Engineered Safeguards (ES) bus poweredpressurizer heaters OPERABLE.
---------------------------
NOTE ----------------------
OPERABILITY requirements on pressurizer heaters do not apply inMODE 4.APPLICABILITY:
MODES 1, 2, and 3,MODE 4 with RCS temperature
> 248 OF.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Pressurizer water level not A.1 Restore level to within 1 hourwithin limits, limits.B. Required Action and B.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition A not ANDmet.B.2 Be in MODE 4 with RCS 24 hourstemperature
< 248 OF.C. Capacity of ES bus C.1 Restore pressurizer heater 72 hourspowered pressurizer capacity.
heaters less than limit.D. Required Action and D.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition C not ANDmet.D.2 Be in MODE 4, 12 hoursANO-13.4.9-1Amendment No. 215 Pressurizer Safety Valves3.4.103.4 REACTOR COOLANT SYSTEM (RCS)3.4.10 Pressurizer Safety ValvesLCO 3.4.10Two pressurizer safety valves shall be OPERABLE.
,1%11 l1. Only one pressurizer safety valve is required to be OPERABLE inMODE 3, and in MODE 4 with RCS temperature
> 248 OF.2. The lift settings are not required to be within limits for entry intoMODE 3 or the applicable portions of MODE 4 for the purpose ofsetting the pressurizer safety valves under ambient (hot) conditions.
This exception is allowed for 36 hours following entry into MODE 3provided a preliminary cold setting was made prior to heatup.3. Not applicable in MODE 3, and in MODE 4 with RCS temperature
> 248 OF during hydrostatic tests in accordance with ASME Boiler andPressure Vessel Code, Section II1.4. The provisions of LCO 3.0.3 are not applicable in MODE 3, and inMODE 4 with RCS temperature
> 248 OF.APPLICABILITY:
MODES 1, 2, and 3,MODE 4 with RCS temperature
> 248 OF.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One pressurizer safety A.1 Restore valve to OPERABLE 15 minutesvalve inoperable in status.MODES 1 or 2.B. Required Action and B.1 Be in MODE 3. 6 hoursassociated Completion Time of Condition A notmet.ORTwo pressurizer safetyvalves inoperable inMODES 1 or 2.ANO-13.4.10-1Amendment No. 215 Pressurizer Safety Valves3.4.10CONDITION REQUIRED ACTION COMPLETION TIMEC. Required pressurizer C.1 Be in MODE 4 with RCS 18 hourssafety valve inoperable in temperature
< 248 OF.MODE 3 or MODE 4 withRCS temperature
> 248 OF.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each required pressurizer safety valve is In accordance withOPERABLE in accordance with the Inservice Testing the Inservice Program.
Following
: testing, as-left lift settings shall Testing Programbe within +/- 1%.IIANO-13.4.10-2Amendment No. 215 LTOP System3.4.113.4 REACTOR COOLANT SYSTEM (RCS)3.4.11 Low Temperature Overpressure Protection (LTOP) SystemLCO 3.4.11 An LTOP System shall be OPERABLE with high pressure injection (HPI)deactivated and the core flood tanks (CFTs) isolated and:-------------------------
NOTES -----------------------
: 1. HPI deactivation and CFT isolation not applicable during ASMESection XI testing.2. HPI deactivation not applicable during fill and vent of the RCS.3. HPI deactivation not applicable during emergency RCS makeup.4. HPI deactivation not applicable during valve maintenance.
: 5. CFT isolation is only required when CFT pressure is greater than orequal to the maximum RCS pressure for the existing RCS temperature allowed by the pressure and temperature curves provided inLCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits."a. Pressurizer level such that the unit is not in a water solid condition andan OPERABLE electromatic relief valve (ERV) with a setpoint of< 563.8 psig; or--- -------------------------
NOTES ---------------------
: 1. Pressurizer level not applicable as allowed by Emergency Operating Procedures.
: 2. Pressurizer level not applicable during system hydrotest.
: b. The RCS depressurized and the RCS open.APPLICABILITY:
MODE 4 with RCS temperature
< 248 OF,MODE 5,MODE 6 when the reactor vessel head is on.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Pressurizer level not A.1 Restore pressurizer level to 1 hourwithin required limits, within required limits.ANO-13.4.11-1Amendment No. 215}}

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