RS-14-338, Byron/Braidwood Nuclear Stations, Updated Final Safety Analysis Report (Ufsar), Revision 15, Chapter 5 Reactor Coolant System and Connected Systems
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B/B-UFSAR 5.3-1 5.3 REACTOR VESSEL 5.3.1 Reactor Vessel Materials
This section is for purposes of reac tor pressure vessel fabrication.
5.3.1.1 Material Specifications Material specifications are in accordance wi th the ASME Code requirements and are giv en in Subsection 5.2.3.
5.3.1.2 Special Processes Used for Manufactu ring and Fabrication
- a. The vessel is Seismic Ca tegory I and Q uality Group A. Design and f abrication of the reactor vessel is carried out in stric t accordance with ASME Code,Section III, Class 1 r equirements. The head flanges and nozzles are manufa ctured as forgings.
The cylindrical portion of the vessel is made up of several forged shells. The hemispherical heads are made from dished plates.
The reactor vessel parts are joined by weldin g, using the sin gle or multiple wire submerged arc.
- b. The use of severely sens itized steel as a pressure boundary material has been prohibited and has been eliminated by either a s elect choice of material or by programming t he method of assembly.
- c. The control rod drive mechanism head adaptor threads and surfaces of the guide studs are chrome plated to prevent possible g alling of the mated parts.
- d. At all locatio ns in the reactor vess el where stainless steel and Inconel are joined , the final joining beads are Inconel weld metal in order to prevent cracking.
- e. The location of full penetration weld seams in the upper closure head and vessel bottom head are restricted to areas th at permit accessibility during inservice inspection.
- f. The stainless steel clad surfaces are sampled to ensure that composition and de lta ferrite requirements are met. g. The procedure qualificat ion for cladding low alloy steel (SA508 Class 2) re quires a special evaluation to ensure freedom fr om underclad cracking.
B/B-UFSAR 5.3-2 5.3.1.3 Special Methods for Nondestructi ve Examination The examination requirem ents detailed in the following are in addition to the examination requ irements of Section III of the ASME Code.
The reactor vessel nondestructive examination (NDE) program is given in Table 5.3-1.
5.3.1.3.1 Ultrason ic Examination
- a. In addition to the d esign code straight beam ultrasonic test, angle beam inspecti on of 100% of plate material is performed during fabrication to detect discontinuiti es that may be undetected by longitudinal wave examination.
- b. In addition to ASME Section III nondestructive examination, all full pe netration welds and heat affected zones in the reactor vessel are ultrasonically examined during fabricati on. This test is performed upon comp letion of the welding and intermediate heat treatm ent but prior to the final postweld heat treatment.
- c. The reactor vessel is examined after hydrostatic testing for information.
5.3.1.3.2 Penetrant Examinations The partial penetration welds for the co ntrol rod drive mechanism head adapt ors and the bottom i nstrumentation tubes are inspected by dye p enetrant after the roo t pass in addition to code requirements.
Core support block atta chment welds were inspected by dye penetrant after first layer of weld metal and after each 1/2 i nch of weld metal. All clad surfaces and other vessel and head internal surfaces were i nspected by dye penetrant after the hydrostatic test.
5.3.1.3.3 Magnetic Particle Examination
All magnetic particle examinations of materi als and welds were performed in accordance with the following:
- a. Prior to the final pos tweld heat treatment - by the prod, coil, or direc t contact method.
- b. After the final postweld heat treatment - by the yoke method.
The following surfaces and welds were ex amined by magnetic particle methods.
B/B-UFSAR 5.3-3 REVISION 15 - DECEMBER 2014 Surface Examinations
- a. Magnetic particle exam ination of all exterior vessel and head surf aces after the h ydrostatic test.
- b. Magnetic particle exam ination of all exterior closure stud surfaces and all nut surfaces after final machining or rolli ng. Continuous circular and longitudinal magne tization were used.
- c. Magnetic particle examin ation of all inside diameter surfaces of carbon and low alloy steel p roducts that have their prope rties enhanced by ac celerated cooling.
This inspection is perfo rmed after forming and machining (if re quired) and prior to cladding.
Weld Examination Magnetic particle examination of the weld me tal buildup for vessel welds att aching the closu re head lifting lugs to the reactor vessel after the first layer and each 1/2 inch of weld metal is deposited. All pressure boundary welds were examined after back chipping or b ack grinding operations.
5.3.1.4 Special Controls for Ferritic and Au stenitic Stainless Steels Welding of ferritic st eels and austenitic stainless steels is discussed in Subsect ion 5.2.3. Subsecti on 5.2.3 includes discussions which indicate t he degree of acceptance with guidelines for c ontrol of ferrite content in sta inless steel metal welds, use of se nsitized stainless ste el, electroslag weld properties, stainless st eel weld cladding of low-alloy steel components and welde r qualification for areas of limited accessibility. Appendix A discusses the deg ree of conformance with regulat ory guides.
5.3.1.5 Fracture Toughness
Assurance of adequat e fracture toughness of ferritic materials in the reactor coola nt pressure boundary (ASME Section III Class 1 Components) is provided by com pliance with the requirements for fracture toughn ess testing included in NB-2300 to Section III of th e ASME Boiler and Pr essure Vessel Code, and Appendix G of 10 CFR 50.
The initial Charpy V-n otch minimum upper shelf fracture energy levels for the r eactor vessel beltli ne (including welds) shall be 75 foot-pounds as re quired by Appendix G of 10 CFR 50.
Materials having a section thick ness greater than 10 inches with an upper shelf of less than 75 fo ot-pounds shall be evaluated with regar d to effects of chemistry (especially copper content), initial upper shelf energy and influence to B/B-UFSAR 5.3-4 REVISION 15 - DECEMBER 2014 ensure that a 50 foot-pound shelf energy as required by Appendix G of 10 CFR 50 is mai ntained throughout the life of the vessel. The specimens sha ll be oriented as required by NB-2300 of Secti on III of the ASME Boi ler and Pressure Vessel Code. The reactor v essel material prope rties for units of the Byron/Braidwood Stations are given in Sectio n 5 of the PTLR.
5.3.1.5.1 Pressurized Ther mal Shock Evaluation Fracture toughness requi rements for protection of reactor vessels against pressurized thermal shock events are given in 10 CFR 50.61. Reference 9 provides the initial assessment of Pressurized Thermal Shock. Su bsequently, evaluations which include surveillance capsule data ha ve been performed in accordance with the re quirements of 10 CFR 5 0.61 for the reactor vessels at Byron / Bra idwood Units 1 and 2.
The evaluations are provided in References 1 and 2 and t he evaluation results are summarized in References 3 and 4 and Tables 5.
3-7 through 5.3-10.
5.3.1.6 Material Surveillance
In the surveillance pr ogram, the evaluation of the radiation damage is based on pre irradiation testing of Charpy V-notch and tensile specimens an d postirradiation testing of Charpy V-notch, tensile and 1/2 thick ness (T) compa ct tension (CT) fracture mechanics test specimens. The program is directed toward evaluation of the effect of radia tion on the fracture toughness of reactor vessel steels based on the transition temperature approach and the fra cture mechanics appr oach. The program conforms with ASTM-E-185 "Recomm ended Practice for Surveillance Tests f or Nuclear Reactor Vesse ls," and 10 CFR 50, Appendix H.
Detailed information on the reactor vessel m aterial surveillance program is provided in Westinghouse repo rts WCAP-9517 for Byron Unit 1, WCAP-10398 for Byr on Unit 2, and WCAP-9807 for Braidwood Unit 1 and W CAP-11188 for Braidwood 2.
The reactor vessel s urveillance program uses six specimen capsules. The capsu les are located in guide baskets welded to the outside of the neu tron shield pads a nd are positioned directly opposite the center portion of the core.
The capsules can be removed when the vessel h ead is removed and can be replaced when the internals are removed. The six capsules contain reactor vessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal working direction of the limiting base material located in the core region of t he reactor vessel and associated weld metal and weld heat-affected zone metal. The 6 capsules contain 54 tensile specimens, 3 60 Charpy V-notch sp ecimens (which include weld metal and weld heat-affected zone m aterial), and 72 CT specimens. Archive material sufficient for two additional capsules will be retained.
B/B-UFSAR 5.3-5 REVISION 1 - DECEMBER 1989 Dosimeters, including Ni, Cu, Fe, Co-Al, Cd shielded Co-Al, Cd shielded Np-237 and Cd s hielded U-238, are p laced in filler blocks drilled to contain them. The dosimet ers permit evaluation of the flux seen by the specimens and the vessel wall. In addition, thermal monitors made of low melting p oint alloys are included to monitor the maximum temperature of the specimens.
The specimens are enclosed in a tight-fitting stainless steel sheath to prevent corr osion and ensure good th ermal conductivity.
The complete capsule is helium leak tested.
Each of the six caps ules contains the following specimens:
Number ofNumber of Number ofMaterial CharpysTensiles CTs Limiting base material*1534
Limiting base material**1534 Weld metal*** 1534 Heat affected zone 15
- Specimens oriented in the major working direction.
- Specimens oriented normal to the major w orking direction.
- Weld metal to be sel ected per ASTM E185.
The following dosimeters and thermal monitor s are included in each of the six capsules:
Dosimeters
Iron Copper Nickel Cobalt-Aluminum (0.15% Co)
Cobalt-Aluminum (Cadmium shielded)
Np-237 (Cadmium shielded)
Thermal Monitors 97.5% Pb, 2.5% Ag (579
°F melting point).
97.5% Pb, 1.75% Ag, 0.75% Sn (590
°F melting point).
B/B-UFSAR 5.3-6 REVISION 12 - DECEMBER 2008 The fast neutron exposure of the specimens occurs at a faster rate than that exper ienced by the ve ssel wall, with the specimens being located between the core and the vessel. Since these specimens experi ence accelerated e xposure and are actual samples from the mater ials used in the v essel, the t ransition temperature shift measur ements are rep resentative of the vessel at a later time in life.
Data from CT fract ure toughness specimens are expected to provide additional information for use in determining allow able stresses for ir radiated material.
Correlations between the calcula tions and the measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Subsect ion 5.3.1.6.1. They have indicated good agreement. The antici pated degree to which the specimens will perturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure will be made by use of data on all capsules withdrawn. For the sc hedule for removal of the capsules for postirradiation testing which follows that of 10 CFR 50 Appendix H, refer to Tab le 4.1 of the PTLR.
5.3.1.6.1 Measurement of Integra ted Fast Neutr on (E>1.0 MeV)
Flux at the Irra diation Samples The use of passive neutr on sensors such as t hose included in the internal surveillance ca psule dosimetry sets dose not yield a direct measure of the en ergy dependent neutron flux level at the measurement location.
Rather, the a ctivation or fission process is a measure of the in tegrated effect that t he time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accur ate assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the senso rs may be developed fr om the measurements only if the sensor c haracteristics and t he parameters of the irradiation are well known. In partic ular, the following variables are of interest:
- 1. The measured specific ac tivity of each sensor
- 2. The physical characteris tics of each sensor
- 3. The operating history of the reactor
- 4. The energy response of each sensor
- 5. The neutron energy spectrum at the sensor location In this section the pr ocedures used to deter mine sensor specific activities, to develop react ion rates for indivi dual sensors from the measured specific ac tivities and the operating history of the reactor, and to derive key fast neutron expo sure parameters from the measured reaction rates are described.
B/B-UFSAR 5.3-7 REVISION 9 - DECEMBER 2002 5.3.1.6.1.1 DETERM INATION OF SENSOR REACTION RATES The specific activit y of each of the radiometric sensors is determined using established ASTM procedures.
Following sample preparation and weighing, the specific activity of e ach sensor is determined by means of a hig h purity ger manium gamma spectrometer. In the case of the survei llance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the i ndividual wires; or, as in the case of U-238 and Np-237 fission monitor s, by direct count ing preceded by dissolution and chemical separation of cesium from the sensor.
The irradiation history of the reactor o ver its operating lifetime is determined f rom plant power genera tion records. In particular, operating da ta are extracted on a monthly basis from reactor startup to the end of the capsule irradiation period.
For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly hi stogram describing rea ctor operation has proven to be an adeq uate representation for use in radioactive decay corrections for the reacti ons of interest in the exposure evaluations.
Having the measured spec ific activities, the operating history of the reactor, and the physical characteristic s of the sensors, reaction rates referenced to full power operation are determined from the following equation:
where: A = measured specific activity (dps/gm)
R = reaction rate averaged over the irradiation period and referenced to operat ion at a core power level of P ref (rps/nucleus).
N o = number of target element atoms per gram of sensor.
F = weight fraction of t he target isotope in the sensor material.
Y = number of product at oms produced per reaction.
P j = average core p ower level during irradation period j (MW). P ref = maximum or reference core power level of the reactor (MW).
[]d j t t j ref j j o e e C P P Y F N A R=1 B/B-UFSAR 5.3-7a REVISION 15 - DECEMBER 2014 C j = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period. = decay constant of th e product isotope (sec
-1).
t j = length of irradi ation period j (sec).
t d = decay time following i rradiation period j (sec).
and the summation is c arried out over the to tal number of monthly intervals comprising the total irradiation period.
In the above equ ation, the ratio P j/Pref accounts for month by month variation of power level within a give n fuel cycle. The ratio C j is calculated for e ach fuel cycle a nd accounts for the change in sensor reaction rates caused by variat ions in flux level due to changes in core power spati al distributions from fuel cycle to fuel cycle. S ince the neutron flux at the measurement locations within the surveil lance capsules is dominated by neutrons produced in the periph eral fuel assemblies, the change in the relative power in these assemblies from fuel cycle to fuel cycle can have a significant impact on the activation of neutron sensors. For a single-c ycle irradiation, C j = 1.0. However, for multiple-cycle irradiat ions, particularly those employing low le akage fuel management, the additional C j correction must be utilized in order to provide accurate determinations of the decay corrected reacti on rates for the dosimeter sets contained in the surveillance capsules.
5.3.1.6.1.2 Corrections to React ion Rate Data Prior to using the measu red reaction rates in the least squares adjustment procedure d iscussed in Section 5.
4.3.6.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium is otopes over the course of the irradiation.
In addition to the c orrections made for the presence of U-235 in the U-238 fission sensors, corre ctions are also made to both the U-238 and Np-237 sensor reaction rat es to account for gamma ray induced fission reacti ons occurring over the course of the irradiation.
5.3.1.6.1.3 Least Square s Adjustment Procedure Least squares adjustment methods provide the capability of combining the measurement da ta with the neutron transport calculation resulting in a Best Estimate neu tron energy spectrum with associated uncert ainties. Best Estimat e for key exposure parameters such as (E > 1.0 eV) or dpa/s along with their uncertainties are th en easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results red uces the uncertainty in the calculated spectrum and acts to remove bi ases that may be present in the analytical technique.
B/B-UFSAR 5.3-7b REVISION 15 - DECEMBER 2014 In general, the least squares methods, as ap plied to pressure vessel fluence evaluatio ns, act to reconcile the measured sensor reaction rate data, dosimetry re action cross-sec tions, and the calculated neutron energy spec trum within th eir respective uncertainties. For example, relates a set of measured reaction rates, R i , to a single neutron spectrum g , through the mult igroup dosimeter cross-section, ig , each with an uncertainty . The use of least squares adjustm ent methods in L WR dosimetry evaluations is not new. The American So ciety for Testing and Materials (ASTM) has a ddressed the use of ad justment codes in ASTM Standard E944, "A pplication of Neutron Sp ectrum Adjustment Methods in Reactor Surve illance" and many indu stry workshops have been held to discuss the vario us applications.
For example, the ASTM-EURATOM Symposia on Reactor Dosimetry holds workshops on neutron spectrum unfolding and adjustment tech niques at each of its bi-annual conferences.
Th primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the me asurement. The analyt ical method alone may be deficient because it inherently con tains uncertainty due to the input assumptions to the cal culation. Typically these assumptions include para meters such as the t emperature of the water in the peripheral fuel assemblies, by-pass region, and downcomer regions, component d imensions, and p eripheral core source. Industry cons ensus indicates th at the use of calculation alone results in overall uncerta inties in the neutron exposure parameters in the ra nge of 15-20% (1). By combining the calculated resu lts with available measurements, the uncertainties associated with the key neutron exposure parameters can be re duced. Specifically ASTM Standard E 944 states; "The algorithims of the adjustment cod es tend to decrease the variances of the a djusted data compared to the corresponding input values. T he least squares adjustment codes yield estimates for the output data with minimum variances, that is, the "best estimates". This is the pri mary reason for using these adjustment procedures". ASTM E 944 provides a comprehensive listing of available adjustment codes.
The FERRET least squares adjus tment code (Reference 5) was initially developed at the Hanford Eng ineering Development Laboratory (HEDL) and has ha d extensive use in both the Liquid Metal Fast Breeder (LM FBR) program and t he NRC Sponsored Light Water Reactor Dosimetry Improvement Program (L WR-PV-SDIP). As a result of participation in several cooperati ve efforts associated with the LWR-PV-SDIP, the FE RRET approach was adopted by Westinghouse in
()()g i g ig ig g R i R++/-=+
B/B-UFSAR 5.3-7c REVISION 15 - DECEMBER 2014 the mid 1980's as the preferred approach for the evaluation of LWR surveillance dosimetry. The least squares m ethodology was judged superior to the previously employ ed spectrum averaged cross-section approach that is totally depen dent on the accuracy of the shape of the calculated neutron spectrum at the measurement locations.
The FERRET code is e mployed to combine the results of plant specific neutron transport cal culations and multiple foil reaction rate measurements to determ ine best estimat e values of exposure parameters ( (E > 1.0 MeV) and dpa) along with associated uncertainties at the measurement locations.
The application of the least squares met hodology requires the following input:
- 1. The calculated neutr on energy spectrum and associated uncertainties at the measurement location.
- 2. The measured reactio n rate and associa ted uncertainty for each sensor contained in the multiple foil set.
- 3. The energy dependent dos imetry reaction cross-sections and associated uncerta inties for each sensor contained in the multiple foil sensor set.
For a given application, the c alculated neutron spectrum is obtained from the results of p lant specific ne utron transport calculations applicable to the irradiation p eriod experienced by the dosimetry sensor set. T his calculation is performed using the benchmarked transport calcul ational methodol ogy described in Section 5.3.1.6.2. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the s pecific irradiation hist ory of the sensor set to perform the radio active decay c orrections. The dosimetry reaction cross-sections and uncertainties ar e obtained from the SNLRML dosimetry cross-secti on library (Reference 6). The SNLRML library is an evalua ted dosimetry reac tion cross-section compilation recommended for use in LWR e valuations by ASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 7 06 (IIB)". There are no additional data or data libraries built into the FERRET code system.
All of the required input is supplied externally at the time of the analysis.
The uncertainties asso ciated with the me asured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares proce dure in the form of variances and covariances. The as signment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.
B/B-UFSAR 5.3-8 REVISION 15 - DECEMBER 2014 5.3.1.6.2 Calculation of Integrated Fast Neutron (E
> 1.0 MeV)
Flux at the Irra diation Samples A generalized set of guideli nes for performing fast neutron exposure calculations within t he reactor configuration, and procedures for a nalyzing measured irradi ation sample data that can be correlated to t hese calculations, has been promulgated by the Nuclear Regulatory Commission (NRC) in R egulatory Guide 1.190, "Calculational and Dosimetry Meth ods for Determining Pressure Vessel Neutron Fluence" [Reference 7].
Since different calculational models e xist and are conti nuously evolving along with the associated model inputs, e.g., cross-se ction data, it is worthwhile summarizing the key model s, inputs, and procedures that the NRC staff finds accep table for use in d etermining fast neutron exposures within the rea ctor geometry.
This material is highlighted below.
Calculation and Dosimetry Measurement Procedures The selection of a particular ge ometric model, the corresponding input data, and the over all methodology used to determine fast neutron exposures within the rea ctor geometry are based on the needs for accurately determining a solution to the problem that must be solved and the date/resources that are currently available to accomplish this task. Based on these constraints, engineering judgment is applied to each problem based on an analyst's thorough understan ding of the problem, detailed knowledge of the plant, and due consideration to the strengths and weaknesses associated with a given calculati onal model and/or methodology. Based on these conditions, Reg ulatory Guide 1.190 does not recommend using a sin gular calculational technique to determine fast neutron exposures. Inste ad, Regulatory Guide 1.190 suggests that one of the following neu tron transport tools be used to per form this work.
- Discrete Ordinates Tra nsport Calculation
- 1. Adjoint calculations benchmarked to a reference-forward calculation, or stand-al one forward calculations.
- 2. Various geometrical mo dels utilized with suitable mesh spacing in order to accurate ly represent the spatial distribution of the material compositions and source.
- 3. In performing discre te ordinates calcula tions, Regulatory Guide 1.190 also suggests that a P 3 angular decomposition of the scattering cr oss-sections be used, as a minimum.
- 4. Regulatory Guide 1.190 also recommen ds that discrete ordinates calcul ations utilize S 8 angular quadrature, as a minimum. 5. Regulatory Guide 1.1 90 indicates that the latest version of the Evaluated Nuclear Dat a File, or ENDF/
B, should be used for determining the nuc lear cross-sections; however, cross-sections based on earlier or equ ivalent nuclear data sets that have been thor oughly benchmark ed are also acceptable.
B/B-UFSAR 5.3-8a REVISION 15 - DECEMBER 2014
- Monte Carlo Transpor t Calculations A complete descripti on of the Westinghou se pressure vessel neutron fluence methodol ogy along with the S ER documenting NRC staff approval of the me thod and computer co des are provided in Reference 8.
Plant-Specific C alculations The most recent fast (E
> 1.0 MeV) neutron fluence evaluations for each of the Byron and Braidwood reactor pres sure vessels were based on a 2D/1D synthesis of ne utron fluxes that were obtained from a series of plant-and cycle-specific forward discrete ordinates transport ca lculations run in R-, R-Z, and R geometric models. The set of calculations, which assessed dosimetry as part of the reactor ve ssel surveillance program and pressure vessel neutron fluences, were conducted in a ccordance with the guidelines that are specified in Regulatory Guide 1.190.
B/B-UFSAR 5.3-8b REVISION 13 - DECEMBER 2010 5.3.1.7 Reactor Vessel Fasteners The reactor vessel closu re studs, nuts, and washers are designed, fabricated, and examined in accordance with the requirements of ASME Section III.
The closure s tuds are fabri cated of SA-540, Class 3 Grade B23 material. The closure stud materi al meets the fracture toughness req uirements of ASME Sect ion III, and 10 CFR 50 Appendix G. Repr esentative closure h ead bolting material properties for the B yron and Braidwood S tations are given in Tables 5.3-3a and b. The gu idelines for materials and inspections for vessel c losure studs are dis cussed in Appendix A. Inservice nondest ructive examinations are performed in accordance with the station ISI program.
The studs, nuts, and washers are removed from the refueling cavity and stored at convenient locations on the containment operating deck prior to removal of the r eactor closure head and refueling cavity flooding. Th erefore, the rea ctor closure studs are never exposed to the borated refue ling cavity water.
Additional protectio n against the poss ibility of incurring corrosion effects is ensured by the use of a manganese base phosphate surfacing trea tment. (For Byron Unit 2, out of service studs may remain installed in the reactor flange when the refueling cavity is flooded.)
B/B-UFSAR 5.3-9 REVISION 15 - DECEMBER 2014 The stud holes in the reactor fl ange are sealed with special plugs before removing the reac tor closure thus preventing leakage of the b orated refueling wat er into the stud holes. (For Byron Unit 2, out of service s tuds remaining i nstalled in the reactor flange do no t have these speci al plugs installed, therefore, prior to returning the stud to servic e, the out of service stud is removed and the stud hole inspected per existing procedures.)
5.3.2 Pressure-Tempe rature Limits 5.3.2.1 Limit Curves
Startup and shutdown o perating limitations are based on the properties of the co re region materi als of the r eactor pressure vessel. Actual material property test data are used. The methods outlined in Appe ndix G to Section XI of the ASME Code are employed for the shell regio ns in the analysis of protection against nonductile fai lure. The initial operating curves are calculated assuming a period of reactor op eration such that the beltline material wi ll be limiting. The heatup and cooldown curves are given in Figures 2.1, 2.2 and Table 2.1 of each station's Pressure T emperature Limits Report (PTLR). Beltline material properties chan ge with radiation exposure, and this change is measured in terms of the adj usted reference nil ductility temperature which includes a refer ence nil ductility temperature shift (RT NDT). Predicted RTNDT values are derived b ased on predicted neutron fluence at the a ssumed vessel wa ll flaw locations and the methodology provided in Regulatory Guide 1.9 9, Revision 2. The expected neutron fluence for rea ctor vessel wall locations of 1/4 T (thickness) and 3/4 T are determined. T hese reactor vessel wall locations represent the tips of the code reference flaw when the flaw is assumed at the inside diameter and outside diameter locations, respectively.
The methodo logy provided within Regulatory Gu ide 1.99, Revision 2 is used to calculate RT NDT based on the effec ts of neutron fluenc e and the effects of chemical composit ion of the vesse l wall material (specifically, copper and nickel). For a selected time of operation, this shift is assigned a sufficie nt magnitude so that no unirradiated ferrit ic materials in other components of the reactor coolant system will be limiting in the analysis.
The operating curves i ncluding pressure-temp erature limitations, are calculated in accord ance with 10 CFR 50, Appendix G, and ASME Code Section XI, Append ix G requirements.
In addition, Byron Units 1 and 2 and Braidwood Units 1 and 2 have received exemptions from the 10 CFR 50, Appendix G, flange region requirements. T he exemption allows for removal of t he pressure limitations that are gov erned by the limiting RT NDT of the closure head flange or vessel fl ange. The pressure-temperature curves in the PTLR account for this exem ption. Changes in fracture toughness of the core region p lates or forgings, weldments and associated heat affect ed zones due to ra diation damage will be monitored by a surve illance program which conforms with ASTM E-185, "Recommended Practice for Surveillance
B/B-UFSAR 5.3-9a REVISION 7 - DECEMBER 1998 Tests for Nuclear Reactor Vess els," and 10 CFR 50, Appendix H.
Byron and Braidwood Stations have received per mission from the NRC to integrate the r eactor vessel surveill ance programs per 10CFR50, Appendix H, Section III.C. This allows the surveillance programs to be integrated for B yron Units 1 and 2, and Braidwood Units 1 and 2, respectively. The evaluation of the radiation damage in this surve illance program is based on preirradiation testing of Charpy V-notch and tensile specimens and postirradiation testing of C harpy V-notch, t ensile, and 1/2 T compact tension specimens.
The postirradiation testing will be carried out d uring the lifetime of the reactor vessel.
Specimens are irradi ated in capsules B/B-UFSAR 5.3-10 REVISION 15 - DECEMBER 2014 located near the core midheight and removabl e from the vessel at specified intervals.
The results of the radia tion surveillance pr ogram will be used to verify that the RT NDT predicted from the effects of the fluence, or copper and nickel content is appro priate and to make any changes necessary to correct the fluence, or copper and nickel content if RTNDT determined from the surveillance program is greater or less than the predicted RT NDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests were calculated in accordance with 10 CFR 50, Appendix G.
The surveillance program withdrawal summary is contained in Table 4.1 of the PTLR document for each unit, respec tively. Changes to the withdrawal summary m ay be made as part of an update to the PTLR under the provisions of 10 CFR 50
.59. The schedule for removal of the capsu les for post irradiation t esting follows that of 10 CFR 50 Appendix H, as specified in Section 5.3.1.6.
Regulatory guides are di scussed in A ppendix A.
5.3.2.2 Operating Procedures The transient conditions that are considered in the design of the reactor vessel a re presented in Subs ection 3.9.1.1. These transients are r epresentative of the ope rating conditions that should prudently be cons idered to occur during plant operation.
The transients selected form a conservative ba sis for evaluation of the RCS to ensure the int egrity of the RCS equipment.
Those transients listed as upset conditi on transients are listed in Table 3.9-1.
None of these transi ents will result in pressure-temperature c hanges which exceed the heatup and cooldown limitations as described in Sub section 5.3.2.1 and in the Pressure Temperature Limits Report (PTLR).
5.3.3 Reactor Vessel Integrity
5.3.3.1 Design
The reactor vessel is cylindrical with a welded hemispherical bottom head and remova ble, bolted, fla nged, and gasketed, hemispherical upper head.
The react or vessel flange and head are sealed by two hollow metal lic O-rings. Seal leakage is detected by means of two leakoff paths: one b etween the inner and outer ring, and one outside the outer O-ring. The vessel contains the core, c ore support structures, control rods, and other parts directly associated with t he core. The reactor vessel closure head cont ains head adapte rs. These head adapters are tubular members, at tached by partial penetration welds to the underside of the cl osure head. The upper end of these adapters conta in acme threads for the assembly of control rod drive mechanisms or instrumentation adap ters. The seal
B/B-UFSAR 5.3-10a REVISION 7 - DECEMBER 1998 arrangement at the upp er end of these ad apters consists of a welded flexible canopy seal. Inlet and outlet nozzles are located symmetrically ar ound the vessel. Outlet nozzles are arranged on the vessel to facili tate optimum layout of the reactor coolant system e quipment. The i nlet nozzles are tapered from the coolant loop ve ssel interfaces to the vessel inside wall to reduce lo op pressure drop.
B/B-UFSAR 5.3-11 REVISION 8 - DECEMBER 2000 The bottom head of t he vessel contai ns penetration nozzles for connection and e ntry of the nuclear inco re instrumentation.
Each nozzle consists of a tubular memb er made of either an Inconel or an Incone l-stainless steel co mposite tube. Each tube is attached to the inside of the bottom head by a partial penetration weld.
Internal surfaces of the vessel which are in contact with primary coolant are we ld overlay with 0.125 inch minimum of stainless steel or Inc onel. The exterior of the reactor vessel is insulated with canned stain less steel refle ctive sheets.
The insulation is a minimum of 3 inches thick and contoured to enclose the top, sides, and bottom of the ve ssel. All the insulation modules are removable but the acc ess to vessel side insulation is limited by the surrounding concrete.
The reactor vessel is designed and fabricated in accordance with the requirements of ASME Section III.
Principal design param eters of the reactor vessel are given in Table 5.3-2. The vessel is shown in Figure 5.3-1.
Cyclic loads are intro duced by normal power changes, reactor trip, and startup an d shutdown operations.
These design base cycles are selected for fatigue evaluation and constitute a conservative design en velope for the pro jected plant life.
Vessel analysis result in a usage factor that is less than 1.
The design specifications require analysis to prove that the vessel is in compliance with the fatigue and stress limits of ASME Section III.
The loading and trans ients specified for the analysis are based on the mo st severe conditions expected during service.
The heatup and cooldown rates imposed by plant operating limits are provided in the Pressure and Temperature Limits Report (PTLR). Th ese rates are reflec ted in the vessel design specifi cations. 5.3.3.2 Materials of Construction
The materials used in the fabric ation of the rea ctor vessel are discussed in S ubsection 5.2.3.
5.3.3.3 Fabricat ion Methods The fabrication methods used in the construc tion of the reactor vessel are discu ssed in Subsecti on 5.3.1.2.
5.3.3.4 Inspection Requirements The inspection methods u sed in conjunction w ith the fabrication of the reactor v essel are described in Subsection 5.3.1.3.
B/B-UFSAR 5.3-12 REVISION 8 - DECEMBER 2000 5.3.3.5 Shipment and Installation The reactor vessel was shipped in a hori zontal position on a shipping sled with a vessel-lifting truss assembly. All vessel openings were sealed to prevent the entrance of moisture and an adequate quantity of desicca nt bags was plac ed inside the vessel. These were plac ed in a wire mesh basket attached to the vessel cover. A ll carbon steel su rfaces were painted with a heat resistant paint before shipment except for the vessel support surfaces and the top surface of the ex ternal seal ring.
The closure head was also ship ped with a shipping cover and skid. An enclosure attached to the vent ilation shroud support ring protected the c ontrol rod mechanism housings. All head openings were sealed to prevent the entrance of moisture and an adequate quantity of desicca nt bags were pla ced inside the head. These were placed in a wir e-mesh basket at tached to the head cover. All carbon stee l surfaces were painted with heat-resistant paint before shipment.
A lifting frame was provided for handling the vessel head.
5.3.3.6 Operating Conditions
Operating limitations are pres ented in Subsection 5.3.2 and in the Technical Specificat ions. The procedure s and methods used to ensure the integrity of t he reactor vessel under the most severe postulated condit ions are described in Subsection 3.9.1.4.
5.3.3.7 Inservice Surveillance The internal surface of the reactor ve ssel is capable of inspection periodica lly using visual a nd/or nondestructive techniques over the ac cessible areas. D uring refueling, the vessel cladding is capable of being in spected in certain areas such as the primary coolant outl et nozzles and, if deemed necessary, the core ba rrel is capable of being removed, making the entire inside vess el surface accessible.
The closure head is examined v isually in accordance with the requirements of ASME Section XI. Optical devices permit a selective inspection of the cladding, control rod drive mechanism nozzles, and the gasket seating surface. The knuckle transition piece, which is t he area of highest stress of the closure head, is acces sible on the outer surface for visual inspection, dye penetrant or magnetic pa rticle, and ultrasonic testing. The closure st uds can be inspected periodically using visual, magnetic particle an d/or ultrasonic techniques.
B/B-UFSAR 5.3-13 The full penetration welds in the following areas of the installed irradiated reactor v essel are available for visual and/or nondestructiv e inspection:
- a. Vessel shell - f rom the inside surface.
- b. Primary coolant nozzles - from the inside surface.
- c. Closure head - f rom the inside and o utside surfaces.
- d. Closure studs, n uts, and washers.
- e. Field welds between the reactor vessel, nozzles, and the main coo lant piping.
- f. Vessel flange seal surface.
The design considerations which have been incorporated into the system design to permit the above inspection are as follows:
- a. All reactor internals are completely removable.
The tools and storage sp ace required to permit these inspections are provided.
- b. The closure head is stor ed dry on the reactor operating deck durin g refueling to f acilitate direct visual inspection.
- c. All reactor vessel studs, nuts, and washers can be removed to dry stora ge during refueling.
- d. Removable plugs are provided in the primary shield. The ins ulation covering the nozzle welds may be removed.
The reactor vessel presents ac cess problems because of the radiation levels and remote un derwater accessibi lity to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufactu ring procedures in p reparation for the periodic nondestructive tests which are requ ired by the ASME inservice inspection c ode. These are:
- a. Shop ultrasonic examinat ions are performed on all internally clad surfaces to an acceptance and repair standard to a ssure an ade quate cladding bond to allow later ultrasonic te sting of the base metal from inside surface. The size of clad ding bonding defect allowed is 1/4-in ch by 3/4-inch in the region bounded by 2T (T = wall thick ness) on both sides of each full penet ration pressure boundary weld. Unbounded are as exceeding 0.442 in 2 (3/4-inch diameter) in all o ther regions are rejected.
B/B-UFSAR 5.3-14 REVISION 15 - DECEMBER 2014 b. The design of the reactor vessel shell is a clean, uncluttered cylindrical surf ace to permit future positioning of the t est equipment with out obstruction.
- c. The weld deposited cla dding surface on both sides of the welds to be inspected is specifically prepared to ensure meaningfu l ultrasonic examinations.
- d. During fabricati on, all full pen etration pressure boundary welds are ultra sonically examined in addition to Code examinations.
- e. After the shop hydrostat ic testing, se lected areas of the reactor v essel are ultras onically tested and mapped to facilitate the ins ervice inspe ction program.
The vessel design an d construction ena bles inspection in accordance with ASME Section XI.
5.3.4 References
- 1. Calculation Note CN-AMLRS-10-7, "Braidwood Units 1 and 2 Measurement Uncertainty Recapture (MUR)
Uprate: Reactor Vessel Integrity Evaluations."
- 2. Calculation Note CN-AMLRS-10-8, "Byron Units 1 and 2 Measurement Uncertainty Recapture (MUR)
Uprate: Reactor Vessel Integrity Evaluations."
- 3. Braidwood Pressure a nd Temperature Limits Reports (PTLRs) for Units 1 and 2.
- 4. Byron Pressure and T emperature Limits Re ports (PTLRs) for Units 1 and 2.
- 5. Schmittroth, E.A., "FERRET D ata Analysis Cod e", HEDL-TME 40, Hanford Engineering Develo pment Laboratory, Richland, Washington, Sept ember 1979.
- 6. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Comp endium", July 1994.
- 7. Regulatory Guide 1.190, "Calculational and D osimetry Methods for Determining Pressu re Vessel Neutron Fluence," United States Nuclear Regulat ory Commission, Of fice of Nuclear Regulatory Research, March 2001.
- 8. Andrachek, J.D., "Methodolog y Used to Develop Cold Overpressure Mitigat ing System Setpoints and RCS Heatup and Cooldown Limit Curves", WCAP-14040-A, Revisi on 4, May 2004.
- 9. Babcock & Wilcox Rep ort No. 77-1159832-0 0, "Pressurized Thermal Shock Evaluati on in Accordance with 10 CFR 50.61 for the Reactor Vessels in Byron U nits 1 and 2 and Braidwood Units 1 and 2", dated January 13, 1986.
REVISION 15 DECEMBER 2014 Normal Cooldown360340 320 260 Time to reach 140°F No SFP Heat Load = 42.3 Hours Min SFP Heat Load = 46.7 Hours300 280 240 -I220200 180 160 Minimum SFP Heat Load 140 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Time (hours after shutdown)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 5.4-6 REACTOR COOLANT TEMPERATURE VS TIME (NORMAL COOLDOWN)
REVISION 15 DECEMBER 2014 Single Train Cooldown360 340 320 300 Time to Reach 200°F No SFP Heat Load = 50.3 Hours 280 220 A 200 260 240 180 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 Time (hours after shutdown)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 5.4-7 SINGLE RHR TRAIN RC TEMPERATURE VERSUS TIME
B/B-UFSAR 5.3-1 5.3 REACTOR VESSEL 5.3.1 Reactor Vessel Materials
This section is for purposes of reac tor pressure vessel fabrication.
5.3.1.1 Material Specifications Material specifications are in accordance wi th the ASME Code requirements and are giv en in Subsection 5.2.3.
5.3.1.2 Special Processes Used for Manufactu ring and Fabrication
- a. The vessel is Seismic Ca tegory I and Q uality Group A. Design and f abrication of the reactor vessel is carried out in stric t accordance with ASME Code,Section III, Class 1 r equirements. The head flanges and nozzles are manufa ctured as forgings.
The cylindrical portion of the vessel is made up of several forged shells. The hemispherical heads are made from dished plates.
The reactor vessel parts are joined by weldin g, using the sin gle or multiple wire submerged arc.
- b. The use of severely sens itized steel as a pressure boundary material has been prohibited and has been eliminated by either a s elect choice of material or by programming t he method of assembly.
- c. The control rod drive mechanism head adaptor threads and surfaces of the guide studs are chrome plated to prevent possible g alling of the mated parts.
- d. At all locatio ns in the reactor vess el where stainless steel and Inconel are joined , the final joining beads are Inconel weld metal in order to prevent cracking.
- e. The location of full penetration weld seams in the upper closure head and vessel bottom head are restricted to areas th at permit accessibility during inservice inspection.
- f. The stainless steel clad surfaces are sampled to ensure that composition and de lta ferrite requirements are met. g. The procedure qualificat ion for cladding low alloy steel (SA508 Class 2) re quires a special evaluation to ensure freedom fr om underclad cracking.
B/B-UFSAR 5.3-2 5.3.1.3 Special Methods for Nondestructi ve Examination The examination requirem ents detailed in the following are in addition to the examination requ irements of Section III of the ASME Code.
The reactor vessel nondestructive examination (NDE) program is given in Table 5.3-1.
5.3.1.3.1 Ultrason ic Examination
- a. In addition to the d esign code straight beam ultrasonic test, angle beam inspecti on of 100% of plate material is performed during fabrication to detect discontinuiti es that may be undetected by longitudinal wave examination.
- b. In addition to ASME Section III nondestructive examination, all full pe netration welds and heat affected zones in the reactor vessel are ultrasonically examined during fabricati on. This test is performed upon comp letion of the welding and intermediate heat treatm ent but prior to the final postweld heat treatment.
- c. The reactor vessel is examined after hydrostatic testing for information.
5.3.1.3.2 Penetrant Examinations The partial penetration welds for the co ntrol rod drive mechanism head adapt ors and the bottom i nstrumentation tubes are inspected by dye p enetrant after the roo t pass in addition to code requirements.
Core support block atta chment welds were inspected by dye penetrant after first layer of weld metal and after each 1/2 i nch of weld metal. All clad surfaces and other vessel and head internal surfaces were i nspected by dye penetrant after the hydrostatic test.
5.3.1.3.3 Magnetic Particle Examination
All magnetic particle examinations of materi als and welds were performed in accordance with the following:
- a. Prior to the final pos tweld heat treatment - by the prod, coil, or direc t contact method.
- b. After the final postweld heat treatment - by the yoke method.
The following surfaces and welds were ex amined by magnetic particle methods.
B/B-UFSAR 5.3-3 REVISION 15 - DECEMBER 2014 Surface Examinations
- a. Magnetic particle exam ination of all exterior vessel and head surf aces after the h ydrostatic test.
- b. Magnetic particle exam ination of all exterior closure stud surfaces and all nut surfaces after final machining or rolli ng. Continuous circular and longitudinal magne tization were used.
- c. Magnetic particle examin ation of all inside diameter surfaces of carbon and low alloy steel p roducts that have their prope rties enhanced by ac celerated cooling.
This inspection is perfo rmed after forming and machining (if re quired) and prior to cladding.
Weld Examination Magnetic particle examination of the weld me tal buildup for vessel welds att aching the closu re head lifting lugs to the reactor vessel after the first layer and each 1/2 inch of weld metal is deposited. All pressure boundary welds were examined after back chipping or b ack grinding operations.
5.3.1.4 Special Controls for Ferritic and Au stenitic Stainless Steels Welding of ferritic st eels and austenitic stainless steels is discussed in Subsect ion 5.2.3. Subsecti on 5.2.3 includes discussions which indicate t he degree of acceptance with guidelines for c ontrol of ferrite content in sta inless steel metal welds, use of se nsitized stainless ste el, electroslag weld properties, stainless st eel weld cladding of low-alloy steel components and welde r qualification for areas of limited accessibility. Appendix A discusses the deg ree of conformance with regulat ory guides.
5.3.1.5 Fracture Toughness
Assurance of adequat e fracture toughness of ferritic materials in the reactor coola nt pressure boundary (ASME Section III Class 1 Components) is provided by com pliance with the requirements for fracture toughn ess testing included in NB-2300 to Section III of th e ASME Boiler and Pr essure Vessel Code, and Appendix G of 10 CFR 50.
The initial Charpy V-n otch minimum upper shelf fracture energy levels for the r eactor vessel beltli ne (including welds) shall be 75 foot-pounds as re quired by Appendix G of 10 CFR 50.
Materials having a section thick ness greater than 10 inches with an upper shelf of less than 75 fo ot-pounds shall be evaluated with regar d to effects of chemistry (especially copper content), initial upper shelf energy and influence to B/B-UFSAR 5.3-4 REVISION 15 - DECEMBER 2014 ensure that a 50 foot-pound shelf energy as required by Appendix G of 10 CFR 50 is mai ntained throughout the life of the vessel. The specimens sha ll be oriented as required by NB-2300 of Secti on III of the ASME Boi ler and Pressure Vessel Code. The reactor v essel material prope rties for units of the Byron/Braidwood Stations are given in Sectio n 5 of the PTLR.
5.3.1.5.1 Pressurized Ther mal Shock Evaluation Fracture toughness requi rements for protection of reactor vessels against pressurized thermal shock events are given in 10 CFR 50.61. Reference 9 provides the initial assessment of Pressurized Thermal Shock. Su bsequently, evaluations which include surveillance capsule data ha ve been performed in accordance with the re quirements of 10 CFR 5 0.61 for the reactor vessels at Byron / Bra idwood Units 1 and 2.
The evaluations are provided in References 1 and 2 and t he evaluation results are summarized in References 3 and 4 and Tables 5.
3-7 through 5.3-10.
5.3.1.6 Material Surveillance
In the surveillance pr ogram, the evaluation of the radiation damage is based on pre irradiation testing of Charpy V-notch and tensile specimens an d postirradiation testing of Charpy V-notch, tensile and 1/2 thick ness (T) compa ct tension (CT) fracture mechanics test specimens. The program is directed toward evaluation of the effect of radia tion on the fracture toughness of reactor vessel steels based on the transition temperature approach and the fra cture mechanics appr oach. The program conforms with ASTM-E-185 "Recomm ended Practice for Surveillance Tests f or Nuclear Reactor Vesse ls," and 10 CFR 50, Appendix H.
Detailed information on the reactor vessel m aterial surveillance program is provided in Westinghouse repo rts WCAP-9517 for Byron Unit 1, WCAP-10398 for Byr on Unit 2, and WCAP-9807 for Braidwood Unit 1 and W CAP-11188 for Braidwood 2.
The reactor vessel s urveillance program uses six specimen capsules. The capsu les are located in guide baskets welded to the outside of the neu tron shield pads a nd are positioned directly opposite the center portion of the core.
The capsules can be removed when the vessel h ead is removed and can be replaced when the internals are removed. The six capsules contain reactor vessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal working direction of the limiting base material located in the core region of t he reactor vessel and associated weld metal and weld heat-affected zone metal. The 6 capsules contain 54 tensile specimens, 3 60 Charpy V-notch sp ecimens (which include weld metal and weld heat-affected zone m aterial), and 72 CT specimens. Archive material sufficient for two additional capsules will be retained.
B/B-UFSAR 5.3-5 REVISION 1 - DECEMBER 1989 Dosimeters, including Ni, Cu, Fe, Co-Al, Cd shielded Co-Al, Cd shielded Np-237 and Cd s hielded U-238, are p laced in filler blocks drilled to contain them. The dosimet ers permit evaluation of the flux seen by the specimens and the vessel wall. In addition, thermal monitors made of low melting p oint alloys are included to monitor the maximum temperature of the specimens.
The specimens are enclosed in a tight-fitting stainless steel sheath to prevent corr osion and ensure good th ermal conductivity.
The complete capsule is helium leak tested.
Each of the six caps ules contains the following specimens:
Number ofNumber of Number ofMaterial CharpysTensiles CTs Limiting base material*1534
Limiting base material**1534 Weld metal*** 1534 Heat affected zone 15
- Specimens oriented in the major working direction.
- Specimens oriented normal to the major w orking direction.
- Weld metal to be sel ected per ASTM E185.
The following dosimeters and thermal monitor s are included in each of the six capsules:
Dosimeters
Iron Copper Nickel Cobalt-Aluminum (0.15% Co)
Cobalt-Aluminum (Cadmium shielded)
Np-237 (Cadmium shielded)
Thermal Monitors 97.5% Pb, 2.5% Ag (579
°F melting point).
97.5% Pb, 1.75% Ag, 0.75% Sn (590
°F melting point).
B/B-UFSAR 5.3-6 REVISION 12 - DECEMBER 2008 The fast neutron exposure of the specimens occurs at a faster rate than that exper ienced by the ve ssel wall, with the specimens being located between the core and the vessel. Since these specimens experi ence accelerated e xposure and are actual samples from the mater ials used in the v essel, the t ransition temperature shift measur ements are rep resentative of the vessel at a later time in life.
Data from CT fract ure toughness specimens are expected to provide additional information for use in determining allow able stresses for ir radiated material.
Correlations between the calcula tions and the measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Subsect ion 5.3.1.6.1. They have indicated good agreement. The antici pated degree to which the specimens will perturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure will be made by use of data on all capsules withdrawn. For the sc hedule for removal of the capsules for postirradiation testing which follows that of 10 CFR 50 Appendix H, refer to Tab le 4.1 of the PTLR.
5.3.1.6.1 Measurement of Integra ted Fast Neutr on (E>1.0 MeV)
Flux at the Irra diation Samples The use of passive neutr on sensors such as t hose included in the internal surveillance ca psule dosimetry sets dose not yield a direct measure of the en ergy dependent neutron flux level at the measurement location.
Rather, the a ctivation or fission process is a measure of the in tegrated effect that t he time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accur ate assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the senso rs may be developed fr om the measurements only if the sensor c haracteristics and t he parameters of the irradiation are well known. In partic ular, the following variables are of interest:
- 1. The measured specific ac tivity of each sensor
- 2. The physical characteris tics of each sensor
- 3. The operating history of the reactor
- 4. The energy response of each sensor
- 5. The neutron energy spectrum at the sensor location In this section the pr ocedures used to deter mine sensor specific activities, to develop react ion rates for indivi dual sensors from the measured specific ac tivities and the operating history of the reactor, and to derive key fast neutron expo sure parameters from the measured reaction rates are described.
B/B-UFSAR 5.3-7 REVISION 9 - DECEMBER 2002 5.3.1.6.1.1 DETERM INATION OF SENSOR REACTION RATES The specific activit y of each of the radiometric sensors is determined using established ASTM procedures.
Following sample preparation and weighing, the specific activity of e ach sensor is determined by means of a hig h purity ger manium gamma spectrometer. In the case of the survei llance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the i ndividual wires; or, as in the case of U-238 and Np-237 fission monitor s, by direct count ing preceded by dissolution and chemical separation of cesium from the sensor.
The irradiation history of the reactor o ver its operating lifetime is determined f rom plant power genera tion records. In particular, operating da ta are extracted on a monthly basis from reactor startup to the end of the capsule irradiation period.
For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly hi stogram describing rea ctor operation has proven to be an adeq uate representation for use in radioactive decay corrections for the reacti ons of interest in the exposure evaluations.
Having the measured spec ific activities, the operating history of the reactor, and the physical characteristic s of the sensors, reaction rates referenced to full power operation are determined from the following equation:
where: A = measured specific activity (dps/gm)
R = reaction rate averaged over the irradiation period and referenced to operat ion at a core power level of P ref (rps/nucleus).
N o = number of target element atoms per gram of sensor.
F = weight fraction of t he target isotope in the sensor material.
Y = number of product at oms produced per reaction.
P j = average core p ower level during irradation period j (MW). P ref = maximum or reference core power level of the reactor (MW).
[]d j t t j ref j j o e e C P P Y F N A R=1 B/B-UFSAR 5.3-7a REVISION 15 - DECEMBER 2014 C j = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period. = decay constant of th e product isotope (sec
-1).
t j = length of irradi ation period j (sec).
t d = decay time following i rradiation period j (sec).
and the summation is c arried out over the to tal number of monthly intervals comprising the total irradiation period.
In the above equ ation, the ratio P j/Pref accounts for month by month variation of power level within a give n fuel cycle. The ratio C j is calculated for e ach fuel cycle a nd accounts for the change in sensor reaction rates caused by variat ions in flux level due to changes in core power spati al distributions from fuel cycle to fuel cycle. S ince the neutron flux at the measurement locations within the surveil lance capsules is dominated by neutrons produced in the periph eral fuel assemblies, the change in the relative power in these assemblies from fuel cycle to fuel cycle can have a significant impact on the activation of neutron sensors. For a single-c ycle irradiation, C j = 1.0. However, for multiple-cycle irradiat ions, particularly those employing low le akage fuel management, the additional C j correction must be utilized in order to provide accurate determinations of the decay corrected reacti on rates for the dosimeter sets contained in the surveillance capsules.
5.3.1.6.1.2 Corrections to React ion Rate Data Prior to using the measu red reaction rates in the least squares adjustment procedure d iscussed in Section 5.
4.3.6.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium is otopes over the course of the irradiation.
In addition to the c orrections made for the presence of U-235 in the U-238 fission sensors, corre ctions are also made to both the U-238 and Np-237 sensor reaction rat es to account for gamma ray induced fission reacti ons occurring over the course of the irradiation.
5.3.1.6.1.3 Least Square s Adjustment Procedure Least squares adjustment methods provide the capability of combining the measurement da ta with the neutron transport calculation resulting in a Best Estimate neu tron energy spectrum with associated uncert ainties. Best Estimat e for key exposure parameters such as (E > 1.0 eV) or dpa/s along with their uncertainties are th en easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results red uces the uncertainty in the calculated spectrum and acts to remove bi ases that may be present in the analytical technique.
B/B-UFSAR 5.3-7b REVISION 15 - DECEMBER 2014 In general, the least squares methods, as ap plied to pressure vessel fluence evaluatio ns, act to reconcile the measured sensor reaction rate data, dosimetry re action cross-sec tions, and the calculated neutron energy spec trum within th eir respective uncertainties. For example, relates a set of measured reaction rates, R i , to a single neutron spectrum g , through the mult igroup dosimeter cross-section, ig , each with an uncertainty . The use of least squares adjustm ent methods in L WR dosimetry evaluations is not new. The American So ciety for Testing and Materials (ASTM) has a ddressed the use of ad justment codes in ASTM Standard E944, "A pplication of Neutron Sp ectrum Adjustment Methods in Reactor Surve illance" and many indu stry workshops have been held to discuss the vario us applications.
For example, the ASTM-EURATOM Symposia on Reactor Dosimetry holds workshops on neutron spectrum unfolding and adjustment tech niques at each of its bi-annual conferences.
Th primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the me asurement. The analyt ical method alone may be deficient because it inherently con tains uncertainty due to the input assumptions to the cal culation. Typically these assumptions include para meters such as the t emperature of the water in the peripheral fuel assemblies, by-pass region, and downcomer regions, component d imensions, and p eripheral core source. Industry cons ensus indicates th at the use of calculation alone results in overall uncerta inties in the neutron exposure parameters in the ra nge of 15-20% (1). By combining the calculated resu lts with available measurements, the uncertainties associated with the key neutron exposure parameters can be re duced. Specifically ASTM Standard E 944 states; "The algorithims of the adjustment cod es tend to decrease the variances of the a djusted data compared to the corresponding input values. T he least squares adjustment codes yield estimates for the output data with minimum variances, that is, the "best estimates". This is the pri mary reason for using these adjustment procedures". ASTM E 944 provides a comprehensive listing of available adjustment codes.
The FERRET least squares adjus tment code (Reference 5) was initially developed at the Hanford Eng ineering Development Laboratory (HEDL) and has ha d extensive use in both the Liquid Metal Fast Breeder (LM FBR) program and t he NRC Sponsored Light Water Reactor Dosimetry Improvement Program (L WR-PV-SDIP). As a result of participation in several cooperati ve efforts associated with the LWR-PV-SDIP, the FE RRET approach was adopted by Westinghouse in
()()g i g ig ig g R i R++/-=+
B/B-UFSAR 5.3-7c REVISION 15 - DECEMBER 2014 the mid 1980's as the preferred approach for the evaluation of LWR surveillance dosimetry. The least squares m ethodology was judged superior to the previously employ ed spectrum averaged cross-section approach that is totally depen dent on the accuracy of the shape of the calculated neutron spectrum at the measurement locations.
The FERRET code is e mployed to combine the results of plant specific neutron transport cal culations and multiple foil reaction rate measurements to determ ine best estimat e values of exposure parameters ( (E > 1.0 MeV) and dpa) along with associated uncertainties at the measurement locations.
The application of the least squares met hodology requires the following input:
- 1. The calculated neutr on energy spectrum and associated uncertainties at the measurement location.
- 2. The measured reactio n rate and associa ted uncertainty for each sensor contained in the multiple foil set.
- 3. The energy dependent dos imetry reaction cross-sections and associated uncerta inties for each sensor contained in the multiple foil sensor set.
For a given application, the c alculated neutron spectrum is obtained from the results of p lant specific ne utron transport calculations applicable to the irradiation p eriod experienced by the dosimetry sensor set. T his calculation is performed using the benchmarked transport calcul ational methodol ogy described in Section 5.3.1.6.2. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the s pecific irradiation hist ory of the sensor set to perform the radio active decay c orrections. The dosimetry reaction cross-sections and uncertainties ar e obtained from the SNLRML dosimetry cross-secti on library (Reference 6). The SNLRML library is an evalua ted dosimetry reac tion cross-section compilation recommended for use in LWR e valuations by ASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 7 06 (IIB)". There are no additional data or data libraries built into the FERRET code system.
All of the required input is supplied externally at the time of the analysis.
The uncertainties asso ciated with the me asured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares proce dure in the form of variances and covariances. The as signment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.
B/B-UFSAR 5.3-8 REVISION 15 - DECEMBER 2014 5.3.1.6.2 Calculation of Integrated Fast Neutron (E
> 1.0 MeV)
Flux at the Irra diation Samples A generalized set of guideli nes for performing fast neutron exposure calculations within t he reactor configuration, and procedures for a nalyzing measured irradi ation sample data that can be correlated to t hese calculations, has been promulgated by the Nuclear Regulatory Commission (NRC) in R egulatory Guide 1.190, "Calculational and Dosimetry Meth ods for Determining Pressure Vessel Neutron Fluence" [Reference 7].
Since different calculational models e xist and are conti nuously evolving along with the associated model inputs, e.g., cross-se ction data, it is worthwhile summarizing the key model s, inputs, and procedures that the NRC staff finds accep table for use in d etermining fast neutron exposures within the rea ctor geometry.
This material is highlighted below.
Calculation and Dosimetry Measurement Procedures The selection of a particular ge ometric model, the corresponding input data, and the over all methodology used to determine fast neutron exposures within the rea ctor geometry are based on the needs for accurately determining a solution to the problem that must be solved and the date/resources that are currently available to accomplish this task. Based on these constraints, engineering judgment is applied to each problem based on an analyst's thorough understan ding of the problem, detailed knowledge of the plant, and due consideration to the strengths and weaknesses associated with a given calculati onal model and/or methodology. Based on these conditions, Reg ulatory Guide 1.190 does not recommend using a sin gular calculational technique to determine fast neutron exposures. Inste ad, Regulatory Guide 1.190 suggests that one of the following neu tron transport tools be used to per form this work.
- Discrete Ordinates Tra nsport Calculation
- 1. Adjoint calculations benchmarked to a reference-forward calculation, or stand-al one forward calculations.
- 2. Various geometrical mo dels utilized with suitable mesh spacing in order to accurate ly represent the spatial distribution of the material compositions and source.
- 3. In performing discre te ordinates calcula tions, Regulatory Guide 1.190 also suggests that a P 3 angular decomposition of the scattering cr oss-sections be used, as a minimum.
- 4. Regulatory Guide 1.190 also recommen ds that discrete ordinates calcul ations utilize S 8 angular quadrature, as a minimum. 5. Regulatory Guide 1.1 90 indicates that the latest version of the Evaluated Nuclear Dat a File, or ENDF/
B, should be used for determining the nuc lear cross-sections; however, cross-sections based on earlier or equ ivalent nuclear data sets that have been thor oughly benchmark ed are also acceptable.
B/B-UFSAR 5.3-8a REVISION 15 - DECEMBER 2014
- Monte Carlo Transpor t Calculations A complete descripti on of the Westinghou se pressure vessel neutron fluence methodol ogy along with the S ER documenting NRC staff approval of the me thod and computer co des are provided in Reference 8.
Plant-Specific C alculations The most recent fast (E
> 1.0 MeV) neutron fluence evaluations for each of the Byron and Braidwood reactor pres sure vessels were based on a 2D/1D synthesis of ne utron fluxes that were obtained from a series of plant-and cycle-specific forward discrete ordinates transport ca lculations run in R-, R-Z, and R geometric models. The set of calculations, which assessed dosimetry as part of the reactor ve ssel surveillance program and pressure vessel neutron fluences, were conducted in a ccordance with the guidelines that are specified in Regulatory Guide 1.190.
B/B-UFSAR 5.3-8b REVISION 13 - DECEMBER 2010 5.3.1.7 Reactor Vessel Fasteners The reactor vessel closu re studs, nuts, and washers are designed, fabricated, and examined in accordance with the requirements of ASME Section III.
The closure s tuds are fabri cated of SA-540, Class 3 Grade B23 material. The closure stud materi al meets the fracture toughness req uirements of ASME Sect ion III, and 10 CFR 50 Appendix G. Repr esentative closure h ead bolting material properties for the B yron and Braidwood S tations are given in Tables 5.3-3a and b. The gu idelines for materials and inspections for vessel c losure studs are dis cussed in Appendix A. Inservice nondest ructive examinations are performed in accordance with the station ISI program.
The studs, nuts, and washers are removed from the refueling cavity and stored at convenient locations on the containment operating deck prior to removal of the r eactor closure head and refueling cavity flooding. Th erefore, the rea ctor closure studs are never exposed to the borated refue ling cavity water.
Additional protectio n against the poss ibility of incurring corrosion effects is ensured by the use of a manganese base phosphate surfacing trea tment. (For Byron Unit 2, out of service studs may remain installed in the reactor flange when the refueling cavity is flooded.)
B/B-UFSAR 5.3-9 REVISION 15 - DECEMBER 2014 The stud holes in the reactor fl ange are sealed with special plugs before removing the reac tor closure thus preventing leakage of the b orated refueling wat er into the stud holes. (For Byron Unit 2, out of service s tuds remaining i nstalled in the reactor flange do no t have these speci al plugs installed, therefore, prior to returning the stud to servic e, the out of service stud is removed and the stud hole inspected per existing procedures.)
5.3.2 Pressure-Tempe rature Limits 5.3.2.1 Limit Curves
Startup and shutdown o perating limitations are based on the properties of the co re region materi als of the r eactor pressure vessel. Actual material property test data are used. The methods outlined in Appe ndix G to Section XI of the ASME Code are employed for the shell regio ns in the analysis of protection against nonductile fai lure. The initial operating curves are calculated assuming a period of reactor op eration such that the beltline material wi ll be limiting. The heatup and cooldown curves are given in Figures 2.1, 2.2 and Table 2.1 of each station's Pressure T emperature Limits Report (PTLR). Beltline material properties chan ge with radiation exposure, and this change is measured in terms of the adj usted reference nil ductility temperature which includes a refer ence nil ductility temperature shift (RT NDT). Predicted RTNDT values are derived b ased on predicted neutron fluence at the a ssumed vessel wa ll flaw locations and the methodology provided in Regulatory Guide 1.9 9, Revision 2. The expected neutron fluence for rea ctor vessel wall locations of 1/4 T (thickness) and 3/4 T are determined. T hese reactor vessel wall locations represent the tips of the code reference flaw when the flaw is assumed at the inside diameter and outside diameter locations, respectively.
The methodo logy provided within Regulatory Gu ide 1.99, Revision 2 is used to calculate RT NDT based on the effec ts of neutron fluenc e and the effects of chemical composit ion of the vesse l wall material (specifically, copper and nickel). For a selected time of operation, this shift is assigned a sufficie nt magnitude so that no unirradiated ferrit ic materials in other components of the reactor coolant system will be limiting in the analysis.
The operating curves i ncluding pressure-temp erature limitations, are calculated in accord ance with 10 CFR 50, Appendix G, and ASME Code Section XI, Append ix G requirements.
In addition, Byron Units 1 and 2 and Braidwood Units 1 and 2 have received exemptions from the 10 CFR 50, Appendix G, flange region requirements. T he exemption allows for removal of t he pressure limitations that are gov erned by the limiting RT NDT of the closure head flange or vessel fl ange. The pressure-temperature curves in the PTLR account for this exem ption. Changes in fracture toughness of the core region p lates or forgings, weldments and associated heat affect ed zones due to ra diation damage will be monitored by a surve illance program which conforms with ASTM E-185, "Recommended Practice for Surveillance
B/B-UFSAR 5.3-9a REVISION 7 - DECEMBER 1998 Tests for Nuclear Reactor Vess els," and 10 CFR 50, Appendix H.
Byron and Braidwood Stations have received per mission from the NRC to integrate the r eactor vessel surveill ance programs per 10CFR50, Appendix H, Section III.C. This allows the surveillance programs to be integrated for B yron Units 1 and 2, and Braidwood Units 1 and 2, respectively. The evaluation of the radiation damage in this surve illance program is based on preirradiation testing of Charpy V-notch and tensile specimens and postirradiation testing of C harpy V-notch, t ensile, and 1/2 T compact tension specimens.
The postirradiation testing will be carried out d uring the lifetime of the reactor vessel.
Specimens are irradi ated in capsules B/B-UFSAR 5.3-10 REVISION 15 - DECEMBER 2014 located near the core midheight and removabl e from the vessel at specified intervals.
The results of the radia tion surveillance pr ogram will be used to verify that the RT NDT predicted from the effects of the fluence, or copper and nickel content is appro priate and to make any changes necessary to correct the fluence, or copper and nickel content if RTNDT determined from the surveillance program is greater or less than the predicted RT NDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests were calculated in accordance with 10 CFR 50, Appendix G.
The surveillance program withdrawal summary is contained in Table 4.1 of the PTLR document for each unit, respec tively. Changes to the withdrawal summary m ay be made as part of an update to the PTLR under the provisions of 10 CFR 50
.59. The schedule for removal of the capsu les for post irradiation t esting follows that of 10 CFR 50 Appendix H, as specified in Section 5.3.1.6.
Regulatory guides are di scussed in A ppendix A.
5.3.2.2 Operating Procedures The transient conditions that are considered in the design of the reactor vessel a re presented in Subs ection 3.9.1.1. These transients are r epresentative of the ope rating conditions that should prudently be cons idered to occur during plant operation.
The transients selected form a conservative ba sis for evaluation of the RCS to ensure the int egrity of the RCS equipment.
Those transients listed as upset conditi on transients are listed in Table 3.9-1.
None of these transi ents will result in pressure-temperature c hanges which exceed the heatup and cooldown limitations as described in Sub section 5.3.2.1 and in the Pressure Temperature Limits Report (PTLR).
5.3.3 Reactor Vessel Integrity
5.3.3.1 Design
The reactor vessel is cylindrical with a welded hemispherical bottom head and remova ble, bolted, fla nged, and gasketed, hemispherical upper head.
The react or vessel flange and head are sealed by two hollow metal lic O-rings. Seal leakage is detected by means of two leakoff paths: one b etween the inner and outer ring, and one outside the outer O-ring. The vessel contains the core, c ore support structures, control rods, and other parts directly associated with t he core. The reactor vessel closure head cont ains head adapte rs. These head adapters are tubular members, at tached by partial penetration welds to the underside of the cl osure head. The upper end of these adapters conta in acme threads for the assembly of control rod drive mechanisms or instrumentation adap ters. The seal
B/B-UFSAR 5.3-10a REVISION 7 - DECEMBER 1998 arrangement at the upp er end of these ad apters consists of a welded flexible canopy seal. Inlet and outlet nozzles are located symmetrically ar ound the vessel. Outlet nozzles are arranged on the vessel to facili tate optimum layout of the reactor coolant system e quipment. The i nlet nozzles are tapered from the coolant loop ve ssel interfaces to the vessel inside wall to reduce lo op pressure drop.
B/B-UFSAR 5.3-11 REVISION 8 - DECEMBER 2000 The bottom head of t he vessel contai ns penetration nozzles for connection and e ntry of the nuclear inco re instrumentation.
Each nozzle consists of a tubular memb er made of either an Inconel or an Incone l-stainless steel co mposite tube. Each tube is attached to the inside of the bottom head by a partial penetration weld.
Internal surfaces of the vessel which are in contact with primary coolant are we ld overlay with 0.125 inch minimum of stainless steel or Inc onel. The exterior of the reactor vessel is insulated with canned stain less steel refle ctive sheets.
The insulation is a minimum of 3 inches thick and contoured to enclose the top, sides, and bottom of the ve ssel. All the insulation modules are removable but the acc ess to vessel side insulation is limited by the surrounding concrete.
The reactor vessel is designed and fabricated in accordance with the requirements of ASME Section III.
Principal design param eters of the reactor vessel are given in Table 5.3-2. The vessel is shown in Figure 5.3-1.
Cyclic loads are intro duced by normal power changes, reactor trip, and startup an d shutdown operations.
These design base cycles are selected for fatigue evaluation and constitute a conservative design en velope for the pro jected plant life.
Vessel analysis result in a usage factor that is less than 1.
The design specifications require analysis to prove that the vessel is in compliance with the fatigue and stress limits of ASME Section III.
The loading and trans ients specified for the analysis are based on the mo st severe conditions expected during service.
The heatup and cooldown rates imposed by plant operating limits are provided in the Pressure and Temperature Limits Report (PTLR). Th ese rates are reflec ted in the vessel design specifi cations. 5.3.3.2 Materials of Construction
The materials used in the fabric ation of the rea ctor vessel are discussed in S ubsection 5.2.3.
5.3.3.3 Fabricat ion Methods The fabrication methods used in the construc tion of the reactor vessel are discu ssed in Subsecti on 5.3.1.2.
5.3.3.4 Inspection Requirements The inspection methods u sed in conjunction w ith the fabrication of the reactor v essel are described in Subsection 5.3.1.3.
B/B-UFSAR 5.3-12 REVISION 8 - DECEMBER 2000 5.3.3.5 Shipment and Installation The reactor vessel was shipped in a hori zontal position on a shipping sled with a vessel-lifting truss assembly. All vessel openings were sealed to prevent the entrance of moisture and an adequate quantity of desicca nt bags was plac ed inside the vessel. These were plac ed in a wire mesh basket attached to the vessel cover. A ll carbon steel su rfaces were painted with a heat resistant paint before shipment except for the vessel support surfaces and the top surface of the ex ternal seal ring.
The closure head was also ship ped with a shipping cover and skid. An enclosure attached to the vent ilation shroud support ring protected the c ontrol rod mechanism housings. All head openings were sealed to prevent the entrance of moisture and an adequate quantity of desicca nt bags were pla ced inside the head. These were placed in a wir e-mesh basket at tached to the head cover. All carbon stee l surfaces were painted with heat-resistant paint before shipment.
A lifting frame was provided for handling the vessel head.
5.3.3.6 Operating Conditions
Operating limitations are pres ented in Subsection 5.3.2 and in the Technical Specificat ions. The procedure s and methods used to ensure the integrity of t he reactor vessel under the most severe postulated condit ions are described in Subsection 3.9.1.4.
5.3.3.7 Inservice Surveillance The internal surface of the reactor ve ssel is capable of inspection periodica lly using visual a nd/or nondestructive techniques over the ac cessible areas. D uring refueling, the vessel cladding is capable of being in spected in certain areas such as the primary coolant outl et nozzles and, if deemed necessary, the core ba rrel is capable of being removed, making the entire inside vess el surface accessible.
The closure head is examined v isually in accordance with the requirements of ASME Section XI. Optical devices permit a selective inspection of the cladding, control rod drive mechanism nozzles, and the gasket seating surface. The knuckle transition piece, which is t he area of highest stress of the closure head, is acces sible on the outer surface for visual inspection, dye penetrant or magnetic pa rticle, and ultrasonic testing. The closure st uds can be inspected periodically using visual, magnetic particle an d/or ultrasonic techniques.
B/B-UFSAR 5.3-13 The full penetration welds in the following areas of the installed irradiated reactor v essel are available for visual and/or nondestructiv e inspection:
- a. Vessel shell - f rom the inside surface.
- b. Primary coolant nozzles - from the inside surface.
- c. Closure head - f rom the inside and o utside surfaces.
- d. Closure studs, n uts, and washers.
- e. Field welds between the reactor vessel, nozzles, and the main coo lant piping.
- f. Vessel flange seal surface.
The design considerations which have been incorporated into the system design to permit the above inspection are as follows:
- a. All reactor internals are completely removable.
The tools and storage sp ace required to permit these inspections are provided.
- b. The closure head is stor ed dry on the reactor operating deck durin g refueling to f acilitate direct visual inspection.
- c. All reactor vessel studs, nuts, and washers can be removed to dry stora ge during refueling.
- d. Removable plugs are provided in the primary shield. The ins ulation covering the nozzle welds may be removed.
The reactor vessel presents ac cess problems because of the radiation levels and remote un derwater accessibi lity to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufactu ring procedures in p reparation for the periodic nondestructive tests which are requ ired by the ASME inservice inspection c ode. These are:
- a. Shop ultrasonic examinat ions are performed on all internally clad surfaces to an acceptance and repair standard to a ssure an ade quate cladding bond to allow later ultrasonic te sting of the base metal from inside surface. The size of clad ding bonding defect allowed is 1/4-in ch by 3/4-inch in the region bounded by 2T (T = wall thick ness) on both sides of each full penet ration pressure boundary weld. Unbounded are as exceeding 0.442 in 2 (3/4-inch diameter) in all o ther regions are rejected.
B/B-UFSAR 5.3-14 REVISION 15 - DECEMBER 2014 b. The design of the reactor vessel shell is a clean, uncluttered cylindrical surf ace to permit future positioning of the t est equipment with out obstruction.
- c. The weld deposited cla dding surface on both sides of the welds to be inspected is specifically prepared to ensure meaningfu l ultrasonic examinations.
- d. During fabricati on, all full pen etration pressure boundary welds are ultra sonically examined in addition to Code examinations.
- e. After the shop hydrostat ic testing, se lected areas of the reactor v essel are ultras onically tested and mapped to facilitate the ins ervice inspe ction program.
The vessel design an d construction ena bles inspection in accordance with ASME Section XI.
5.3.4 References
- 1. Calculation Note CN-AMLRS-10-7, "Braidwood Units 1 and 2 Measurement Uncertainty Recapture (MUR)
Uprate: Reactor Vessel Integrity Evaluations."
- 2. Calculation Note CN-AMLRS-10-8, "Byron Units 1 and 2 Measurement Uncertainty Recapture (MUR)
Uprate: Reactor Vessel Integrity Evaluations."
- 3. Braidwood Pressure a nd Temperature Limits Reports (PTLRs) for Units 1 and 2.
- 4. Byron Pressure and T emperature Limits Re ports (PTLRs) for Units 1 and 2.
- 5. Schmittroth, E.A., "FERRET D ata Analysis Cod e", HEDL-TME 40, Hanford Engineering Develo pment Laboratory, Richland, Washington, Sept ember 1979.
- 6. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Comp endium", July 1994.
- 7. Regulatory Guide 1.190, "Calculational and D osimetry Methods for Determining Pressu re Vessel Neutron Fluence," United States Nuclear Regulat ory Commission, Of fice of Nuclear Regulatory Research, March 2001.
- 8. Andrachek, J.D., "Methodolog y Used to Develop Cold Overpressure Mitigat ing System Setpoints and RCS Heatup and Cooldown Limit Curves", WCAP-14040-A, Revisi on 4, May 2004.
- 9. Babcock & Wilcox Rep ort No. 77-1159832-0 0, "Pressurized Thermal Shock Evaluati on in Accordance with 10 CFR 50.61 for the Reactor Vessels in Byron U nits 1 and 2 and Braidwood Units 1 and 2", dated January 13, 1986.
REVISION 15 DECEMBER 2014 Normal Cooldown360340 320 260 Time to reach 140°F No SFP Heat Load = 42.3 Hours Min SFP Heat Load = 46.7 Hours300 280 240 -I220200 180 160 Minimum SFP Heat Load 140 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 Time (hours after shutdown)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 5.4-6 REACTOR COOLANT TEMPERATURE VS TIME (NORMAL COOLDOWN)
REVISION 15 DECEMBER 2014 Single Train Cooldown360 340 320 300 Time to Reach 200°F No SFP Heat Load = 50.3 Hours 280 220 A 200 260 240 180 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 Time (hours after shutdown)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 5.4-7 SINGLE RHR TRAIN RC TEMPERATURE VERSUS TIME