RS-16-248, Redacted Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Chapter 4, Reactor
Text
B/B-UFSAR
Attachment 4.4A
Additional Information On the Plant Specific Application of t he Westinghouse Improved Thermal Design Procedure To Byron/Braidwood
B/B-UFSAR 4.4A-1 4.4A Additional Infor mation On the Plant Sp ecific Application of the Westinghouse Improved Thermal Design Procedure To Byron/Braidwood The NRC Safety Evaluation Re port on WCAP-9500 entitled Reference Core R eport 17x17 Optimize d Fuel Assembly noted the specific plants using the Westinghouse Improved Thermal Design Proc edure (ITDP) must supply additional informati on on the plant sp ecific application of the ITDP to p erform thermal-hydra ulic analyses.
Thus, Byron/ Bra idwood specific resp onses to NRC information requests a re provided below.
4.4A.1 Request 1 Provide the sensitiv ity factors (S i) and their range of applicability.
4.4A.2 Response 1 The sensitivity factors (S i) and their range of applicability are given in T able 1 of Reference 2 for Byron/Braidwood. Please note that these values are the same as those used in WCAP-9500 with the exception of the range for vessel flow.
The range on flow for Byron/
Braidwood has been exten ded down to 273270 gpm (70%
flow) with no change in the correspond ing sensitivity factor being required.
4.4A.3 Request 2 If the S i values used in the B yron/Braidwood analyses are different than those use d in WCAP-9500, then the applicant must reevaluate th e use of an uncertainty allowance for application of equation 3-2 of WCAP-8567, "Improved Thermal Design Procedure" and the linearity assumption must be validated.
4.4A.4 Response 2
The S i values used in Byron/Br aidwood analys es are the same as those used in WCAP-9500. Therefore, reevaluating the use of an uncertainty a llowance for application of equation 302 of WCAP-8567, "Improved Thermal Design Procedu re" and the li nearity assumption is not required.
4.4A.5 Request 3 Provide and justify the varian ces and distributions for input parameters.
B/B-UFSAR 4.4A-2 4.4A.6 Response 3 The distribution assumed for the input parameters such as pressurizer pressure, core average temperature, reactor power, and R CS flow are normal, two-sided 95+%
probability distributions.
The variances of the se parameters fo r Byron/Braidwood are consistent with the varian ces calculated in the
generic response.
Specifically, the uncertainties for pressurizer pressure and cor e average temperature are identical to the generic res ponse since the sensors, process racks, and compu ter and readout devices are standard Westinghouse su pplied NSSS equipment.
Variances in reactor pow er and reactor coolant system flow are calculated based on equation 4 and equation 8 respectively in Refere nce 1. As can be seen from the equations, both prim ary and secondary side parameters are measured for power and flow calorimetrics. The
error allowances for the parameters measured by Westinghouse sup plied equipment are identical to those used in the generic submittal (Reference 1). Two input parameters are measu red by non-Westi nghouse supplied instruments. Th ese are feedwater temperature and feedwater pressure. As expe cted, the error allowances for these instruments vary slightly from those used in Reference 1. The error allowances for feedwater temperature and pressure were statistically combined (as described in Reference 1) to get the total channel allowance for each parameter.
The feedwater pressure error allowance w as calculated to be less than the error allowance used in Reference 1.
Therefore, the error contribution to the reactor power and flow uncertainties from fe edwater pressu re is less than that used in the generic response.
Similarly, the error s for feedwater temperature were combined to get the total channel allowance. The total allowance was found to be slightly hig her than that used to calculate RCS flow uncert ainty in Reference 1.
However, the error allowance from feedwater temperature is very small relati ve to the other co ntributing errors and in fact this small additio nal error is absorbed in the statistical comb ination. Therefore, the flow uncertainty calculated in Re ference 1 is applicable for Byron/Braidwood.
As stated in Reference 1, the flow cal orimetric can be performed one of several way
- s. Commonwealth Edison plans to do a precis ion flow calorimetric at the beginning of the cycle and normalize the loop elbow
B/B-UFSAR 4.4A-3 taps. For monthly surve illance to assure plant operation consistent with th e ITDP assum ptions, the loop flows will be read o ff the plant proce ss computer. The total flow uncertainty associated with this method was calculated in Reference 1 and is applicable to the Byron/Braidwood units.
It is to be noted that the total channel allowance for feedwater temperature was calc ulated to be l ess than the error assumed for the re actor power uncertainty calculation in Reference 1.
Therefore, the power uncertainty for Byro n/Braidwood is bounded by the uncertainty calculated in the generic response.
4.4A.7 Request 4 Justify that the nor mal conditions used in the analyses bound all permitted modes of plant operation.
4.4A.8 Response 4
This item was addressed in R eference 1 and is applicable to the Byron/B raidwood units.
4.4A.9 Request 5
Provide a discussion of what code uncertainties, including their values, are included in the DNBR analyses.
4.4A.10 Response 5 The uncertainties included in the ITDP DNBR analyses for Byron/Braidwood are given in Table 1 of Reference 2. As a result of these values bei ng different from those used in WCAP-9500, the Design DNBR Limits also differ. The calculation of the D esign limit DNBRs for the Typical and Thimble cells are given in Reference 2, Tables 2 and 3 respectively.
Since the Design DNBR Limits given in Table 2 and 3 are different fr om those originally given, Section 4.4 has been rev ised to inco rporate the Reference 2 values.
4.4A.11 Request 6 Provide a block diag ram depicting sens or, processing equipment, computer and readout devices for each parameter channel used in the uncertainty analysis.
Within each element of the b lock diagram identify the accuracy, drift, range, span, operating limits, and setpoints. Identify the ove rall accuracy of each channel transmitter to final output and specify the minimum acceptable accur acy for use with the new procedure. Also identify th e overall accura cy of the
B/B-UFSAR 4.4A-4 final output value and maximum accuracy requirements for each input channel for this final output device.
4.4A.12 Response 6 Block diagrams are n ot provided in t his response.
However, as in the gen eric response, a table is provided in Reference 2 givin g the error breakd own from sensor to computer and readout devices.
This table is abbreviated though, giving only the error breakdowns for instruments that differ from those in Ta ble 4, "Typi cal Instrument Uncertainties," of Ref erence 1. As noted earlier, these instruments are those that measure fee dwater temperature and pressure.
4.4A.13 Request 7
If there are any changes to the THINC-IV correlation, or parameter values outside of previously demonstrated acceptable ranges, t he staff requires a reevaluation of the sensitivity factors and of t he use of equation 3-2 of WCAP-8567.
4.4A.14 Response 7 For Byron/Braidwood, the THINC-IV code and WRB-1 DNB
Correlation are the same as that used in WCAP-9500.
Therefore, reevaluat ing the sensitivity factors and the use of equation 3-2 of WCAP-8567 is not required.
References
- 1. Westinghouse letter, NS-EPR-2577, E.
P. Rahe to C. H. Berlinger (NRC), March 31, 1982, proprietary.
- 2. General Electric Compa ny letter transmitting improved thermal design info rmation to the NRC (to be written), proprietary.
B/B-UFSAR 4.5-1 4.5 REACTOR MATERIALS Section 4.5 provides a discussion of the materials employed in the control rod drive system and the r eactor internals.
A more detailed evaluation of the re actor materials and reactivity control sys tems indicating the degree of conformance with the recommendatio ns of the applicable R egulatory Guides is presented in the Final Safet y Analysis Report as follows:
- a. control rod drive mechan ism and reactor internals:
Chapter 3.0, b. control rod drive mechan ism testing:
Chapters 3.0, 14.0, and the Techni cal Specifications, c. control rod drive mech anism and reactor internals materials: Chapter 5.0,
- d. safety injection sys tem: Chapter 6.0,
- e. instrumentation for reac tor control and protection:
Chapter 7.0, and
- f. failure of the control rod drive mechanism cooling system and chemi cal and volume c ontrol system:
Chapter 9.0.
4.5.1 Control Rod System Structural Materials 4.5.1.1 Materials Specifications
All parts exposed to reactor coolant are mad e of metals which resist the corrosive action of the water.
Three types of metals are used exclusively:
stainless steels, nickel-chromium-iron, and cobalt based alloys. In the case of stainless steels, only austenitic and m artensitic stainless s teels are used. For pressure boundary parts, martens itic stainless s teels are not used in the heat treat ed conditions which ca use susceptibility to stress corrosion cracking or accelerated corrosion in the Westinghouse pressurized wat er reactor water chemistry.
- a. Pressure Boundary All pressure contain ing materials co mply with Section III of the ASME Boil er and Pressure Ve ssel Code, and are fabricated from austenitic (Type 304) stainless steel. b. Coil stack assembly The coil housings require a magnetic material. Both low carbon cast steel and ductile iron have been successfully tested for this application. On the
B/B-UFSAR 4.5-2 basis of cost and perfor mance, ductile iron was selected for the control rod d rive mechanism (CRDM).
The finished housings are zinc plated or flame sprayed to provide c orrosion resistance.
Coils are wound on bobbins of molded Dow Corning 302 material, with doubl e glass insulated copper wire.
Coils are then vacuum im pregnated with silicon varnish. A wrapping of mica sheet is secured to the coil outside diameter.
The result is a well insulated coil capable of sustained operation at 200°C. c. Latch assembly Magnetic pole pieces are fabricated from Type 410 stainless steel. All nonmag netic parts, except pins and springs, are fabricated from Type 304 stainless steel. Haynes 25 is used to fabricate link pins.
Springs are made from nickel-chromium-iron alloy (Inconel-750). Latch arm tips are clad with Stellite-6 to provide improv ed wearability. Hard chrome plate and Ste llite-6 are used selectively for bearing and wear surfaces.
- d. Drive rod assembly The drive rod assemb ly utilizes a Type 410 stainless steel drive rod and disconnect rod a ssembly. The coupling is machined from ty pe 403 stain less steel.
Other parts are Type 304 sta inless steel with the exception of the springs, which are nickel-chromium-iron alloy, and the locking button, which is Haynes 25; and the belleville washers which are Inconel 718.
Several small parts (s crews and pins) are Inconel 600.
Material specifications for Cl ass 1 components of the CRDM are as follows:
CRDM, upper head SB-166 or SB-167 and SA-182 Grade F304 Latch housing SA-1 82, Grade F304 or SA-351 Grade CF8 Rod travel housing SA-182, Grade F304 or SA-336 Class F8 Cap SA-479, Type 304 Welding materials Stainless Steel Weld Metal Analysis A-8
B/B-UFSAR 4.5-3 4.5.1.2 Austenitic Stain less Steel Components
- a. All austenitic stainless steel materials used in the fabrication of CRDM comp onents are p rocessed, inspected and tested to avoid sensitization and prevent intergranula r stress cor rosion cracking.
The rules covering these controls are stipulated in Westinghouse process specifications.
As applicable, these process specifications supplem ent the equipment specifications and purchase order requirements of every individual austenitic stainless steel component regardless of the AS ME Code Classification.
Westinghouse practice is that austenitic stainless steel materials of p roduct forms with simple shapes need not be corrosion te sted provided that the solution heat treatment is followed by water quenching. Simple shapes are defined as all plates, sheets, bars, pipe and tubes, as wel l as forgings, fittings and other shaped products whi ch do not have inaccessible cavities or chamb ers that would preclude rapid cooling when water que nched. When testing is required the tests are p erformed in accordance with ASTM A 262, Prac tice A or E, as amended by Westinghouse Process S pecification 84201 MW.
If, during the c ourse of fabrication the steel is inadvertently exposed to the sensitization temperature range, 800
°F to 1500°F the material may be tested in accordance with ASTM A 262, as amended by Westinghouse Process Specificatio n 84201 MW to verify that it is not susceptible to intergranular att ack, except that testing is not required for:
- 1. Cast metal or weld m etal with a ferrite content of 5 percent or more, 2. Material with a carbon c ontent of 0.03%
or less that is subjected to tem peratures in the range of 800°F to 1,500
°F for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- 3. Material exposed to special processing provided the processing is proper ly controlled to develop a uniform product an d provided t hat adequate documentation exists of service expe rience and/or test data to demonstrate that the processing will not result in increased susceptibility to intergranular st ress corrosion.
If it is not verified th at such material is not susceptible to intergr anular attack, the material
B/B-UFSAR 4.5-4 will be re-solution anne aled and water quenched
or rejected.
- b. The welding of austenitic stainless steel is controlled to mitigate the occurrence of microfissuring or hot cracking in the weld.
Available data indicates tha t a minimum delta ferrite level expressed in F errite Number (FN), above which the weld metals commonly use d by Westinghouse will
not be prone to hot cracking, lies som ewhere between 0 FN and 3 FN. The undi luted weld dep osits of the starting welding materials a re required to contain a minimum of 5 FN.
4.5.1.3 Other Materials The CRDMs are cleaned prior to delivery in a ccordance with the guidance of ANSI N45.2.1. Westi nghouse personnel do conduct surveillance to ensure that manufacturers an d installers adhere to appropriate requirements.
Haynes 25 is used in small q uantities to fabrica te link pins.
The material is ordered in the solution trea ted and cold worked condition. Stress cor rosion cracking has not been observed in this application over the last 15 years.
The CRDM springs are made fr om nickel-chromium-iron alloy (Inconel-750) ordered to MIL-S-23192 or MIL-N-24114 Class A #1 temper drawn wire. Op erating experience has shown that springs made of this materia l are not subject to stress-corrosion cracking.
4.5.1.4 Cleaning and Cleanliness Control
The CRDMs are cleaned prior to delivery in a ccordance with the guidance of ANSI N45.2.1. Mea sures are applied, as appropriate, to apply packaging req uirements to procureme nt orders, to review supplier packaging procedures, to apply prop er cleaning requirements, marking and iden tification and to provide protection to equipment from physical or wea ther damage, to apply special handling precautions and to define sto rage requirements.
Westinghouse quality assurance procedures are described in "Westinghouse Water Reactor Divisions Quality Assurance Plan," WCAP-8370, Revision 8A u pdated per letter NS-TMA-2039, From T. M.
Anderson to W. P. Haass, February 8, 1979.
4.5.2 Reactor Internals Materials
4.5.2.1 Materials Specifications All the major material for the reactor i nternals is Type 304 stainless steel. Parts not fa bricated from Type 304 stainless steel include bolts and dowel pins, which are fabricated from
B/B-UFSAR 4.5-5 REVISION 11 - DECEMBER 2006 Type 316 stainless steel, and radial support key bolts, which are fabricated from Inconel-750.
Material specifications for reac tor vessel internals for emergency core cooling systems are listed in Table 5.2-4.
There are no other materials u sed in the reactor internals or core support structu res which are not ot herwise included in ASME Code,Section II I, Appendix I.
4.5.2.2 Controls on Welding
The discussions provided in Subs ection 4.5.1 a re applicable to the welding of reactor i nternals and core support components.
4.5.2.3 Nondestructive Examination of Wr ought Seamle ss Tubular Products and Fittings
The nondestructive exa mination of wrought seamless tubular products and fittings is in acco rdance with Sect ion III of the ASME Code.
4.5.2.4 Fabrication and Processing of Au stenitic Stainless Steel Components
The discussions provided in Subsection 4.5.1.4 are applicable to the cleaning of reactor internals and core s upport structures in accordance with ANSI N45.2.1.
B/B-UFSAR 4.6-1 REVISION 7 - DECEMBER 1998 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6.1 Information for Control Rod Drive System (CRDS)
Figure 4.2-8 pro vides the layout of the CRDS. The CRDS is a magnetically operated jack with no hydra ulic system associated with its functioning.
The control r od drive mec hanism consists of four separa te subassemblies.
- a. The pressure vessel which includes the l atch housing and rod travel housings.
- b. The coil stack assembly which includes three operating coils: stationary gripper coil, movable
gripper coil and lift coil.
- c. The latch assembly which includes the guide tube, the stationary and the movab le pole piec es and the stationary and movab le gripper latches.
4.6.2 Evaluation of the CRDS The CRDS has been analyzed in detail in a fa ilure mode and effects analysis (Refere nce 1). This study, and the analyses presented in Chapter 15.0, d emonstrates that the CRDS performs its intended safety function, reactor tr ip, by putting the reactor in a subcritic al condition when a safety system setting is approached, with any assumed credible failure of a single active component. T he essential elements of the CRDS (those required to ensure reactor trip) are isolated fr om nonessential portions of the CRDS (the rod control system).
Despite the extremely low probability of a c ommon mode failure impairing the ability of the rea ctor trip system to perform its safety function, analyses have been performe d in accordance with the requirements of WASH-1270.
These analyses, documented in References 2 and 3, have demonst rated that acc eptable safety criteria would not be exceeded even if the C RDS were rendered incapable of functioning during a reactor tran sient for which their function w ould normally be expected.
The design of the control ro d drive mechanism is such that failure of the contr ol rod drive mechanism c ooling system will, in the worst cas e, result in an individu al control r od trip or a full reactor trip.
4.6.3 Testing and Veri fication of the CRDS The CRDS was extensively tested prior to its ope ration. These tests may be subdivided into five catego ries: (1) prototype
B/B-UFSAR 4.6-2 REVISION 8 - DECEMBER 2000 tests of components, (2) prototy pe CRDS tests, (3) production tests of components following manufacture and prior to installation, (4) onsite preoper ational tests, a nd (5) initial startup tests.
In accordance with Table 14.2-65, the reactor trip system operation was verified in a startup test. This test ensured that the system operated in accordance with the safety analysis report, design requireme nts, and plant insta llation. A final test was performed in which a manual reactor trip was initiated, (after fuel load but pri or to initial criticalit y) to verify that all rods would fully insert.
The rod cluster control assembli es were dropped and the drops were timed. The time from b eginning of deca y of stationary gripper coil voltage to dashpot entry shall be less than or equal to 2.7 seconds for each rod, the Technical Specification limit. In compliance with Tables 14.2-66 and 14.2-66a, all rods falling outside the two-sigma limit were retested a minimum of three times each. Rods we re dropped into represen tative flow condi-tions. In addition, the CRDS is subject to periodic ins ervice tests.
These tests are conducted to verify the oper ability of the CRDS when called upon to function.
4.6.4 Information for Combined Perfo rmance of Reacti vity Systems
As is indicated in Chapt er 15.0, the only po stulated events which assume credit for re activity control systems other than a reactor trip to render t he plant subcritical are the steam line break, feedwater line break, and loss-of-coolant accident. The reactivity control sys tems for which cre dit is taken in these accidents are the reactor trip system and the safety injection system (SIS). Note that no cred it is taken for the boration capabilities of the ch emical and volume cont rol system (CVCS) as a system in the analys is of transients present ed in Chapter 15.0.
The adverse boron dilution possi bilities due to the operation of the CVCS are investigated in C hapter 15.0. Prior proper operation of the CVCS has been presumed as an initial condition to evaluate transients, and appropriate Techni cal Specifications have been prepared to ensure the correct operation or remedial action. 4.6.5 Evaluation of Combined Performance
The evaluations of the steam line break, feedwater line break, and the loss-of-coolant accident, which presume the combined actuation of the reactor trip system to the CRDS and the SIS, are presented in Chapter 15.0.
Reactor trip signals and safety injection signals for these events are generated from functionally diverse sensors and actuate diverse means of reactivity control, i.e., control rod insertion and injection of soluble poison.
B/B-UFSAR 4.6-3 REVISION 7 - DECEMBER 1998 Nondiverse but r edundant types of equipm ent are utilized only in the processing of the incoming sensor signals into a ppropriate logic, which ini tiates the protective ac tion. In particular, note that protection from eq uipment failures is provided by redundant equipment and periodic testing. Eff ects of failures of this equipment have been extensively investiga ted as reported in Reference 4. The failure mode and effects ana lysis described in this reference v erifies that any single failure will not have a deleterious effect on the engine ered safety feat ures actuation system. 4.6.6 References
- 1. Shopsky, W. E., "Failu re Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Control Syst em," WCAP 8976, August 1977.
- 2. "Westinghouse Anticipa ted Transients Without Trip Analysis," WCAP-8330, August 1974.
- 3. Gangloff, W. C.
and Loftus, W. D., "An E valuation of Solid State Logic Reactor Pr otection in Antici pated Transients," WCAP-7706-L (Proprietary) and WCAP-7706 (Nonproprietary), July 1971.
- 4. Eggleston, F. T., Rawlins, D.
H. and Petrow, J. R., "Failure Mode and Effects Analysis (FMEA) of the Engineering
Safeguard Features Actua tion System," WCAP-8584 (Proprietary) and WC AP-8760 (Nonpropriet ary), April 1976.