ML20113A858
| ML20113A858 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/14/1996 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20036D299 | List: |
| References | |
| CAW-96-942, NUDOCS 9606260036 | |
| Download: ML20113A858 (28) | |
Text
. - _
Westinghouse Energy Systems Nuclear Technology Division Electric Corporation m 355 Pittsburgh Pennsylvania 15230-0355 March 14,1996 CAW-96-942 bocument Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. William T. Russell, Director APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
" Wolf Creek Nuclear Operating Corporation (WCNOC) - Cycle 9 Reload Design Changes -
Configuration Change Package (CCP) - 5672 Rev.1 & Rev. 0"
Dear Mr. Russell:
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-%-942 signed by the owner of the proprietary information, Westinghouse Electric Corporation. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Wolf Creek Nuclear Operating Corporation.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should refert ce this letter, CAW-96-942, and should be addressed to the undersigned.
Ve truly yours, N.
Liparulo, Manager TJK/bbp Regulatory and Engineering Networks Attachment Kevin Bohrer/NRC(12H5) cc:
CAWW2 9606260036 960620 PDR ADOCK 05000482 P
PDR l
CAW-96-942 l
l AFFIDAVIT 1
1 COMMONWEALTH OF PENNSYLVANIA:
l ss l
COUNTY OF ALLEGHENY:
i Before me, the undersigned authority, personally appeared Nicholas J. Liparulo, who, being l
by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
i b
i
\\J Nicholas J. Liparuto, Manager Regulatory and Engineering Networks t
l l
Sworn to and subscribed D
before me this 14 day of M daeE
.1996 flo'ana!Dnl Denise K Henccrsen, Notary Public Monroede Cam,/Jegwy County f%ComrncdonDves Oc' 8,19'A l
gauat
-WlO Momuer, Pormyivwu wist tu.ews Notary Public l
l 195&TJK-12129f,
J CAW-96-942 (1)
I am Manager, Regulatory and Engineering Networks, in the Nuclear Services Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.
(2)
I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in tnat connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
1957C-TJK 2:031296
. CAW-96-942 l
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
i (c)
Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the % estinghouse system which include the following:
(a)
He use of such information by Westinghouse gives Westinghouse a j
competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)
It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
1957C TE3212%
. CAW-96-942 l
l (c)
Use by our competitor would put Westinghouse at a competitive disadvantage l
by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving 1
Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Wes6nghouse in the world market, and thereby give a market advantage to the competition of those countries.
(0 The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
(iv)
The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in " Wolf Creek Nuclear Operating Corporation (WCNOC) -
Cycle 9 Reload Design Changes", Configuration Change Package (CCP) #5672 Rev. I and Rev. 0 (Proprietary), for the Wolf Creek Generating Station, being transmitted by the Wolf Creek Nuclear Operating Corporation (WCNOC) letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk, Attention Mr. William T. Russell. The proprietary information as submitted for use by the Wolf Creek Nuclear Operating Corporation for the Wolf Creek Generating Station is expected to be applicable in other licensee 1957C-TJK-4:tm296
_ _. ~ - _. _ _ _ _ _ _ _. - _. - _. _ _.. _ _ _ _... _. _. _ _ _. _
~
5-CAW-96-942 l
l submittals in response to certain NRC requirements for justification of reload design methodologies.
This information is part of that which will enable Westinghouse to:
(a)
Establish applicable reload and safety analysis, i
(b)
Perform and provide reload designs.
Further this information has substantial commercial value as follows:
(a)
Westinghouse plans to sell the use of similar information to its customers for purposes of reload design methodology.
(b)
Westinghouse can sell support and defense of safety analyses and reload design methodology.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar reload design methodologies and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
1 l
l 1957C-TES.031296
I l ' CAW-96-942 i
In order for competitors of 'Vestinghouse to duplicate this information, similar j
technical programs wculd have to be performed and a significant manpower effort, i
having the requisite.*21ent and experience, would have to be expended for developing reload design methodompes.
Further the deponent sayeth not.
I l
1957C-TJK 6:031296
l Westinghouse Non-Proprietary Class 3 l
l ENCLOSURE 2 i
Wolf Creek Nuclear Operating Corporation Wolf Creek Nuclear Station Configuration Change Package (CCP) - 5672 Rev.1 & Rev. O Westinghouse Non-Proprietary Class 3 l
NSDNT122/SAPilo l
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..aet y
w
h$c.ErnTo 3 S4 71 '-/ r Wolf Creek Generating Station a
t Cycle 9 N1) CLEAR OPERATING CORN'N Reload Safety Evaluation i
Both Region 10 and 11 feature the Westinghouse EM grids. The EM grid feature pan of the licensing basis in other plants and meets all fuel assembly and fuel rod desig The IFM grid is described in the " VANTAGE 5 Reference Core Report", WCAP-10444-P-A (Reference 10). Application of the IFM grids to the VANTAGE SH fuel assembly is addressed in Addendum 2-A (Reference 11). Principle design features for these fuel types are describe below.
i The EM grid in the VANTAGE SH fuel assembly is an adaptation of the exisung VANTAGE 5 EM grid design to a 0.374 inch O.D. standard fuel rod. EM's are located in the three upp spans between the VANTAGE SH Zircaloy structural grids to promote flow muung in the hottest fuel assembly spans. The EM grids are fabricated of Zircaloy in the same manner as the l
VANTAGE SH Zirealoy grids but are not intended to be structural members. The EM grid envelope is slightly smaller than the VANTAGE SH Zircaloy grid. Each EM grid cell provides four (4) point fuel rod support. His simplified cell arrangement allows the EM grid to accomplish it's flow miving objective with minimal pressure drop.
1
)
The Region 10 and 11 fbel assemblies are the Wesunghouse VANTAGE SH with EM Performance + fuel design. He significant featun:s incorporated in the Performance + fuel include; 4
Addition of a protective bottom grid.
i Extmling the length of the end plugs.
Relocation of grids 1 and 2.
e Use of a variable pitch plenum sprmg.
Rotated mid-grids.
High burnup bottom grid.
Rese design features were subject of several Enginming reviews. The Westinghouse design review for Performance + fuel (DR-93-01) serves as the top level safety review for the fuel design changes. A summmy of the Engineenng reviews for Performance + fuel follows.
1.
He fuel rod design is termed PERFORMANCE +. The fuel rod has a longer bottom end plug, longer top end plug with extemal gripper, var (agleyitch plenum sprin, shorter fuel tube and plenum. DQk)ttom End Plug [ ]"$d th]fr'on(
]
e Top End Plug length [
]ffom[
e configuration changed g a
accommodate the extegnt gripper design. The Fuel Tube length [
]Tr5m ti
[_
] His decrease was performed to accommp the longer end plugs. The plenum spnng length changed from[
] It should be noted that i
the overall length of the fuel rod did not change. De design review for PERFORMANCE + fuel addressed all the affected design criteria associated with the bottom nozzle, protective grid and fuel rod design changes. Letter PDT-94-016 5
~_
\\,- N
-~c D S 4 ? L 'ii I
Wolf Creek Generating Station j
Cycle 9 1
NUCl EAR OPERAT1NG CORPORATION Reload Safety Evaluation i
1 documents its applicability to the Region 10 fuel and jusufies the use of an e review for the remaining changes. This conclusion remains valid for the Region 11 fuel.
j 2
A Protective Bottom Grid Assembly was added to provide an additional debris barrier and to provide additional fretting resistance. His grid is referred to as Grid P.
2 3.
A cast debris Siter bottom nozzle with a pitch change was added. The bottom nozzle is 4
now a two piece cast and has a pitch change from[
]'T$e thimble tube holes also moved from the "Y" corner to accommodate the change.
4 The new design reduced fuel rod to bottom nozzle gap to[
]
5.
Rotated mid-grids. An engineenng review (ER-93-33) was conducted after the flow testing was completed on the 17x17 V5H fuel assemblies. He data collected during flow j
testing included test assemblies with and without alternately rotated mid-grids. The i
excessive vibratior.s present in the standard configuration was virtually elimiwM when the alternate vaned pids were rotated by 90 degrees about the axis of the fuel assembly.
This review provided the basis for acceptance of rotating attemate vaned mid-grids.
SECL-93-221 (Reference 15) addressed the rotation of altematirs gri:is for fuel assemblies with and without intermediate flow mixers (IFMs). h evaluation in SECL-93-221 addressed all aspects of the rotated grid design and i termind that there would be m wiay impact due to the design change. Westmghouse's determination was based on the fact that the basic grid design was unchanged and rotation would have no impact on fluid turbulence and direction. However, departure from nucleate boiling tests performed at the Columbia University Heat Transfer Research Facility indicated that rotation of altemating gnds would reduce the thermal performance in fuel which also included IFMs in the assembly design. Upon these findings, Westinghouse notified the Nuclear Regulatory Comminion (Reference 16) of the discrepancy and notified the utilities that were incorporating both rotated grids and IFMs into their fuel design. With the exception of thermal performance, all other areas evaluation performed in SECL 221 remam valid for WCGS's Cycle 9 fuel assembly design. A complete investigation of the impact of the fuel's reduced thermal performance at WCGS was performed (References 17,18). De WCNOC evaluation determmed that the thennal penalty associased with the rotated mid-grids could be accommodated in the generic margm for the Cycle 8 design. This conclusion remains valid for the Cycle 9 design. Based on the conclusions of the WCNOC evaluation and the evaluation provided in SECL-93-221 applied to all other design change concerns, implementation of altemately rotated grids in fuel assemblies with IFMs has no safety impact on WCGS.
6.
Relocated grids 1 and 2. Engineenng Review (ER-94-01) was conducted to address this design change. PDT-94-047 provided a data package for the engineering review. It described the proposed change as moving the bottom grid [
]aifd the
GSWHI Wolf Creek Generating Station Cycle 9 I
NUCt. EAR OPERATING CORPORATION Reload Safety Evaluation
- a. e.
second grid [
]for the 0.374 inch rod design. Information was presented on drawing tolerances and irradiation growth allowance, resistance to vibra thermal hydraulic impact and manufacturing impact of the design change. FD-94-045 documents the results of the engineering review. The proposed design change was accepted based on the completion of three action items: 1) Assess whether the changes have any effect on the cross flow velocities: 2) Dependent upon action item 1, revise the VIBAMP analysis, and: 3) Review the interface drawings for required changes. Letter i
FA-94-072 from A. Meliksetian states, that the cross flow velocities based on the
' original' design are applicable and bound the cross flows with the proposed modifications implemented." Therefore, relocation of grids I and 2 has no impact on safety for Cycle 7.
The top nozzle has a revised spring tang. The slot chamfer in zone 4 of the nozzle drawing was re-dimensioned.
4 8.
The spring force was increased for the Bottom Grid and longer grid insens were introduced.
Other core components that will be included with the Region 11 fuel include Westinghouse Integral Fuel Bumable Absorbers. In a chemical shim core, excess reactivity is controlled by boric acid dissolved in the moderator. Above a certain concentration, the moderator temperatu coefficient may be positive due to the reduction in boron number density associated with a reduction in water density. With no other control, the moderator coefficient at HZP conditions at BOL could be greater than the Technical Specification limit. The use of bumable absorbers for partial reactivity control results in a lower soluble boron concentration and assures a moderator 4
temperatore coefficient within the Technical Specification limits.
The absorbers used in the Region 11 fuel assemblies are the Westinghouse Integral Burnable Absorber (IFB A) design. 'I)ese IFBA rods contain UO2 pellets coated with a thin layer of
[.
[The IFBA stack height for the Region 11 fuel is 132 inches. There are a total of 8000 IFBA rods in Cycle 9, distributed throughout the core as shown in Figure 2.
Tab' I compares perunent design parameters for the various fuel regions in Cycle 9. Fuel rod design evaluations for the Wolf Creek Cycle 9 were performed using the NRC approved models described in Reference 9, and the NRC approved extended bumup desigt,ethods in References 12 and 13 to demonstrate that all of the fuel rod design bases are satisfiet Dunng fuel receipt inspection of fuel assembly L75 on 12/15/95, a shallow scar near the top end plug on the peripheral fuel rod in cell K-1 was noted. The scar is approxunately 0.1 inch long, oriented at a 45 degree angle to the horizontal, and begms e 2., S.3t approximately 0.25 inches below the top end plug weld zone. The depth of the scar was determined to be 0.0026 inches.
The impacu of this scar on the fuel stress analysis and fuel perf)rmance was assessed in Reference 22, and it was concluded that the scar will have no effect on the function of the rod.
The light duty experienced by the cladding in this region of the fuel rod would permit significant 7
rom er os.oos.ot rzv x INTTIATING DOCUMENT ENGINEERING DISPOSITION A
DISP.
y NO.:
N/A PAGE REY: N/A MA CCPIDCP Number: 06872 Revision: 0 Some of the fuel pellets may be coated with a thin layer of[ %,CThis[ ] coating is a,t Fuel Bumaele Absorber or IFBA.
The IFBA coating prov distnbutions and, in some cases, soluble the pnmary means of controlling power tent of the Reactor Coolant System. The helium back fill pressure for non-lFBA fuel rods is typically [
u psi, but may vary. The reduccon in backflll pressure for the IFBA rod hepsi. The helium bac offset the increased rod pressure at the end of life due to the production and release of helium from the{ng on the fuel pellets.
The purpose of the fuel rod is to sustain a nuclear reacton and generate heat.
It is this heat which is transferred by the reactor coolant system to the steam generators which produce the steam necessar the turbine generators for the produchon of electncety To guarantee adequate reactmty to sustain the nuclear reacDon, the pellets are ennched to a nominal level with more fissile uranium isotopes (U") which are mixed with the more naturally occumng vanety (U"). Many dessgn constraints are imposed on the fuel rod itself to assure the integnty of the rod against rupture, distortion and its continued ability to generated heat.
l i
l 5.2.
Incore Control Components Given the design objectve '.o safely and efficiently operate the core, additional tools besides the desi l
fuel assembly and the cere are available to be used at the designer's discretion These tools are generally
)
referred to as incore contret components. The purpose is implied by the name to provide controlled opersbon of the core. These componsnts are of two types:
j 1.
Vanable control components.
A hardware specific vanable control component is the Rod Cluster Control Assemblies (RCCA).
l l
Although it is not a hardware component, Boron in the coolant also provides vanable control of the reactmty in the core and assumptions on boration enter into the design of the core.
2.
Stationary control components.
Stationary control components allow the designer to tailor the power distnbubon, ensure a neutron 1
source to intbate the nuclear reachon chain, suppress local power generation, take core measurements l
or divert fluid flow.
These components conset of Bumable Pomon Rod Assemblies. Thimble Plug Assembles and Secondary Source Rod Assembles.
Regardless of the type of component, the guide thimble tubes are purposefully left open to receive these components. The camponents are al rod-like in shape are inserted into the guide thimble tubes of those assemblies where some control may be needed or desirable. Access is obtained through the top of the fuel assembly.
The incore detector instrumentabon tube is a unique component which does not er:tsr the fumi assembly through the top nozzle. The instrument tube, located in the center of the fuel assembly is dedicated for the l
purpose of accommodating incore detector instrumentabon The incere detector instrumentation tube is J
inserted up through the bottom nczzle of the fuel assembly into the instrumentabon tube. It is then connected to j
the vessel instrumentabon lines that pass down through the lower pienum support columns and out the lower hIad penetracons.
i 5.2.1. Rod Control Cluster Assemblies (RCCA)
This RCCA consats of a group of individual neutron absorter rods that are permanently attached to a common spicer assembly. The absorber matenal used in the rods is suver-inclurrsacmium (Ag-in-Cd) alloy or hafnium
~
Westingriouse Energy Systems j
Sectr!C Corporat!on n as i
% anew ia ma am i
May 17,1995
?,
95 SAP G 0029
{
Mr. W. Brad Norton i
Manager, Nuclear Engineenng Wotf Creek Nuclear Operaing Corpcraten Wolf Creen Generacng Station P. O. Box 411 i
Burlington. Kansas 66839 i
4 4
i
Dear Mr. Norton:
{
WOLP CREEK NUCLEAR OP,ERATING CORPORATION f,
WOLF CREEK GENERATING STATION LRCL TRANSMf77AL - WCGS CVCi.E 9 i
Please find anached, for input into your rescad desigt safety analyses, the L CheckNst (LRCL) for Wolf Creek Generating Stadon (WCGS) Cycle 9. This i bemg provided in accordance with Secton 4.4.6 of the Demgn interface Proc Wolf Creek Nuclear Operagng Corporanon (WCNOC) and W4.v.h.
l The LOCA Current Umrts are in essence the same as those tranemstted d reload with the following two changes:
1)
The minimum Fuel Rod Internal Pressures for Small Break LO Lower Rod internal Pressures may be more conservanve only in the absence of ro burst which occurs at very low clad temperatures.
2)
An additional limit has been added. Rod Power Census, due to the explicit evaluati of the 1% core wide Zlrealoy water reacdon requirements of 10CFR50.46. This rod power census is very conservatrve, and it should not be violated at any point along clave.
Ana#y. Watt Creek dentified in the Reload Demgn Irvtlallantion Checkist that a CCP and changeout has occumed. The LOCA analyms pose had not pnmously been notfled of ttis change or whelhnr it affects the WCNOC calculated ECCS f, L..ance used in the LCCA malysas. It wnB be assumed that WCNOC made appropriate assessments that no impact ECCS flows occurred as a result of this change.
m-
/
A N-N. M I GvO
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O___ Alof. Al6
Mr. W. Stac Norton 2
gg, The attached LRCL has been designated Westinghouse Propn roovest that you handle rt accordingty. Should you have any questio result, we comact me.
ems, please
~
Sincere #y,
,., _. a--
Jeffrey L Slater Project Eroneer Domestic Sh & Customer Projects JLS/sh Attachment cc:
G. J. Neises Surfington 1A J. S. Hseu Burtington 1A l
J y - ?f-C i- /
3' 2
PA3e. Az. o0 114
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WOLF CREEK CYCLE 9 i
LOCA RELOAD CHECKLIST GENERAL PARAMETERS This section includes a list of some of the general parameten used in the LOCA analyses, values listed are the limiting values supported by analyses or evaluanons.
NOMINAL OPERATING CONDmONS Rated Core Power 35d5 MW:
Thermal Design Flow 93.200 gptmloop Vessel Average Temperamre (operating window) 588.4'F (Nominal) 570.7'F (Reduced)
Design Core Bypass Flow Fraction
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RCS Pressure n50 2 50 psia Steam Generaar Tube Pluggmg 10 %
(mann=== in any or all generators) i i
l W
4 3
Past 49 of Alb
WOLF CREEK CYCLE 9 LOCA RELOAD CHECEUST 1.0 FLTL DESIGN CtJERENT RgsERENCE RELOAD N NCE PARAMETERS N
v4tyg 1.1 Fuel Pelles Tesipermare Table 1 Rd1.2 r,M A.V #A_gW e one.pran 2MA in-9r. iso
[fifa l.2 Fuel Rod lasernal Pressure Tahis 2 Rd t u,,
f 1.3 Stack Height Shrtakage
,*, 6
, a, e I
Rd 1. 2 J_
g 1.4 Fuel Type 1717 V5M R d 1. 2. 5
/7//7 vf' V
/,
5
/71/7 fth4 4$~X,a/orn/
2.8 CORE DESIGN CUREENT REFERING REIAAD REFERENCE PARAMETERS LDUT VALLs 2.1 LOCA F,(I)
Fim I Rd1.2 2.2 Emhalpy Rise Penhag Factor (fan) 1 65 Itat.1. 2 2.3 Hot Assembly Pombag Factor (P.BAR)
IA69 Raf.1. 2 Any devisses e a Assi assamtfy 6se thus dansribed try the PL1L TYPE perummer a its curreur timer rusuas in aa
-w psammer reqmreg a sedsry evemasess. This incanden. as is as tisemes a es fouowus t 3 mm asemanges vnh sensinas sent or Zhed aller roes or vasemass l} mew W 46 M M IM
- 3) ans at asw meursais u%e dN of k
Ib 4
- w R~ o
a t
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WOLF CREEK CYCLE 9 LOCA RELOAD CHECKLIST GN N CE RELOAD REFERENCE LBUT
2.4B Refmace Peakmg 2 50 Rd i Factor (Raf F,)
~
2.4C Fracuce of Hot Rod Integrasad Power for Each Third of ths Rod
- i. Raf P. (0-4 ft) 02581 Raf. I ii. Raf P (4-4 ft) 0.4835 Raf.1 iii. Raf P,(4-12 ft) 0.2584 Raf.I 2.3 Anal Offset @ 100% Power
+13% / 20%
Raf. t. 2 2.6 Rod Power Census Arm 2 Ref. I 2.7 Post LOCA Cancal Boros A zure 3 Raf.3 (ARO. No XI. Most Reacuve Thus in Cars Lifs. 64-212T) 2.8 RC3 Besse Concesuranos N
Raf.4 (Maa. 9 IUP (ppus))
3
,qN - U 06/
A ccC A/6
WOLF CREEK CYCLE 9 LOCA RELOAD CHECKLIST TABLE 1: FUEL AVERAGE Tk.MPERATURE LOCAL ROD POWER i KWE PEAK FUEL 'EMPERATURE (*F)
(BEST ESTIMATE)
(PLUS LHCERTABTQ
%, C.
i TABLE 2: FUEL ROD LVITJLNAL PRESSURES ROD AVERAGE POWER ROD INT 7.RNAL PRESSLM (KW/FT)
(PSIA) a, 6 b
p rJ kr O&s RuO
I WOLF CREEK CYCLE 9 LOCA REAI) CHEN 3.0 References i
1.
SEC SAI 3972 CO. Wolf Creek (SAF) Renmag BASH Analysis for V5H
)
2.
SEC SAI 3898-CO.
3.
l SEC SAI 4413-CO. " Wolf Creek Cycle 8 Reload".
}
4 SEC-TSA-3958-CO " Wolf Creek (SAP) Hot Leg Switchover Time for Pow 5.
THPL-89-477 " VANTAGE 5H Add.adism to Fuel Parameters DM j
RSAC Evaluation", and PDT-88 015, Vantage 5 Hybrid T/M Informanon for LOCA Aa -=.nts.
l I
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77s 9F>29, "Wdf Geek Un!) Qch. 9 f.es.a.d' p,,y,.,,-,y.R d.D./~ j'ez im -J m -i.ra
&n.uf ~.
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5 WOLF CREEK CYCLE 9 FIGURE 1: K(Z)
N sii 0.8 w9g 64.
De Q,f ElL 1[ZL REVATION
.E 2.5 1.0 0.0 4 d
2.5 1.0 6.0 h 0
2.31 0.925 12.0 h m.
z I
.8 0.4 i
E o* 0.2
~
0 0
2 4
6 8
10 12 Core Elevadon (ft) gy u a m f
u.9ga wo
J WOLF CREEK CYCLE 9 i
FIGURE 2: ROD POWER CENSUS i
y 1
xm u
v kam W
3o ch.
W
- g i
i
_m.
Wx 4
r Percent Rods Below Relative Power 9
A N-f r-26 / Tu 4 e er A/6
1 WOLF CREEK CYCLE 9 M
U FIGURE 3: POST-LOCA BORON. CONC.
2o 2,300 U
z O
~
g (2000, 2235) 2,200 2<
wI Qw 25 2,100 I
CL,I a
K 0 2,000 M
(1000,1978) dg Post LOCA (ppm) =0.257 x RCS (ppm) + 1721 J
d 1,900 0
1,000 1,500 2,000 a.
PRE-TRIP RCS BORON CONC. (PPM)
Y'N6$l to
NE. *DetmarJeg MYSIS FUIL AN TEptmAm bag i op' 2 GENERIC 17X17 VW4 ZZRLc psyg.! PEA ' Lb k-s q4 T235 (ggggg ; amMJP (tam /uTul R ( W/PT) Putt avgRAeg Trupgn,N (PLUS LSICERTAINT!gg) (0E4.7) 4, L t l l l l l l l t TMSE RESULTS OERIVED FIN 31 TM POLLOW13AR PAS ISat(S): han ID.Dete: ItAZf 00SS,10/30/M JM4GS98.10/30/ M J3847043.10/30/94 J3847002.10/30/94 { J3547140.10/30/ M TABLE p#tP4AfD VIA FTPOST R488: Run 10.Date: J3547188.10/30/M Am 4 F/ 9 % 2 7 7 i. I AIU~9T Vol EcvG II A H e 4' A ' 6 c, w a.:.: - o
l a., t, rust. Argueu reatm4TLuis emagazc o.:r74 on 1.EX traA 'adE18 ARLEast i e i Ttsa (MiamS ) 44, j p sulass (sam /trTU) i LDCAL j / POWER teot/PT) PWL AYg3Am 7EbptRAftat (Pus tacERTA!str!ES) ~ ( 53.p) 1 Ct & 3 i TMIE RESULTS WRIVED PW Tes PE&strDe PAD Est(S): Alm ID.Dete: 15150310.08/05/W AESSE35.5/W/gg .ammana'".05/ W /95 deesseBS.05/05/95 deeBSS88.05/W/W AESS710. W/ W/95 A SS8733. 5 /08/05 Jg588751.05/05/98 TABLE PREPAAED YTA PTFOST Maf: Aim ID aste: Jg588817.05/ts/05 TABLE E 2 ~ N -T5-lAI i s A /\\/ - 7 T C E / 2 e v. 2,.4 A IZ. o F 410
~. -. - i s . PUEL TElletRATuutt AssaLYS23 P1EL IRB INftmeAL 84Essas g'g Pam 1 0F t i MMR2C 0.274 m t.ER Ip3A saw agggggg , 71 5 (M) Q; G e I 01247 (IAS/IffU) i AVERAW L23SAR N IE/MI W IlfTEEe4L PREERRE (N!885 LSCERTA24fft23) {pgg3 t b i l i i ) i i j i I l 1 1 j flMn IS.anteT M B E IIE m aL T5 glERI VE B Pa s s n g PeL& a m e e m m as( 3 ): { .seess7?s.es/m/gg massess.es/as/g5 j ASS 4873.88/5/05 61.M/W/M mes4427.05/05/M mesasso.os/es/es 1, Dats PENEAfartc7. m artry,mpeas.g#LgXees.Se; l l i TABLE E 4 5+ _ f5'- / i / Afd-lT 0(oi K**e a a n fl' - - - - p c a
i e s i FutL f tMPCAAfunt ANALYSIS FUEL AVERA44 TEMPERATURS = SAIF 17X17 v3M t SX IPEA - ?!=E emi q, c BURAL'8 iWWD/WTU) LOCAL a POWER tKw/PT> PUEL AVERAGE TEMPERATURE ^ (OtG.P1 Q, b THESE RESULTS DERIVED PRoll THE FOLLOWINE PAD Mat (5): Ruri !D.Dete: SA!P1371,0s/14/gg TABLE PREPAAID VIA FTPOST sum: Rurt ID. Oste: J1112722,00/14/9g Talle lB g 14 2,,,pItfof 4Ib
.e .~ p '., PugL fggg>gAAfuat ApeALYS!$ SA!F t?It? VW4 1.52 IPSA .bsb Pugh AvgAAeg freestAATURE \\ , 72W (Masas) 0 (, l BufMJP ' taf0/ttfU) LOCAL POWER t KW/FT) PutL AyggAgg ygggggAyygg ( plWS L80 CERTA!9ff!ES) i0te.P) Q& l l l 1 l l l 7 MESS RESUL7S OSS!v50 Pagil TM POLLOWIM PAD RISf(S): Itun ID.Dete: SAIF1271,00/14/OS J1112987.00/14/05 J1113821.08/14/95 J1113084.08/14/95 J1113877.00/14/SS J1112SB2,00/14/SS J1112707,08/14/98 J1113707.00/14/98 TASLE PRSPASSB VIA F7 POST 8531: Sun 10,0ste: J1112722.00/14/03 J';,b[x 23 l \\ l ff. Y w l l j 6 %e A 'I$ f A '$ S
t . rust festaArung ANALYSIS paea t oF i FutL SueFACE TgwtAATUAE bb a SAIP 17X17 V9M 1.1x IPSA r!w !MounS1 q' SUSHJ9
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eowin e rw/PT) PUtL SuRPAct TEWERAWRE (OSS.P) 0 l 1 o THESE RESULTS DERIVED PRINI TM POLLOWING PAS EBHS): aun 10,0ste: SAIF1371.08/14/98 TABLE MtEPAARD v!A PTPOST MBf: nun 10. Oste: J1112722,00/14/98 g3 A M. T s~ CG I 5'O A i(o ,i Alb ~ I /b P' Y}}