ML17342B404
| ML17342B404 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/25/1986 |
| From: | Rossi C Office of Nuclear Reactor Regulation |
| To: | Rahe E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| Shared Package | |
| ML17342B405 | List: |
| References | |
| TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM TAC-45853, TAC-45854, TAC-48192, TAC-48193, TAC-62166, NUDOCS 8608280060 | |
| Download: ML17342B404 (16) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 E. P.
Rahe, Jr., Manager Nuclear Safety Department Westinghouse Electric Corporation Box 355 Pittsburgh, Pennsylvania 15230-0355 AUG 25 1988
Dear Mr. Rahe:
SUBJECT:
ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT WCAP 9561, ADDENDUM 3, REVISION 1
The Nuclear Regulatory Comission (NRC) staff has completed its review of Topical Report WCAP 9561, Addendum 3,'Revision 1, "Thimble Modeling in Westinghouse ECCS Evaluation Model," which was submitted with your letter dated July 24, 1986.
We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed.
The evaluation defines the basis for acceptance of the report.
We do not intend to repeat our review of the matters described in the report and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved.
- Our acceptance applies only to the matters described in the report.
In accordance with procedures established in NUREG-0390, it is requested that Westinghouse publish an accepted version of this report, proprietary and non-proprietary, within three months of receipt of this letter.
The accepted version shall incorporate this letter and the enclosed evaluation after the title page.
The accepted version shall include an -A (designating accepted) following the report identification symbol.
Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, Westinghouse and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or sub~it justification for the continued effective applicability of the topical report without revision of their respective documentation.
Sincerely,
Enclosure:
As Stated Division of PWR Licensing-A Office of Nuclear Reactor Regulation
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SAFETY EVALUATION ON CHANGES IN THE 1981 WESTINGHOUSE ECCS EVALUATION MODEL WITH BART Intloduction In a meeting held on June 23, 1986, Westinghouse met with representatives of the staff to discuss changes in their large break ECCS Evaluation Models.
These changes were necessitated by a modeling:change in the WREFLOOD code and an input error to the BART code.
Westinghouse estimated that inclusion of the modeling change and correction of the input error could result in an increased peak clad temperature of up to 120'F for ECCS analyses that were performed with the 1981 Evalua'tion Model with BART.
The corrections and remedial actions are describ'ed in Topical Report'CAP 9561 Addendum 3, Revision 1, "Thimble Modelinq in Westinghouse ECCS Evaluation Model," which was submitted in a letter dated July 24, 1986 (Reference 1).
The-model change in the WREFLOOD, code was required because the water volume which flows into the control rod guide thimbles during the core reflooding period following a large break LOCA was previously neglected.
This model change was found to have the effect of increasing the calculated peak cladding
'emperature by up to 20'F.
The BART change is an error in input which caused systematically -low values of hot assembly bundle power to be used by the code.
This error was found to have the effect of increasing peak cladding temperature by approximately 100'F.
To offset these
- changes, Westinghouse will include a portion of the heat transfer model calculated by BART in the peak cladding temperature calculation of the LOCTA code.
The combined effect of the thimbles on flooding rate and the error in the hot assembly power was determined to be offset by the benefit of including a portion of the heat transfer model.
A net benefit was obtained for plants with high peak cladding temperatures approaching 2200'F and a small penalty was obtained for plants with lower peak cladding temperatures.
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Thimble Fillin This issue affects the following Westinghouse ECCS Evaluation Models:
- 1) 1978;
- 2) 1981; and 3) 1981 with BART.
This safety evaluation applies only to the 1981 model with BART.
Westinghouse does not intend to correct the 1978 or 1981 models since they will not be used in the future.
The control rod guide thimbles are hollow tubes within the fuel rod bundles which replace 25 fuel rods in 17x17 fuel assemblies and 21 fuel rods in 15x15 fuel assemblies.
Control. rods are operated within 1/3 of the thimbles.
The thimbles may also contain burnable poison rods or in-core neutron detectors.
The thimbles contain small flow holes at the bottom to allow water to escape during control rod insertion.
During a large LOCA, the fluid in the thimbles will flash to steam.
During the reflooding period, the flow holes will allow water to reenter.
In evaluating core reflooding, thimble refilling was not considered by Westinghouse in the MREFLOOD -code.
This tends to be nonconservative since water which would otherwise enter the coolant channels would instead flow into the thimbles.
Although nonconservative, this simplifying assumption was considered to be reasonable in view of the relatively restricted flow into the thimbles compared to flow through the core.and considering the effect of the thimble hole plugging devices, During analysis of the effect of removal of the thimble plugging devices, it was found.that plug clearances were sufficiently large and core flow was sufficiently low during reflood that the assumption of no flow in the thimbles warranted reconsideration.
Subsequent work by Mestinghouse assessed the effect of the assumption of flow through the thimbles during the reflood phase - a phase in which core flow rates are significantly lower than during blowdown and during which refill of the thimbles would tend to be at the same rate as that of the core.
These subsequent studies indicated that a modeling assumption of flow in the thimbles
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during reflood would result in a slightly higher peak clad temperature calculation than would result from the assumption of no flow through the thimbles.
For the 1981 Westinghouse ECCS Evaluation Model with BART, thimble
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refilling has now been included in a conservative manner in which all thimbles are assumed to be empty at the beginning of the reflooding period and fill at the same rate as the core.
This is accomplished by including the total thimble volume in an existinq model of the WREFLOOD code which has been approved by the staff (Reference 2).
The effect of this conservative modeling of thimble filling during reflood is small (10 to 20 degrees F) because the more significant phemonena of liquid entrainment in the core and steam binding in the coolant loops would be unchanged.
Hot Assembl Bundle Power The hot assembly bundle power error in'BART resulted from a confusion between similar inp'ut requirements for BART which is utilized to. calculate heat trans-fer coefficients and LOCTA which is utilized to calculate peak cladding temperatures.
The LOCTA code evaluates the thermal behavior of a single pin and the fluid conditions in the adjacent hydraulic channel.
The hydraulic channel is defined by the hydraulic diameter which is a function of the wetted perimeter.
The wetted perimeter includes both heated rod surfaces and unheated thimble surfaces.
Tomccount for the effect of the unheated surfaces in computing channel enthalpy rise, the total number of fuel pins plus thimbles is input to LOCTA and utilized to calculate a total surface area and an average heat flux for use in the coolant enthalpy rise calculation.
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The BART code evaluates an entire fuel bundle including the thimbles.
In cal-culating the coolant enthalpy rise the BART code correctly utilizes a heated diameter in defining the coolant channel adjacent to a fuel rod.
Only the fuel rod perimeter is utilized to derive the heated diameter and not the thimble perimeter.
The use of an average heat flux for both rods and thimbles is therefore not required in BART.
Only the number of rods should be input rather than both rods and thimbles as in LOCTA.
The same input was utilized for both codes however.
This produced an under-prediction in enthalpy rise in BART and caused an over-prediction of the fuel rod heat transfer coefficient to be cal-culated and transferred to LOCTA.
The higher heat transfer coefficient caused LOCTA to calculate a peak cladding temperature that was too low by
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approximately 100'F.
The error is corrected by inputting the number of fuel rods into BART rather than the number of rods plus thimbles.
The staff concludes the correction is acceptable.
Heat transfer from the fuel rod surface to the fluid is a combination of con-vection and radiation.
Evaluation of data from the FLECHT reflooding heat transfer experiments has shown that radiation represents a significant fraction of the total heat flux.
Radiation heat transfer is a function of the fuel rod surface temperature to the fourth power and increases rapidly at elevated tem-peratures.
The radiation models'n the current BART code were reviewed and approved by the staff as discussed in Reference 2.
In transferring the fuel rod heat transfer coefficients from BART to LOCTA. for calculation of peak cl'ad-ding temperature, current Westinghouse ECCS evaluations have not included a
portion of the radiation heat transfer coefficient, Westinghouse deleted this I
heat transfer mode because it was thought to have only a small effect on fuel rod cooling and its deletion would provide additional. conservatism in their ECCS Evaluation Model with BART.
As discussed in Reference 1, the effect of including this portion of radiation heat transfer compensates for the iden-tified hot assembly bundle power and control rod thimble changes in the 1981 ECCS Evaluation Model with BART.
Since the staff previously approved the radiation heat transfer model (Reference 2), incorporation of this portion at this time is acceptable.
Conclusion As stated
- above, the NRC staff concludes that the changes to the 1981 Westinghouse ECCS Evaluation Model with BART, as described in Reference 1,
meet the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 and are, therefore, acceptable.
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References 1.
Westinghouse letter, NS-NRC-86-3147, from E.
P.
Rahe to J. Lyons,
- NRC, "Review of WCAP-9561-P, Addendum 3, Revision 1," July 24, 1986.
2.
NRC letter, from C. 0.
Thomas to E.
P.
- Rahe, Westinghouse Electric Cor-
- poration, "Acceptance for Referencing of Licensing Topical Report WCAP-
- 9561, BART'A-1:
A Computer Code for Best Estimate Analysis of Reflood Transients,"
December 21, 1983.
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Generic Letter No.'ubject Date of Issuance Issued To LIST ECENTLY ISSUED, GENERIC L ERS
'L 86-15 INFORMATION RELATING TO COMPLIANCE WITH 10 CFR 50. 49, "EQ OF ELECTRICAL EQUIPMENT IMPORTANT TO SAFETY" GL 86-14 OPERATOR LICENSING EXAMINATIONS 09/22/86 08/20/86 ALL LICENSEES AND HOLDERS OF AN 4PPLI CATION FOR AN OPERATING LICENSE ALL POWER REACTOR LICENSEES AND APPLICANTS GL 86-13 POTENTIAL INCONSISTENCY BETWEEN PLANT SAFETY ANALYSEB AND TECHNICAL SPECIFICATIONS 07/23/86 ALL POWER REACTOR LICENSEES WITH CE AND BLW PRESSURIZED WATER REACTORS GL 86-12 CRITERIA FOR UNIQUE PURPOSE 07/03/86 EXEMPTION FROM CONVERSION FROM THE USE OF HEU FUEL ALL NON-POWER REACTOR LICENSEEB 4UTHORIZED TO UBE HEU FUEL GL 86-1 1 DISTRIBUTION OF PRODUCTS IRRADIATED IN RESEARCH REACTORS Ob/2S/86 ALL NON-POWER REACTOR LICENSEES GL 86-10 IMPLEMENTATION OF FIRE PROTECTION REQUIREMENTS 04/28/86 ALL POWER REACTOR LICENSEE AND APPLICANTS GL 86-09 TECHNICAL RESOLUTION OF GENERIC ISSUE NO. B-59 (N-1)
LOOP OPERATION IN BWRS AND PWRS 03/31/86 ALL BWR AND PWR LICENSEES AND 4PPLICANTS GL 86-08 AVAILABILITYOF SUPPLEMENT 4 03/2S/86 TO NUREG-0933~
"R PRIORITIZATION OF GENERIC S4FETY ISSUES" ALL LICENSEES, APPLICANTS AND CONSTRUCT ION PERMIT HOLDERS GL 86-07 GL 86-06 TRANSMITTAL OF NUREG-1 190 REGARDING THE SAN ONOFRE UNIT 1
LOSB OF POWER AND W4TER HAMMER EVENT IMPLEMENTATION OF TMI 4CTION 05/29/Bb ITEM II.K.3.5, "RUTOMATIC TRIP OF REACTOR COOLANT PUMPS" ALL LICENSEES AND 4PPLICANTS RLL LICENSEES AND APPLICANTS OF CE DESIGN NSSS
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MEMORANDUMFOR DocKET No. 50 250/251 John Phillios Rules and Procedures Branch Division of Rules and Records Office of Administration I
October 22, 1986 I
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DISTRIBUTION
,Docket File w..o enclosure PADIII2 Rdg w/o enclosure DMcDonald w/enclosure DMiller w/enclosure FROM:
SUBJECT:
Office ot Nuclear Reactor Regulation TURKEY POINT NUCLEAR PO)fER PLANTS,. UNITS 3 AND 4 One signed original of the Federal Register Notice identified below is enclosed for your transmittal to the Office of the Federal Register for publication. Additional conformed copies (
6
) ot the Notice are enclosed for your use.
Notice of Receipt ot Application for Construction Permit(s) and Operating License(s).
Notice ot Receipt of Partial Application for Construction Permit(s) and Facility License(s): Time for Submission of Views on Antitrust Matters.
Notice of Consideration of Issuance of Amendment to Facility Operating License.
Notice of Receipt of Application for Facility License(s); Notice of Availability of Applicant's Environmental Report; and Notice ot Consideration ot Issuance ot Facility License(s) and Notice of Opportunity for Hearing.
Notice of Availabilityof NRC DraftiFlnal Environmental Statement.
Notice of Limited Work Authorization.
Notice of Availabilityof Safety Evaluation Report.
Notice of Issuance of Construction Permit(s).
Notice of tssuance of Facility Operating License(s) or Amendment(s).
Order.
Exemption.
Notice of Granting Exemption.
Environmental Assessment.
Notice ot Preparation of Environmental Assessment.
g oiaac Please arran e for oblication of this notice on I'Ionda, October 27 ~ 1996 in order to dalidate the date of November 10 1986 on the last a e. If for an reason this ublication date cannot be mdt ls. call Debbie Miller, x28362 Division of PHP, Licensing-A Office of Nuclear Reactor Regulation nclosure As stated contact:
D ~ MCOonal d Phone:
OFFICE>
SURNAME>
DATE>
lO/
86 NRC FORM 3IS 110/eo) NRCM 0240 OFFICIAL RECORD COPY.
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