ML20148J494
| ML20148J494 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/29/1974 |
| From: | Heider L YANKEE ATOMIC ELECTRIC CO. |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8011240179 | |
| Download: ML20148J494 (5) | |
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YANKEE ATOMI YLEftRIC COMPANY y
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Attention: Directorate of Licensing
Reference:
1.
Yankee Atomic Electric Company Proposed Change No. 115 Core XI Refueling. License No. DPR-3 (Docket No. 50-29) 2.
AEC letter dated May 13, 1974 requesting reports concerning Proposed Change No. 115.
3.
Telephone conversation with Mr. Fred Burger on April 26, 1974.
Dear Sir
- i Reference No. 1 transmitted revisions and additions to sections of the Technical Specifications. Three of the references in the document were Yankee Atomic Electric Company reports that had not been published at the time of submittal. Two of these reports are now available and the publi-cation dates have been added to the references No. 1 and No. 6 on the attached revised draft page 400:05.
The third report was intended to be an internal engineering report describing the use of the FLASH-4 model as applied to Yankee Rowe. The equations and assumptions used are described in the FLASH-4 manual (Reference 5, Page 400:05 of Reference 1).
Since this information is now available, we are deleting the reference describing the model application. We are amenable to a meeting to discuss this report or any other aspects of Reference 1.
As discussed in the referenced telephone conversation and tequested in your letter, we are enclosing the following reports:
1.
A. E. Ladieu, "TEMTRA-A Digital Computer Program for Thermal Transient Analysis of a Nuclear Reactor Core Channel",
YAEC-1031 (August 1969).
2.
A. E. Ladieu, "A Thermal-Hydraulic Analytical Model Using COBRA-3C", YAEC-1058 (May 1974).
I THIS DOCUMENT CONTAINS
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S/29/74.
DRAFT 16.
-R. C. Martinelli and D. B. Helson, " Prediction of Pressure Drops During Forced Circulation Boiling of Water", Trans. ASME, 70, 695 (19h8).
17.
J. R. S. Then, " Prediction of Pressure Drop Laing Forced Circulation Boiling -of Water", Intern. J. Esat and Mass Transfer, 7, 709.(19ih).
13.
E.' Quandt, " Analysis of Gas-Liquid Flew Fatterns", A. ~ Ch. E. Preprint h7, 6th National Heat Transfer C:,nference, Scsten (1963).
~19 F. E. Tippets, " Critical Heat Pluxes and Flow Patterns in High Pressure-Boiling Water Flows", AS'E Paper 62-WA-162. (1962).
- 20.. L. S. Tong, " DUB Prediction f:r an Axially Non-Unifom Heat Flux-Distribution", WCAP-538h (1?63).
21.
L. S. Tong, R. W. Steer, A. H. Wenzel, M. Bogaardt, and C. L. Spigt,
" Critics.1 Heat Flux on a Heater Rod in the Center of Snooth and Rough Square Sleeves, and in Line Contact with an Unheated Wall", ASME Paper, 67-WA/HT-29 (Novenber, 1967).
22.
R. O. Sandberg, " CAT-II-An IEM 7C90 Code for Predicting Thernal and
. Hydraulic Transients in an Oper-lattice Core", WCAP-2059 (1962).
23.
A. E. Ladieu, "A Thernal-Hydraulic Analytical Model Using COBPl3C" YAEC-1058 ()tay-1974).
+
2h.
Change No. 97 to licens? No. Dr?.-3, Docket 50-29
- 25.. Rosal, E. R., J. O. Ce=ak, L. S. Tong, " Rod Bundle Axial Non-Unifom Feat Flux DNB Tests and Data", UCAP-7h11-L Rev. I, May 1970, Westinghouse Troprietary Class II.
26.
" Technical Report on Densificnion of Light Water Reactor Fuels", Regulatory Staff, U.S. Atenic Energy Cennission, Novenber lh, 1972.
27.
S. Glasstone and A. Sesonski, " Nuclear Reactor Engineering", D. Van Nostrand, ;
Co., 1967.
28.
L. S. Tong, H. Chelener, and E. A. McCabe, Jr., " Hot Channel Factors for Flow Distribution and Mixing in Core Thernal Design", WCAP-2211 (1963).
29.
D. S. Powe and C. W. A.gle, " Crossflow Mixing Between Parallel Flow Channels During Boiling - Part II'- Measurement of Flow and Enthalpy in Tso Parallel Channels", EDTL-371 Pt. 2 (Decenber,1957).
4
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-t h00 :03 S/29/74 DRAFT for ute.of.conssrvativ317 large 'versus small coefficient values att treated
~en an eventiby event basis.
The. values used in the transient analysis are given in Table: LOC-3 The valuas'of moderator tiemperature and pressure ~
fec lbac::' eccf Cicient praconted 'includa the effects of calculational-uncer-tainties.: As these. coefficients vary with' changes in coolant _ temperature and '
- inserted rod : worth. the most conservative, but. applicable,.value is used. The
- nominal value of ths fuel. temperature coefficient at rated power is presented
~
in Table hCC-3 and the manacr of application of the' uncertainties is provided
-in the: discussion-of each event..
The effective neutron lifetime and-delayed neutron fractions are functions of fuel.burnup. The calculations have been performed with a neutron lifetime of 18 microseconds (minimum value during core life) and with the delayed neutron fractions shown in Table h00-h.
ANALYTICAL METHODS Summaries of the principal cenputer codes used in the' transient analyses are given below. Specialized codes in which the modeling has been developed to simulate only One accident, such as the SATAN-7 code used in the Loss-of-Coolant Accident analysis and which consequently have a direct bearing or the analysis
- of the accident itself, are summarized in. their respective accident analytas sections. The codes used in the analyses of each transient are indicated in the d$saussion of each accident.
Sincie leop Svstems Analysis Frocra:a (GDID1 II) l The GFMINI-II program sinulates the transient response of a pressurized water reactor system to specified perturbations in th'e process parameters. A multi-loop system is sinulated by a lenped parameter single loop model con-taining the reactor vescel, het and cold leg piping, steam generator (tube and shell sides) and the pressurizer (including heaters, spray, relief and safety valves).
The core model inelades a point neutron kinetics simulation with moderator, fuel and scram rod reactivity feedback effects incorporated.
The steam generator secondary side utill:es a homogeneous saturated mixture model for the thermal transients. The reactor protective systen simulation includes reactor and turbine trips en neutron flux, high pressurizer pressure and level, low ficw and low pressurizer pressure.
The centrol systems simulated include steam dump and bypass, feedwater control and p:mssurizer pressure and level control.
Details of the GOD'I-II nodel are discussed in ?.eference 1.
Multi-Loon Systens Analysis Fregram (FLASH-h)
The FLASH _ (Reference 2 ) digital computer program was initially deve' ped to,
-calculate. coolant flows, inventories, pressures and temperatures in a water reactor system during a loss of coolant accident in order to provide a basis for predicting the consequences of such an accident and the effectiveness of the safety systems designed to mitigate these consequences.
Subsequent improvements wore made to the code (References 3. b and 5) culminating in the FLASH-h computer program.
The FLASH-h code solves the conservation
(:
equatiens of energy, nass and mcmentum for a general nodal representation of a water reactor plant daring transients.
Because of this generality, a multi-loop pressurized water reactor nodel was developed to simulate the response of the plant when the conditions in cach leep ara not identical during the transions.
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Core "inetics Pr:rr,.r. (CEIC-KIU)
The CHIC-T.IH pregram is used to analyze fast and intermediate reactivity or flow tra.1sienr.s in a witar-cocied he teroceneous nuclear reactor.
The code calculates
.the pen r, tempe -'" -a e anc internsi pressure surges when control rod motion, inl:t.emperatura, it.le t f ic. and system pressure are known functions of t.Lte.
The validity of.hi,e analytical tool has been verified by cenparisons with SPEET Ti ex;2rinentai transiento.
The details of the codel are provided in Raforence 7.
Hot Channel An:1*rsis Procram (COBRA-3C)
The COERA-3C program is discussed in Section 102.
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h00:05 d
o 5/29/74 DRAFT REFERE3CES 1.
T. R.. Hencey, "G3 INI-II, A Modified Version of the GDIINI Computer Precram", YAIC-10sS,. (April 1974).
2'.
S. G.-Margo 11a,'et, sl., " FLASH: : A Program for Digital Simulation of the Loss. of Coolant Accident";.WAPD-DI-53h, May 190o.
I L3.
J. A. Redfield, - et. a1., " FLASH-2 : A FOPCRAN 'IV Program for the Digital Sinulation of a Maltincdo Reactor Plant Nrin=. :Less,of Coolant",
WAPD-DI-665, April 1967.
h.
J. H.. Murphy, et. al., " FLASH-3 : A FORTRAN IV Program for the Simulation of Pteactor Plant Transients in Space and Time", WAPD-TM-800, July 1968.
t 5
T. A'. : Persching, et. al., " FLASH-h : A Fully Implicit FCETRAN IV Program
[
for the Digital Sinulati:n of Transients in a: Reactor Plant",
WAPD-DI-Sh0, March 1969 6.
J.- A. Redfield, CHIC-KIN -- A FORTRAN Program for Intermediate and Fast-
"ransients in a Water Moderated Reactor", WAPD-TM-479, January 1965.
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AEC DIST 'iUTION FOR PART 50 DOCKET MATET~ q (TEMPORARY FORM)
CONTROL NO: 4916 FILE: Oh FROM:
DATE OF DOC DATE REC'D LTR MEMO RPT OTHER Y:nkee Atomic Electric Co.
W:stborough, Mass. 01581 5-29-74 6-3-74 X
L. E. Heider TO:
D. L.
3 signed SENT LOCAL PDR X
CLASS UNCLASS PROP INFO INPUT NO CYS REC'D DOCKET NO:
XXX XXX 40 50-29 DESCRIPTION:
ENCLOSURES:
Ltr notarized 5-29-74, re our 5-13-74 ltr, (1) A Digital Computer Program for Thermal r questing proposed change #115, Suppl #1 to Transient Analysis of a Nuclear Reactor ths Opr Lie (chge to Tech Specs) regarding Core Channel Core XI Re fueling.......w/atchmt.........
(2) A Thermal-Hydraulic Analytical Model Usin; trrns the following:
Cobra IIIC DO NOT REMOVE (3) Gemini-II, A Modified Version of the Gemir Computer Program PLANT NAME: Yankee ggg
( 3 Orig & 37 cys rec'd )
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