ML18142A048
| ML18142A048 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 08/07/1984 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18142A047 | List: |
| References | |
| CAW-84-58, NUDOCS 8409060277 | |
| Download: ML18142A048 (51) | |
Text
rTTACH~'P.ff 11 Westinghouse Water Reactor Nuclear Technology Division Electric Corporation Divisions Box 355 Pittsburgh Pennsylvania 15230 August 7 , 1984 CAW-84-58 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regul atory Conmi ssi on Washington, D.C. 20055 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
REFERENCE:
Duke Power Comi:sny letter to NRC dated March 1984
Dear Mr. Denton:
The proprietary material for which withholding is being requested in the reference letter by Virginia Electric and Power Comi:sny is further identified in an affidavit signed by the owner of the proprietary information, Westinghouse Electric Corp:>ration. 1he affidavit, which accomµ3nies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addre~ses with specificity the considerations listed in paragraph (b)( 4) of 10CFR Section 2. 7CJJ of the Commission's regulations.
1he proprietary material fer which withholding is being requested is of the same technical type as that proprietary material previously submitted with application fer withholding AW-76-31.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Virginia Electric and Power Comi:sny.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-84-58, and should be addressed to the t11dersigned.
~~J Robert A. Wiesemann, Manager Regulatory & Legislative Affairs
/pj cc: E. C. Shomaker, Esq.
Office of the Executive Legal Director, NRC
e AW-76-31 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared Robert A. Wiesemann, who, being by me duly sworn according-to law, de-poses and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation ("Westinghouse") and that the aver-ments of fact set forth in this Affidavit are true and correct to the best of his knowledge, infonnation, and belief:
Robert A. Wiesemann, Manager Licensing Programs
- 1 Sworn to and subscribed befo m this - :.f day of --w--+---1976.
Nc-tary P
.:.iC 1.~ __ . , ~ . UJUNJ'f IIY QWMISSIOH EJU>IRE.S APR. 15, 1978
- e AW-76-31 (1) I am Manager, Licensing Programs, in the Pressurized Water Reactor Systems Division, of Westinghouse Electric Corporation and as s~ch, I have been specifically delegated the function of reviewing the proprietary infonnation sought to be withheld from public dis-closure in connection with nuclear power plant licensing or rule-making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Water Reactor Divisions.
(2)
- I am making this Affidavit in conformance with ttie provisions of 10 CFR Section 2.790 of the Comnission's regulations and in con-junction with the Westinghouse application for withholding ac-companying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse Nuclear Energy Systems in designating infonnatiQn as a trade secret, privileged or as confidential corm1ercial or financial infonnation.
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the C01TV11ission's regulations, the following is furnished for consideration by the Corrmission in detennining w~ether the in-fonnation sought to be withheld from public disclosure should be withheld.
(i) The infonnation sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
- e. e AW-76-31
~-
\
(ii) The 1nfonnation is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Westinghouse has a rational basis for detennining the types of infonnation customarily he_ld in confidence by it and, in that connection, utilizes a system to detennine-when and whether to hold certain types of infonnation 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, infonnation is held in confidence if it
.falls in one or more of several types, the release of which aright result in the loss of an existing or potential com-petitive advantage, as follows: -*
(a) The infonnation 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 consti-
. tutes a competitive economic advantage over other companies.
(b) It consists of supporting data, including test data, relative to a proces_s (or component, st~ucture, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved m;arketability .
e (c) Its-use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assarance of quality, or licensing a similar product.
(d) It reveals cost or price 1nfonnation, production cap-0 acities, budget levels, or cotrmercial strategies of Westinghouse, its customers or suppliers*.
(e) It reveals aspects of past, present, or future West-inghouse. or customer funded development plans and pro-grams of potential cotrmercial value to Westinghouse.
(f) It contains* patentable ideas, for which ~atent pro-tection may be desirable.
(g) It is not the property of Westinghouse, but must be .
treated as proprietary by Westinghouse according to.
agreements with the owner.
There are sound policy reasons behind the Westinghouse system which include the following:
(a) The use of such infonnation by Westinghouse gives Westinghouse a competitive advantage over its com-petitors. It is, therefore, withheld from disclosure*
to prn+~~t the Westinghouse competitive position.
-s- AW-76-31 (b) It_is infonnation which is marketable in many ways.
The extent to ~hich such infonnation is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the 1nfonnation.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d) Each component of proprietary infonnation pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary infor-mation, any one component may be the key to the entir'!!
puzzle, thereby depriving Westinghouse of a competitive advantage.
(e) Unrestricted disclosure would jeopardize the position
- of prominence of Westinghouse in the -world market, and thereby give a market advantage to the competition in those countries.
(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
- e AW-76-31 (iii) The information is being transmitted to the Cormnission in confidence and, under the provisions of 10 CFR Section 2. 790, it is to be received in confidence by.the Comnission.
(iv) The information is not available in public sources to the best of our knowledge and belief.
(v) The proprietary information sought to be withheld i*n this submittal is that which is appropriately marked in th~ attach-ment to Westinghouse letter No. NS-CE-1142, Eicheldinger to Eisenhut dated July 27, 1976 concerning reproductions of view-graphs used in the Westinghouse presentation to the NRC during the meeting on July 27, 1976 on .the subject of Westinghou~e Reload Safety Evaluation Methodology.
This infonnation enables Westinghouse to:
(a) Justify the design for the reloa"d core
'(b) Assist its customers to obtain licenses (c) Meet contractual requirements (d) Provide greater flexibility to customers assuring them of safe and reliable operation.
e*.
AW-76-31 Further,_this infonnation has substantial corrrnercial value as follows:
(a) Westinghouse sells the use of the infonnation to its customers for purposes of meeting NRC requirements for licensing doc~ntation.
(b) Westinghouse uses the infonnation to perfonn and justify analyses which are sold to customers.
(c) WestinghQuse uses the infonnation to sell nuclear fuel and related services to its customers.
Public disclosure of this infonnation is likely to cause. sub-
- stantial hann to the competitive position of Westinghouse jn selling nuclear fue1 and related services.
Westinghouse retains a marketing advantage by virtue of the knowledge, experience and competence it has gained through long involvement and considerable investment in all aspects of the nuclear power generation industry. In particular Westinghouse has develo~ed a unique understanding of the factors and parameters which are variable in the process of design of nuclear fuel and which do affect the in service perfonnance of the fuel and its suitability for the purpose for which it was provided.
e
-a- AW-76-31 In all cases that purpose is to generate energy in a safe and efficient manner while enabling the operating nuclear gener-ating station to meet all regulatory requirements affected by the core loa~ing of nuclear fuel. Confidence in being able to accomplisn this comes from the exercise of judgement based on experience.
Thus, th.e essence of the competitive advantage in this field lies in an understanding of which analyses should be performed and in the methods and models used to perfonn these analyses.
A substantial. part of this competitive advantage will be lost if the competitors of Westinghouse are able to use the results of the Westinghouse experience to nonnalize or verify their own process or if they are able to claim an equivalent under-standing by demonstrating that they can arrive at the same~or similar conclusions. Its use by a competitor would reduce his expenditure of resources or*improve his competitive
- position in the design and licensing of a similar product.
Thi~ infonnation is a product of Westinghouse design technology.
As such, it is broadly applicable to the sale and licensing of fuel in pressurized water reactors. The development of this infonnation is the result of many years of Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this process
I e AW-76-31 would require the investment of substantially the same amount of effort and expertise that Westinghouse possesses and which was acquired over a period of more than fifteen years and by the investment of millions of dollars.
I Further the deponent sayeth not.
e PAGE 1 ATTACHMENT 1 III. VEPCO RETRAN CONTROL SYSTEM MODELS DESCRIPTION/2UALIFICATION
PAGE 2 III. VEPCO RETRAN CONTROL SYSTEM MODELS DESCRIPTION/QUALIFICATION Vepco's RETRAM models make extensive use of the RETRAN cont%ol system modeling capability. The cont%ol system featu%e is used in the following
- 1. modeling ce%tain featu%es of the %eacto% p%otection system. Thes~ use signals which a%e gene%ated by the ope%ation of analog compute% elements on va%ious r
p%ocess ~ignals (e.g., the tempe%atu%e and ove%powe%
delta-T t%ips).
- 2. modeling ce%tain aspects of the %eacto% plant con-
- t%ol systems which may significantly influence the cou%se of a t%ansient (examples a%e the p%essu%iZe%
p%essu%e cont%ol system, the tu%bine gove%ne% valve Celect%ohyd%aulic) cont%ol system and the seconda%y steam dumps).
- 3. special submodels which calculate time-dependent bounda%y conditions 0% fo%cing functions which involve sevezal sequential a%ithmetic ope%ations.
The only application of this type which Vepco cU%%%ently makes is to a model to desc%ibe the t%an~po%t and mixing of bo%on in the RCS following a safety injection.
The pa%ag%aphs below desc%ibe the va%ious models, thei% development, use
PAGE 3 and qualification, where appropriate. Each model is also presented in terms of a block diagram which shows the interrelationships between variables and operations and also describes the interface between the control model and the rest of the system model.
Figures III-1 and III-2 show the overtemperature delta-T reactor trip and the overpower delta-T reactor 'trip, respectively. Normally, no credit is taken for the overpower delta-T trip feature, and the trip is disabled with a long delay on the correspo~ding trip card. The overtemperature delta-T logic calculates a delta-T setpoint based on measured average temperature and pressure. The final control block in the sequence differences the actual delta-T with the calculated setpoint. When the difference becomes positive, a reactor trip signal is generated (after an appropriate time delay to account for signal processing delays, etc.). The calculated setpoint conservatively reflects the various processing and setpoint errors. The model has been qualified by comparison of calculated steady-state trip setpoints to hand calculations, and by comparing the calculated time to trip during rod withdrawal transients with FSAR results and with alternate calculations.
Figure III-3 presents the pressurizer pressure control model used by Vepco.
The model represents a proportional-plus-integral controller, the output of which drives the pressurizer heaters and spray. The linear variation of spray valve position with controller output is modelled by a weighted
- summer. Spray flow rate is calculated from the valve position and the loop flow fraction, since the driving force for the spray is the dynamic head of
PAGE 4 reactor coolant in the cold leg. The controller output is also used to trip the pressurizer he.aters on and off, and to open and close one of the two pressurizer power operated %elie£ valves (the other valve is controlled directly from pressurizer pressure). The controller gain and time constant are taken from plant operating documents. The reference pressure is adjusted up or down during safety analyses as appropriate, to reflect steady state pressu~e measurement errors.
An example of a comparison of-a RETRAM calculated pressure response with the pressure control ~ystem assumed to be functional to FSAR results is shown in Figure 5.10 of the topical report. Comparisons with Vepco-generated results using an alternate method are presented and discussed in Section V of this supplement.
Figures III-4 and III-5 illustrate how the pressurizer pressure and steam pressure, respe~tively, are filtered before passing the signals to the reactor trip and engineered safeguards (safety injection) systems. The lead and lag time constants are best estimate values, taken from plant setpoint documentation.
Figure III-6 illustrates how the control system function generator feature is used to generate power feedback reactivity. This method of representing the reactivity feedback is used in situations where power is varying slowly enough that a defined relationship between power and fuel temperature exists. In most cases the independent variable is taken as the neutron power. For steam line break calculations, where the system returns to power
e e PAGE 5 from a subcritical condition, using neutron power as the independent variable could lead to calculational instabilities in the vi9inity of the initial power 'jump' following a return to power. For this reason, the heat flux is used as the input variable for steambreak calculations. For transients where the neutron power is varying rapidly (e.g., rod withdrawal from subcritical) the power reactivity concept is not applicable, and doppler feedback is_represented as a function of fuel temperature.
Figure III-7 shows how main st~am line isolation valve closure following a steam line break is modelled. This model allows the initial opening of the break and the closure of the isolation valve to be modelled at the same junction. The upper integrator simulates the opening of the break in 0.01 seconds. The lower integrator recloses the break path upon reciept of a signal from the trips which model the engineered safety features. The closure time is the maximum allowable value from the technical specifications.
Figure III-8 shows how control blocks are arranged to calculate a region-weighted moderator temperature for use in steam line break calculations. Since. point kinetics is used, consistent with vendor methodology, a radial moderator temperature weighting factor is used to approximate the effects of the coldest water entering the core region containing, a stuck rod. The function generator allows representation of a nonlinear variation of reactivity with moderator temperature.
Figure III-9 represents the core average heat flux calculation performed in
PAGE 6 the two loop model. This heat flux is expressed in texms of fraction of the rated fuil powex value, and is used fox editing purposes, and to dxive the powex reactivity feedback calculation descxibed in Figuxe III-6 during steamline bxeak calculations.
l few of the accidents which may xequixe RETRAH analysis axe affected by the tuxbine gove~nox valve Cox electxohydxaulic control-EHC) system. A simple contxol system model is used to desc~ibe the effects of this system on steam flow to the turbine; this model is shown in Figuxe III-10. The model assumes that steam flow is constant with decxeasing pressuxe until the govexnox valves xeach a full open position. Thereaftex, steam flow is assumed to decrease lineaxly with pxessuxe.
Certain best estimate calculations Ce.g. the analysis of the Hoxth Anna
"'""" cooldown event discussed in Section 5.3.3 of the topical xepoxt), xequixe a xepxesentation. of the secondaxy steam dump system. Figuxe III-11 shows the arrangement of contxol blocks used to calculate steam dump flow area as a function of average temperatuxe. Following a tuxbine trip, the steam dumps xapidly trip open to provide load xejection capability fox the system. The valves then modulate closed as the measured average coolant tempexatuxe decreases and appxoaches the no-load value. Values fox the no-load xeference temperatuxe, Txef, the filtex time constants T1 and T2 and the pxogxam fox dump capacity vs (Tavg - Txef) axe all taken from cuxxent plant
~ setpoint documents. Fox the Horth Anna cooldown event, initial post-txip cooldown rates calculated with this model agreed well with observed txends.
e PAGE 7 The RETRAH submodel for calculating the mixing and transport of high boron
'concentration water from safety injection into and around the primary coolant loops is shown in Figure III-12. The model shown is appropriate for full flow conditions in all loops. Pipe-like regions of the system are treated with delay control blocks. Plena are treated with a first order lag. The delay times and time constants are calculated from the nodal fill times for the various regions. Time dependent core boron concentrations obtained with this model agree reasonably well with results obtained with hand calculations and simpler, RCS-average mixing assumptions.
e PAGE 8 FIGURE III-1 OVERTEMPERATURE DELTA-T TRIP Y=1------
I I
V 100 .------.T::cip
. ---- *. bT Rated. ------------------->11 SUM 1->i:f
--Th----> l1SUMl-->I 1/ 1-->I X/ 1----------' .>l-1 l>O.O I .--> 1-1 I 1+t1SI I Y I I I Tc *----* *--* I I I I I
I
._I I
I I .----- .. 5 ----- .-----.K2 ----- ----- -----
1 '----->11 SUMl-->I 1/ 1-->11 SUMl~->l1+t3Sl-->l-1SUMl-->l1 SUMI ___________
'-------->11 l1+t2SI .>1-T::ce:fl l1+t4Sl.->IK1 I .>11 I
* 1* * . I . ____ . I I I I I 1_1 1 __ I I I
Gain= K3 I I
P------>11 SUM 1------>I-P:z:e:f ---------'
AT H 100 C1+t3S) r::cip i:f -------- > K1 - K2 ----------(Tavg - T::ce:f) + K3 CP-P::ce:f)
AT Rated C1+t4S)
e PAGE 9 FIGURE III-Z OVERPOWER DELTA-T TRIP Y=1----.
I I
V 100 . ----. T:rip
.----.~T Rated .------------------>11SUMl-->i£ Th->11SUMl-->I 1/ 1-->I X/ I__________ .--->l-1 l>O.O Tc->l-1 I l1+ts I I Y I I I
I I
.----- ** 5 ----- * -----. KS ----- * -----. I Th->11 SUMl-->I 1/ 1-->I d/ 1-->I '1/ l-->l-1SUMl-->I 1 SUMI_'
Tc->11 I l1+ts I I I dt I I 1+t3SI .->IK4 I 1-1 I I . -----* I .. ->. _____
I I I I 1_1 I I I I . -------. 1<6 I
'---->11 SUM I _____________ I 1------>I-T:re£ I
.ATx100 t3S l':rip i£ -------- > K4 - KS Tavg - K6 CTavg - T:re£)
~ T Rated 1+t3S
e PAGE 10 FIGURE III-3 LOGIC FOR PRESSURE CONTROL SYSTEM
.---> IT:r:ip Ho. 4(-4)
I IHeate:r:s on if <A~. off if >A'-*
I I
I IT:r:ip Ho. 9 C-9)
BIS TABLES---> 1---> IOpen Powe:r: Ope:r:ated Relief Valve t2 I I l>B,. Close if <B'-*
I I I
I I I T:r:ip Ho. 8 C-8 >
I I
!Open Power Operated Relief Valve 111 if >c; close if <C*
I .....
I 1/T Controller IPressurel---->I 1.SUMl---->I INT 1---->11. SUM l->-
Unit------> 1 I . ->I-PREF I I I I . --> 11. I I I . I 1.0------' I I
'I I
Gain=Full Spray/
I .-----.Max=1.0 ------ Loop Flow Spray Valve '-->11.SUMl---------->11.0MULl------------->IFilll Controller--> I -Yi** I . --> 11. 0 I I I t I -->Spray
~~-*Min=0.01 I I 1 !Junction C19 >
I I I I Gain=-1 Cold Leg Flow ~~' I
'-->11. 1-->IFill !Spray
- 1. 0---> I MUL I It I Intake 12 1(18)
- Setpoints A, Band Care best estimate values taken from plant setp~int documents.
- The parameter "Y" in the summe:r: block fo:r: the spray valve cont:r:oller is a measu:r:e of the diffe:r:ence between the pressu:r:e at which' the sp:r:ay valves begin to open and the :r:eference pressu:r:e. "T" is the pressu:r:e cont:r:oller :r:eset time con~tant.
e PAGE 11 Figu~e III-4 LOW PRESSURE TRIP SIGNAL
. I
I r
PRESSURIZER------------>! 1+T1S 1-------> TO REACTOR TRIP PRESSURE I I OH LOW PRESSURE I 1+T2S I T1 = LEAD TIME CONSTANT T2 = LAG TIME CONSTANT S = LAPLACE TRANSFORM VARIABLE TIME CONSTANTS ARE TAKEN FROM PLANT SETPOINT DOCUMENT LOW PRESSURE TRIP SETPOIHT IS THE SAFETY ANALYSIS VALUE (INCLUDES UNCERTAINTIES)
SEE ALSO SINGLE LOOP MODEL TRIP DESCRIPTION IM SECTION I.
PAGE 1Z FIGURE III-5 LOW STEAM PRESSURE SIGNAL I
I FAULTED LOOP-----------~>I 1+T1S I~------> TO SAFETY INJECTION STEAM PRESSURE I I LOGIC I 1+T2S I T1 = LEAD TIME CONSTANT TZ = LAG TIME CONSTANT S = LAPLACE TRANSFORM VARIABLE TIME CONSTANTS ARE TAKEN FROM PLANT SETPOIHT DOCUMENT LOW PRESSURE SET~OINT IS THE SAFETY ANALYSIS VALUE (INCLUDES UNCERTAINTIES)
SEE ALSO TWO LOOP MODEL TRIP DESCRIPTION IN SECTION 1.
PAGE 13 FIGURE III-6 POWER REACTIVITY FEEDBACK FUNCTION CORE AVERAGE HEAT FLUX----------->I I OR I FNG I NORMALIZED-----> X INITIAL------->! 1---->POWER --->TO POWER ~ POWER REACTIVITY KINETICS (GAIN FACTOR) ($) TABLES TABLE OF POWER REACTIVITY---------
VS. POWER IM$
SEE ALSO "DOPPLER POWER COEFFICIENT" DESCRIPTION IM SECTION IV-
"IHPUT OPTIONS"
PAGE 14 FIGURE III-7 SIMULATION or MAIM STEAM ISOLATION VALVE CLOSURE FOR STEAMLIHE BREAK CALCULATIONS
.------.GAIN=100 I IMAX=1.0 .-----.MAX=1.0 TRIP TO INITIATE------->! IMT 1------------------>1+1 I VALVE BREAK CO OR 1 > I I I SUM I ----> AREA
.---->l-1 I TABLE I .MIH=O.O I
HIGH STEAM FLOW/ I LOW TAVE MAX=-1.0 OR -->TIME--->MAIH STEAM------>I GAIN= 1/T1 HIGH STEAM YLOW/ DELAY ISOLATION I INT T1=CLOSURE TIME LOW PRESSURE SIGNAL CO OR 1)
SEE ALSO "MAIN STEAM ~SOLATIOH VALVES" IM SECTION II - COMPONENT MODELS AND TWO LOOP MODEL TRIP DESCRIPTION IM SECTION I - VOLUME AND FLOW PATH NETWORK DESCRIPTION.
rHIS LOGIC APPLIES ONLY TO THE "INTACT" LOOP DURING A MAIN STEAM LIME BREAK.
e PAGE 15 FIGURE III-8 MODERATOR TEMPERATURE DEFECT CALCULATION (TWO LOOP MODEL)
.------.GAIN1 ------
T113--->I I I I l<--T115 T114--->I SUM 1----------->I SUM !<------------------ SUM l<--T116 I REACTIVITY I I VS I IMOD TEMP I
.--v---. ------
' I '------>I I TO I SUM 1------------>I FMG !---->REACTIVITY
. ,. TABLES li\
.--1---. GAIN2 T213--->I I I I l<--T215 T214--->I SUM 1----------->I SUM !<------------------ SUM l<--T216 TXXX = MODERATOR TEMPERATURE IN VOLUME XXX GAIM1 = RMWF/4 GAIN2 = (1-RMWF)/4 RMWF = RADIAL MODERATOR TEMPERATURE WEIGHTING FACTOR SEE ALSO THE GENERALIZED DATA TABLE DESCRIPTION FOR MODERATOR TEMPERATURE DEFECT IN SECTION IV - INPUT OPTIONS, AND THE TWO LOOP MODEL CONTROL VOLUME DESCRIPTION IM SECTION I - VOLUME AND FLOW PATH NETWORK DESCRIPTION.
e PAGE 16 FIGURE III-9 CORE HEAT FLUX CALCULATION CTWO LOOP MODEL) 2101--->I I I I l<--2103 2102--->I SUM 1----------->I SUM I<------------------ SUM l<--2104
.--V---. GAIN I I I SUM 1-----> CORE AVERAGE HEAT FLUX FOR MINOR EDITS 2201--->I I I I l<--2203 2202--->I SUM 1----------->I SUM I<------------------ SUM l<--2204 2XXX = POWER TO WATER FROM CONDUCTOR XXX, BTU/HR GAIN= CONVERSION FACTOR, BTU/HR TO FRACTION OF RATED POWER SEE ALSO TWO LOOP MODEL HEAT CONDUCTOR DESCRIPTION IM SECTION 1
PAGE 17 FIGURE III-10 SIMULATION OF ELECTROHYDRAULIC TURBINE CONTROL SYSTEM
!GENERAL DATA TABLE I ISTEAM FLOW VS STEAM I
- !PRESSURE FOR FULL-OPENI
!THROTTLE CONDITIONS 1------>I I
.-->I FNG 1--. -------
I '-->I 1---->STEAM FLOW STEAM PRESSURE------------ .-->I MIN I FILL TABLE
I TIME------------------------->I 1--'
.-->I FNG I
I
!GENERAL DATA TABLE 1----
ISTEAM DEMAND VS TIME I SEE ALSO THE MAIN STEAM FLOW FILL TABLE DESCRIPTION IN SECTION IV.
PAGE 18 FIGURE III-11 STEAM DUMP CONTROLLER - BEST ESTIMATE ANALYSES
.-------.GAIN=.5 -------- --------
THOT-->I 1--------~>1+1 1---->I 1+T1S TCOLD->I SUM I TREF-->l-1 SUM I I ------
1 1+T2S FILTERED CTAVE-TREF)
I V
TABLE OF DUMP CAPACITY I VS CTAVE-TREF)------->I FMG I
.------.GAIN= 100.
I IMAX=1.0 '-------->I MUL TURBINE TRIP------>IIMT 1-------------------------->I SIGNAL CO OR 1) I I V
TO VALVE AREA TABLE THIS FEATURE IS NOT USED IN SAFETY ANALYSES, WHICH TAKE MO CREDIT FOR THE LOAD REJECTION CAPABILITY ASSOCIATED WITH STEAM DUMP. THE FEATURE IS USED IM SOME BEST ESTIMATE ANALYSES, SUCH AS THE ANALYSIS OF THE NO~TH AHMA COOLDOWM EVENT DISCUSSED IM SECTION 5.3.3 OF THE TOPICAL REPORT.
PAGE 19 FIGURE III-12 BOROM TRANSPORT MODEL Gain=100000000 Gain=1/60
.------. Max=1.0 -----
Integ:rated-->11 SUM 1------------->1 MUL Safety Inj. ---->IIMT 1--Volume ->I-Vpu:rgl .->I Flow, gpm I . .Min=O.O I I 1.0-' I 1---------------~------------I Gain=1/VBIT
.-->IIMT 1--->I XPO l---->IC2SUMl--->Bo:ron cone. exiting I .>I I . ->IC 1 I Bo:ron Injection
- Tank I "e"-' .___ I I I I 1. 0--' V V
.------*-----------------------------------<-------------* I I
.'>.--*-. 61.---.
IMULl--->ISUMl-->IDLYl-->ILAGl-->IDLYl->IDLYl->ILAGl6
- -->. __ . -->. __ . *--* *--* I * - - *
' 1 2 3 4 I 5 I* Bo:ron I *----------v-->at co:re I .---. ' .---. midplane
'-IDLYl<-ILAGl<-IDLYl<-ILAGl<-IDLYl<-ILAGl<----IDLYl7 13 12 11 10 9 8 Region Numbers: 1- cold leg mixing zone 7- hot leg 2- cold leg/downcome:r ~- steam gene:rato:r inlet plenum 3- bottom plenum 9- steam gene:rato:r tubes 4- bottom co:re 10- steam gen. outlet plenum 5- top coze 11- cold leg 1 6- outlet plenum 12- pump 13- cold leg 2 (pump outlet to mixing zone)
PAGE 1 ATTACHMENT 3 V. COMPARISON TO ALTERNATE CODE CALCULATIONS (NON-PROPRIETARY)
e PAGE 2 V. COMPARISON TO ALTERNATE CODE CALCULATIONS In the topical report CVEP-FRD-41>. Vepco provided numerous comparisons of transient results obtained with our RETRAN aodels to licensing results obtained by the HSSS/fuel vendor for Vepco's units. The latter were performed pximaxily to support the FSAR's and subsequent reload safety evaluations. This section provides a supplement to those compaxisons in the form of parallel calculations perfoxmed by Vepco using both a standard Vepco RETRAN model and a coxresponding LOFTRAN aodel. The LOFTRAN code is a proprietary code deve~oped and maintained by the Westinghouse Electric Corpoxation for use in performing general non-LOCA accident analyses. *vepco has had access to LOFTRAM for foux years via a special licensing agreement with Westinghouse. A detailed description of the LOFTRAN code is given in Reference V-1.
Vepco safety analysis engineers have undergone extensive training in the use of Westinghouse core design and safety analysis codes, including formal classroom instruction by Westinghouse (see Table V-1> and on-the job-training at Westinghouse and/or Vepco. Part of this training included a formal forty-hour non-LOCA safety analysis course which covered theory, input preparation and applications of LOFTRAN. Surry and North Anna specific models have been assembled in-house by Vepco.
The comparisons shown here were performed with a LOFTRAM model of the Surry reactors assembled by Vepco using the same data base used for development
PAGE 3 of the RETRAN models. Thus the basic plant geometric and thermal parameters are con-sistent for the two models. Initial conditions such as[
a,c were [
J a,c J~/om the use of [ J in the two codes to represent the equations of state for the coolant, etc.). Comparison of steady state conditions for the two codes are provided in Table V-2. Table V-3 provides a description of the three transients used in the comparisons. Discussions of the com-parisons are given in the paragraphs below.
' e I8AINING e WESTINGHOUSE DESIGN AND SAFEIY.ANALVSis CODES AND METHODOLOGY COUR~E NUMBER OF TOT~.L
- COURSE DESCRIPTION LENGTH PBE~ENTATIONf JRAI~IN~
INTRODU~TJON TO THE~ 3 DAYS 2 39 MAN-DA'1' COMPUTER SYSTEM BASIC PWR CORE PHYSICS 5 DAYS 1 35 MAN-DAY!
AND THERMAL HYDRAULICS
~THDOLOGY AND COMPUTER 5 DAYS 3 80 MAN-DAY~
MODELS FOR W DESIGN CODES WESTINGHOUSE DESIGN CODES 5 DAYS 2 60 MAN-DAY!
NUCLEAR DESIGN DEVELOPMENT 5 DAYS 3 115 MAN-DA' OF THE RELOAD SAFETY ANALYSIS CHECKLIST INTRODUCTION TO NoN-LOCA 5 DAYS 3 80 MAN-DAY SAFETY ANALYSIS RoD EJECTION, MAIN STEAMLINE 5 DAYS 2 LIO MAN-DAY BREAK, DROPPED Ron ANALYSIS WESTINGHOUSE THERMAL HYDRAULIC 5 DAYS 1 20 MAN-DA'1 METHODS
. WESTINGHOUSE LARGE BREAK 5 DAYS 2 30 MAN-DA' LOCA CODES (THEORY)
WESTINGHOUSE LARGE BREAK 20 DAYS 3 EO MAN-DA' LOCA CODES CoN-THE-JoB TRAINING)
TOTAL 559 MAN-DA'
e e PAGE TABLE V-2 COMPARISON OF RETRAN/LOFTRlM CALCULATED STEADY STATE CONDITIONS I
a,c Pa:ramete:r RETRAM Value Co:re powe:r, mwt 2489.82 -s Pump heat, mi.rt 12. 15 -c .
Tcold, or 547.11 Ca:fte:r pump) -S*
546.68 (be£o:re pump) -C*
Thot, °F 610.15 -c Tavg, 0 r 578.63 -c Steam Flow, lb/sec 3017.5 -s Steam P:ressu:re, psia 785.0 -s Steam gene:rato:r invento:ry, lbm 313200 -s Feedwate:r enthalpy, btu/lbm 413.69 -c Steam Enthalpy, btu/lbm 1199.7 -c Ave:rage :fuel tempe:ratu:re, 0 r 1405.7 -c L _J
'C' denotes a code calculated pa:ramete:r
'S' denotes a pa:ramete:r specified as input
- ~1 e PAGE 5 TABLE V-3 RETRAM/LOFTRAH Transient Comparisons Case Description 1 Reactor trip from hot full power followed by a turbine trip.
2 Turbine trip from hot full power. Mo credit taken £or direct reactor trip on the turbine trip. Pressurizer sprays, PORV's and steam generator relief valves are assumed'available.
3 Simultaneous trip 0£ all three reactor coolant pumps at hot full power. Mo credit taken £or reactor trip on pump undervoltage or under£requency. Pressurizer sprays, PORV's and steam generator relief valves are assumed available.
e PAGE 6 REACTOR TRIP Figu%es V-1 to V-4 show the xesults fox the xeacto% txip. Figuxe V-1 p%esents the co%e xesponse in te%ms of nuclear power, fuel temperature and co%e heat flux. As the results show, the core neutron and thermal kinetics models for the two codes give results which are[ J a,c Figure V-2 compares the steam generator response in terms of steam pressu%e and primary to secondary heat t%ansfe%, ox heat extraction, rate. The response of the reactor coolant system is shown in Figures V-3 (RCS a~erage temperature) and V-4 (pressurizer water volume and p%essu%e). The RCS average temperature responser L
temperature at a specific location is input Cin this case the cold leg) and the average temperature is then calculated based on the steady state initialization results. Zn Figure V-4, about [
J a,c
e PAGE 7 TURBINE Tit.IP Figures V-5 to V-8 show the comparisons for the turbine trip without direct reactor trip. The results are shown out to the time of reactor trip, and present steam pressure, reactor inlet temperature, reactor power and pressurizer pressure, respectively. Figure v-7 is of interest in that it shows a slight difference in the nuclear power ~esponse. This difference stems from a different treatment of power reactivity feedback in the two aodels. The LOFTRAM model generates power feedback as a function of core heat flux. The lt.ETRAM model, on the other hand, u~es a tabular representation for the power feedback which relates the feedback directly to neutron power. Since the reactivity feedback is more accurately a function of the fuel temperature, [
'!'he [
a,c
] in the two models. Vepco's lt.ETRAM aodels treat the steady state pressure error as a bias in the signal going into the proportio*nal plus integral contoller which controls pressurizer spray and one of the two pressurizer power operated relief valves. Thus spray and one PORV are assumed to open about 30 pc.i below their nominal setpoints.[
a,c J Since the spray and one PORV are actuated
[ J a.,c 1n the lt.ETRAM model, a [ J a,c pressure[ J a,c results. For
PAGE 8 safety analyses zelated to system overpressure and vessel integrity concerns. pressurizer PORV's and spray are assumed not to function. and a,c this modeling[
Jon the results.
PAGE 9 FLOW COASTDOWM Figures V-9 to V-11 show comparisons for the flow coastdown event. Total core flow Cthis is a three-pump coastdown) is shown in Figure V-9. LOFTRAM uses a lumped parameter approach in solving for loop flow (the rate of change of flow is a characteristic of the entire coolant loop), whereas RETRAM solves a momentum equation at every flow junction in the loop. For a,c incompressible flow, the two models give [ J results, as shown.
Figure V-10 presents the nuclear power and core heat flux repsonse, and pressurizer presssure. response for the two codes is presented in Figure
, a,c V-11. The [ .J :following the trip is related to
[ ]~~ray is driven by the dynamic head o:f the reactor coolant flowing through the loops. In the RETRAM model, under flow coastdown conditions, spray :flow is assumed to be proportional to loop
- flow. In the LOFTRAH model, the spray :flow is assumed proportional to the square of the loop flow. Thus under loss of flow conditions LOFTRAM[
a,c J in the transient.
e PAGE 10 CONCLUSIONS TEansient Eesults fEom the Vepco RETRAH aodels have been compaEed to
~ ,c Vepco-geneEated results using the tOFTRlK code. The[ ] in Eesults is [
a,c
-] in the codes.
REFERENCE
- 1. Burnett, T. \*l. T., et al, "LC:-T?l.\; Cl'1rE !iESC':l.IPTIC'r*:," '!CP.P-7907-P-.A. (Hestin~house Proprietary Class 2), \*!C,1.P-7:G7-A (Westinr:,hou~e ~!on-Pro!"'rietary), Jl.pril 1~84 .
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