ML20205G159
| ML20205G159 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 01/20/1987 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19292H011 | List: |
| References | |
| CAW-87-004, CAW-87-4, NUDOCS 8703310386 | |
| Download: ML20205G159 (54) | |
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Westinghouse Water Reactor ' uclear N Techr@h/ Divis6; Electric Corporation Divisions , 333 Pittsburgh Pennsylvania 15230-0355
- January *2, 1907
. CAW-87-004 e]
je Dr. Thomas Murley, Director - ~
Office of Nucleae Reactor Regulatien'
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U.S. Nuclear Reg 11atory Comnission 3' -
1 Washington, D.C. 20555 - i
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APFCICATION FOR WITHHOLDING PROPRIETARY -
INFORMATION FROM PUBLIC DISCLOSCHE '
i
Subject:
Re'eponses to NRC Review Questions on Byror/Braidwood RTD Bypass Elimination
Dear Dr. Murley:
The proprietary material for which withholding is being requested in the enclosed letter by Connonwealth Edison is further identified in an affidavit si hed S by Ene owner of the proprietary information, Westinghouse Electric Corporation. The affidavit, which accompanies this. letter, sets forth the basis on which the infonnation may be withheld frm public disclosure by the Commi.? zion and' addresses with specificity the considerations listed in paragraph (b)(4) of L10CFR Section 2 790 of Qie Commission's regulationsr ;- <
1he proprietary matei41al for which withholding is bei.ng requested is of the game a4 '
technical type as that proprietary material previous 1'y submitted with Application for Withhql. ding Mi-76-060. p,
+1 Accordingly,' this lettdr authorizes the utilization of themccmpanying afficavit by
- Commonwealth Edison. ;
Correspondence with respect to the proprietary aspects of the a) plication for withholding or the Westinghouse affidavit should reference this letter, CAW-87-004, and should be addressed to the undersigned.
Very truly yours, ,
D
~ ncx obe t A. Wiesemann, Manager
/dmr .. ulatory & Legislative Affairs s 9
Enclosure (s) I f
cc: E. C. Shomaker, Faj.
, Office of the General Council,. NRC r [
. 8703310386 970320 -
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PROFP.HTARY INFORMATION WOTICE ,
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t-5 TRANSNITTG i'DEWITH ARE PROFRIETARY AND/0R NON-PR D0QJMENIS ."URNISHED TO THE NRC IN CONNECTION WITH REQ
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PLANT SPECIIIC REVIEW AND APPRWAL.
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l Ir CCER TJ CONFORM 70 THE RWUIRDENTS W 10CFR2.790 W THE C
' EcDLATIOLS CONCERNING THE P30TT.CTION & PROPRIETARY INF IT3 T5E WRbs' 11'E INFDPHa. TIC 6 WIIbH IS PROPRIE "0N A1NED WIDlIN BRAQ2TS AND WHERE THE PROPRIETARY BEEN INFORMATI DII.ETE3 IN THE NON-PROPRIETARY VERS'ONs GLY THE BRACKETS RDi&IN, THE -
INT 0!MTION 1 HAT NAS CONTAINED WITHIN T.iE BRACKETS IN TH H.iVING BEEN DEI.E"ED. THE JUSTIFICATION FDR Q. AIMING THE
' DESIDNATED AS P1.0PRIETARY IS INDICATED IN BODI VERSIONS -
LETTERS (a) TR'(00GH (S) CONTAINED WITHIN PAREmiESES LOCA f IMMEDIATII.Y Fru. WING THE BRACKETS ENG.05ING EADI ITEM OF IN i
IDENIIFIED A3 TA0PRET/.RY OR IN THE MARGIN OPPOSITE THI3E SJCH INFOR
- LWIP ,CLSE 1 ff7ERS REFER 10 THE TYPES & INFOT.MATION WESTINGHOUSE CUS1D I? OLDS IN CCf /IDENCE IDENTIFED IN SECTIONS (4)(ii)(a) through (4)(ii)(g) 0F THE APTIDRIT .CCOMPANYING 7)!IS TRANSMITTAL PURSUANT TO 10CFR l k y e
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l AW-76-60 l
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. s AFFIDAVIT-COMMONWEALT'H OF PENNSYLVANIA: s .
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, ' 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 N
, C.' of W'estinghouse 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, information, and belief:
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., Licens.ing Programs
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. Swornw to and subscribed .
beforemethis[.. day 7
ofJYfc*tixb4N 1976.
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AW-76-60 (1) I am Manager, Licensing Programs, in the Pressurized Water Reactor Systems Division, of Westinghouse Electric Corporation and as such,
.I have been specifically delegated the function of reviewing the proprietary information sought to be withheld frcm public dis-closure in connection with nuclear power plant licensing or rule-making proceedings, and am authorized to apply for its withholding bn behalf of the Westinghouse Water Reactor Divisions.
(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.790 of the Commission'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 Huclear Energy Systems 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 in-formation sought to be withheld from public disclosure should be
, withheld.'
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(i) The information sought to be withheld'from public disclosure is owned and has been held in confidence by Westinghouse.
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(ii) The information is of a type customarily held in confidence by
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Westinghouse and not customarily disclosed to the publi'c.
Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that.
connection, utilizes a system to determine when and whether to .
hold certain types of information in confidence. The ap-plication 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 com-
/
! petitive advantage, as follows:
(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.
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Its use by a competitor would reduce his expenditure e (c) of resources or improve his competitive position in the design, manufacture, shipment, installation, ass 0rance of quality, or licensing a similar product.
(d) It reveals cost or price information, production cap-acities, budget levels, or commercial 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 commercial value to Westinghouse.
/ (f) It contains patentable ideas, for which patent pro-tection may be desirable. .
(g) It is not the property of Westinghouse, but must bo treated as proprietary hy 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 information by Westinghouse gives Westinghouse a competitive advantage over its com-
- - petitors. It is,* therefore, withheld from disclosure to protect the Westinghouse competitive position.
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AW-76-60 (b) It is information which is marketable in many ways. e The extent to iwhich such information is available to competitors diminishes the Westinghouse ability to -
sell products and services involving the use of the information. .
(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 information pertinent to a particular competitive advantage is potentially as valuable as the total ccmpetitive advantage. If
/
I competitors acquire components of proprietary infor-mation, any one component may be the key to the entire ,
puzzle, therehy 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.
9
AW-76-60 (iii) The information is being transmitted to the Commission in <
confidence and, under the provisions of 10 CFR Section 2._790, it is to be received in confidence by the Commission. '
(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 in this sub-mittal is that which is appropriately marked in the attach- c ment to Westinghouse letter number NS-CE-1298, Eiche1dinger to Stolz, dated December 1,1976, concerning information relating-to NRC review of WCAP-8567-P and WCAP-8568 entitled, " Improved
, Thermal Design Procedure," defining the sensitivity of DHB
/ ratio to various core parameters. The letter and attachment are being submitted in response to the NRC request at the October 29, 1976 NRC/ Westinghouse meeting.
This information enables Westinghouse to:
(a) Justify the Westinghouse design.
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(b) Assist its customers to obtain licenses.
(c) Meet warranties. .
(d) Provide greater operational flexibility to customers assuring them of safe and reliable operation.
(e) Justify increased power capability or operating margin for plants while assuring safe and reliable operation. .
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i 'f *y AW-76-60 (f) Optim'ize reactor design and performance while maintaining a high level of fuel integrity. ~
Further, the information gained from the improved thermal design procedure is of significant commercial value as follows: .
(a) Westinghouse uses the information to perform.and justify analyses which are sold to customers.
. (b) Westinghouse sells analysis services based upon the~ ,
. experience gained and the methods developed.
Public disclosure of this information concerning design pro-JI cedures is likely to cause substantial harm to the competitive '
position of Westinghouse because competitors could utilize this information to assess and justify their own' designs without commensurate expense.
The parametric analyses performed and their evaluation represent
- a c6nsiderable amount of highly qualified development effort.
This work was contingent upon a design method development pro- ,
gram which has been underway during the past two years.
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Altogether, a substantial amount of money and effort has been expended by Westinghouse which could only be duplicated by a
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competitor if he were to invest similar suns of money and pro-j vided he had the appropriate talent available.
Further the deponent sayeth not.
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9 t- , WESTINGHOUCE CLASS 3 l
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Response to Question is The Westinghouse Setpoint Methodology calculates all cT the various uncertainties and values pertal.ning to a reactor trip. The basic methodology of error combination is the same as that noted ,lp the Virgli C. Summer setpoint study, performed by Westinghouse. However, there have been some modifications in the uncertaintles used and the i calculations to reflect the use of more than one TH RTD and the removal of the RTD Bypass piping.
For the RTD Bypass piping elimination, an effort was made to remove conservatism from the calculation of temperature uncertaintles, e.g.,
the convolution of TM and Te RTD errors and the convolution of the R/E errors; instead of the ultra conservative arithmetic summation for the two areas. When this decision was made It was felt necessary to revalidate the models of the Overtemperature Delta-T COTDT) and Overpower Delta-T (OPDT) functions with regard to what was
- , the driving function for temperature uncertaintles. In the Virgil C.
Summer setpoint study the assumption was made that Tave is the predominant temperature function. However, through a sensitivity study, explained below, it was determined that Delta-T is the driving V function. This results in a restructuring of the OTDT and OPDT uncertainty calculations which now calculate somewhat smaller values for the protection function Total Allowances.
A careful examination of the uncertalnty calculation reveals the following areas of conservatism in the original versiont
- 1) the temperature streaming uncertaint'y is set at C 3 a.c for
.* the hot leg and C 3 a.= for the cold leg for an average of C 3 a.c. However, based on work performed by Westinghouse,
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- the streaming uncertainty has been reduced to random and systematic components which are different based on the us'e of RTD Bypass piping C 3 a.= or thermowells in l the RCS pjp.ing, C 3 a.c. ,
l 2) the value used for the Power Calorimetric uncertalnty is based on the generic assumption of C 3 a.c. This can be removed entirely based on the shift from Tave to Delta-T.
- 3) the arithmetic summation of the uncertainties for Tw and Tc (SCA and SD). Since the RTDs are independent devices, the uncertainties can be treated as independent avantitles, l.e,
((SCAw + SDw )m + ( SCAc + SDc )a y s ea.
l 4) the arithmetic summation of the R/E uncertainties for Tu and Tc. These devices are calibrated independently of each other, l therefore they can be treated as independent quantitles.
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- 5) the double treatment of R/E uncertainties (one set for Tave and one set for Delta-T) even though there is only one set of R/Es in the circuit. The calculations now place R/E uncertainties in the Delta-T calculation only, based on the sensitivity study. ..
For a plant with the RTD Bypass piping intact, review of the process l block diagrams for the OTDT and OPDT protection functions results in the verification that only one Ts and one Tc RTD feed through.
t corresponding R/Es to a function generator, which calculates Tave and
. Delta-T. Therefore, only one set of RTD and R/E uncertainties (Tave or Del ta-T,) needs to be accounted for in the uncertainty analysis if L lt can be determined that a ilmiting set exists. Evaluating the impact of an error in TH or Tc, or both, on Tave and Delta-T results in the conclusion that, at the extremes for the uncertainties, only one of the two parameters (Tave or Delta-T) can experience a change. Table 1 demonstrates this points I TABLE 1 DEVIATION IMPACT DN Tw Tc Tave Delta-T max + max + max delta no impact max -
max -
max delta no impact max + max -
no impact max delta max -
max + no impact ma< delta some + some -
some., delta some delta some -
some + some celta some delta
, It should be obvious from the table that if both RTDs are off by X *F in the same direction, that Delta-T is not impacted (because the relati ve di f ference between the two RTDs remains the same ), but
, that the Tave value is off by the error (because of the change in the average of the absolute temperature). It should also be obvious that, if both RTDs are off by X oF in opposite directions that the I
indicated Delta-T would be in error but that the average of the two temperatures remains the same. This would allow the conclusion that the worst uncertainty for the two conditions would be for the parameter that had the largest total uncertainty. For the intermediate case where the RTD errors are in opposite directions but not at the extreme values (and probably not of the same magnitude ) a sensitivity study or an evaluation of the gain factors is necessary. -
l Assuming that all other uncertaintles remained constant, the primary di f ferences between Del ta-T and Tave are the Tavg gain factor ( K2 )
and the division by 2.0 to reflect the averaging function. This then Indicates that Delta-T la more significant as long as the value for
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s K2 times the vessel Delta-T remains.less than 2.0. Since K2 times vessel Delta-T is typically in the range of 0.8 to 1.5 this conclusion should always be valid.
A sensitivity study to verify this point was performed as ah' additional degree of conservatism. A slightly modified version of the DTDT calculation (which allowed the calculation of SCA + SD uncertaintles for both Tave and Delta-T ) was used to perform the study. Runs 1 and 10 of Table 2 are where the RTD errors are at the extreme values used (1.8 oF ) and Delta-T is at it's max and Tave is 0.0 or vice versa. Basically the results confirm the logic argument e for the first four cases of Table 1.
The Table 2 summarizes the results of the sensitivity study wheret Delta-T is the error for a single RTD for Delta-T, TAVERAGE is the error for a single RTD for Tave, DT is the convoluted sum of the Tw and Tc RTD errors, Tave is the convoluted sum of the RTD errors
~ multip11ed by K2 times vessel Delta-T and divided by 2.0 to arrive at k -
an average value and CSA is the Channel Statistical Allowance for the
. function.
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TABLE 2 DELTA-T TAVERAGE DT Tave CSA (oF) (oF) (% span) (5 span) (% span)
., _. a,c 1.8 0.0 2.857 s 0.000
, 1.6 0.2 2.540 '
O.129 1.4 0.4 2.222 ,
0.258
- .1. 2 0.6 1.905 7 0.387 1.0 0.8 1.587 0.517 0.8 1.0 1.270 0.646 0.6 1.2 0.952 0.775 0.4 ~ *
- 1.4 0.635 0.904 0.2 1.6 0.317 1.033 0.0 1.8 0.000 1.162 Instrument span = 89.1 oF Vessel Delta-T = 59.4 *F K2 = 0.01370 As can easily be deduced from the above, the Delta-T error has a larger contribution to the uncertainty of the function than the Tave error. The same argument can be made with respect to the t
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- uncertainties due to the R/E. Also, from the above table, it can be seen that the worst case condition for the errors for Delta-T results in the largest CSA value which Indicates that the use of the extreme value for the uncertainty for the RTD in a Delta-T function.w111 be
, conservative. Based on these conclusions, a new model for the OTDT uncertainty analysis was constructed with the following featurest
- 1) a PMA for Delta-T based on the hot leg streaming uncertainty
- instead of the daily Power Calorimetric,
- 2) PMA terms for the Delta-I portion of the trip based on the accuracy of INCORE and the allowed INCDRE/excore mis. match,
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- average assuming N1 Tu and N2 Tc RTDs (this explicitly addresses the changed number of RTDs WQe to the removal of the RTD Bypass piping), .
, . 4) deletion of RTD and R/E uncertainties for Tave,
- 4) utilizat4cn of a convoluted average for the multiple R/Es assuming
_. N1 Tw and N2 Tc R/Es, (again to reflect the RTD Bypass piping
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- 6) calculation of Delta-T span as a function of power.
A similar argument can be made for DPDT, therefore the following
- changes were made to the modelt i 1) a PMA for Delta-T based on the hot leg streaming uncertainty Instead of the daily Power Calorimetric,
, average assuming N1 RTDs for Tu and N2,RTDs for Tc for Delta-T
- 3) deletion of RTD and R/E uncertainties for Tave
- 4) convolution of R/E errors, assuming N1 R/Es for Tw and N2 R/Es
- for Tc for Delta-T, and f
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- 5) calculation of Delta-T span as's fuEction of power.
Tave Low-Low was modified to reflects
- 1) the change in hot les streaming to represent the removal of the .
1 RTD Bypass piping,
! 2) the use of mult1ple Tu and Tc RTDs,
- 3) the convolution of the uncertalntles for multiple RTDs,
- 4) the use of multiple Tw and Tc R/Es, and
- 5) subsequent convolution of the the uncertainties for the R/Es.
The RCS Calorimetric Flow measurement uncertalnty was modified to reflect the change in the Tu streaming and the use of multiple RTDs after removal of the RTD Bypass piping.
! Based on these changes the OTDT, OPDT, Tave Low-Low and RCS i Calorimetric Flow measurement uncertainty see reductions in the total i
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i uncertainty for each, primarily due to the increase in the number of RTDs used for the measurement of Tw and to the overall reduction of the impact of the temperature streaming in the hot leg. The uncertainties used for the instrumentation remain specific't'o the type and manufacture of the hardware and are not a function of the presence, or absence, of the RTD Bypass piping. The only uncertaintles that change as a direct result of the removal of the RTD Bypass piping are the Tw streaming values.
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. WESTINGHOUSE CLASS 3 d
Response to Question 2
" Westinghouse utilizes a qualification methodology entitled "WWCAP E!E?.
Methodology to Qualifying Westinghouse WR0 Supplfed NSS5 Safity Related Electrical Equipment
- and conducts qualification test progrars accordance with this methodology. in
- (These are documented in the varices
- equipment qualification Westinghouse test reports (EQTR) found in WWCAP BEST).
received a " Safety Evaluation Report" (SER frcm teh NRC in 1982 which states ".....it is concluded that these repcr)ts cc: ply (8587 ESE7) with.
Section 50.49...."the NRC environmental requirements as codified by 1CCFREC, In essence, the NRC has blessed the Westinghouse qualification methodology and test reports and specifically stated that the requirements of 10CFR50.49 have been net. It should be noted that 10CFR50.49 only applies to equipment which is located in a harsh environment.
- With respect to the 'RTD Bypass Elimination" hardware being supplied
- :
Byrca and Braidwood, it has been qualified to the WWCAP E587 Methodolc;y.
The WCAP SEB7 reports which documented this are idintified belcw:
- ' EQTR-13 Process Reatect-lon System EQTR-7 RTO .
' ~
EQTR-7 is for RTDs located in a harsh environment and meets JCCFR50..tg EQTR-13 is alid environment equipment testing and thus 10CFR50.49 is n::
applicable."
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- WESTINGHOUSE CLASS 3
- t _
Response to Question 3:
RTD BYPASS ELIMINATION FOR , ,
,' BYRON UNITS 1 AND 2 BRAIDWOOD UNITS 1 AND 2 DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE CONPENSATION PROCEDURE s.,
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DEFINITION OF AN OPERABLE CHANNEL The RTD Bypass Elimination modification uses the average of 3 RTDs in each hot leg to provide a representative temperature measurement. In the event one or
. more of the RTDs fails steps must be taken to compensate for the loss of that RTD's input to the averaging function.
Single RTD Failure Hot Leg: All three hot leg RTDs must be operable during the period following refueling from cold to hot zero power and from hot zero power to full power.
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Duringtheheatupperiodtheplantoperatorswillbe{
3
., a j any hot leg can then tolerate Once g a single RTD failure and still remain operable. If the situation arises where a single hot leg RTD failure occurs a bias value must be applied to the averageoftheremainingtwovalidRTDs.]..
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.,s,s No reanalysis will be necessary to evaluate this situation. The plant will be allowed to operate for the balance of the fuel cycle with this single RTD failure in one of the hot legs. If another single RTD subsequently fails in a different hot leg the same bias application will apply.
The plant may operate with a failed hot leg RTD at any power level during that same fuel cycle. It is permissible to shutdown and,startup during the cycle
, without requiring that the failed RTD be replaced. ,
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- . j In order to eliminate any control system concerns, the Tavg and AT signal associated with the loop containing the failed hot leg RTD will be defeated as an input to the control system. This will prevent the control system from using a Tavg or AT at power levels less than 100% which may be offset due to
! the fixed bias. If another hot leg RTD fails in a different loop the utility 1
i should operate using manual control. Manual control is recommended because
- only one control channel at a time can be defeated. If automatic operation is continued the control system will most likely auctioneer the biased channel ;
- because it will be the highest Tavg due to the positive (or zero) bias i application. This means the control system will perceive a higher Tavg than
- l. is real at reduced power and the plant will operate at depressed l
- temperatures. While this is not necessarily undesirable it does reduce the
!". total plant megawatt output. The use of automatic control can be considered 3 .
. based on utility power requirements.
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!' Cold Leg: If the active cold leg RTD fails that RTD should be disconnected i from the 7300 cabinets. The installed spara RTD should then be connected in j the failed RTD's place.
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Double RTD Failure: Inoperable Channel x t
I . Hot Leg or Cold Leg: If two or more of the three hot leg RTDs or both cold leg RTDs fail in the same protection channel then that channel is considered ,
inoperable and should be placed in trip. Operation with a single valid hot. !
i leg RTD is'not presently analyzed as part of the licensing basis. .
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PROCEDURE FOR OPERATION WITH A HOT LEG RTD OUT OF SERVICE The hot leg temperature measurement is obtained by averaging the measurements ,
from the three thermowell RTDs installed on the hot leg of each loop. ,
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In the event that one of the three RTDs fails, the failed RTD will be
- disconnected and the hot leg temperature measurement will be obtained by averagingtheremainingtwoRTDmeasurements.[
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', The bias adjustment corrects for [ ,
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To assure
] m., c. that the measured hot leg temperature is maintained at or above the true hot leg temperature, and therefore, to avoid a reduction in safety margin at reduced power,[ *
, a., c.
An RTD failure will most likely result in an offscale high or low indication and will be detected through the normal means in use today (i.e., Tgyg and ATdeviationalarms). Althoughunlikely,theRTD(oritselectronics channel) can fail gradually, causing a gradual change in the loop temperature s707e Id/1202a6 3 4
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measurements. ,
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l The detailed procedure for correcting for a failoi hot leg RTD is presented below:
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5707e:1W120208 6
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p, APPENDIX CALCULATION OF HOT LEG TEMPERATURE BIAS ..
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eles 5707e 1d/120244 7
RESPONSE TO QUESTION 4.
i The sentence is correct as stated. Four loop plants, much as Byron /Braidwood, de not have dedicated RTDs for control system usage but rather derive control signals from protection channels.
?
f RESPONSE TO QUESTION 5..
The calculation of delta-T and T... is still performed by the I presently installed 7300 electronics. The new 7300 electronics are installed to average only the three hot leg RTD inputs. The single hot leg temperature signal that results from this averaging process is then provided to the existing electronics and along with T.. . is used to calculate delta-T and T.w.. These two signals are sent to the protection system and also through isolators to the control
' system. 'In this way, the protection system calculates T... and delta-T and then shares this data with the centrol system. -
This relationship between the control and protection system and their processing of delta-T and T... presently exists at all Westinghouse four loop plants utilizing 7300 analog electronics and is not a new 4
Y. concept. What is new is the use of the three RTDs to provide an average hot leg temperature which is in turn fed into the existing electronics.
RESPONSE TO QUESTION 6. ,y
)'..
p The RCS Flow eeuoted in the Technical Sp'ecifications is Minimum
-Measured Flow (P91F) eehich is higher than Thermal Deslen Flow (TDF) plus measurement uncertaintles. Pelf is used in ITDP for Condition 1 and 2 transients. TDF plus measurement uncertaintles.la used ;
for Condition 3 and 4 transients. The Technical Speelfications j place the Ilmit on the higher of the two flows to assure adesquate measured flow. Since the Technical Specifications are based on PetF, which is still higher than TDF plus the new measurement uncertainty, no change to the Technical Speelfication flow value is necessary. The measurement uncertainty has changed from 2.2 5 flow to 2.0 5 flow, primarily due to the use of three TM RTDs.
i 5
9 k
l
l' RESPONSE TO QUESTION 7.
i Answer 7a:
- I The Technical Speciffcet1ons presently allow a manImum tfue isterval between operator cognfrance of a channel failure and the placing of that channel in trip. The operator will be made aware of an RTD ,
failure famediatel !
Deviation Alams. y via the presently installed delta-T and Tevg He will then defeat the appropriate b1 stable within the allotted Technfcal speciffcatfon tfue fnterval.
Answer 7b:
Once the channel has been placed in trfp. efforts will be initiated to
,' return that channel to service. There is a detailed administrative procedure which fs used to control protection channel repair efforts.
The actual modifications to the electrontes will take 2-4 hours. The seministrative procedures, which must be followed in the perfomance of this work, will require approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> . Therefore time the channel will be in trip is approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />., It theistotal importanttonotethat(1)theabovesituationexistsforanRTDfailure in the present system, and duration of the electronics (mo)dification and therefore in a safe 2 the p condition.
Answer 7c: 5 N , .,
ThestepsinvolvedintheprocessareasfolJows:
1.
place the system with failed RTD channel fn the test mode condition.
i
- 2. Iderttify and dfsconnect the failed R7D.
- 3.
Res'ca'le' the svuuning amplifier averaging card for a two RfD fnput condition. '
- 4. Set the bias on the sunning amp 11fler averagfng card to the required level. -
- 5. put the system back to nomal operation.
RESPONSE TO QUESTION 8.
No.
Test data taken on the RTD/Thernovell/5 coop arrangement (se WCAP-11323. section 2.1) and plant data for the 7300 e actronics response bounding time at Byron 1 indicate that the values in Table 2.1-1 are No modiffcation to the table is anticfpated.
will cove. r the control system fateractfon. Questfon.35 g
--.__- ~
e RESPONSE TO QUESTION'9.
The statement on page 13 of WCAP-11323 is correct, but some. shat spi sleading in the reference to the latest Westinehouse RTD scoss-cal procedure. The orleinal version of the report was pub 18 shed for use shortly af ter the Westinghouse resolution of the RTD calibration problem. This problem was resolved by the use of a multiple temperature cross-calibration of all the RTDs in.the I primary eide at Isothermal'sondItIons. Use of this technIague results in the determination of the Individual RTD sallbration surves to'oelthin E 3*e.a. The RCS Flow Calorimetric Measurement uncertainty presented in WCAP-11323 is based on the assumption that the Narrow Range RTDs are salibrated using this technique. This technique is no longer considered to be new since it has been in use for 18 months. The primary benefit of the use 4 of this procedure is C 3*m.... This change Is morth about l C 3*m.= compared to the previous analyses.
i RESPONSE TO QUESTION 10.
l
! Flest, it should be noted that the statement in Section 4.2 of
. WCAP-11323 is incorrect. Byron /Braldwood currently have RdF RTDs Installed. .For RTD Bypass Ellmination, RdF RTDs all) be Installed in the thermowells, therefore there is no RTD vendor change nor i change in the basic calibration uncertainty since the use of the cross-calibration procedure mentioned in (9) above. The orleinal l
version of the report was pubilshed Mlth a different plant as the i basis where a change-over from Rosemount to RdF RTDs was made.
1 Westinghouse uses different uncertaintles for the two different
~
- RTDs. Increasing the number of TM RTDs from one to three results in a Flow Measurement uncertelnty change of about E 3*e.e. . ..
RESPONSE TO IUESTION 11.
1 i
! The TH streamine allowance has been modified from I i
2*e.e. Proper treatment of the revised -
uncertaintles results in s' slight increase in the Impact of the streamine allowance. This increase is so small it is not visible through the uncertainty round-off process.
1 .
I l
1
RESPONSE TO QUESTION 12. ,
Answer 12a: _
. Please refer to the procedure provided as an answer to Question.3 fo,r, more detail. Briefly, a unique bias is calculated for each hot leg .
RTD.
Answer 12b:
Yes. The answer to Question 7 provides more detail, as well as the answer to Question 3. The failed RTD signal will be manually , defeated, the summing card re-scaled to divide by 2 and a bias applied to the susuiing card output. The plant will be able to operate the balance of the fuel cycle with the failed RTD. It must be replaced during the next refueling outage.
- Answer 12C: -
The operabl11ty of a protection channel is covered under the existing plant Technical Specifications, 3.3.1 for RPS and 3.3.2 for ESFAS. The definition of operable Ie noted in SectIon 1, Item 1.19. The loss of an RTD seculd result in failure to satisfy the '
operabl11ty requirement and the plant enould then have to satisfy the ACTION re,quirements. -
l Answer 12d:
If two of the three hot leg RTDs fail, that channel will be considered inoperable and placed in trip. On page 2 of the answer to Question 3 this situation is discussed in more detail.s, I
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i RESPONSE TO QUESTION 13.
The equation used to determine the CSA value for use of Special Test Equipment or a DVM at the input to the racks is Eq. 1 on page 2 of WCAP-11168 Rev. 1. The equation used to determine $he CSA value for Indication utilizing the plant process computer.ls Eq. 2 on page 2 of WCAP-11168 Rev. 1. The equation used to determine the Calorimetric Flow Measuremen't. Uncertainty (with the actual values Inserted) is noted at the top of page 11 of WCAP-11168 Rev.
I 1. All signs for dependent effects are noted. The Byron specific values of Table 3.1-3 of WCAP-11323 may be substituted into the equation of page 11. For additional clarification the following definitions are provideds M&TE = SMTE or RMTE as defined in WCAP-11168 Rev. 1 dependlng on whether the term is used for a transmitter or a rack uncertainty.
- R/E = the resistance to voltage (or resistance to current)
- conversion module in a temperature channel using an RTD as the sensor.
. RODT = RDOUT as defined in WCAP-11168 Rev. 1.
BIAS = an uncertainty which is treated as a blas in the I statistica) sense, l.a., is not random and is one-sided.
PMA = Process Measurement Accuracy, a non-Instrument effect, e.g.,
streaming in the Hot Leg, or densit'y shift in the Cold Lee.
.' PEA = Process Element Accuracy, a metering device error, usually associated with the use of a v'entur'ly orifice, or elbow.
{
The instrument uncertainty breakdown i s provided as part of the -
revised tables supplied as Attachment A. .
RESPONSE TO QUESTION 14 The statement on page 14 Is incorrect in that the RTD vendor has not been changed for the RTD, Bypass Elimination. RdF RTDs will be Installed in the thermonells, the same manufacturer of the current RTDs used at the plant. The Nominal Trip Setpoints noted in the Technical Specifications remain as the current values.' The Allowable Values and specific values for Z change as a direct result of the Impact of changing the number of RTDs used in the uncertainty calculations. The methodology used for these emiculations is essentially the same as used in the Setpoint Study performed for this plant.
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RESPONSE TO QUESTION 15. ;
, i Subsequent to insta11stion of the new R1D system, the need to modify.
~
, c control system setpoints wjl1 he determined by observing centrol system ~. ,
behavior. If, for example, it is necessary to rodify any of the red . ,
l control system setpoints..setpoints such the had,. lag and filter tire constants on auctionekred, average temperature woult! be modified to
- i provide a more optieren system response. None e,f the parameters listed .-
in Table 2.1-1 however, would be re. quired to be podified e s a result of .
modification of control system setpoints. In particular, the RTD Filter .
Time Constant would not' be modified since it is currently ret at 0.0 and o increasing the setpoint would not be expected to provide a benefit to control system performance.
Also, no rod controi is ' ssumed a in any of the'FSAR transient analyses where reactor trip on overtemperature AT or everpower a T is assumed. . r hence modification of rod control setpoints does rot impact the results - " >
of these analyses where the value of Table 2.1-1 are assumed. .
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.. l TABLE 3.1-3 l
~
' bALOR! METRIC RCS FLOW MEASUREMENT UNCERTAINTIES INSTRUMENT ERROR FLOW UNCERTAINTY COMPONENT q "4 !
FEEDWATER FLOW -
C VENTURI THERMAL EXPANSION COEFFICIENT
- TEMPERATURE -
MATERIAL '
DENSITY TEMPERATURE PRESSURE DELTA P
- 1, FEEDWATER ENTHALPY TEMPERATURE
~ *
,. PRESSURE STEAM ENTHALPY PRESSURE *(
j : .
MOISTURE NET PUMP HEAT ADOITION f,. 64DT LEG ENTHALPY TEMPERATURE STREAMING, RANDOM STREAMING, SYSTEMATIC PRESSURE COLO LEG ENTHALPY )
TEMPERATURE ',.
f
. PRESSUPE
< 1 l *
~
COLD LEG SPECIFIC VOLUME '
TEMPERATURE ,
PRESSURE .
RTD CROSS-CAL SYSTEMATIC ALLOWANCE -
SIAS. VALUES -
FEEDW,TER PRESSURE DENSITY
- is
- ENTHALPY STEAM PRESSURE ENTHALPY j ENTHALPY = HOT LEG h PEESSURIZER PRESSURE
' ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG FLOW BIAS TOTAL VALUE - -
- ,*E,+,++ INDICATE SETS OF DEPENDENT PARAMETERS
(
SINGLC LOOP UNCERTAINTY (WITHOUT EI AS VALUES)
N LOOP UNCERTAINTY (WITHOUT BI AS sVALUES )
N LOOP UNCERTAINTY (WITH BIAS VALUES) - -
e -s I
.r
. f, . .
_' TABLE 3.1-2 FLOW CALORIMETRIC SENSITIVITIES
'I .
FEEDWATTR FLOW
- *Esb h Fa TEMPERATURE =
MATERIAL = ,
DENSITY *
- . TEMPERATURE =
PRESSURE = !
l i DELTA P = .
~
- , FEE'DWATER ENTHALPY .
' ~
- TEMPERATURE = ,
PRESSURE =
.g MS ,
= 1198.0 BTU /LBM -
" hF . = . 419.4 BTU /LBM Dh(SG) = 778.6 BTU /LBM STEAM ENTHALPY m + Id PRESSURE =
MOISTURE =
. HOT LEG ENTHALPY TEMPERATURE = '
= .
PRES.SURE - - ,
- :.s
~ hH = 613.2 BTU /LBM hC = 537.5 BTU /LBM DhtVESS) = 75.7 BTU /LBM Cp(TH) = 1.42$ 97U/LBM-DEGF COLD LEG ENTHALPY .
1 -
m + E,4 TEMPERATURE =
PRESSURE =
~ 1.225 BTU /LBM Y CptTC) = DEGF COLD LEG SPECIFIC VOLUME
- e + Isb TEMPERATURE =
PRESSURE =
~ ~
. e
_. _ . _ _ , _ _ . . _ _ . _ _ , _ _ , _ - _ . _ . . _ _ _ _ _ . . _ , . , _"" , , , . , . . . _ _ , , _ , , _ , _ _ _ ..,,__.,_,-,..-__..___,....,_m_, _..., _ _ _ .
.s -
s i
l
- l : TABLE 3.1-1 I
FLOW CALORIMETRIC INSTRUMENTATION, UNCERTAINTIES i 1.TC (2 SPAN) FW TEMP FW PRES FW DP STM PRESS TH RCS PRESS
. td.:
C
. SCA =
M&TE=
SPE =
STE =
SD =
. R/E'=
. RCA = _
RTE =
~
.Y- RD = -
RDOT=
BIAS =
s
-. CSA = l
~
.. ~/
~
M 3 1 1 **
e OF INST 3 -
. .. DEG F PSIA N DP PSIA DEG F DEG F , PSIA 1500, 120. 1500.. 100. 100. 3000.0 INST SPAN = 567.
_1+
INST UNC. -
4 (RANDOM) =
INST UNC.
(BIAS) = -
=4 c. 948. 848. 600.0 542.2 2250.
NOMINAL
- Number of Feedaster, Hot Lee and Cold Leg RTDs used for measurement in each loop and the number of Wide Range ROS Pressure channels used overall.
.y ,. ,.,_r,._,._ _ _ - _ . . _ , _ . , , _ - _ _ . , _ , . . . - , . _ . . . . , , , , . , . , _ - - . - .w_, ,, . . _ _ _ . . ,% , . . _ _ - _ . , ,
1 . . .
/* ,
l .-
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TABLE 3.1-4 COLD LEG ELBOW TAP FLOW UNCERTAINTY INSTRUMENT UNCERTAINTIES'
}
5 DP SPAN 5 FLOW ~
+ E C.,
0
, := .
j . PMA =
PEA =
8 i SCA =
SPE = -
1 1
l,- STE- =
So =
RCA =
~
M&TE= ,
8 RTE =
RD =
s ID = t. .
A/D = '.
RDOT=
BIAS =, ,,
FLOW CALORIM. BIAS =
FLOW CALORIMETRIC =
INSTRUMENT SPAN = \'t
"', _ + %4 SINGLE LOOP ELBOW TAP FLOW UNC =
=
N LOOP ELBOW TAP FLOW UNC N LOOP RCS FLOW UNCERTAINTY
=-
(NITHOUT BIAS VALUES) w -
N LOOP RCS FLOW UNCERTAINTY (tJITH BI AS VALUES ) = 1.9 I
RESPONSE TO OUESTION 16 The RTD bypass elimination modification removes the valves, piping, snubbers, and supports associated with the RTD bypass system and replaces them with thermowell-mounted fast response RTD's which are installed directly in the reactor coolant pipe. The following actual hardware changes and activities will take place on Byron Unit 1.
Byron Unit 1 Reactor Coolant Cold Legs:
- 1. Sever and remove two inch pipe from existing bosses for installation of the thermowells.
- 2. Install four (4) thermowells and complete welding.
- 3. Provide holes for four (4) new bosses.
- 4. Install four (4) new bosses and weld.
- 5. Cut or drill thru four (4) new bosses.
- 6. Install four (4) thermowells and weld.
- 8. Remove remainder of two inch pipe, supports, and deleted components.
Reactor Coolant Hot Legs:
- 1. Removal of one inch pipe from existing twelve (12) bosses.
- 2. provide holes in flow scoops in twelve (12) bosses.
- 3. Install and weld twelve (12) thermowells.
- 5. Removal of remaining one inch pipe, supports, and deleted components.
Reactor Coolant Crossover Legs:
- 1. Sever three inch pipe at four (4) nozzles.
- 3. Removal of approximate two hundred feet of three inch pipe and supports.
Electrical Work:
- 1. Removal existing cables from existing conduit and re-pull thru new conduit runs.
A major portion of the installation for Byron Unit 2 was completed during construction prior to fuel load.
The following modification work remains on Byron Unit 2.
Byron Unit 2 Reactor Coolant Cold Legs:
- 1. Sever and remove two inch pipe from existing bosses for installation of the thermowells.
-2. Remove remainder of two inch pipe, supports, and deleted components.
- 3. Cap the old bypass bosses.
Reactor Coolant Hot Legs:
- 1. Removal of one inch pipe from existing twelve (12) bosses.
- 2. Removal of remaining one inch pipe, supports, and deleted component.
- 3. Cap the old bypass bosses.
Reactor Coolant Crossover Legs: Same as Unit 1.
Electrical Work: Same as Unit 1.
J RESPONSE TO OUESTION l'1 (a) The estimated occupational radiation exposure for each RTD bypass modification is shown below. The estimate is based on anticipated
- exposure times for each major subtask and actual area dose rate
! measurements made in July of 1985.
Byron Unit 1 Sub Task Man-Hours Dose-Estimate (person-Rem) 1)- Reactor Coolant Cold Legs 1,200 60
- 2) Reactor Coolant Hot Legs 1,920 96
- 3) Reactor Coolant Crossover Legs 1,280 64
- 4) Electrical 600 15 TOTAL FOR UNIT 1 5,000 235
Byron Unit 2 Sub Task Man-Hours Dose-Estimate (Person-Rem)
- 1) Reactor Coolant Cold Legs 560 28 i
- 2) Reactor Coolant Hot Legs 1,440 72
- 3) Reactor Coolant Crossover Legs 1,280 64
- 4) Electrical 600 15 TOTAL POR UNIT 2 3,880 179 (b) Typical general area dose rates are anticipated to range 20-100 mR/hr with an average dose rate of 50 mR/hr.
(c) Maximum dose rates should range from 1.5 to 2 R/hr on contact and 200 to 400 mR/hr general area at RTD manifolds.
RESPONSE TO QUESTION 18 Upon removal of the RTD bypass piping system a significant amount of dose associated with valve and manifold maintenance,_ISI, and snubber inspection will be eliminated. The amount of dose associated with steam
-generator maintenance, reactor coolant pump maintenance, and loop isolation valve maintenance will also be reduced. As a result, over the life of the plant, an estimated 817 person-rem will be avoided for each unit. Therefore, a net exposure savings of 582 person-rem for Byron Unit 1 and 638 person-rem for Byron Unit 2 should be realized.
RESPONSE TO OUESTION 19 ALARA preplanning will be utilized to identify potential exposure concerns during RTD bypass modification and effect solutions. Proper tools and equipment will be available for each subtask. Temporary shielding will be utilized whenever and wherever practical. The use of mockups and special training will be utilized to assure that the personnel are familiar with the l job effort. A reactor coolant pipe from Marble Hill is being used as the mockup where teams of workers are being trained so that they will have a full understanding of the job effort. Job observations and constant communication between job supervisors, H.P. technicians and ALARA personnel will be used to identify additional exposure reduction methods. These techniques will ensure worker doses are maintained, ALARA. '
l l
l
e RESPONSE TO OURSTION 20 The types of radioactive waste expected are manifolds, valves, three inch diameter pipe, two inch diameter pipe, insulation, instrument orifice tags, and swage lock connectors for the RTD loop material. Total volume expected is 210 cubic feet per unit.
The radiation and contamination expected is:
.2 R/hr to 2.0 R/hr contact and 30 mR/hr to 2.55 R/hr smearable.
If reasonable measures ne taken to minimize cross contamination, hanger material surfaces can probably be decontaminated to less than 100 dpm/100 cm2 Currently, Commonwealth Edison is considering options for disposal of the RTD waste. These are:
- 1. Burial at low-level waste site Richland, WA, a US ecology burial site.
- 2. On site decontamination by CBCo or vendor with disposal as non-radioactive material.
- 3. Off-site decontamination by vendor.
RESPONSE TO QUESTION 21 Commonwealth Edison is expecting high dose rates, and high contamination levels. Therefore multiple dosimetry, time keeping and the special. training and ALARA described in the response to question 19 will be utilized.
2743K
T ATTACHMENT C REVISIONS TO TECHNICAL SPECIFICATION VALUES 2743K
g TABLE 2.2-1
~
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 3
- g .
'; c-o SENSOR d
- TOTAL ERROR e FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWA8LE VALUE
[ 1. Manual Reactor Trip N. A. N.A. N.A. N.A. N.A.
- 2. Power Range, Neutron Flux
! a. Nigh Setpoint 7.5 4.56 0 1109% of RTPa 1111.1% of RTP*
1
- 3. Power Range, Neutron Flux, 1. 6 0.5 0 15% of RTP* with $6.3% of RTP* with
- liigh Positive Rate a time constant a t.ime constant
- y 12 seconds 12 seconds j 4. Power Range, Neutron Flux, 1.6 0.5 0 $5% of RTP* with 16.3% of RTP* with liigh Negative Rate a time constant a time constant 12 seconds 12 seconds i 5. Intermediate Range, 17.0 8.4 0 Neutron Flux $25% of RTP* $30.9% of RTP*
j 6. Source Range, Neutron Flux 17.0 10.0 0- $105 cps $1.4 x 105 cps
.l 7. Overtemperature AT 9.7 5. 8(o See See Note 1 See Note 2 i
(2 7. '1)* (5.38), Note 5 *
- 8. Overpower AT 4.8 1. 2. 1. 2 See Note 3 See Note 4
! (43[- (1.3[6
- 9. Pressurizer Pressure-Low 5.0 2.21 1.5 11885 psig 11871 psig
- 10. Pressurizer Pressure-liigh 3.1 0.71 1.5 $2385 psig '12396 psig
- 11. Pressurizer Water Level-liigh 5.'0 2.18 1.5 192% of instrument 193.8% of instrument !
! span span
- I
- RTP = RATED lilERMAL POWER
g TABLE 2.2-1 (Continued)
B z
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
.; g SENSOR q TOTAL ERROR m FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWA8LE VALUE
- 12. Reactor Coolant Flow-Low 2.5 1.50 0.6 >90% of loop mini-
~
of loop mini-
[ (i .'ll)# mum measured flow
- mum measured flow" i
4
- 13. Steam Generator Water 88.9 (89.2)* 7oj)
[
Level Low-Low
- a. Unit 1 27.1 18.28 1.5 >40.8% of narrow >39.1% of narrow range instrument range instrument
- span span
- b. Unit 2 17.0 14.78 1. 5 >17% of narrow >15.3% of narrow range instrument range instrument >
q> span span u.
- 14. Undervoltage - Reactor 12.0 0.7 0 >5268 volts - >4728 volts -
Coolant Pumps each bus each bus
- 15. Underfrequency - Reactor 14.4 13.3 0 157.0 Hz 156.5 Hz Coolant Pumps
- 16. Turbine Trip -
i a. Emergency Trip ileader N.A. N.A. N.A. 2540 psig. 1520 psig -
Pressure
- b. Turbine Throttle Valve N.A. N.A. N.A. 11% open 11% open Closure 1
- 17. Safety Injection Input N.A. N.A. N.A. M.A. N.A.
from ESF
- 18. Reactor Coolant Pump N.A. N.A. N.A. N.A. N.A.
Breaker Position Trip
!
- Minimum measured flow = 97,600 gpa '
WThe value. in Pa"**'3e" '*Y Pelies until 1be. YTD hp ss manife)ds are e);m,noted .
i
.]
g TABLE 2.2-1 (Continued) y . TABLE NOTATIONS E NOTE 1: OVERTEMPERATURE AT
% 1 '
AT (3 Il ('~ * (1
+ 3(P - P') - f (AI))
t taS} O 1 + Ts TsS) -
n.
fu Where: AT = Heasured AT by RTD NorrWe44 Instrumentation, I[,'
3
= Lead-lag compensator on measured AT, I I t, 12
=
Time constants utilized in lead-lag compensator for AT. In = 8 s, 12=3s,
'* 1 4 y , ,,3
Lag compensator on measured AT, ta
Time constants utilized in the lag compensator for AT, ta = 0 s,
= Indicated AT at RATED THERMAL POWER, AT, Ki = 1.164, K2 = 0.0265/'F, .
I
=
The function generated by the lead-lag compensator for T,y0 dynamic compensation,
=
t, ts Time constants utilized in the lead-lag compensator for T,yg, 1 4 = 33 s, 13 =4s, T = Average temperature. *F, 1
3, 3
=
Lag compensator on measured T,yg, .'
g TA8LE 2.2-1 (Continued) ,
B z
t e TA8LE NOTATIONS (Continued) ,
c: NOTE 1: (Continued)
- d T. = . Time constant utilized in the measured A ,
T,,, lag compensator, t. = 0 c,
~eo 1
T' <
N 588.4*F (Nominal T,,,at RATED THERMAL POWER),
1 K3 = 0.00134,
=
P Pressurizer pressure, psig, i
P' =
2235 psig (Nominal RCS operating pressure),
S =
Laplace transform operator, s 8, 7
j " and f (al) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
4 l (1) for qt - q between b -=% and +10%, f (al) = 0, where qt and qb are Percent j
RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt+qb IS total TilERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of q gbexceeds +10%, the AT Trip setpoint
' shall be automatically reduced by 2.0%-of its value at RATED THERMAL POWER.
l NOTE 2:
(3 9,%of AT span.The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint 9
)
j *The. value. in parenthe.ses only arY lies until ihe. RTD bypos3 <wanifoMs arc eN; naled.
_ TABLE 2.2-1 (Continued) '
E . TABLE NOTATIONS (Continued)
B NOTE 3: (Continued) .
g i
Z K. =
1
. 0.00170/*F for T > T" and K. = 0 for T 5 T",
w T =
- As defined in Note 1, T" =
Indicated T, .
at RATED THERNAL POWER (Calibration temperature for AT instrumentation,,1 588.4*F), ,
1 S =
As defined in Note 1. and f 2(AI) =
0 for all AI.
NOTE 4:
y 2.6% of AT span.she channel's maximum Trip Setpoint shall not exceed its computed T i e
r p Setpoint. by more than NOTE 5:
The sensor error for temperature is i.q ( 1 . '2. [ o M &
Pr* W "E d hl(I d 4:-
l i . -
l I "
I 1
e I
i U
gh M rcOC se.s o 4
0FFM
- M O'
- D P" "
J
LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux, High Rates (Continued)
The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DN8R to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.
Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core
- protection during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED
! THERMAL POWER unless manually blocked when P-10 becomes active. ,
Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power i
distribution, provided that the transient is slow with respect toYpiping L
transit delays from the core to the temperature detec'.frs (about 4 seconds),
l and pressure is within the range between the Pressurizer High and Low Pressure
! trips. The Setpoint is automatically varied with: (1) coolant temperature to l correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop l temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.
With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater i
than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
l
+hermal delays associakd win 4herm0WEM- rh0VWM 4empernhire dekchr.s (abo
- G secouls) and l
l SYRON - UNITS 1 & 2 8 2-5 l
l..-- . --... _ _ _ _ --_.___ -_ _ _ _ . _ _ _ . _ _ _ _ _ - _ - _ .
i
-themal d eln i kom +be.
core. b h oop fempewfum-LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtempera AT trip, and provides is automatically a backup to the High Neutron Flux The varied with: .
trip Setpoint induced changes in density and heat capacity of water, an e of temperature for dynamic compensation fortpiping delays from the cor loop temperature detectors, (kW/ft) is not exceeded. to ensure that the allowable heat generatio the consequences of various size steam breaks as reported in Core Response to Excessive Secondary Steam Releases." , " Reactor Pressurizer Pressure each with its own trip setting to provide for a High a thus limiting the pressure range in which reactor operation is permitted The Low Setpoint trip protects against low pressure which could lead to D .
tripping the reactor in the event of a loss of reactor coolant pressure .
(a power level of approximately 10% of RATED THERM
-impulse chamber pressure at approximately 10% of full power equivale on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety overpressure. valves to protect the Reactor Coolant System against sy Pressurizer Water Level through the pressurizer safety valves.The Pressurizer High Water Lev High Water Level trip is automatically blocked by P-7.(a power level ofOn d approximately 10% of RATED THERMAL POWER with a turbine impulse chamber power, automatically reinstated by P-7. pressure at approximately 10%
i i
1
- BYRON - UNITS 1 & 2
, 5 2-6
.- ----,- - - _ n,,,,a- ,_ , - -, --n--..-n , . - , ---,----.-,r-.- - - - - . - , - - ,
.?
TABLE 3.3-2 "l ^
g REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES
, z e
i c:
- FUNCTIONAL UNIT RESPONSE TIME d 1. Manual Reactor Trip- N.A.
FJ e 2. Power Range, Neutron Flux <0.5 second*
l 3. Power Range, Neutron Flux, ,
liigh Positive Rate- N.A. '
- 4. Power Range, Neutron Flux, '
- liigh Negative Rate ;gD.5second"
- 5. Intermediate Range, Neutron Flux -
M.A.
~
/
) }{ 6. Source Range, Neutron Flux N.A. 7, o
<4;9 !seconds *#
l 4 7. Overtemperature AT
- 8 ., Overpower AT N.A.
- 9. Pressurizer Pressu.re-Low -<2.0 seconds 1 (Above P-7) .
i
- 10. Pressurizer Pressure-liigh <2.0 seconds 4
- 11. Pressurizer Water Level-High N.A. -
(Above P-7) '
j ^ Neutron detectors are exempt from' response time test'ing. Response time of ti>e . neutron. flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
& ~
we we eu a y w b qw l .
TABLE 4.3-1 h REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
. TRIP e ANALOG ACTUATING MODES FOR 55 CHANNEL DEVICE WHICH d CHANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE g FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED
- 1. Manual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3* , 4 * , 5* i
- 2. Power Range, Neutron Flux
- a. liigh Setpoint S D(2, 4), Q N.A. N.A. 1, 2 M(3, 4),
Q(4, 6),
R(4, 5) .
- b. Low Setpoint S R(4) Q N.A.
N.A. 1###, 2
$ 3. Power Range, Neutron Flux, N.A. R(4) Q N.A. H.A. 1, 2 l o liigh Positive Rate I d>
- 4. Power Range, Neutron Flux, N.A. R(4) Q N.A. N.A. 1, 2 liigh Negative Rate
- 5. Intermediate Range, S R(4, 5) Q N.A. N.A. 1###, 2 Neutron Flux
- 6. Source Range, Neutron Flux S R(4, S, 12) Q(9) N.A. N.A. 2##, 3, 4, 5-
- 7. Overtemperature AT S R(13) Q N . A'. N.A. 1, 2
- 8. Overpower AT S R Q N.A. N.A. 1, 2
- 9. Pressurizer Pressure-Low S R Q**- N.A. N.A. 1 (Above P-7)
- 10. Pressurizer Pressure-liigh S R Q N.A. N.A. 1, 2
- 11. Pressurizer Water Level-liigh 5 R Q N.A. H.A. 1 (Above P-7) i
~
3
- This ' nom. only applies unWI %RTb bypss manifolds we dimincdecl. '
w i
TABLE 4.3-1 (Continued)
TABLE NOTATIONS
- With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal. n
- These channels also provide inputs to ESFAS. The Operational Test Frequency for these channels in Table 4.3-2 is more conservative and, therefore,
% controlling.
MBelow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) If not performed in previous 7 days.
(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to ?%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels
, the provisions of Specificatiotr 4.0.4 are not applicable for entry.into MODE 2 or 1.
(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(8) With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window.
(9) Surveillance in MODES 3*, 4*, and 5" shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Surveil-lance shall include verification of the Boron Oilution Alarm Setpoint
, of less thal or equa! to an increase of twice the count rate within a
. 10-minute pariod.
, (10) Setpoint verification is not applicable.
(11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifiqation of the Undervoltage and Shunt trips.
(12) At least once per 18 months during shutdown verify that on a simulated Boron Dilution Doubling test signal CVCS valves 1120 and E open and 112B and C close within 30 seconds.
(13) CHANNEL CALIBRATION shall include the RTO bypass loops flow rate.
BYRON - UNITS 1 & 2 3/4 3-12
)
- TABLE 3.3-4 (Continued) 1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWA8tE g FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE Q
m 8. Loss of Power s .
l ' e- a. ESF Sus Undervoltage N.A. N.A. N.A. 2870 volts '
o* 12730 volts
- w/1.8s delay w/51.9s delay j b. Grid Degraded ,,
Voltage N.A. N.A. N.A. 3604 volts >3728 volts w/310s delay w/310 1 30s delay i '
i
- 9. Engineered Safety i Feature Actuation tg System Interlocks
+
i;> a. Pressurizer Pressure, 4 gg P-11 N.A. N.A. N.A. $1930 psig $1936 psig j b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
l c. Low-Low T,,g, P-12 N.A. N.A. N.A.
1550*f 1 547.O (597.62*)F k
- d. Steam Generator Water See Item 5.b. above for all Steam Generator Water Level Trip Level, P Setpoints and Allowable Values.
(High-Nigh) l l
4 J
i -
l
TABLE 3.3-4 (Centinu-d)
TABLE NOTATIONS
" Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are t, > 50 seconds and T2 < 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.
- The time constant utilized in the rate-lag controller for Steam Line Pressure -
Negative Rate - High is greater than or equal to 50 seconds. CHANNEL CALIBRA-TION shall ensure that this time constant is adjusted to this value.
~4k This value. only applies unbl the KTD bypasa manifoldJ Gre c}imimIed.
BYRON - UNITS 1 & 2 3/4 3-29
m _ _ _
.. POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
- c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
FhwillbemaintainedwithinitslimitsprovidedtheConditionsathrough
- d. above are maintained.N The combination of the RCS flow requirement (390,400 gpm) and the requirement on Fg guarantee that the DNBR used in the safety analysis will be met.
A rod bow penalty is not applied to the final value of F H f r the following reason:
Fuel rod bowing does reduce the value of the DNBR. However, predictions with the methods described in WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation," July 1979 for the 17x17 Optimized Fuel Assemblies indicate that the fuel rod bow reduction on DNBR will be less than 3% at 33,000 MWD /MTU assembly average burnup.
At higher burnups, the decrease in fissionable isotopes and.the buildup of fission product inventory more than compensate for the rod bow reduction in DNBR.
There is a 11% margin available between the 1.32 and 1.34 design DNBR limits and the 1.47 and 1.49 safety analysis DNBR limit. Use of the 3% fuel rod bow DNBR margin reduction still leaves a 8% margin in DNBR between design limits and i
safety analysis limits.
The RCS flow requirement is based on the loop minimum measured flow rate of 97,600 gpm which is used in the Improved Thermal Design Procedure described in FSAR 4.4.1 and 15.0.3. A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators. Potential fouling of the F feedwater venturi, which might not be detected, could bias the results from the S precision heat balance in a non-conservative manner. Therefore, a penalty of g 0.1% is assessed for potential feedwater venturi fouling. A maximum measurement l g uncertainty of W 2X has been included in the loop minimum measured flow rate to
. account for potential undetected feedwater. venturi fouling and the use of the d RCS flow indicators for flow rate verification. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring O
- ~ g and trending various plant performance parameters. If detected, action shall be taken, before performing subsequent precision heat balance measurements, i.e.,
either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.
Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup.
Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumentation are not included in the flow measurement uncertainty analysis. This requirement does not apply for the instrumentation whose drift effects have been included in the uncertainty analysis.
i *The valve in PQMee3*.I only an$es unHl %c. RTD byosJ manifdds om.chmindtl.
BYRON - UNITS 1 & 2 B 3/4 2-4
I ATTACISEBIT D Revisions to WCAP-11323 (Proprietary)
Pages 16 through 21 Revisions to WCAP-11324 (Non-Proprietary)
Page 16 2743K
M TABLE 3.1-1 FLW CALORINETRIC INSTRUMENTATION WCERTAINTIES .
(% SPAN) FW TEMP FW PRES FW'd/p STM PRESS T H T, RCS PRESS . a,c
~
SCA =
"~"
I~ -
MTE =
SPE = .
STE =
SD = .
R/E =
RCA = -
RTE a RD =
RDOT =
. BIAS = .
CSA =
~ _
- OF INST USED 3 3 1 1 **
- F . 4. m d/, p.i. or er ,.4.
insi srut - es7. asco, ano, tsco. 200, too. soon.
INST UNC. - -
(RANDON)= , .
a.e INST UNC. -
=
(BIAS) ,
NCWINAL = *
, ea:h leep aid ,the nusber of wide range RCS pressure channels used l '
overall. .
ers7c.tc/c m ae 16 l
. . - _ _ _ _ _ _ . _ . _ _ . _ _ . _ . _ . _ . _ _ . _ . . . _ _ _ _ _ _ _ _ _ , . . . _ . . _ . . . . _ _ _ _ . . . . _ _ . _ _ . _ _ _ - . _ . _ . . _ . . . . _ . _ _ . . , _ _