ML20217A813

Requests Withholding of WCAP-12632, RTD Bypass Elimination Licensing Rept for Turkey Point Units 3 & 4, from Public Disclosure

ML20217A813
Person / Time
Site:	Turkey Point
Issue date:	11/05/1990
From:	Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Murley T NRC
Shared Package
ML17348A745	List: ML20217A813 ML20217A821
References
CAW-90-085, CAW-90-85, NUDOCS 9011260185
Download: ML20217A813 (25)

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Westinghouse Energy Systems (g8;8lgg"f Electric Corporation ibn 355 kEvN E T,"INO CAW-90-085 Document Control Desk US Nuclear Regulatory Commission Washington, DC 20555 Attention:

Dr. Thomas Murley, Director APPLICATILN FOR WITHHOLDING PROPRIETARY INFORMATiON FROM PUBLIC DISCLOSURE

Subject:

WCAP-12632,Rev 1 "RTD Bypass Elimination Licensing Report for lurkey Point Units 3 and 4" (Proprietary)

Dear Dr. Murley:

The proprietary information for which withholding is t'eing requested in the enclosed letter by Florida Power and Light Company is further identified in Affidavit CAW-90-085 signed by the owner of the proprietary information, Westinghouse Electric Corporation.

The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10CFR Section 2.790 of the Commission's regulations.

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Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Florida Power and Light Company.

Correspondence with respect to the proprietary aspects of the application for I

withholding or the Westinghouse affidavit should reference this letter, CAW-90-085, and should be addressed to the undersigned.

V ry;t uly yours,

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Mobert A. Wiesemann, Mana99r Enclosures Regulatory & Legislative Atfairs cc:

C. M. Holzle, Esq.

Office of the General Counsel, PC V. Wilson, Nuclear Reactor Regulation 9011260185 901115

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j CAW 90-085 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, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

'hch(L j JAWA(O Robert A. Wiesemann, Manager Regulatory and Legislative Affairs L

Swcrn to and subscribed l.

befve me this L day of/vm/do,1990, I 1 $ [/ l l DY /G l Notary Public [ 4 i l x ww;n;;iin.a-~w a tm-

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? ' CAW 90-085 (1) I am Manager, Regulatory and Legislative Affairs, in the Nuclear and Advanced Technology Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit. (2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidtvit. (3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial er financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the folicwing is furnished for consideration by the Commission in determining whether ihe information sought to be withheld frca public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse, i

j j - CAW 90-085 (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and t!ie substance of that system constitutes Westinghoure policy and provides the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (orcomponent, structure, tool, method,etc.)wherepreventionof its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage )ver 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. l

i . CAW-90-085 (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (g) It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner. t 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 competitors. It is, therefore, l withheld from disclosure to protect-the Westinghouse competitive position. l I i w w

^ \\ 5-CAW-90 085 l I (b) It is information which is marketable in many ways. The extent. to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (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 competitive advantage. if competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. L (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to'the competition of those countries. 1 (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and L maintaining a competitive advantage. e

. CAW-90-085 (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission. (iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (v) The proprietary informetion sought to be withheld in this submittal is that which is appropriately marked in "RTD Bypass Elimination Licensing Report for Turkey Point Units 3 and 4", WCAP-12632, (Proprietary) for Turkey Point Units 3 and 4, being transmitted by the Florida Power and Light Company (FPL) letter and Application for Withholding Proprietary Information from Public Disclosure, J. Goldberg, FPL, to Document Control Desk, to the Attention Dr. Thomas Murley, Director, Office of NRC, July, 1990. The proprietary information as submitted for use by Florida Power and Light Company for the Turkey Point Units 3 and - 4 is expected to be applicable in other licensee submittals in j response to certain NRC requirements for justificatior, of actions to remove the existing Resistance Temperature Detector (RTD) Bypass Elimination system and replace with fast response thermowell mounted RTD's in the reactor coolant loop piping.

7-CAW 90-085 This information is part or that which will enable Westinghouse to: (a) Provide documentation of the analyses, methods, and testing for reaching a conclusion relative to the removal of existing Resistance Temperature Detector (RTD) Bypass system and the replacement of fast response thermowell mounted RTD's. (b) Support the continued validity of Loss of-Coolant Accident '.0CA) and non-LOCA safety analysis initial condition assumptions. (c) Establish the effects of the fast response thermowell RTD system on instrumentation and Reactor Coolant uncertainties. (d) Assist the customer to obtain NRC approval for operation with RTD Bypass Elimination. Further this information has substantial commercial value as follows: (a) Westinghouse plans to sell the use of similar information to its customers for purposes of satisfying NRC requirements for licensing documentation. (b) Wastinghouse car sell support and defense of the RTD Bypass Elimination technology to its customers in the licensing process. l l l. L

. CAW-90 085 Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analytical documentation and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar inhnical programs would have to be i performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests. 1 l Further the depenent sayeth not. i y 1 i

g,,',". 2 ( L Westingnouse Energy Systems w ea m a a m ea Electric Corporation '*'**D D*5*" b 355 PMSb/gh Pemsylvan,a 15230 0355 May 2. 1990 FPL 90 606 NS 0PLS 0PL !!-90 338 Mr. S. T. Hale Engineering Project Manager Florida Power & Light Company P. O. Box 14000 700 Universe Blvd Juno Beach, Florida 33408 Attention: Mr. R. L. Wade FLORIDA POWER & LIGHT COMPANY TURKEY POINT UNITS 3 & 4 Safety Evaluations for Overpower AT, Overtemperature AT and Underfreauency Reactor Coolant Pumo Trio

Dear Mr. Hale:

Please find the attached safety evaluations for the Overpower AT, Overtemperature AT and the Underfrequency Reactor Coolant Pump Trip. These evaluations were performed to support the current safety analysis. [ With the installation of Rosemount transmitters for monitoring pressurizer l pressure, the Containment High-1 channel has been utilized in calculating the Pressurizer Pressure Low SI trip function.- This was done in order to l accommodate the large environment eerors associated with the Rosemount transmitters.. The use of Containment High 1 is inconsistent with standard Westinghouse design practices. Westinghouse is currently investigation an internal recommendation with regards to utilization of protection channels for accident mitigation which do not directly measure the parameter of interest (i. e. taking c edit for Containment high - 1 for Pressurizer Pressure Low SI). When the resuli., cf the investigation are available they will be forwarded. I If you have any questions, please contact the undersigned. Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION .h. h ) D. J. Richards, Manager Florida Power & Light Project

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SECL NO. 89-1164 } Customer Reference No(s). Westinghouse Reference No(s). ~ WESTINGHOUSE NUCLEAR SAFETY SAFETY EVAt.UATION CHECK LIST 1.)NUCLEARPLANT(S): Turkey Points Units 3 & 4 2.) SUBJECT (TITLE): Revised ffAI) Penalty Functions for the Overoower and Overtemocrature AT SetDoints l 3.) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59(b) has been preparet to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2. Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed. CHECK LIST - PART A - 10CFR50.59(a) (1) Yes 1_ No A change to the plant as described in the FSAR7 Yes _, No 1 A change to procedures as described in the FSAR7 Yes,_,No _.1. A test or experiment not described in the FSAR? Yes.l No A change to the plant technical specifications l (See Note on Page 2) l

4) CHECK LIST - PART B - 10CFR50.59(a) (2) (Justification for Part 8 answers I

l sust be included on Page 2.) (4.1) Yes __ No 1. Will the probability of an :ccident previously evaluated in the FSAR be inc.sased? (4.2) Yes No L. Will the consequences of an accident previously evaluated in the FSAR te increased? (4.3). Yes No 1 ' Nay the possibility of an accident which is different than any already evaluated in the FSAR be created? (4.e) Yes No 1 Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased? . (4.5) Yes No 1. Will the consequences of a malfunction of equ pment important to safety previously I evaluated in the FSAR be increased? (4.6) Yes No 1 Nay the possibility of a malfunction of equipment important to safety different than any already. evaluated in the FSAR be created? (4.7) Yes ___ No 1 Will the margin of safety as defined in the bases to any technical specification be reduced? I 1-PAGE 1 L

l s SECL NO. 89 1164 NOTES: If the answers to any of the above questions are Unknown, indicated under 5.) REMARKS and explain below. If the answer to any of the above questions in Part A (3.4) or Part B cannot be answered in the negative, based on written safety evaluation, the change review would require an application for license amendment as required by 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90. 5.) REMARKS: The following sumkrizes the justification upon the written safety evaluation (1) for answers given in Part A (3.4) and Part B of this SECL The effect of eliminatina tho ffAI) nanalty function for the Overnower AT reacuor trin has buen eva' uated based on the ':urrent CAOC sand for the i Turkey Point units. he ava' uation considered tan effect of the nanalty function elimination on Condition 11 accidonts. .t was confirmed that the resultino overnower conditions did not vie' daL inear nower dannity that gggjjl contribute to fuel centerline meltina. FSAR and Technica' jngfification chances recuired to imolement this chance are inc'luded. (1) Reference to document (s) containing written safety evaluation: FOR FSAR UPDATE Section: 14.1.2 Page(s): 14.1.2 1 Table (s): None Figure (s): None Reason for/ Description' of Change: In nerformina the Turkov Point Satooint study it was datormined that the overnower and overtamnerature reactor trin netnoints cou' d not be suonorted by the curront safety analysis. En order to reduce the channel I uncerta'lnties an ava' untion was nerformed _iustifyina the elimination of the f!AD nana' tv funct' on for the overnower A" setnoint and reduction of the s' one of the ffA.) for the overtemocrature AT setnoint. L SAFETY EVALUATION APPROVAL LADDER: N3OfkO Prepared by (Nuclear Safety): $A Date: g g CoordinatedwithEngineer(s): Date:- CoordinatingGroupManager(s): A Date: e l Nuclear Safety Group Mana Y ( N/ Date: ( id PAGE 2

s SECL 89 1164 Turkey Point Units 3 & 4 Safety Evaluation Supporting Revised f(AI) Penalty Functions for the Overpower and Overtemperatt.re AT Reactor Trip Setpoints

1.0 Background

in performing the Turkey Point setpoint study it was determined that the uncertainty allowances between the safety analysis and nominal values of the overpower AT (OPAT) and overtemperature AT (OTAT) reactor trip setpoints were insufficient. The purpose of the overpower and overtemperature protection system is to define a region of permissible core operation in terms pressure, and axial power of power, temperature, reactor coolant system (RCS)he limits of this region shape; and to trip the reactor automatically when t are approached. This region of permissible operation is defined by three boundaries: the thermal overpower limit, the thermal overtemperature limit, and the locus of conditions where the steam generator safety valves are open. The thermal overpower limit protects the core against excessive fuel centerline temperature, whic) could cause fuel melt. The thermal overtemperature limits protect the core against DNB and hot leg boiling. This safety evaluation describes the evaluation performed to justify the relaxation of the slope of the f(AI) penalty function for the OTAT trip and elimination of the f(AI) senalty function for the OPAT trip. A detailed description of tie design basis for the OPAT and OTAT trip functions can be found in WCAP-8746, 2.0 Non-LOCd Analysis overnower AT The OPAT reactor trip is designed to ensure operation within the fuel temperature design basis. Experience has shown that this can be accomplished by preventing the core average power from exceeding a prescribed limit of 110% of nominal power. This is achieved via the OPAT trip by correlating core thermal power with the coolant temperature difference across the vessel. Since tu prescribed overpower limit may not be adequate for highly skewed axial tv,wer distributions, a penalty function that lowers the setpoint is factorkd into the OPAT trip channels. This term is a function of the axial flux difference, AI, and is known as the f(AI) penalty function. If it can be demonstrated that the peak linear heat generation remains below the design limit during ANS Condition 11 overpower events without penalizing the overpower setpoint, the f(AI) penalty function for the OPAT setpoint is not necessary and can be eliminated. i An evaluation of analyses previously performed was conducted to support elimination of the f(AI) function based on the current constant axial offset control (CAOC) strategy used at the Turkey Point units. The evaluation examined Condition 11 overpower transients that could produce potentially limiting linear heat generation rates. These events were analyzed with the assumption that the OPAT trip setpoint (without f(AI)) provides a I reactor trip at 118% of the nominal full power. The events considered were: PAGE 3 ^ l t

__ _ _ _ _ _ _ _ _. ~ s SECL 89 1164 control bank malfunction, core cooldown, and boration/ dilution system 1 malfunctions. These limiting transients are analyzed to determine the core power level and power distribution using static nuclear core models. An analysis was performed showing that a heat generation rate of less than 22 kw/ft does not violate the fuel centerline temperature design basis for any of the fuel types used in the Turkey Point core. The linear heat generation rates from the limiting transients noted above were compared to the 22 kw/ft value, confirming that the overpower conditions did not yield any linear power densities that would violate the fuel centerline design basis. Therefore, the f(AI) functic.) can be elimi + ed in the OPAT setpoint with no adverse effect on core protection. Overtemnerature AT The thermal overtemperature trip is designed to ensure plant o)eration within the DNB design basis and the hot leg boiling limit. Since bot 1 of these limits are functions of coolant temperature and pressure as well as core thermal power, the OTAT trip is correlated with the core AT, veul average temperature and primary system pressure. This is accomplished in the following manner. First the core DNB limits are determined for a range of reactor operating conditions. The core DNB limits are represented as the locus of points of core thennal power, primary system pressure and coolant inlet temperature that define a DNBR at the DNBR limit. These conditions are calculated assuming a reference axial power shape characterized by a chopped cosine with a peak to average ratio of 1.55. Similar to the OPAT reactor trip, a compensating term which is a function of Al is factored into the OTAT trip setting to offset the effects of core axial power distributions more severe than the reference power shape on DNB. Essentially the f(AI) penalty function allows the power distribution effects to be separated from the core-wide parameters. The f(AI) function for the OTAT setpoint is determined in the following way. For each of a set of five standard asynnetric axial power distributions and core inlet temperatures, the power level that results in l DNBR at the limit value is determined by the THINC computer code. The standard asynnetric axial power distributions are calculated in a fashion that l bounds all ANS condition I and II DNB events. A f(AI) penalty function is calculated that will ensure the OTAT setpoint is reached before limiting l core power, pressure and power condition! are reached. Note that certain l constraints limit the range over which tu core DNB limits must apply. The OPAT reactor trip places a limit on the maximum power level that needs to 1 be considered, the high and low pressurizer reactor trip limits the pressure range that needs to be considered and the steam generator safety valves place a physical upper limit on the coolant temperature that needs to considered. l Frn this analysis it was determined that an f(AI) penalty function slope of 1.5 on both the positive and negative wings is sufficient for core protection for Turkey Point Units 3 and 4. This is less restrictive than the current Technical Specification slope of 2.0 on the negative wing and 3.5 on the positive wing. PAGE 4 -w -+ ---r- ,-m -'n ~

.o SECL 89 1164 3.0 Conclusions An evaluation was performed to establish that the f(AI) penalty function can be removed for the OPAT setpoint and that the f(AI) penalty function slope can be revised and relaxed to 1.5 for the OTAT setpoint without compromising core protection. This evaluation is applicable to reactor cores containing standard (LOPAR), optimized (OFA), or debris resistant fuel (DRFA) assemblies for both Turkey Point Units 3 and 4. Note that these conclusions will be reconfirmed on a cycle specific basis for future fuel reloads as part of the normal reload design process. Justification for the answers provided in section 4 of tie Safety Evaluation Checklist are addressed below. 1. Will the probability of an accident previously evaluated in the FSAR be increased? The OPAT and OTAT reactor trips protect the core from fuel centerline melting and DNB during ANS Condition I and Il events. The particular values of the setpoints do not affect the probability that an event will occur. 2. Will the consequences of an accident previously evaluated in the FSAR be increased? Elimination of the f(AI) pena'ty function for the OPAT setpoint i l does not increase the consequences of any accident previously analyzed in the FSAR. An evaluation was 1erformed that demonstrated that none of the l non LOCA transients require tie f(AI) penalty function of the OPAT reactor trip for mitigation. The evaluation established that the OPAT reactor trip continues to prevent the fuel centerline melting design basis from being violated. A separate evaluation established that the proposed reduction in the slope of the f(AI) penalty function for the OTAT setpoint does not increase the consequences of any accident previously analyzed the FSAR. The OTAT setpoint continues to trip the reactor core before an operating condition that would violate the DNB design basis is reached. 3. May the possibility of an accident which is different than any already evaluated in the FSAR be created? The recommended changes in the f(AI) penalty function for the OPAT and the OTAT reactor trip setpoints will not create the possibility that a different accident than that which is already analyzed in the FSAR l will occur. As stated above, the OPAT and OTAT reactor trips protect the core from fuel centerline melting and DNB during postulated ANS Condition I and II events. The particular value of the reactor trip setpoints do not affect the possibility that any event will occur. I PAGE 5

SECL 89 1164 o '4. Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased? The recommended changes in the f(AI) penalty function will not increase the probability of a malfunction to any equipment important to safety previously analyzed in the FSAR. The reactor core relies on these trips for protection during ANS Condition-I and !! events. Setpoint changes do not impair the ability of the protection to perform its intended function. Analysis has shown that the reactor core remains protected for the reconnended penalty function modifications. Therefore, the proposed change will have no impact on any equipment important to safety. 5. Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased? The recommended changes in the f(AI) penalty function for the OPAT and the OTAT reactor trip setpoints will not increase the consequences of any malfunction of equipment important to safety different than that which was previously evaluated in the FSAR. Analysis has shown that the reactor core is in no way impacted by the p*oposed changes to the setpoints. The consequences of any equipment malfunction remain the same. l 6. May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created? i ~ The recomended changes in the f(AI) penalty function for the OPAT L and the OTAT reactor trip setpoints will not create the possibility that a malfunction of equipm6nt important to safety different than that i previously evaluated in the FSAR will be created. As stated above, the reactor core relies on the OPAT and OTAT reactor trips for protection during ANS Condition I and II events. Analysis has shown that the reactor core remains protected with the proposed changes. Therefore, the proposed changes will have no impact on any equipment important to safety. 7. Will the margin of safety as defined in the bases to any technical specification be reduced? The OPAT and OTAT reactor trips are designed to ensure reactor operation within the DNS and fuel centerline melting design basis. The analyses described above demonstrate that with the recommended changes to the f(AI) penalty function the DNS and fuel centerline design bases j-continue to be met. No margin of safety is reduced. 1 i c PAGE 6 l ~ ) ~ T'___ _ - _ -___ _ L L __ J _ _ -. _ J'.J L.._-.,-----

\\ O. SECL 89 1164 l Technical Specification Modifications i ( i l \\ i l l l l' PAGE 7 '**6-san = w * =, aw......-

4 RIACTOR C00Lurt TNttP,1THE Overtesperature & T,1 AT, (Kg = 0 0101 *(T-3?&) 4 0 000633 (P-1223) - ! LtO) a 47, = ladisated of at rated power. F 'T= Avarage temperature, h F= Pressurtser pressure, paig - f (A g)= a fumaties et the Lati' sated 4t!!stesse between top and bettes detesters of the power-range svelaar tes shaabers; with gatas to be saleeted based es sensured'&astrustas raspenas dustag stattg 1 ' tests suah that For (t are thI th) with?4 + 10 yetMet and =14 persent dere gg and th pertest power is the top and bestem halves et the sore I respettively, su g + gb to total eers power to perseet el rated g power, f (ag) = 0. Fev each persest that the magnitude et (g$e automatically redu gh) entests + 10 pe the Delte-T trip settstat shall t ert p reest et its value at 1staria power. I, of (g, = g ) esteeds =J & For eseh perseat that the magnitud g pe , the De:.ta-T trip setteist. hall le automatteally reduted b sett et its value at intests power. I (Thr ee ' Operation) = 1 093 Q, < --- ~ --- - n - l L N l L j i r i l l l. I l This amendment effective as of date of issuance for Unit 3 and date of startup. Cycle 10. for Unit 4. 3.3-1 %nement Nos. 99 and 93 w.m u,.. eeoo m

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[.. ) LIMnte sArrty systen grTTIss i BAsrs OverseverAT The Ove, power AT trip prevents ' power density anywhere in the core free encendi las of the design power density. This erovides asswance of fuel [ integri (e.g. ne fuel pellet melling and less than N cladding strain) possible everpower conditions limits the required range for 6ver-under a temperature &T trip and provides a bac to the Nigh Neutron Fi m trip. The setpointisavtaastIcellyvariedwith: ( ) coolant temperature frect for . temperature indu.ced change Rat t#: tie 2;'u:: *.s in density and. heat,

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- - ren. rate of gofto.,e, re fer na.ic n. sor l Pressuriner pressure i .ista.in es. of.,resswiser,ress.we.nneis re are t.e in.,en,.nt l ies, e. a w m its own t,i set gtop,evi for a nigt. and tow resswe trip thus lietting the presswo r,ange in which reseter operatten is permitted The Low Setpoint trip protectc against low presswo which could lead to 05 by l tripping the reacter in the event of a less of reacter coolant pressure. I l ) On decreasing power the Law Setpoint trie is avtamatically blocked by p-7 i (a power level of appreminatal 15 of RATED THERMAL p0WER with tw6tne first stage pressure at appreminate 15 of full power epivalent); and on increasing power, autenatically reinstate by p 7. The High Setpoint trip functions in conjunction with the presswiser safety valves to protect the Reacter Coolant $ystaa against systes overpressure. pressuriner Water Level N pressuriser Water Level Nigh trip is provided to prevent water relief through the pressuriser safety valves. On decreasing power the presswicer High Water Level trip is avtaastically blocked by p=7 (a power level of appreminately 11E of Raft 0 THEm4L p0WER with a tertine first stage presswo at approximately 1 5 of full natically reinstated by p-7. power equivalent) and on increasing power, auto-Resctor Coolant Flaw The teacter Coolant Flow Lew trip provides core protection to prevent Om by eitigating the conseguences of a less of flew resulting free the less of one er more reacter seelant pumps. On increasing power above po? RATED THERMA' POWER er a turbine fi(a power level of appreateetely 15 of rst stage presswo at appreminately is TUREIY POINT

• UNITS 3 & 4 8 2*5 AmeMENT N05.

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1sdtastet average tasp eratura as asstaal seaditasas and = sated power. F %= 0 for destaasing average sosparature 0.5 see./Y ter tsareasing,evarage stepera,tura E 0 00048 for T egual he er aste thas T*g 0 ter T less g than t' E= Rate et ahanga et tarytrature, f/ sat ds w Oy1g<aII.o su.=u an s ??assuriser tsw Pressuriser prasaura - etuat. te er greater thaa 1833 pais. Righ Pressurtser pressura - agua.':ts or less t>.as 2383 prit. Kjgh Prasasurtser vatar 's.a!, - e ud to or less shar. 9tf. ef !di . sa at.e. Reatter Caelatt Flev Law roastar seelaat tiev - etual te er grea:ar thaa SC* s! atraal tatisated flev. Law reaatar asolant pump satar fraguessy awd to or graatar than 361 Bs. todarveltage sa vaatter seelaat pump metar hs - agual to or greater than 602 et aossai valtage. Steam canarators Low-law staas genenator sacar level - egual to er greater _ than 132 et marrow range tastruseat scala. This amendment effective as of date of issuance for Unit 3 and date of startup. Cycle 10. for Unit 4. 2. '3 - 3 , he:deent Nos. % wi et j i

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SECL 89 1164 7 . FSAR Modifications i<'- .n t ,'\\ \\ ~. E i i 1 . f 1 PAGE 8 .T 3 = \\ '?

's 14.1.2 UNCONTROLLED RCCA WITHDRAWAL AT POWER An uncontrolled RCCA withdrawal at power results in an increase in core heat flux. Since the heat extraction from the steam generator reasins constant, there is a not increase in reactor coolant temperature. Unless terminated by manual or automatic action, this power mismatch and resultant coolant temperature rise would eventually result in DNS. Therefore, to provant the possibility of damage to the cladding, the Reactor Protection System is designed to terminate any such trans-1ent with an adequate margin to DitB. The automatic featuras of the Reactor Protection System which prevent core damage ir a rod withdrawal accident at povar include the followings a) Nuclear power range instrumentation actuates a reactor trip if two out of the four channels exceed an overpower setpoint. l b) Reactor trip is actuated if any two out of three AT chanels exceed an overtemperature AT setpoint. This setpoint is automatically i,

• arted with power distribution, taperature and pressure to protect against IMB.

c) Reactor trip is actuated if any two out of three AT channels. exceed j an overpower AT setpoint. Thia aggtging, i= t-tin.:,1, ...;.2 v.$ ;- :: 21....'..; = :r r. in 9 -'^ - -- - L 1 - - '

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d) A' high pressure reactor-trip, actuated fram any two out of three pressure channels, is set at a fixed point. This set pressure will l ba less than the set pressure for the pressuriser safety valves. i i 14.1.2-1 gav 4 7/86 i nn.--n - -nn..-------~ - - - - = - - - --]}}