ML20052B597

Forwards Westinghouse Analyses Supporting Retention of RPI Tech Specs for Cycle 3 Operation & Beyond.Proposed Tech Spec Change to Increase Overall Limit,Necessitated by Retention of Existing Tech Specs,Evaluated

ML20052B597
Person / Time
Site:	Beaver Valley
Issue date:	04/26/1982
From:	Carey J DUQUESNE LIGHT CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
TAC-48018, NUDOCS 8205030353
Download: ML20052B597 (31)

Text

{{#Wiki_filter:a b 'A@ M Telephone (412) 45s-6000 foTo',4 April 26, 1982 Shippingport, PA 15071 @ 04 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission .IOQ j Attn: Mr. Steven A. Varga, Chief y Q Operating Reactors Branch No. 1 / .j r[' %4 +,3 Division of Licensing l v Washington, D. C. 20555 -l d o ' s

Reference:

Beaver Valley Power Station, Unit No, h. j 7 Docket No. 50-334, License No. DPR-66 \\( ' Cycle 3 Reload Safety Evaluation 'y >i m _g \\' ; l Gentlemen: Our February 23, 1982 letter informed the Commission that Duquesne Light Company had completed its review of the safety and operational aspects of the design of the reload to be installed in Beaver Valley Unit No. 1 for the third cycle of operation. In addition, the Company submitted a number of requests for amendment of the Technical Specifications for Cycle 3 and beyond which were factored into the safety review of the Cycle 3 design. With one exception (F at high power), the requests for amendment of the xy Technical Specifications were not the result of the Cycle 3 design. Since Fxy is cycle dependant, and is a function of the core loading patterr, the Company proposed a Technical Specification which would substitute the submission of a Peaking Factor Report each Cycle rather than change Technical Specifications to accomodate changes in Fxy each cycle which would be necessary to assure that the FAC would be met. On April 15, 1982 and again on April 20, 1982, the Staff contacted our office and requested that the Company submit further information related to these proposed amendments to the Technical Specifications. On April 21, 1982, the Company submitted a discussion of the factors related to the demonstrated accuracy of the Rod Position Indication System which led the Company to conclude that the existing Technical Specification 3.1.3.1 should be retained foc Cycle 3 and subsequent Cycles. By retaining this Specification, however, it became necessary to request amendment of Technical Specification 3.2.3 to increase the overall limit on FAH from 1.5355 to 1.55 and to explicitly state the rod bow penalty as a function of burnup using the expression [1-RBP(BU)] where RBP (BU) is specified in Figure 3.2-4. ))oo/ s / / 820503o353

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Cycle 3 Reload Safety Evaluation April 26, 1982 Page 2 The original specification used 1.5355 as the multiplier to implicitly account for the maximum Rod Bow Penalty regardless of burnup. In addition, the 0.2 multiplier was changed to 0.3 to provide additional operating margin at lower power levels, since at the 100% power level the term (1-P) becomes zero. The margin to DNB is maintained by the reduction in the core safety limits per Figure 2.1-1. We contracted with Westinghouse to perform an evaluation of this proposed amendment to the Technical Specifications and a copy of the Westinghouse report is enclosed as Attachment A for your information. The change in FAH requires a reduction in the Core Limits as described in Figures 2.1-1 and 2.1-2 of the Technical Specifications in order to maintain DNBR greater than 1.30 at all power levels. In addition, to provide consistency with the Specification for FAH, the 0.3 multiplier was also factored into the equation for Fxy. This change in F is xy meaningful only at lower power levels and the change in the full power F limit is unrelated to the change in FAH which has been proposed. x At(achmentC, Rod Misalignment Analysis, indicates that considerable margin to the FAC calculated FQ will exist at all times for the proposed new Technical Specification on rod misalignment. The changes in FAH, the F Report and the Core Limits are proposed to permit tbe (Peaking Factor) x use of wider limits for the Rod Position Indication Specifications 3.1.3.1. While the changes in FAH, Fxy and Core Limits are necessary for the change in Specification 3.1. 3.1, f these changes (with the exception of the F full power Limit) x are not necessary to validate the Cycle 3 Reload Safbty Evaluation. We anticipate requiring these changes to accomodate future Cycles and, therefore, we consider it prudent to continue to pursue these changes at this time. The Staff has also questioned our proposed change to the negative rate flux trip setpoint as described in Table 2.2-1 of the Technical Specifications. This change results from a Westinghouse reevaluation of the flux rate trip setpoints in which Westinghouse concluded that the maximum allowable value is less than or equal to 6.3 percent of RTP with a time constant of greater than or equal to 2 seconds. The proposed specification is within this allowable margin. Westinghouse has submitted additional information related to the above questions as a result of Staff requests on April 20, 1982. A copy of the Westinghouse response is enclosed as Attachment B.

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 ' Cycle 3 Reload Safety Evaluation April 26, 1982 Page 3 As we stated in our April 21, 1982 letter, we believe that there are good reasons to support our application to retain the existing i-1. RPI Technical Specifications for Cycle 3 operation and beyond. To provide the Staff with further information relating to the scope of analysis performed by Westinghouse to support this request, we are enclosing as Attachment C the text of the Westinghouse report related to this change. Our April 21, 1982 letter provided a copy of the generic Westinghouse proposed Technical' Specification which has been modified to recognize site specific conditions at Beaver Valley. This i-Technical Specification was submitted as information to the Staff and has not undergone our standard review process. We continue to believe that the appropriate solution to the RPI problem is to provide greater misalignment li. nits to accomodate known characteristics of the instruments. We are willing to discuss any aspects of this problem with the Staff to achieve an appropriate resolution of this problem. } 1 Should you have further questions on this subject, do not hesitate to contact my office. Very truly yours, i - J. J. Carey Vice President, Nuclear i i Enclosures (3) cc: Mr. R. C. Haynes, Regional Administrator USNRC Region 1, King of Prussia y Mr. D. A. Beckman, Resident Inspector l Beaver Valley Power Station, Unit No. 1 USNRC Document Management Branch [ t r i ,,,rw ,,,,r, - - - - - -

l '3-ATTACHMENT "A" " $ 0Use '/,'i.iCf Il&,Ct0F ".iliC C0fp0fallon Di;jSlons 3.shu;h Penn (sa,.a i$230 January 22, 1982 82DL*-G-002 Mr. J.D. Sieber Duquesne Light Company KEYWORDS: Beaver Valley Power Station DLW-Fall P.O. Box 4 TECH-SPEC-CHANGE Shippingport, PA 15077

Dear Mr. Sieber:

i DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT N0. 1 Ftli TECHNICAL SPECIFICATION CHANGE The purpose of this letter is to document the results of our safety analyses that will allow a change in the Beaver Valle , g duced power. U' Technical on To dofeMaaJmr wa r".e. u., ee nor tm is r ~ ' oidinf; t4ew=effBM(9We6] i ii > t,tst KJainablp*D This change will also minimize the chance-i naving icense a nsertion Technical Specification limit change l in order to satisfy peaking factor criteria at low po d banks as

• it.

Other W L r p An evaluation has been completed for Beaver Valley Unit 1 which shows that the Ftli Technical Specification limit may be modified by changing the allowable slope as follows: N < 1.55 [1.0 + 0.3 (1-P)[1-RBP (BU)] F fg where P = fraction of rated thermal power and RBP (BU) = Rod Bow Penalty as a function on region average burnup (Figure A of the attachment). Note that the power multiplier is changed from 0.2 to 0.3, and the current FtH I Tech. Spec. limit of 1.5355 is replaced with a 1.55 limit with rod bow l accounted for by the burnup dependent [1-RBP (BU)] rod bow penalty. The Fxy relationship has also changed the power multiplier from 0.2 to 0.3.

,."

• r. J.D. Sieber Jer uary /2, l'M2 The core limits and axial offset DilB limits (fl and fi-1 loop, one loop I'olated) for an increased allowable Fall at reduced power levels have been evaluated to determine the effect of this change.

The results are as. follows: 1. Core Limits - Both the il and fi-1 (one loop isolated) proposed core limits are more limiting in the areas bounded by the quality and D!iBR limits. However, no change to K1 through K5 will be required. Therefore, the new core limits will require no accident reanalysis. See t T ecagd. 2. Axial Offset DNB 1.imits - The proposed axial offset DNB envelopes result in a more limiting f(AI) function than currently in the Technical Specifications. Therefore, a change to the f(al) function in the Technical Specifications is required. See attached g rom Beaver Valley Tech. Specs. Additional changes to the Technical Specifications required to implement this modification to the Fall limit are also attached. These changes are en M f the Beaver Valley Unit 1 og If you should have any questions regarding this justification for the proposed FtJi Tech. Spec. change, please let us know. Very truly yours, M-%\\h B.D. McKenzie Froject Engineer NFD Fuel Projects /kph Attachment cc: A.R. D'Amico WR0 w/ attachment l R.D. Scherer WR0 w/ attachment

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e-c s TABLE 2.2-1 (Continued) h REACTOR TRIP SYSTEM IfiSTRUMEtiTATI0ti TRIP SETP0ItiTS M" fl0TATION (Continued) F Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops Q (no loops isolated) (1 loop isolated) e e

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= 1.18 K) = 0.99 K) = 1.1 g c-5 E K = 0.01655 K = 0.01655 K = 0.01655 &,g_4 2 2 2 K = 0.000801 K = 0.000801 K = 0.000801 {[ 3 3 3 k b. and f ( AI) is a function of the indicated difference between top and bottom detectors pg % m a of the power-range nuclear ion chambers; with gains to be selected based on measured gg 3 instrument response during plant startup tests such that: 4 ;; [ ,n

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(i) for q (wherb.q-qNndq are percent RATED T t L POWER in the top and bottom halves of the cbre respectively, and qt*9b is total THERMAL POWER in percent of RATED TilERMAL POWER). y 2 -q exceeds -23 percent, for each percent that the magnitude of (qthe AT trip setpoint shall be automat (ii) its value at RATED THERMAL POWER. y 4 gg for each percent that the magnitude of (q -q exce'eds + 13 percent, the AT trip setpoint shall be automaticalfy' rebu)ced by 1.9 (iii) / percent of its value at RATED THERMAL POWER.

f ' ',. i i i. it i s-Prt;cscd Technical Spacificatien Ch:ngcs 1 e ~ for Eeaver Valley Unit 1 f}ES_ __, N The curves are based on an enthalpy hot channel factor, Fty, of 1.55 and a reference cosine with a peak of 1.55 f r axial power shape. An N allowance is included for an increase in F g at reduced power based on 3 the expression: 0.3 N F H = 1.55 [1 + pr'T (1-P)] where P is the fraction of RATED THErJ%L POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trip will reduce the setpoint to provide protection consistent with core safety limits. 2.1.2 REACTOR CCJLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping and fittings are designed to ANSI B 31.1 and the valves are designed to ASA 16.5 which permit a maximum transient-pressure of 120% (2985) psig of component design pressure. The Safety Limit of 2735 psig is therefora consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation. BEAVER VALLEY - UNIT l' B 2-2

f Prc;,ased Technical Specificati'.n (banges for E. ur Valley Unit 1

f. i P DIS;F.1BUT10!! LIMITS f.UitVElLIA!;CE REQUIRi P.E!!TS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F shall be evaluated to detemine if F (Z) is within its 0

• Y limit by:

Using the movable incore detectors to obtain a power distribu-a. tion map at any THEP. MAL POWER greater than 5% of PATED THERilAL POWER. Increasing the measured F component of the power distribution b. map by 3% to acr.ount for Ednufacturing. tolerances and further increasing the value by 5% to account for measurement uncertainties. c. Comparing the F computed (F ) obtained in b, above to: 1. The F limits for RATED THERMAL POWER (FRTP) for the xy xy appropriate measured core planes given in e and f below, and 2. The relationship: 0.3 R F, =F [),jpg[)_p)) x x l where F is the limit for fractional THERIML POWER

• Y RTP and P is operation expressed as a function of Fxy the fraction of RATED THERMAL POUER at which F was xy measured.

d. Remeasuring F according to the following schedule: P 1. When F is greater than the F limit for the appropriate measured core plane but less than the F relationship, xy additional power distribution maps shall be taken and RTP F compared to F and F a) Either within 24 hours after exceeding by 20" of RATED THERMAL PONER or greater, the THERIGL POWER C at which F was last determined, or ry HEAVERVALL(Y-U:lT1 3/4 2-6 , Amendment flo. 2

3... Fr cposed it..ical Specification C at. es for Ecaver Valley Unit I ~ POWER pl5TRIEUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F H LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the folicwing relationship: g l.55 0.3 F,gg S W [1.0 + M (1-M D _WA F SPN>>)' E'l b" '"" / e" THERMAL POWER where P = TiATED THERML POWER

• frepoA*Ag (obkned fro, Figure A APPLICABILITY:

MODE 1 ACTION: N 1(ith F exceeding its limit: AH a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 surs and reduce the Power Rance Neutron Flux-High Trip Setpoints to 1 55% of RATED THERMAL POWER within the next 4 hours, b. Demonstrate thru in-core mapping that F is within its limit H within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours, and Identify and correct the cause of'the out of limit condition c. prior to increasing THERMAL POWER; subsequent POWER OPERATION N may proceed provided that F is dcmonstrated through in-core g mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THEPE L POWER prior to exceeding this ' THERMAL POWER and within 24 hours after attaining 9.5 or preater RATED THEPMAL POWER. I BEAVER VALLEY - UNIT 1 3/4 2-8 1-F Ar,endment 70, 20 i. 1.

u ATTACHMENT "B" \\ Water Rea0t0i[ {f \\ W85tingh0BSS APR 221981 Electric Corporation EMslons

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s JDS l tw 3312 Pmsbu@ Perm,%fia 15230 CENTRAL FILE April 21, 1982 { 1 Hr.. J.D. Sieber 82DL*-G-022 Duquesne Light Company XEYWORDS: Beaver Valley Power Station P.O. Box 4 CYCLE-3 TECH-SPEC-CHANGES Shippingport, PA 15077 D:ar Mr. Sieber: DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 1 TECHNICAL SPECIFICATION CHANGES On April 19, 1982 changes that were submitted to thewe discussed with the NRC various technical specification Reload Safety Evaluation. NRC simultaneously with the Cycle 3 with our discussions with you and the NRC regarding the various technic sp:ct f1 cation changes. 1. Exx A change to the Beaver Valley Unit 1 Technical Specification values by means of a Peaking Factor Report. values of Fxy as a funct has been approved for J.M. Farley linits 1 and 2.A change of this type It is antici-pated that Fxy will change from cycle to cycle and this one change for each reload. change would eliminate the necessity of making a Tech Spec The values of Fxy recomended for Cycle 3 are: Fxy61.71 for all core planes containing D-Bank i For unrodded core planes ^ Fxy a l. 68 up to 2.4 ft. elevation Fxy a 1.6B from 2.4 ft. elevation to 7.8 ft elevation Fxy

• 1.65 above 7.8 ft. elevation These values would be found in the Peaking Factor Report described above.

and the results show that all limits are met.These values have been used in t -i

V April 21, 98 . Mr. J.';. Sieber Ar, additional change to the Fxy Tech Spec (4.2.2.c.2) is recoi ende d which changes the multiplier from 0.2 to 0,3 so that the limit rea s: t.- Irr* Fxy [1+0.3(1-P)) Fxy a This change has also been factored into the safety evaluation f and no limits ware violated. ( 2. fsH_ Part P6wer Multiplier change to 0.3 a. Core DNB limits were generated for Beaver Valley Unit 1 usin following allowance for F H at reduced power: S F H = 1.55 [1 + 0.3 (1-P)] - where P is the fraction of As a result of the changes in Core DNB limits, A the Core Safety Limits (Figure 2.1-1 and 2.1-2 of the thernal power. Specifications) changed.in the safety evaluation for Cycle 3 a h that all limits are met, b, Rod Bow Penalty Change This change removes the restriction of a constant rod bow pen (1%) and replaces it with a rod bow penalty dependent upo region aver ge burnup.to acconnodate the effects of fuel rod bowing th conservatism due to lack of significant data. values of g Additional in-plant operational data and DNB tests (as desc contained in References 1 and 2) designed to quantify the effects of ro bow have provided a basis for reducing the penalty applied to F[H to account for rod bow effects. Ref.1-Westinghouse submittal NS-TMA-1760, " Fuel J) J.F. . Rod Bowing"), T.N. Anderson,1 to Stolz (NRC, Apri1 19, 1978. Ref. 2-J.F. Stolz (NRC) letter to T.M. Anderson .di.f. dated April 5, 1979 Negative Flux Rate Trip 3, Westinghouse has re-evaluated the setpoint of the flux rate. As f - trip setpoints with respect to the dropped rod transient. .: :3 ' a result of this re-evaluation it has been de Negative Rate trip s5tpoints should be as follows for both trips: Allowable Value Nominal Value 66.3% of RTP with a 6 5.0% of RTP with a time constant 2 2 seconds time constant 12 seconds

r. J.D. Sie'.e - Ap.-il 21,1982 f you have any questions, please do not hesitate to contact ree.

Very truiy yours, MN W Brian D. McKenzie Project Engineer NFD Fuel Projects c: A.R. D' Anico WR0 R.D. Scherer WR0

P. 1-ATTACHMENT "C" N " C+ d' %511hghruse '7C'ef P.'.ECl0F 00C1fic C0lp01Biion Di/jslGas g,33i7 P.M. gt, rem nWn:a 15230 January 22, 1982 - ~ Nr. J.D. Sieber Duquesne Light Company KEYWORDS: Beaver Valley Power Station DLW-ROD-MISALIGN P.O. Box 4 Shippingport, PA 15077 Dear Mr. Sieber DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STAfl0N UNIT NO. 1 R0D MISALIGNMENT ANAL.YSES Attached for your use and information is a report which details the calculations and analyses done to assess the peaking factor penalties associated with undetected rod misalignments of + 32 steps. This report was prepared to support Tech. Spec, changes for incorporating a rod misalignment into Cycle 3 and future cycles of + 32 steps as it had been incorporated into Cycle 2. The only Tech. S'pec. change required is to remove the note from the Beaver Valley Technical Specification which states that the + 32 t misalignment is "For Cycle 2 only". See pages of Tech Specs. If you have any questions on the attached report, please call. Very truly yours, V.% hh B.D. McKenzie Project Engineer NFD Fuel Projects /kph cc: A.R. D'Amico WR0 w/ attachment R.D. Scherer WR0 w/ attachment Attachment

[,. Beaver Valley Unit 1 Cycle 3 Rod Misalignment Analysis

Background

During the operation of Cycle 2 of Beaver Valley Unit 1, Tech. Spec. changes were made to permit indicated rod misalignments of up to 1 16 steps.- (Previously, 1 12 steps). Also, the error allowance of the Analog Rod Position Indication System (ARPI) was increased from + 12 steps to 1 16 seeps. Consequently, the largest permissable undetected rod misalignment increased from 1 24 steps to 1 32 steps. As worded, this Tech. Spec. change was applicable to Cycle 2 only. Reference 1 details the analysis done by Westinghouse to assess the Cycle 2 peaking factor penalties associated with the change. Similarly, this report assesses the peaking factor penalties for the Cycle 3 core as justifica-tion for continued use of the Cycle 2 Tech. Spec. in Cycle 3. Methodology The methodology employed in this analysis is similar to that used in Cycle 2. To support the 1 32 Tech. Spec., rod misalignment calculations were performed to determine the peaking factor penalties resulting from 1 32 step misalignments of single D and C Bank rods. Together with fine mesh 2D-TURTLE, a TURTLE benchmarked I )' ( ) model was employed,'using 2x2 meshes per assembly 3D-PALADON radially and 20 meshes axially. Calculations were made at various power levels with control rods positioned at the appropriate insertion limits (Figure 1). Single D and C bank rods were misaligned by 32 steps and the resulting effects on peaking factors determined. The Beaver Valley Unit I control rod pattern is shown in Figure 2. The potentially mis-aligned rods are indicated on an eighth core basis. '1633F_: 6

T ~ s, Rod Misalignment FaH Table i shows the calculeted rod misalignment FLH as a function of power level. Calculations were done at BOL with both equilibrium xenon and no xenon conditions considered. The FaH limit values also shown correspond to: Limit FAH (P) = 1.55 [1 + 0.3(1-P)] s ~ where P is the fraction.cf rated thermal power. (The 0.3 multiplier is being proposed as a Tech. Spec. change for Cycle 3.) Figure 3 provides a plot of the Table 2 data. As in Cycle 2, at significant power levels where DNB is a concern, the rod misalignment FaH is smaller than the design basis FAH limit. For power levels below

• 20%, the rod misalignment FAH values exceed the proposed Tech. Spec. limit by _2.5% or less.

For these low power levels, however, the core limits are vessel exit boiling limited and are not a function of peaking factor. Consequently, there is no safety Concern. Rod Misalignment Fxy Worst case Fxy(Z) values were collected for each. axial mesh from the three-dimensional misalignment calculations. Figures 4 and 5 show the maximum calculated Fxy(Z) values for power levels of 100% and 45%, respectively. ' Also shown are the axial and power dependent Fxy limits which are applicable to Cycle 3 and which will be proposed as a Tech. Spec. change in the Cycle 3 RSE. These limts are: I d.

1 ~ e 1. For unrodded core planes: RTP Fxy < l.68 up to 2.4 ft. elevation RTP Fxy < l.68 from 2.'4 ft. to 7.8 ft. elevation RTP fxy <.1.65 above 7.8 ft. elevation 2. RTP Fxy < l.71 for all core planes containing D Bank The power dependence of the Fxy limits will become: Limit RTP Fxy (P) = Fxy [1 + 0.3 (1-P)] where P is the fraction of Rated Thermal Power (RTP). As Figures 4 and 5 indicate, the rod misalignment Fxy's lie below the ~ proposed Fxy limits for the 45% and 100% power levels. These Fxy values include an 8% uncertainty factor. For very low power levels and deep rod insertion, the Fxy limits could be exceeded. At low powers, how-ever, the F limit is very large, 4.64 below 50% power, and. power q peaking is not a concern. Rod Misalignment F (Z) Associated with the limiting rod misalignment Fxy (Z) values discussed above are the limiting F (Z) values shown in Figure 6. These F 9 g values were conservatively generated for steady state conditions by syn-thesizing the limiting Fxy(Z) with the largest axial relative power for each axial plane. The F (Z) values for the reference case, i.e., the 9 ^.

I. HFP case with D-Bank inserted to the full power insertion limit, are also shown in Figure 6. HFP equilibrium xenon conditions were assumed for both the reference case and the misalignment case. The plotted values include the 1.05 and 1.03 Tech. Spec. uncertainty factors. Note that the misalignment penalty occurs at the top of the core (in the rodded region) and that considerable margin to the LOCA F envelope q exists at all core elevations. For load follow operation, the limiting points from the Cycle 3 FAC analysis are also shown in Figure 6. These points were generated using the Tech. Spec.'Fxy values which are conservative by 5 to 10% with respect to as calculated values (including uncertainties) and by 16 to 21% with respect to best estimate Fxy's. All of the limiting .F x Power points occurred during full power time steps. Furthermore, g in the upper part of the core where the misalignment penalty is the largest, the limiting points occurred for 0-Bank ins <tions of < 24 steps. Typically, D-Bank is inserted only 12 steps for these points. Consequently, upward rod misalignments would fall within range of the standard misalignment (24 step) Tech. Spec which Beaver Valley used prior to the Tech. ' Spec. change in Cycle 2. Therefore, for upward misalignments, no additional F x Power increase will result from the q proposed Tech. Spec. Downward rod misalignments result in only small (* 3%) F increases in the vicinity of the misalignment. There is q sufficient conservatism in the analysis to acccmodate these small increases. Review of Cycle 3 Safety Analysis The Cycle 3 Reload Safety Analysis Checklist was reviewed to determine the impact of the proposed +32 step mi'salignment Tech. Spec. The results obtained are as follows: l 1.1.1 Moderator Coefficient - No impact. l

1.1.2 Doppler Coefficient - tio impact. 1.1.3 Beff and Prompt tieutron Lifetime - t;o impact. 1.1.4a f;ormalized Trip Reactivity vs. ?osition - Substantial margin to the Cycle 3 limit curve exists to accommodate the mis-alignment. 1.1.4b Trip Reactivity Worth - The estimated change in trip worth due to the misalignment is <.04%Ap. The available mar-gin'to the limit is >.8% ap. 1.1.4c Shutdown Margin - No impact. l.2 Peaking Factor Basis a. FAH - Margin remains to the 1.55 HFP limit after the misalignment penalty is applied. b. FAH vs. Power. - A proposed Tech. Spec. revision changes the FAH multiplier from 0.2 to 0.3. With this change, margin to the limit exists except at very low power levels (< 20%). Westinghouse Thermal-Hydraulics has confirmed that there is no safety concern since the core limits at these low pcwers are not peaking factor dependent. 1.2.2, Overpower kw/ft - Substantial margin to the limit is avail-able to accommodate the increase resulting from the mis-alignment. 1.4.2 Power Peaking Factors - See F discussion in previous sec ' q tion. L i(.VGH B

l.4.3 Rod Insertion Limits -'f;o impact. Boron Dilution During Refueling - No impact'. 2.2.1 2.3.1 Rod Withdrawal From Subcritical - The available margin (s 30%) to the 100 pcm/in limit can accommodate the mis-alignment. 2.3.2 Single Rod Withdrawal - Greater than 4% margin to the FaH limit is available after a misalignment penalty is applied. 2.3.3 Rod Withdrawal at Power - Substantial margin to the current limit is available. 2.6.1 Rod Out of Position - Greater than 4% margin to the FaH limit is available after a misalignment penalty is applied. 2.6.2 Dropped RCCA Substantial margin to the FaH vs. Dropped Rod Worth limit is available to accommodate the misalignment. 2.7 Rod Ejection - For the HZP rod ejection, the worst case is an upward misaligned C-Bank rod'with the other C-and 0-Bank rods positioned at the HZP insertion limits. Margin to the F and ejected rod worth limits is available to acconno-g date the misalignment. For the HFP rod ejection, rod mis-alignments result in only small peaking factor and ejected rod worth increases (s2%). Adeouate margin is available to the HFP F and ejected rod worth 1imits. q 2.8 Steamline Break - No impact. Rods are tripped to N-1. q x Power vs. Core Height (LOCA) - See discussion in pre-2.9.3 F vious section. t i .P w .x

T b: +G Conclusion The analysis performed to support the +32 rod misalignment Tech. Spec. has shown that the peaking factor increases resulting from mis-alignments of +32 steps are small at significant power levels and are within the available ma~rgin and conservatisms employed in the Cycle 3 design. At very low power levels, FaH and Fxy limits may be exceeded. However, for the reasons discussed above, exceeding these peaking factor limits at low powers does not pose a safety concern. Routinely, plants are operated with rods essentially fully withdrawn. This mitigates large upward rod misalignments which analysis has found to be much more limiting than downward rod misalignments. This, com-bined with the. fact that actual rod misalignments are very rare in West-inghouse plants, makes the occurrence of large misalignments very unlikely. While this report specifically addresse'd only N-Loop operation, the above conclusions are valid for N-1 Loop operation as well since-the peaking factor characteristics of the N and N-1 loop cores are very similar. 1633Fi6

o 4 - 0.eie renc e s

1.. Letter f rom M. M. Judkis (Westinghouse) to J. Seiber (Duquesne Light) containing Cycle 2 Rod Misalignment Analysis (DLWP0-66).

-2. Camden, T. M., et. al.; "PALADON - Westinghouse Nodal Computer Code",-WCAP-9485A (Proprietary) and WCAP-9485A (Non-Proprietary), l. December (1978). 3. Ankney, R. D., "PALAD0N - Westinghouse Nodal Computer Code", l WCAP-9485 Supplement 1 (Proprietary) and WCAP-9486 Supplement 1 (Non-Proprietary), September 1981. ,e 9 1633F:6'

e

1.,.

~ TABLE 1 }f: R0b M!5ALIGi? MENT FAH Ai;D OUAD?t'.T TILT Eq. Xenon Re#erence Misalignment Power I Misalignment Misalignnant' Quadrant Quadrant Level FAH FaH FAH Tilt-Tilt 1 (%) Limit No Xenon Eq. Xenon (%) (%) 5% 1.99 2.04 1.93 .0 4.9 -45% 1.81 1.72 'l.66 0 1.5 100% 1.55 1.52 1.50 0 1.1 F 5 t e i r i r 1633F:6

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/.

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