ML20079L159

Proposed Tech Spec,Providing for Use of Vantage 5H Fuel in Subsequent Plant Operating Cycles

ML20079L159
Person / Time
Site:	Beaver Valley
Issue date:	10/15/1991
From:	DUQUESNE LIGHT CO.
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ML20079L152	List: ML20079L151 ML20079L159 ML20079L176
References
NUDOCS 9111060130
Download: ML20079L159 (32)

Text

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21 5AFE" . M '5 t

BASES . .

211 REA** 3 CORE The resteicti:ns of this safety limit prevent overheating of the fuel arc p ssible clacc'9; perf ora' ion which woulo result in the release of fission pro-du:ts to the rea: tor :colant. Overheating of the fuel cladding is preventet by restricting Nei operation to within the nucleate boiling regime ehere the '

heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temoerature.

Operation above the upper boundary of the nucleate boiling regime could re-suit in excessive cladding temperatt.res because of the onset of ceparture f rom nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coef-ficient. DNB is not a directly aseasurable parameter during operation and there-  !

fore THERKAL POWER and Reactor Coolant Temperature and Pressure have been relatec 1 4/ /f I to DNB through thefr4/]t9srW correlation. The N DNB correlation has been oevelopec to' precict the DNB flux and the location of%N8 for axially uni-I forfr anc non uniform heat flux distributions. The local DOB heat flux rati ,

DNBR, defined as the ratio of the heat flux that would caus e DNB at a particular core location to the local heat flux, is indicative of the sargin to DNB. l WRB l i 'a l of h NJA d/ i sm pe t'4n, yo ~la p a-IJ5df /t n 1 s a e s, nd nt ip adj(r si t s ed o .3. Ini v u  ;

o e n a rc t ob 6i y t 9 p con c i n 1 e t j

{ 1 no o ur nd s os 6 a a to i e g t D f ,

r og _ ng, or t, ns / /_

The curve of Figure 2.1-1 shows the loci of points of THERMAL POWER, Reacto-Coolant System ressure and hverage ter.perature for which the minimum DNBR is no less than ,or the average enthalpy at the vessel exit is equal to the I c.ocvEeIy- Whk Db)At8 >g n . 6 2. " ' ,f Thepurgesarfbase{og } , of a:$$ g ru J e.-

1at,halyhy*tchannelffcor,1.F. ..." _ M...' i 8 7...

??. t.'..t.je. e S k *!.A..*. S...._.*.M.I.a.[r %_._t _,

N is included for an increase in F at reduced power based on the expressionf p novidtd in #t. CM & @lkrMM Hm Ts Af AM r ROL & .

2 - us: a c.3 + m

'AH dcrc P i:%ofrectien cf RAKO THEEL r%1 Eft- _. __

/ mThese imiting l heat flux conditions are higher than those calculated for

/the range of all control rods fully withdrawn to the saximum allowable control

/ rod insertion assuming the axial power imbalance is within the limits of the .

/ f(AI) function of the Overtemperature AT trip. When ths axial power imbalance l is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trip will reduce the setpoint to provide protection consistent with -

core safety limits. _ _.

g SuU N mxf BEAVER VALLEY - UNIT 2 B 2-1 A lto'.ntco etsege f20 POSED f'R ADom o70000 p PDR

BASES 2.1.1 REACTOR CORE INSERT 1 The DNB design basis is cs follows: there must be at least a 95 porcent probability that the minimum DNBR of the limiting fuel rod during condition I and II events is greater thun or equal to the DNBR limit of the DNB correlation being used (the WRB-1 corrolatin* in this application) . The correlation DNBR limit is based on tl entire applicablo experimental data set such that thoro is a 95 porcent probability with 95.porcent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.i7 for the WRB-1 corrolation).

Incorporating the peaking factor uncertainties in the correlation limit results in a DNBR d9 sign limit value of 1.21. This DNBR value must be mot in plant safety analyses using nominal values of the input

, paramotors that were included in the DNBR uncertainty evaluation.

In iddition, margin has boon maintained in the design by mooting a safety analysis DNBR limit of 1.33 in performing safety analysos.

INSERT 2

• The Thermal-Hydra.ulic and non-LOCA analysos that were conducted for i

Unit 1 bounds the Unit 2 analyses (i.e., F(N oH) of 1.61). The LOCA and Core Design licensing basis is 1.55.

I

LIMITING SAFET* SYSTEM SETTINGS flASES specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EM NTS.

The method: logy to derive the trip setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating witnin the allowances of these uncertainty segnitudes. Rack drift in excess of the Allowable Value exhibits the bchavior that the rack has not met its allowance. Peing that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the ellowance that is more than occasional, may be indicative of more serious problems and should warrant further investigntion.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the autosatic protective instrumentation channels and provides manual reactor trip capability.

Power Ray e, Neutron Flux The Power Range, Nsutron Flux channel high setpoint provides reactor core protection against rsractivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low setpoint provides redund-ant protection in the power range for a power excursion beginning from low power.

The trip associated with the low setpoint may be manually bypassed wiien P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERML POWER) and is automatically reinstated when P-10 becomes inactive (three ef the four channels indicate a power level below approximately 10 percent of RATED THERML POWER),

Power Ranoe, Neutron Flux, High Rates The Power Range Positiva Ratt trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip ::ceplements the Power Range Neutron Flux High and Lov trips to ensure that the critai'ia are apt for rod ejectic from artial power.

\ CJJ, dis q' w Ds)R 2 l

The Power Range Negative Rate rip previots protection to ensure that the minimum DNBR is maintained above for control rod drop accidents. At high I power a multiple rod drop accident could cause local flux peaking which, when I in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR 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 tri for those control rod drop accidents for which DNBRs will be greater than l

• \ .tAL DW82Y, l

BEAVER VALLEY - UNIT 2 B 2-3 l

PROPOSEb 1

REACT!VITY CONTROL SYSTEMS ROD DROP TIME L191 TING CONDITION FOR OprRATION [ _.

3.1.3.4 The. individual full length (thutdow and control) rod drop time from the fully withdrawn position shall be < seconds from beginning of decay l of stationary gripper coil voltage to 3ashpot entry with:

a. T,yp > 541'F, and

b. All reactor coolant pumps opert. ting.

APPLICARIt.!TY: MODE 3.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

I SURVIILLANCf PIOUIREMEHIi.

4.1.3.4 The rod drop time of full length rods shall be demonstrated through steasurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any raintenance on or mo61fication to the contrcl rod drive system which could affect the drop time of those specific rods, and l

C. At least once per la months.

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l l

BEAVER VALLEY - UNIT 2 3/4 1-23 MO$0S Eb

3/4.2 POWER O!STRIBUTION LIMITS g[g // .w DN82 S w WES _ _ _ _ _

/

The specifications of this section provide assur.ince i fuel integrity during Contiition ! (Nomal Operation) and !! (Incidents of tNocerate Frequency) events by; (a) maintaining the minimum DNBR in the core 3 @ during normal I operation and in short terin transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed ce-sign criteria. In addition, limiting the peak linear power density during Con-dition ! ever.ts provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is tot exceeded.

The definitions of hot channel factors as used in thest speci'ications are as follows:

F9 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear powwr along the rod with the highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX O!FFERENCE ( AFD),

The limits on AXIAL FLUX O!FFERENCE assure that the gF (Z) upp. Lound envelope times the normalized axial peaking factor is not exceeded during either normal c,peration or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrin xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup cunditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by tne appropriate fractional T14ERMAL POWER level. The periodic updating of the target flux difference value is necessa.'y to reflect core burnap considerations.

Although it is intended that the plant will be operated with the AXIAL FLUX O!FFERENCE within the target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target hand at reduced TilERMAL POWTR levels.

This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time l

BEAVER VALLEY - UNIT 2 8 3/4 2-1 NC903E&

l

. POWER DI$TRIBUTION LIM!TS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUClF.AR ENTHALPY HOT CHANNEL FACTORS nF (Z)

ANDFg(Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

d. The axial power distribution, expressed in tems of AX1AL FLUX DIFFERENCE is maintained within the limits. .

The relaxation in F as a ft.netion of THERMAL POWER allows changes in the radial power shape for all t,ernissible rod insertion limits.

Fhwillbe maintained within its limits provided conditions a thfu d above, are maintained.

When an gf measurement is taken, both experimental scror and manuf acturing ,

tolerance must be allowed for. $X is the appropriate experimental error allow-

, ance for a full core map taken with the incore dete.ar flux mapping system and-3% is the appropriate allowance for manufacturing tolerance.

Thespecified*,imitofFhcontainsan85allowanceforuncertaintieswhicn moans that normal, full power, three loop operation will result in Fh less than or equal to the design limit specified in the CORE OPERATING LIMITS REPORT.

el

/- fpf,/od ow r uce theglue ft D rati . C dit/ fs a la[,e l

(o s du on t hVgen c' gi Th c ig r s a ng . NS , a c tel' off s y to ow na es 1. f . th 9

< t se i oc s ab up 2 00 /MT .

1. T s r i s low g:

. ig ini f1 vs .2

/ /:

of 2.- ri pac (, . s. 59

/4 .

0 1 Dif 11 1

of oef cie 5 .0 of .038 s. 59 /

The radial peaking factor F,y (Z) is measured periodically to provide assurance that the hot channel' factor, F F,y limit for Rated Thereal Power (F P)q (I),-remains provided in the CORE within OPERATING its limit. The LIMITS REPORT was determined free expected power control maneuvers over the

-full range of burnup conditions in the core.

3/4.2.4 '0UADRANT POWER TILT RATIO The Quadrant Power Tilt Ratio limit assures that the radial power distri-l bution satisfies the design values used in the powar c:apability analysis.

L -BEAVER VALLEY - UNIT 2 8 3/4 2-4

, ff0fosen

POWER 0!sTR18U?!0N L!MI?$

4 BfM1 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CNANNEL FACTORS gF (Z)

N AND F Insert 2.

ruel rod bowing reduces the value of DNB ratio. Margin has toen maintained between the DNBR value used in the saf ety arialyses and the design limit to offset the rod bow penalty arid ether penalties which may apply.

l 8tAvu VAu.ty'- usitf 2 J

POWER DISTRIBUTION LlHITS BASES 3/4,2.4 QUADRANT POWER TILT RATIO (Continued)

Radial power distribution measurements are made dJring startup testing and periodically during power operation.

The limit of 1.02 at whic.) corrective action is required provides ONB and linear heat generation rate orotection with x y plane power tilts.

The two-hour time allowance for operation with a tilt condition greater than 1.02 but lest than 1.09 is provided tc allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainly on F isg reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0, 3/4.2.5 0NB PARAMETER $

The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a h h h$

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument DAJA/E M readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurnment of the RCS total flow rate is adequate to detect flow dog mdation and ensure correlation of the flow indication channels with i measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

l BEAVER VALLEY - UNIT 2 8 3/4 2-5 M0posth

, '3/4.4 REACTOR COOLANT SYSTEM EASES 3/4.4.1 RE ACTOR COOLANT LOOPS AND COOLANT CIRCUL ATION The plant is designed to operate with all reactor coolant loops in opera-tion and maintain DNBR above during all normal operations and anticipated l transients. In MODES 1 and 2, ith one reactor coolant loop not in operation, this specification requires that the plant be in at least ICT STANDBY within tkL n DA)8o2 limlb In H0DE 3, a single reactor ecolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assu.*ed in the analysis for the c>ntrol-rod bank withdrawal from a suberitical condition, two operating coolant loops lire recuired to meet the DNB design basis S r this Condition 11 event when the rod control system is capable of contrrl bank rod withdrdwal.

In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient her,t removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor .;,olant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with or.e or more RCS cold legs less than or equel to 350*F are provided to prevent RCS pressure tran-s' u ts, caused by energy additions from the secondary systes, w'.iich could exceed the limits of Appendix G to 10 CF3 Part 50, The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by ristrict-ing starting of the RCPs to when the secondary water temperature of each steam generatur is less than 50'F above each of the RCS cold leg temperatures.

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l BEAVER VALLEY - UNIT 2 B 3/4 4-1 PROPosEb

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- . .. - . -. - - _ . - - =

. ATTACHMENT B Beaver Valley Power Station, Unit No. 2 Proposed Technical specification Change No. 57 VANTAOC 5H FUEL UPGRADE A. DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would address changes similar to those incorporated in the BV-1 Technical Specifications to provide for the use of VANTAGE SH fuel in subsequent plant operating cycles.

The proposed changes incorporate an increased control rod drop time and replace referenua to the DNB limit and W-3 R-Grid correlation noted in the Bases with the safety analysis DNBR limit and WRB-1 correlation respectively.

B. BACKGROUND The mechanical design of the upgcaded fuel assembli s will continue to use the reconstitutable top nozzles, dobrls filter bottom nozzles, snag resistant grids, standardized fuel pr.ots, enriched Integral Fuel Burnable Absorbers (IFBAs) and axial blankotu that were previously incorporated in the cycle 2 coload fuel design. Additional VANTAGS SH design features being incorporated here include the removal of thimble plugs, and the addition of zircaloy grids and reduced thimble tubo diancter.

Attachment D provides a report " Plant Safety Evaluation for Beaver Valley Power Station Unit 2 VANTAGE SH Fuel Upgrado" which sumnarizco the safety evaluations that were performed to confirm the acceptable use of these tptions for the safe operation of the plant. This Plant Safety Evaluation (PSE) will serve as a reference safety evaluation / analysis report for the transition from the present core to a core containing the upgradod fuel features and will be used as a basic reference document to support future reload safety evaluations using upgraded fuel designs.

C. JUSTIFICATION The Cycle 4 reload fuel will incorporate upgraded Westinghouse VANTAGE SH fuel design features in accordance with the PSE.

These design changes are currently part of the licensing basis in other plants including BV-1 and root all fuel aasombly and fuel rod design criteria. No changes to the nuclear design philosophy or methoda are required because of the upgraded fuel design. The reload design philosophy includes the evaluation of the reload core key safety parameters which comprise the nuclear design dependent input to the UFSAR safety evaluation for each reload cycle. The key safety parameters are evaluated for each reload cycle and if one or more of the parameters fall outside the l

bounds assumed in the safety analysis, the affected transients will be re-evaluated and the results documented in the cycle specific reload safety eve.luatior..

i

ATTACliMENT B, continued Proposed Technical Specification Changa No. 57 Page 2 D. SAFETY ANALYSIS The VANTAGE 5 11 fuel assembly design incorporatos the use of zircaloy grids with standard diamotor Westinghouse fuel rods.

The zircaloy grid material is thickor than the current inconel grid design due to the differenco in material strength properties. The use of thimble tubos with a reduced diamator  ;

identical to those uuod in the 17 x 17 Optimited Fuel Assembly and VANTAGE 5 fuel assembly designs is required with the zircaloy grida due to the increased grid metal thickness. The reduced i thimble diamotor will increase the design rod drop timo from the current maximum of 2.2 seconds to 2.7 seconds.

Thorofore, specification 3.1.3.4 has boon revised to incorporate the 2.7 l nocond rod drop time. This slower rod drop timo will affect tho l results of the limiting UFSAR transients affected by rod drop i timo sJch as Loss of Forced Reactor Coolant Flow, Locked Rotor, RCCA Bank WILhdrawal from Subcritical and Rod Ejection. Those and the other applicable UFSAR accidents have boon re-evaluated '

using the slower. rod drop time. Demonstration fuel assemblics l with zircaloy grids have been used in Westinghouse cores, including Bonver Valloy Unit 1, and zircaloy grids have been used in many hostinghouso roload cores sine.o the early 1980's, Thorofore, the uso of zi- aloy grids ha. boon proven based on a successful and safo operating history.

The DNB design basis has boon modifi<d to address beth the 17 x 17 standard and VANTAGE 511 fuel assemblios using the NRB-1 DNB correlation," New Westinghouse Correlation WRB-1 for Predicting Critical lleat Flux in Rod Imndlos with Mixing Vane Grids" WCAP-8762-P-A, and MINI-RTCI " MINI Revised Thermal Design Proceduro (MINI RTDP)" WCAP-12178-P. The WRB-1 DNB corralation is based entirely on rod bundle data and takes credit for a significant improvement in the accuracy of the critical heat flus, predictions over previous DNB correlations. With the MINI-RTDP methodology, peaking factor uncertainties are combined statistically with the DNB corrolation uncertainties to obtain the overall DNBR uncertainty factor which is used to dufine the design limit DNBR that satisfios the DNB design critorion. This criterion states that the probability that DNB will not occur on the most limiting fuel rod is at least 95% (at 95% confidence level) for any Condition I or II event. Tho 95/95 limit DNBR using the WRB-1 DNB correlation for the 17 x 17 standard and VANTAGE 5 11 fuel assemblies is 1.17. The design limit DNBR for typical and thimble cells is 1.21. The limit DNBR is

' conservatively increased in the DNB safety analyses to provide DNB margin to offset the offects of rod bow and any other DNB penalties that may occur while providing flexibility in the design and operation oi~ the plant. The safety analysis limit DNDR with 9% margin is equal to (Design limit DNBR)/(1.0 .09)al.33. The current maximum rod bow penalty

ATTACHMENT B, continued proposed Technical Specification Change No. 57

. pago 3 is 1.3% DNDR, this penalty is also applicable to VANTAGE SH fuel annomblies based on the similaritics betwoon the 17 x 17 standard and VANTAGE SH fuel ancomblios including fuel rod diameter. fuel rod pitch and grid spacing. Thereforo, adequate margin .a the safety limit DNDR is available to cover any rod bow penalties and, in addition, satisfy all current thermal hydraulic design criteria. Worst case fabrication tolerances along with fuel rod and assembly growth are used te datormino thu Initial s uol rod to nozzle gro9th gaps in the evaluation of fuel rod performanco summarized in Section 2.4 of the PSE. This is in complianco with condition 1 of thu VANTAGE 5 NRC Safety Evaluation Report.

Thimble plug assemblies have been used to limit the core bypaes flow in guide thimble tubos that are not in control rod positions or containing other core components. Evaluations have been performed assuming the complete removal of thimble plug assemblies from the coro. The effect of thimble plug removal on the distribution of outlet loss coefficients has boon ovaluated and it was demonstrated that the variations in the outlet loss coefficient are within the bounds of the sensitivity utudios.

Safety evaluations for thimble plug removal have boon performed to show that the licensing analysos will bound plant oporations with or without thimble plugs rfmoved from the core.

The Cycle 4 reload and future reload cores will contain fuel assemblies that incorporate reconstitutable top nozzles, debris filter bottom nozzica, snag resistant grids, standardized fuel pollets, integral fuel burnable absorbors and axial blankota as well as the VANTAGE SH zircaloy grids. Thase donign changes are currently part of the licensing basis in other plants and meet all tuel assembly and fuel rod design critoria.

The decris filter bottom nozzle is designed to inhibit debris from entering the active fuel region of the coro and thereby I improves fuel performance by minimizing debris related fuel failures. This is a low profile bottom nozzle design made of Ptainless stool with a reduced end plate thickness and Icq height. This low profile design is structurally and hydreulically equivalent to the existing bottom nozzle design.

The reconstitutable top nozzle differs from the current design in that a groove is provided in each thimble thru-hole in the nozzle plate to facilitato attachment and removal, and the nozzle plate thickness is reduced to provide additional space for fuel rod growth. Also, a long tapered fuel rod bottom and plug is used to facilitate removal and reinsortion of the fuel rods.

The standardized fuel pellets are a refinement to the current pellet design with the objective of improving manufacturability while maintaining or improving performance. This design incorporates a recluced pellet length, modification to the previous dish size and the addition of a chamfer.

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l ATTACHMENT B, continu d Proposed Technical Specification Change No. 57

- Page 4 The snag-resistant grids contain outer grid straps which are modified to help prevent assembly hangup from grid strap interference during fuel assembly renoval. This was accomplished by changing the grid strap corner geometry and adding guide tabs on the outer grid strap.

The integral fuel burnable absorber (IFBA) coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin zirev.ium diboride or enriched zirconium diboride coating (less than one mil) on the pellet cylindrical surface along the central portion of the fuel stack length. The enri:hed IFBA properties are the same mechanically and chemically (except for the small reduction in density) as the natural boron material. A safety evaluation for the application of enriched boron in the current IFBA design was performed and concluded that the enriched boron did not adversely affect core safety considerations. IFBA's provide power peaking and muderator temperature coefficient control.

The axial blanket consists of natural uranium dioxide pellets at each end of the fuel stack to reduce neutron leakage and to improve uranium utilization. The axial blanket pellets are of the same design as the enriched and IFBA pellet designs except for an increase in length. The length difference in the axial blanket pellets will help prevent acciCental mixing with the enriched and IFBA pellets.

The spent fuel pool criticality analysis is applicable to the upgraded fuel features including the use of VANTAGE SH fuel. The analysis is conservative since the fuel assemblies were modeled without taking credit for flux reduction due to neutron absorption in the grids.

The transient analyses in Section is of the UFSAR have been re-analyzed where required or re-oveluated to include the increased rod drop time and revised fuel assembly design parameters to ensure the safety analyr!s limits are satisfied.

UFSAR changes to reflect the revised accident analyses are attached to the Plant Safety Evaluation. The design bssoa for the Locked Rotor event has boon revised in a manner similar to that requested by the NRC staff for BV-1. It was found that 18%

of the fuel rods could experience DNB with minimum DNBR's less than the safety analysis DNBR limit. This was calculated based on a fucl rod power census which is conservative for Cycle 4 operation and is expected to bound future cycles. For the cases with and without offsite power, the peak reactor coolant pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits. The peak clad temperature calculated for the hot spot remains less than 2700*F and the amount of zirconium-water reaction is small.

Therefore, it is concluded that the integrity of the primary coolant system is not endangered and the core will remain intact with no consequential loss of core cooling capability. The radiological consequences of the Locked Rotor event are being j analyzed to show compliance with the siting guidelines of 10 CFR j

ATTACHMENT B, continucd Proposed TAchnical Specification Change No. 57 l -

Pago 5 100 using methodology consistent with Standard Review Plan Section 15.3.3-15.3.4. The radiological analysis will assume the instantaneous release of 18% of the fuel rod gap activity to the reactor coolant system. The analysis results will be similar to those found ur BV-1 where the projected donos are a small fraction (i.e., loss than 10%) of the 10 CFR 100 guidelinos, and will be loss than 10 CFR 100 guidelinec even if the accident should occur coincident with an lodino spike. Radiation exposure to personnel assigned to the control room will also be ovaluated and the results compared to the limits provided in Gonoral Design Crittria 19. Additional UFSAR changes will be duvoloped to address the paramotors used and the analysis results found.

These changes will be incorporated in a futuro updato. Au long as the analysis results are verified to be within the limits no further action is intendod, If the results are found to be outsido the limits, wo will notify the NPC accordingly. The analysis results will be maintained in a filo and may be reviewed at NRC request.

The VANTAGE SH fuel assembly design han boon approved by the NRC and has boon used in other plant roload cores. The other upgraded fuel features described in the Plant safety Evaluation have been implomonted in other Westinghouse reload cores in accordance with 10 CFR 50.59 and do not requiro prior NRC approval, however, the change in rod drop timo requires a Technical Specification change.

The proposed Technical Specification change and supporting Plant Safety Evaluation are provided for NRC review and approval to document the safety analysis and ovaluations performcd to ensure the proposed changes are consistant with accepted methodology and required safety limits. The attached UFSAR changes are provided for background information and will be incorporated into the UFSAR in a future update following approval of the proposed Technical Specification change. The proposed Technical Specification Change and Bases changes have boon ovaluated in accordance with approved methodology and have boon shown to satisfy the applicabin acceptance critoria. Therefore, based on the above, the proposed changes have boon determined to bo safe and will not reduce the safety of the plant.

E. NO SIGNIFICANT HAZARDS EVALUATIOM The no significant hazard considerations involved with the proposed amendmant have been ovaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:

The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating licenso for a facility licensed under paragraph 50.21(b) oc paragraph 50.22 or for a testing facility involves no significant nazards consideration, if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increaso in the probability or consequences of an accident previously evaluated; or

. ATTACHMENT B, continued Proposed Technical Specification Change No. 57

. Page 6 (2) Create the possibility of a now or differont kind of accident from any accident previously evaluated; or (3) Involvo a significant reduction in a margin of safety.

The foll.owing evaluation is provided for the no significant hazards considoration stanuards.

1. Does the chango involve a significant increase in the probability c' consequencon of an accident previously evaluated?

The Cycle 4 reload fuel design will incorporate additional Westinghouse VANTAGE SH fuel design features as well as reconstitutable top nozzles, dobris filter bottom nozzles, anag resistant grids, standardized fuel pellets, enriched Integral Fuel Burnable Absorbors (IFBAs) and axial blankots that were incorporated previously in the cycle 2 reload dusign. As a result of those changes, the maximum control rod drop time has boon changed from 2.2 seconds to 2.7 seconds.

The VANTAGE SH and the Standard 17 x 17 fuel assembly are hydraulically equivalent. Implementation of the VANTAGE SH tuel design will not significantly change the core physics enaracteristics. The proposed changes have boon assessed as described in a Plant Safety Evaluation (PSE) to confirm the acceptable use of those options from a core design and safety analysf1 stand point. No increase in the probability '

of occurrence of any accident was identified. Extensive re-analysos were undertaken to demonstrate compliance with the revised Tachnical Specifications. The methods used to perform the analyses have been previously approved by the NRC. The results, which include transition coro effects, show changes in the consequences of accidents previously evaluated, however, the results are 311 clearly within NRC acceptance critoria and demonstrace the capability to operate the plant safely. The major components that determine the structural integrity of the fuel assembly are the grids.. Mechanical testing and analysis of the VANTAGE SH zircaloy grid and fuel assembly have demonstrated that the VANTAGE 5H structural integrity under seismic /LOCA loads will provide margins comparable to the standard 17 x 17 fuel assembly design and will moet all design bases.

The design bases for the Locked Rotor event has been revised in a manner similar to that requestod by the NRC staff for BV-1. It was found that 18% of the fuel rods could experience DNB with minimum DNBR's loss than the safety analysis- limit. This calculation is based on a fuel rod power census which is conservative for cycle 4 operation and is- expected to bound future cycles. For both the cases with and without offsito power, the peak reactor coolant pressure

l ATTACHMENT B, continued Proposed Technical Specification Chango No. 57

- Page 7 reached during the transient is lous than that which would cause stress to exceed the faulted condition stress limits.

The peak clad temperature calculated for the hot spot remains loss than 2700*F and the amount of zirconium-water reaction is small. Therefore, it is concluded that the integrity of the primary coolant system is not endangered and the core will remain intact with no consequential loss of coro cooling capability. The radiological consequences of the Locked Rotor event are boing analyzod to show compliance with the siting guidelinos of 10 CFR 100 using methodology consistent with Standard Review Plan Section 15.3.3-15.3.4. The radiological analysis will assume the instantaneous release of 18% of the fuel rod gap activity to the reat.cor coolant system. The analysis results will be similar to those found for BV-1 where the projected doses are a small fraction (i.e., loss than 10%) of the 10 CFR 100 guidelinos, and will be less than 10 CFR 100 guidelines even if the accident should occur coincident with an iodino spike. Radiation exposure to personnel assigned to the control room will also be evaluated and the results compared to the limits provided in General Design critoria 19.

Additional UFSAR changes will be developed to address the paramotors used and the analysis results found. Theen changes will be incorporated in a futuro updato. As long as the analysis results are verified to be within the li'.dito no further action is intended. If the results are found to be outsido the limits, we will notify the NRC accordingly. The analysis results will be maintained in a file and may be reviewed at NRC request.

The proposed Technical Specification change and Bases changes have been evaluated in accordance with the Plant Safety Evaluation based on approved methodology and have been shown to satisfy the applicable acceptance criteria.

Thorofore, based on the above, the proposed change will not result in an increase in the probability or consequences of a previously evaluated accident.

2. Does the change create the possibility of a now or different kind of accident from any accident previously evaluated?

The proposed changes will not significantly affect the overall method and manner of plant operation and can be accommodated without compromising the performance or qualification of safety related equipment. These changes have been evaluated in accordance with approved methodology and satisfy the criteria set forth in the regulations.

Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

. 4

. ATTACllMENT B, continusd Proposed Technical Specification Chango No. 57

. page il i

The evaluations and analyson to support the proposed Technical Specification change and operation of the plant with VANTAGE SH fuo.\ show some changes in the consequences of previously analyzed accidents. In some cason, an increase in ovent consequences occure, however, in all casos the analysis results will be within the plant design and NRC safety acceptance critoria limits. Plant operation will be maintained within required limits, thorofore, the proposed changes will not involve a significant reduction in a margin of safety.

F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activition ascociated with this licenso amendment request satisfies the no significant hazards consideration standards of 10 _CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified.

G. ENVIRONMENTAL EVALUATION The proposed changes have boon ovaluated and it hac boon determined that the changes do not involvo (i) a significant hazards consideration, (ii) a significant chango in the typos or significant increase in the amounts of any offluents that may be released offsite, or (iii) a significant increase in individual or cumulativo occupational radiation exposure. Accordingly, the proposed changos moet the eligibility critorion for categorical l oxclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed changos is not required, t

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. ATTACllMEllT C

.- Boavor Valley Power Station, Unit lio. 2 Proposed Technical Specification change 140. 57

., 6 Typed Pages: B 2-1, i B 2-2 B 2-3*

B 2-4 B

2-5**

B 2-6*

B 2-7 D 2-8*

3/4 1-23 B 3/4 2-1 B 3/4 2-4 8 3/4 2-5 B 3/4 4-1 Added due to shifting from previous page.

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2.1 SAFETY LIMITS

. BASES 2.1.1 REACIQR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the funi cladding is prevented by restricting fuel operation to within l

the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly 'bove the l coolant saturation temperature. l l

Operation above the upper boundary of the nucleate boilire egime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly i measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been l developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The-local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB- at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting fuel rod  ;

during Condition I and II events is greater than or equal to the DNBR 1 limit of the DNB correlation being used (the WRB-1 correlation in this  !

. application). The correlation DNBR limit is based on the entire '

applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation). ,

Incorporating the peaking factor uncertainties in the correlation limit results in a DNDR design limit value of 1.21. This DNBR value must be met in plant safety analyses using nominal values of the input paramotors that were included in the DNBR uncertainty evaluation. In addition, margin has been maintained in the design by meeting a safety analysis DNBR limit of 1.33 in performing safety analyses.

The curve of- Figure 2.1-1 shows the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR. limit,I or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

The curves are conservatively based on an enthalpy hot channel factor, F H The Thermal-Hydraulic and non-LOCA analyses 6g, of 1.62.

that were conducted for Unit 1 bounds the Unit 2 analyses (i.e., F N

g of 1.62). The LOCA and Core Design licensing basis is 1 55. These will bourd actual plant operation which is restricted to an BEAVER VALLEY - UNIT 2 B 2-1 Proposed

SAFETY LIMITS

. BASES 2.1.1 REACTOR _ CORE. cont.inqqd igu 11mst of 1.ss <see the CORE OPE aAr1Na t1MIrS REPORT). An allowance is included for an increase in F g at reduced power based on the expression provided in the Core Operating Limits Report (COLR).

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 AT 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 COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclidos contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The e r. t i r e Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drif' assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value in BEAVER VALLEY - UNIT 2 B 2-2 )

Proposed

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LDilT1H9._DAE.LTY SYSTEM SETT11{GS

. BASES 2.2.1 REACTOR _ _ TRIP SYSTEM INSTRUMEliTAT10N SETPoINTS.__ continued acceptable sinco an allowance has boon made in the safety analysis to accommodate this error. An optional provision has boon included for determining the OPERABILITY of a channel when its trip notpoint is found to excoed the Allowablo Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and consor components in conjunction with a statictical combination of the other uncor".aintion in calibrating the instrumontation. In Equdtion 2.2-1, Z + c + S s TA, the interactive offects of the errors in the rack and the sensor, and the "as measured" values of the orrors are considorod. Z, as specified in Table 2.2-1, in porcent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measuromont. TA or Total Allowance is the difference, in porcent span, betwoon the trip sotpoint and the value used in the analysis for reactor trip. R or Rack Error is the "as measured" deviation, in porcent span, for the l affected channel from the specified trip notpoint. S or Sensor Drift is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the trip sotpoints is based upon combining all of the uncertaintics in the channels. Inhoront to the datormination of the trip sotpoints- are the magnitudou of those channel uncertainties. Sensors and other instrumentation utilized in those channels are expected to be capable of operating within the allowances of those uncertainty magnitudos. Rack drift in excess of the Allowablo Value exhibits the behavior that the rack has not met its allowanco. Heing that there is a small statistical change that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

Ma nua 1._Rc.attsr_..TI.iR The Manual Reactor Trip is a redundant channel to the automatic protectivo instrumentation channels and provides manual reactor trip capability.

Power Rangg2 Neutron Flux The Power Rango, Neutron Flux channel high setpoint providos reactor coro protection against reactivity excursions which are too rapid to be protected by temperature. and pressure protective circuitry. T low sotpoint provides redundant protection in the power range for i power excursion beginning from low power. The trip BEAVER VALLEY - UNIT 2 B 2-3 Proposed

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j k ' . , G. SADEI SYSTEM SETTINGS V

~

.inted with the low setpoint may be manually bypassed when P-10

. s active (too of the tour power range channels indicato a power

..i e invel of c: 9ve approximately 10 percent of RATED THERMAL POWER) and

'a :tomat ically reinstated when P-10 becomes inactive (thre.e of the m HT v channels indicate a power level below approximately 10 percent fjg V TO TH7R AT. POWCR) .

19r_B2D9.?.m .' '.'pn Flux, High Rates

-"i r ..,,

The Power Range Positive Rate trip provides protection against Tid flux increases whicn are charac ristic of rod ejection events

~ am any pc we r 1c'..... Specifically, this trip complements the Power

,, u nge Neutron Flux Migh and Low trips to ensure that the criteria are j met for rod ejection from partial power. _

The Power Range Negative Rate trip provides protection to ensure that ths minimum DNBR is maintained above the design DNBR limit for

PJ control rod drop accidents. At high power a murtiple rod drop

+d accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by nction of the automatic rod control system, could cause an unconservative local DNBR to exist. The rower Range Negative Rate trip will reeent this from occurring by tripping the reactor. No credit is .aken for operation of the Power Range Negative Rate trip for those control rod drop accidents for wnich DNBRs will be greater than the design DNBR limit, i

htermediate and Source Rance, Nucle 31_f_ly).

The Intermediate and Sourca Range, Nuclear Flux trips provide reactor core protection during reactor s.avtup to mitigate tha consequences of an uncontrolled rod cluster cont,01 assembly bank withdrawal fron a suberitical condition. Thesu trips provide rodundant protection to the low setpoint trip of the vower Ranga, Neutron Flux channels. Th Source Range Channels will initiate a -

reactor trip at about 10^g counts per second unless manually blocked when P-6 becomes active. The intermediate range channels will initiate a reactor trip at a cutrent leve.:. 3roportional to approximately 25 percent of RATED TIiERMAL POWER unless manually blocked when P-10 beco.nes active. Although no explicit credit was taken for operatien of the Source Range Channels in una accident analiscs, operability requirements in the Technical Specifications will ensure that the source Range Channels ar e available to mitigate ,

the consequences of an inadvertent control bank withdrawal in MODES

g. 3, 4 and 5.

Overtemperftpre AT The Overtemperature AT . ip providas core protection to prevent DNB for all combinations of pressure, power, cool 1t temperature, and axial power distribution, provided that the transient is slow with respect to piping transit, thermowell, and RTD responsa time delays BEAVER VALLEY - UNIT 2 L 2-4 Proposed 1

. . . .~ - ,

LIMITING SAFETY SYSTEM SETTINGS BASES m . _ _

from the core to the temperature detectorc (about 4 seconds), and pressure is within the rang' between the High and Low pressure reactor trips. This setpoinc includes corrections for tLanges in g density and .ou c capacity of water with temperature and dynamic j compensation for transport, thermowell, and RTD response time delays frcm the core to RTD output indication. With normal axial power distribution, this reactor trip limit is always below the core safety ,

limit as shown on Figure 2.1-1. If axial peaks are greater than dasign, as indicated by the difference between top and bottom power

_ range r.uclett detectois, the reactor trip is automatically reduced according to the notation in Table 2.2-1.

Overpower a_T The overpos .r AT reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, 1

limits the required range for overtemperature AT protection, and provides a backup to the High Neutron Flux Trip. The setpoint i includes corr ections for changes in density and heat capacity of

( water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output

- indjcation. The Overpower AT trip provides protection to mitigate the consequences of various size steam line breaks au reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releasc.

Pressurizer Pressu_r.e r The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High ,

Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and in therefore set lower than the set precsure for these valves (240, psig). The Low Pressure trip .

protects against 1cw pressure whica could lead to DNB by tri' ing the

= reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Pressure trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER or turbine impulse chamber prear"re at approximately 10% of full power equivalent); and an increasing power, automatically reinstated by P-7.

On decreasirg pover, the low setpoint trip is automatically E

blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10 percent cf full power equivalent); and on increasing power, 7 automatically reinstated by P-7.

1 Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by la miting the water level BEAVER VALLEY - UNIT 2 B 2-5 Proposed 6

l e - __- __ - ___________- _-_-_____--__----___ - - _ - - _ - - . - - - - - - - - - - - - - - - - - - - - - - -

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LIlilT11[Q_EMETY SYSTEM SETTllLQS IIASES to a volume rufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. On decreasing power, the pres 3urizer high water level trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 'O percent of fu)1 power equivalent); and on increasing power, automatically reinstated by P-7. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

LQEE_pf Flow The Loss of Flow trips provide core protection to prevent DNB in tc.c event of a loss of one or more reactor coolant pumps.

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90 percent of nominal full loop flow. Above 30 percent (P-8) of RATED TiiERMAL POWER, automatic reactor trip will occur if the flow in any single locp drops below 90 percent of nominal full loop flow.

Etsam Gerlqtator Watqr Level The Steam Generator water uovel Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint p ovides allowance that here will be sufficient water inventory in the steam acnerators at the time of trip to allow for starting delays of the auxitiary feedwater system.

MDiqrvoltane._imd Underf requency - RiactoE Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than or.e reactor coolant pump.

The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. TiL; delays are incorporated in the underfrequency anc undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For undervoltage, the delay is set sa that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underirequency trip setpoint is reached shall not exceed 0.6 seconds.

On decreasing power, the Undervoltage and Underfrequency Reactor Coolant Pump bus trips are automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a BEAVER VALLEY - UNIT 2 B 2-6 Proposed

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LIMITING CAFETY SYSTEM SETTINGS

. BASES turbine impulse chamber pressure at approximately 10 percent of full power equivalent) ; and on increasing power, reinstated automatically by P-7.

Tyrlino Tria A Turbine Trip causes a direct reactor trip when operating above P-9. Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

Safety Iniection Inout f rQJg_ JH If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic a.tuation logic channels wjll iniciate a reactor trip apon any signal which initiates a safety injection. This trip is provided to protect the ccre in the event of a IOCh. The ESP instrumentation channels which initiate a safety injection signal are shown in TABLE 3.3-3.

Reactor Coolant Pumo Breaker Position Trio The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. These trips are blocked below P-7. The open/close position trips assure a rc ctor trip signal is generated before the low flow trip setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position seucings is required to enhance the overall reliability of the Reactcr Protection System.

Reactor Trio System Interlocks The Reactor Trip System interlocks parform the following functions:

P-6 Above thc setpoint P-6 allowa the manual block of the Source Range reactor trip and de-energizing of the high voltage to the detectors. Below the setpoint source range level trips are . automatically reactivitated and high voltage restored.

P-7 Above the setpoint P-7 automatically enables reactor trips on low flow or coolant pump breaker open in more than one primary coolant loop, reactor coolant pump bus undervolatage and underfrequcncy, pressurizer low pressure and pressurizer high level Below the setpoint the abcVe listed trips are automatically blocked.

BEAVER VALLEY - UNIT 2 B 2-7 l Proposed

LIMITINC SAFETY SYSTEM SETTINGS BASES P-8 Above the setpoint P-8 automatically raables reactor trip _

on low flow in one or more primary coolant loops. Below the setpoint P-8 automatically blocks the above listed trip.

P-9 Above the setpoint P-9 automatically enables a reactor trip .

on turbine trip. Below- the setpoint P-9 automatically  !

blor :s a reactor trip on turbine trip.

P-10 Above the setpoint P-10 allows the manual block of the  :

Intermediate Range reactor trip and the low setpoint Power Range reactor trip; and automatically blocks the Source ,

Range reactor trip and de-energizes the Source Range high I voltage power. Below the setpoint the Intermediate Range I reactor trip are automatically reactivated. Provides input l to P-7.

P-13 Provides input to P-7. )

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BEAVER VALLEY - UNIT 2 B 2-8  !

Proposed i

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. . . _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ - _ . _ . _ _ . . . . _ _.-_._...m__ . _ _ _ _ _ - ._ .

. 4 .

. REACTIVITY CONTROL SYSTEMS

- ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4. The individual full length ~ (shutdown and control) rod drop time. from the fully withdrawn position shall be 5 2.7 seconds from l beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T avg 25 l'F, and

b. All reactor coolatt pumps operating.

APPLICABILITY: MODE 3.

ACTIOE:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE REQUIREMENTS 4.1.3.4- The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically .affected individual rods following any maintenance on or modification to the control rod drive system .hich w could affect the drop time of those specific rods, and e At least once per 18 months.

BEAVER VALLEY - UNIT 2 3/4 1-23 Proposed

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141 2 POWER DISTRIBUTION LIMITS a BASES

. x=

The specifications of this coction riovide assurarce of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core 2 the design DNBR limit during normal operation and in short l term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance critoria limit of 2200*F is not exceeded.

The definitions of hot channel factors as used in these specifications are as follows:

F9(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at-core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

FN Nuclear Enthalpy Rise Hot Channel factor, is defined as the AH integral of linear power along the rod with the highest integrated power to the average rod power.

1/4.2.1 AXIM FLUX DI.'TXBENCE iAFD1 The limits on AXIAL FLUX DIFFERENCE assure that the F9(Z) upper bound enve' ope times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution tollowing power changes.

Target -flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be

[ inserted near their normal position for steady state operation at high power Invels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED '1HERMAL POWER is the target flux difference at RATED THERMAL POWER for the l associated core burnup conditions. Target flux differences for other THERMAi POWER levels are obtained by multiplying the RATED THERMAL l POWER- plue by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to j reflect core burnup considerations.

i Although it le intended tha' tne plant will be operated with the AXIAL FLUX DIFFERENCE ' thin t he target band about the target flux difference, during rat a plreit T.$2RMAL POWER reductions, control rod motion will cause the 3FD to deviate outside of the target band at reouced THERMAL POWER levels. This deviation will not affect the xenon redistribution Ecfficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target' band) provided the time BEAVER VALLEY - UNIT 2 B 3/4 2-1 Proposed i

, r. 1 POWER DISTRIBUTION LIMITS .

BASES 2]4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNFL FACTORS Fg(Z) AND F H " "

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained,

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.

The relaxation in FN H s a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. FN ""# " " * *E 6H conditions a through d above, are maintained.

When an Fg measurement is taken, both experimental crror and manufacturing tolerance must be allowed for. 5% is the appropriate experimental error allowance for a full core t , .:aken with the incore detector flux mapping system and 3% is tha appropriate allowance for manufacturing tolerance.

The specified limit of FN contains an 8% allowance for uncertainties which means that normal, full power, three loop operation will result-in F N 1 ss than or equal to the design limit H

specified in the CORE OPERATING LIMITS REPORT.

Fuel rod bowing reduces the value of DNB ratio. Margin has been maintained between the DNBR value used in the safety analyses and the design. limit to offset-the rod bow penalty and other penalties which may apply.

The radial pc.aking reactor Fxy(Z) is measured periodically to provide assurance that the hot channel factor, Fg(Z), remains within its limit. The F xy limit for Rated Thermal Power (F P) provided in- the CORE OPERATING LIMITS REPORT was determined from expected power control maneuvers over the full range of burnup conditions in the core.

BEAVER VALLEY - UNIT 2 B 3/4 2-4 Proposed

, POWER DISTRIBUTION LIMITS l .

BASES 3/4.2.4 QU).DRANT POWER TILT RATIO The Quadrant Power Tilt Ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are nada during st'rtup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does ne'. correct the tilt, the margin for uncertainly on Fg is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt ir, excess of 1.0.

3/4.2.5 DNiLE2BAMETER The limits on the DNB related parameters assure that each of the parameters are maintained within the normal stesdy state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAK assumptions and have been analvtically demonstrated adequate to maintain a minimum DNBR greater than or equal to the design DNBR limit throughout each analyzed l transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance cf thest parameters throught instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

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BEAVER VALLEY - UNIT 2 B 3/4 2-5 Proposed 1

. 3/4.4 REACTOR COOLANT SYE.TRI

• BASES a

3/4.4.1 REACTQB COOLANT LOOPS AND COOLANT CIBCl M TlQH The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design DNBR limit during all l normal operations and anticipated transients. In MODES I and 2, with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank uithdrawal from a suberitical condi'. ion, two operating coolant loops ao required to meet the DNB design basis for this condition II event wr.en the rod control system is capable of control bank rod withdrawal.

In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The operatior. of one Reactor Coolant Pump or one RHR pump providen adequate flow to ensure mixing, pravent stratification and produco gradual reactivity changes during boron s encontration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs lese than or equal to 350*F are provided to prevent KCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of appendix G by restricting starting of the RCPs to when the secondary water temperature ot each steam generator is less than 50*F above each of the RCS cold leg temperatures.

BEAVER VALGEY - UNIT 2 B 3/4 4-1 Proposed I

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ATTACHMENT D Beaver Valley Power Station, Unit No.2 Proposed Technical Specification Change No. 57

.9 2 PLANT SAFETY EVALUATION FOR BEAVER VALLEY POWER STATION UNIT 2 VANTAGE SH FUEL UPGRADE l

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