Text

Proposed Tech Specs Re VANTAGE-5 Fuel Design
	ML20065P204
Person / Time
Site:	Vogtle
Issue date:	11/29/1990
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20065P198	List: ELV-02166, Application for Amends to Licenses NPF-68 & NPF-81,changing Tech Specs for VANTAGE-5 Fuel Design; ML20065P204;
References
	NUDOCS 9012130155
	Download: ML20065P204 (625)
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O Enclosure 1 _

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Vogtle Electric Generating P17.at Units 1 and 2 ReoJest for Technical Specifications Changes VANTAGE-5 fuel Design ,

Basis for Proposed Chanaes '

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I 9012130155 901129 PDR ADOCK 05000424 l P PDC

ENCLOSURE 1 V0GTLE ELECTRIC GENERATING PLANT REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES VANTAGE-5 FULL DESIGN BASIS FOR PROPOSED CHANGES -

s Proposed Chances The proposed Technical Specifications changes are listed in the attached table. The proposed changes are based on the operational and core design benefits provided through use of the VANTAGE-5 fuel design in conjunction with improved computer code methodologies.

Ik111 In order to implement a long-term fuel management strategy planned by Georgia Power Company (GPC) for the Vogtle Electric Generating Plant (VEGP)

Units 1 and 2, it has been decided to use reload fuel assemblies of tha Westinghouse VANTAGE-5 design. This will require a transition from the current LOPAR fueled core to a full VANTAGE-5 fueled core. The transition is expected to be complated by the third loading of VANTAGE-5 fuel for each VEGP unit. This long term strategy includes the implementation of high (7 energy 18 month fuel cycles with high capacity factors, low leakage loading patterns, and extended fuel burnup. In addition, GPC plans to power rerate

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VEGP from 3411 to 3565 MWt; however, the Technical Specifications changes for power rerate are not being pursued in this license amendment request.

The Technical Specifications changes provided in Enclosure 3 are based on the use of VANTAGE-5 fuel specific design features-and use of improved computer code methodologies. The fuel design features include the smaller diameter (OFA) fuel rods, mid-span zircaloy grids, Intermediate Flow Mixer (IFM) grids, natural uranium oxide (UO2) axial blankets, Integral Fuel Burnable Absorbers (IFBA), extended fuel burnup, and reconstitutable top nozzles. The new computer code methodologies relative to the current VEGP safety analyses include the BART/ BASH (large-break LOCA), NOTRUMP (small-break LOCA), and improved THINC-IV (thermal-hydraulics) computer codes, as well as the Revised Thermal Design Procedure (RTDP), WRB-2 DNB correlation, and Relaxed Axial Offset Control (RAOC) strategy. Each

-Technical Specifications change associated with either the change to VANTAGE-5. fuel or change in methodology is discussed below.

i.nclosure 3 provides instructions for incorporating the proposed VANTAGE-5 Fuel' Design Technical Specifications changes. Since VEGP uses combined Units 1 and 2 Technical- Specifications, the instructions for incorporating the proposed changes for each unit will be done in two phases. The first

_ph ase is the proposed Technical Specifications changes involving the first -

fuel loading of VANTAGE-5 fuel in Unit 1 in late September 1991. At this point in time, the proposed Technical Specifications changes discussed

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- below only apply to VEGP Unit 1, while VEGP Unit 2 will continue to operate with the existing Technical Specifications until its initial loading of

! ENCLOSURE 1 Page 2 VANTAGE-5 fuel scheduled for February 1992. Therefore, the second phase fully implements the proposed VANTAGE-5 Fuel Design Technical Specifications changes discussed below for both VEGP Units I and 2.

Enclosure 3 provides the Technical Specifications changes which implement this two phase process.

The Technical Specifications changes for the core safety limits and over-temperature delta-T and over >ower delta-T setpoints (Technical Specifications 2.1 and 2.2) were c1anged as a result of using DNB margin gained through the use of the VANTAGE-5 IFM grid feature, improved THINC-IV code, WRB-1 and WRB 2 DNB correlation, and RTDP. BART/ BASH, NOTRUMP, and RAOC implementation were also factored into these Technical Specifications changes. The Technical Specifications core limits and setpoints changes accommodate the use of higher peaking factors for F delta-H and FQ, deletion of thimble plugs, axial blankets, IFBAs, the VANTAGE 5 rod and lattice geometry, higher steam generator tube plugging, mixed fuel type core DNB and LOCA peak clad temperature transition penalties, and future >ower rerate changes. The higher peaking factors will be revised throug1 the use of the Core Operating Limits Report (COLR).

The Technical Specifications changes to describe the WRB-1, WRB 2, and RTDP O

V correlation / methods are discussed in the BASES changes to the Technical Specifications- (2.1.1 BASES, 3/4.2 BASES, and 3/4.4 BASES). The use of the VANTAGE-5 fuel design requires the use of the WRB 2 correlation. The WRB-1 correlation is used with the RTDP methodology for the LOPAR fuel design.

The RTDP methodology was used to statistically combine the uncertainties in the DNB correlations with the uncertainties in the plant instrumentation, to provide operational margin to increase the maximum reactor coolant system average temperature (TAVG) limit, reduce the pressurizer pressure limit, and reduce the reactor coolant system flow by the reduction in the flow measurement uncertainty (Technical Specifications changes to Sections 3/4.2.5 and 3/4.2.5 BASES). These DNB parameter Technical Specifications changes provide for improved plant operating flexibility.

The Technical Specifications change to Section 3/4.1.3.4 is proposed to increase the control rod drop time from 2.2 to 2.7 seconds. The VANTAGE-5 IFM grid feature slightly increases the core pressure drop. The VANTAGE 5 guide thimble inside diameter is slightly rd.ced compared to VEGP's current LOPAR fuel design. Both of these VANTiGE-5 design changes result

.in an increased control rod drop time. Therefore, the safety analyses performed for VEGP's VANTAGE-5 fuel design inccrporated an-increased Technical Specifications control rod drop time of 2.7 seconds.

The Technical Specifications changes to 3/4.2.1,3/4.2.2,3/4.2 BASES, and 6.8.1.6 are proposed to implement the NRC approved RAOC methodology. The O

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ENCLOSURE 1 Page 3 RAOC methodology provides for increased plant operating flexibility in axial power shape control. The RAOC methodolos 1983), forms the basis for changes to the axial;y, flux difference WCAP and 10216-P FQ A (June surveillance Technical Specifications changes. Since RAOC is being implemented with the use of VANTAGE 5 fuel, the FQ surveillance Technical Specifications changes (includes use of a W(z) function) will replace the current Fxy surveillance Technical Specification (Section 3/4.2.2). The proposed changes to the Core Operating Limits Report (COLR) references listed in Technical Specification 6.8.1.6 include the switch from the current VEGP constant axial offset control methodology to RA0C. The new axial flux difference operating limit envelope associated with RAOC implementation will be revised through the use of the COLR. The Fxy limits currently given in the COLR will be removed and replaced with cycle and burnup-dependent W(z) functions to be implemented with the Fg surveillance Technical Specifications proposed changes.

Technical Specifications changes to 3/4.1.2.5, 3/4.1.2.6, and 3/4.5.4 are proposed to the RWST Minimum Water Temperature to provide additional plant operating flexibility. The changes reduce the minimum RWST water temperature limiting condition for operation and surveillance limit from 540F to 440F and from 500F to 400F, respectively. These proposed changes were only affected by those safety analyses being performed to support the 9 use of VANTAGE-5 fuel; therefore, these proposed changes were included in the VANTAGE 5 safety analyses.

Proposed Technical Specifications changes to the BASES Section 3/4.5.4 will provide additional operational and core design flexibility. The change uses the existing RWST boron concentration limits and applies the Leak Before Break (LBB) methodology to the large Break Loss of Coolant Accident (LBLOCA) to allow for control rod insertion following the LBLOCA.

Allowance for control rod insertion upon a reactor trip during a LBLOCA (all rods inserted less two control rods-- one ejected rod and one stuck rod) eliminates the need for higher RWST boron concentration limits to maintain suberiticality at cold conditions during long term core cooling, post-LOCA conditions. Without LBB methodology application to allow for control rod insertion during post-LOCA conditions, the RWST boron concentration limits in the Technical Specifications would have had to be increased which would place undesirable duty on plant equipment. Enclosure 4 supports the use of the LBB methodology to allow for control rod insertion during the LOCA.

The Technical Specifications changes to 3/4.5.1 provide plant operational flexibility by widening the accumulator water level limits. It is desirable to widen the accumulator water level range to accommodate potential changes in water levels that may be experienced over an 18 month.

operating cycle. These proposed Technical Specifications changes only affect those safety analyses being performed to support the use of VANTAGE-5 fuel; therefore, these proposed changes were included in the G VANTAGE-5 safety analyses.

O ENCLOSURE 1

~Page 4 The Technical Specifications change to 3/4.3.2, Table 3.3 3 is proposed to raise the P ll setpoint from 1970 psig to 2000 psig. The change to the P-ll setpoint does not affect any safety analysis results in the VEGP FSAR.

The change to this setpoint increases the band between the safety injection

.(SI) block and the SI setpoint. The wider band provides improved plant operational flexibility.

A significant hazards evaluation (Enclosure 2) has been performed to support GPC's conclusion that these proposed Technical Specifications changes do not involve significant hazards considerations. Safety evaluations and analyses (Enclosure 4 and Appendices A, B, and C) also have been performed by Westinghouse to support these proposed Technical Specifications changes.

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SUMMARY

AND JUSTIFICATION FOR THE V0GTLE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS CHANGES FOR VANTAGE 5 FUEL ggg iggMg) Descriotion justification 22 2.1 Change to core limits, This change is a result of changes 2 4,2 8 2.2 to the 0101 and OPDT sesociated with the VANTAGE $ fuel 2 9,2 10 letpoints. and to provide operational 2 11 flexibility.

B21 2.1.1 Basis Additivo of the WRB 1 This change reflects the DNB B22- and WR1 il correlation correlation used in onetyses.

B 3/4 2 1 3/4.2 tesis B 3/4 2 2 B 3/4 2 4 5 3/4 6 1 3/4.4 Bests 3/4 1 11 3/4.1.2.5 RWsf minim a solution This change is to ellow 3/4 1 12 3/4.1.2.6 tenperature operationet flexibility.

3/4 1 13 3/4 5 10 3/4.5.4 8 3/4 5 2 3/4.5.4 Basle RWsf Bases This change is to allow operational flexibility.

3/4 1 19 3/4.1.3.4 Revised rod drop time This change is a result of changes to less then or ocpel to associated with the VANTAGE $ fuel.

2.7 seconds The effect of this increase on the safety onelysis bas been considered.

3/4 2 1 3/4.2.1 Axist Flux Difference This change is mode to give the 3/4 2 2 and fo(Z) changes plant operating flexibility end 3/4 2 4 3/4.2.2 also es a result of changes 3/4 2 6 associated with the VANTAGE 5 fuel.

3/4 2 7--

B 3/4 2 1 B 3/4.2 Basis B 3/4 2 2 5 3/4 2 4 3/4 2 13 3/4.2.5 DNB parentter chen0es This change is made to give the B 3/4 2 5 3/4.2.5 Benis plant operating flexibility ord eleo es a result of changes eseccleted with the VANTAGE 5 fuel.

3/4 3 33 3/4.3.2 Pressurizer pressure This change is to provide trip setpoint operationet flexibility.

3/4 5 1 3/4.'$ .1 Accunulator Water Level This change is to provide Range operationet flexibility.

6 21 6.8.1.6 Core Operating Limits This change Is to provide 6 Zie Report operationet flexibility.

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. J Enclosure 2 Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE 5 fuel Design 10 CFR 50.92 Evaluation O 9 l.

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D Enclosure 2 V0GTLE ELECTRIC GENERATING PLANT REQUEST FOR TECHNICAL SPECIFICATIONS CHANGES VANTAGE 5 FUEL DESIGN 10 CFR 50.92 EVALUATION Pursuant to 10 CFR 50.92, each application for amendment to an operating.

license must be reviewed to determine if the proposed change involves a significant hazards consideration. The amendment, as defined below, describing the-Technical Specifications changes associated with implementation

- of. VANTAGE 5 fuel assemblies, has been reviewed and deemed not to involve significant hazards considerations, The basis' for this determination follows.

Background

f-~ In order tofimplement the long term fuel management strategy planned by Georgia

( Power for the VEGP Units 1 and 2, it has been decided i.v.-uss= rsicad fuel assemblies of the Westinghouse VANTAGE 5 design. This strategy includes the implementatio'n o'f high energy eighteen month- fuel cycles with high capacity factors,-low leakage loading patterns, and extended fuel burnup. - Westinghouse VANTAGE-5 fuel-has been-designed to accommodate these operating characteristics-by inclusion of specific design features'and by use of improved methodologies previously approved by the NRC, or under review.- These design features inclu'de Intermediate Flow Mixer (IFM) grids, natural. uranium oxide .(U0 2 ) axial blankets,

~ Integral Fuel. Burnable Absorbers (IFBA), extended fuel burnup, and Reconstitutable Top Nozzles (RTN). . Each feature _ and methodology change is discussed below.

Inte'rmediate Flow Mixer -(IFM) Grids - The IFM grids promote flow mixing within the assembly and provide increased DNB margin. This additional' margin can then be applied to accommodate higher design peaking' factor ' values of F-del _ta H and FQ while still- maintaining operational margin .to new core O

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13 V safety-limits and trip setpoints (Technical Specifications Sections 2.1 and 2.2 and associated BASES). The IFM grids also offer additional structural support for the fuel rods, reduced rod bow and improved seismic stability of the assembly. The IFM grid feature also increases the core pressure drop; thus the control rod drop time (Technical Specifications Section 3.1.3.4) was increased as a result of the increase in core pressure drop.

Axial Blankets - Axial blankets consist of six inches of natural UO2 pellets (instead of enriched uranium) within the fuel rods at each end of the fuel stack, which reduces neutron leakage and improves uranium utilization.

Inteoral Fuel Burnable Absorber (IFBA) - The advantages provided by the IFBA include improved neutron utilization, since the neutron absorber material is a thin boride coating on the fuel pellets themselves. This reduces the need for individual absorber rods which displaces water molecules. Also, burnable absorber rods continue to absorb neutrons later in core life when-this function is no longer necessary, creating a residual penalty which inhibits A econor.ic fuel use. In contrast, the IFBA's capacity to a.bsorb neutrons is U limited and precisely matched to the reactivity depletion of the fuel. The resulting more efficient use of neutrons enables fuel to last longer, reducing fuel cost. Because control of reactivity is not required at the blanketed ends, the IFBA is only added to the fwl in the central portion of the rods.

-Extended Burnuo - Longer fuel cycles result in increased rod growth and production of fission product gas which increases rod internal pressure.. By increasing the fuel rod plenum and providing additional space between the fuel rod and nozzle, the effects of increased internal rod pressure and rod growth resulting from longer cycles and extended fuel burnups to a lead rod average of 60,000 MWD /MTU can be accommodated.

Reconstitutable Too Nozzle IRTN) - The assembly reconstitutable top nozzle provides the capability to replace damaged fuel rods. This feature avoids the discharging of an entire assembly for minor fuel damage. The advantage is increased fuel reliability and reduced fuel costs.

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V Each of these features supports the safe, efficient fuel management scheme planned for the VEGP Units 1 and 2. Margin provided by the VANTAGE 5 fuel design and by recently NRC approved (or under review) methodologies will be used to revise core design parameters that will provide increased operational flexibility (Technical Specifications Sections 3/4.1.2.5,3/4.1.2.6,3/4.3.2, 3/4.5.1,3/4.5.4 and associated BASES). These methodologies include the use of BART/ BASH, NOTRUMP, and the improved THINC-IV computer codes, as well as the Revised Thermal Design Procedure (RTDP), the WRB-2 DNB correlation, and the Relaxed Axial Offset Control (RA00) Strategy.

BART/ BASH - These codes are utilized in the analysis of the large break LOCA.

BART employs rigorous mechanistic models to_ generate heat transfer coefficients appropriate to the actual flow and heat transfer regimes experienced by the fuel rods. The BART code has been coupled with a loop model to form the BASH code. BART provides the entrainment rate for a given flooding rate. The coupling of the BART code with a loop code produces a dynamic flooding transient which reflects the close coupling between core thermal-hydraulics and loop behavior. The BASH code provides a realistic thermal-hydraulic simulation U(7 of the reactor core and RCS during the reflood phase of a large break LOCA by utilizing a sophisticated treatment of steam / water flow phenomena in the RCS.

The use of this methodology provides additional margin in peak clad temperature to support an increase in FQ peaking factor.

NOTRUMP - This code is utilized in the analysis of the small break LOCA.

Features of this code include thermal non-equilibrium in all fluid volumes, flow regime dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple stacked fluid nodes and regime dependent heat transfer correlations. In NOTRUMP, the RCS is nodalized into volumes interconnected by flow paths. The broken loop is modeled explicitly with the intact loops lumped into a second loop. The transient behavior of the system is determined from the governing conservation equations of mass, energy and momentum applied throughout the system. The use of this methodology also provides additional margin in peak clad temperature to support an increase in FQ peaking factor.

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C j _- Imoroved THINC-IV - This code is used te perform thermal-hydraulic calculations for the non-LOCA transients. It calculates coolant density, mass velocity, 'enthalpy, void fractions, static pressure and DNBR distributions along the flow channels within a reactor core under all expected accident conditions. The improved THINC-IV design modeling methodology currently under NRC revir.w (WCAP-12330 P, August 1989) replaces the present THINC-IV model.

Reviser. Thermal Desian Procedure (RTDP) - With this methodology, DNBR uncertginties associated with uncertainties in plant operating parameters, nuclea> and thermal parameters and computer codes aro statistically combined with tne DNBR correlation uncertainties. Since the parameter uncertainties are considered in determining the design limit DNBR values, the plant safety analyses are performed using values of input parameters without uncertainties.

The RTDP methodology was used to provide additional operating margin for the proposed Technical Specifications Section 3/4.2.5 and associated BASES changes 2.1.1 and 3/4.2.

O WRB This is a DNB correlation that is used for fuel assemblies utilizing IFM grids. It takes credit for significant improvement in the accuracy of the critical heat flux predictions over previous ONB correlations as well as for the reduced grid-to-grid spacing of the . VANTAGE 5 fuel assembly mixing vanes.

The WRB-2 DNB correlation is incorporated into the Technical Specifications BASES Sections 2.1.1, 3/4.2, and 3/4.4.

j Relaxed Axial Offset Control- (RAOC) - This strategy was developed to provide l- wider control band widths and more operator freedom than with Constant Axial

!. Offset Control (CAOC). RA0C provides wider control bands particularly at reduced power by utilizing core margin effectively. The wider operating space

-increases plant availability by allowing quicker plant startups and increased maneuvering flexibility without reactor trip or reportable occurrences. The RAOC. strategy-forms the basis for Technical Spacifications changes 3/4.2.1, 3/4.2.2 and the associated BASES, and Section 6.8.1.6.

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Analysis The proposed Technical Specifications changes to reflect the operational and core design benefits provided through use of the VANTAGE 5 design in conjunction with NRC approved (or under review) methodologies are summarized in Enclosure 1 and the changes are provided in Enclosure 3. The effect of these changes on plant safety is discussed in detail in the Safety Evaluation portion of this submittal (Enclosure 4). The implementation of VANTAGE 5 fuel includes conservatism far power rerating to 3565 MWt, additional increased peaking factors, steam generator tube plugging up to ten percent, a reduction in thermal design flow and various other plant operational margins.

Therefore, the NSSS design parametsrs to which the safety analyses have been performed in support of VANTAGE 5 fuel have been generated to account for these future changes. However, these future changes are not being requested in the licensing amendment.

Results Ba.'d on the information presented above and the Safety Evaluation in

( .i aure 4, the following conclusions can be reached with respect to 10 C M 50.92.

1. The VANTAGE 5 fuel related Technical Specifications changes do not involve a significant increase in the probability or consequences of an accident previously evaluated in the VEGP FSAR. The mechanical design changes associated with VANTAGE 5 fuel result in the capability for relaxation of analytical input parameters such that increased margin can be generated without violation of any acceptance criteria. This margin can then be applied towards relaxation of operational limits such as reduced safety injection flow or increased steam generator tube plugging. In each case however, the appropriate design and acceptance criteria are met. No new performance requirements are being imposed on any system or component in order to suppo n the revised analysis assumptions. Subsequently, overall

~p lant integrity is not reduced. Furthermore, the parameter changes are a;sociated with fea+ures used as limits or mitigators to assumed accident g scenarios and are not accident initiators. Therefore, the probability of

() an accident has not increased.

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D The consequences of an accident previously evaluated in the VEGP FSAR are not increased due to the VANTAGE 5 fuel related Technical Specifications changes. Evaluation of the radiological consequences of extended fuel burnup to lead rod average burnups of 60,000 MWD /MTV has been performed for the transition to VANTAGE 5 fuel. These evaluations have confirmed that the doses remain w' thin previously approved acceptable limits as well as those defined by 10 CFR 100. Therefore, the consequences to the public resulting from any accident previously evaluated in the VEGP FSAR has not significantly increased.

2. The VANTAGE 5 fuel related Technical Specifications changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the VEGP FSAR. Mechanical evaluations have been performed on the fuel assemblies, fuel rods and control rod drop times to confirm that their function and reliability are consistent with the originally supplied equipment. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the fuel transition. The presence of VANTAGE 5 fuel assemblies in the core or O the revised analytical assumptions have no adverse effect and do not U challenge the performance of any other safety related system. Therefore, the possibility of a new or different kind of accident is not created.

3. The VANTAGE 5 Technical Specifications changes do not involve a significant reduction in the margin of safety. The margin of safety for fuel related parameters _ associated with the VANTAGE 5 transition are defined in the BASES to those Technical Specifications identified in Enclosure 1. These BASES and the supporting Technical Specifications values are defined by the accident analyses which are performed to conservatively bound the operating conditions. These operating conditions are defined by the Technical Specifications such that the regulatory acceptance limits will not be exceeded. Performance of analyses and evaluations for the VANTAGE 5 fuel transition have cenfirmed that the operating envelope defined by the Technical Specifications continues to be bounded by.the rev.ised analytical basis, which in no case exceeds the acceptance limits. Therefore, the margin of safety provided by the

! analyses in accordance with these acceptance limits is maintained and not l Q reduced.

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Conclusion Based upon the preceding informatit,h, it has been determined that the proposed changes to the Technical Specifications do not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from t.ny accident previously evaluated or involve a significant reduction in a margin of safety.

Therefore, it is conwiuded that the proposed changes meet the requirements of

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10 CFR 50.92 (c) and do not involve a significant hazards consideration.

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Enclosure 3 Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE 5 Fuel Design lastructions for Incorocration O .

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j V0GTLE ELECTRIC GENERATING PLANT REQUEST FOR TECHNICAL $PECIFICATIONS CHANGES VANTAGE 5 FUEL DESIGN l

INSTRUCTIONS FOR INCORPORATION The-proposed amendment to the Technical Specifications would be incorporated as follows:

Phase 1 Attachments la and Ib Effective following the Vogtle 1 Cycle 3 Shutdown (Effective as of Vogtle 1 Cycle 4 Startup)

Reagv6 Pace Insert Pace

!!! and IV* III and IV* .'

Y and VI* V and Vl*

XV and XVl* -

XV and XVI*

, -XXIll and XXIV

XXI!! and XXIV * -

2 1 and 2 2 2 1 and 2 2, 2 2a 0 2 3 and 2 4 2 5 and 2 6 2 7 and 2 8 2 3 and 2 4 2 5 and 2 6 2 7 and 2 8 2 9 and 2 10 2 9 and 2 20 2 11 2 11, 2 12 and 2 13, 2 14 and 2 15, 2 16 and 2 17, 2 18 and 2 19 B 2 1 and B 2 2 8 2 1 and B Z-la, B 2-2 3/41-11and3/4lel2 3/4 1 11 and 3/4 1-12 3/4 1 13 and 3/4 1 14* 3/4 1 13 and 3/4 1-14*

3/4 1 19 and 3/4 1 20* 3/4 1-19 and 3/4 1 20*

3/4 2 1 and 3/4 2 2 3/4 2 1 and 3/4 2 la, 3/4 2 2 3/4 2 3 and 3/4 2 4 3/4 2 3 and 3/4 2 4-3/4 2-5 and 3/4 2 6 3/4 2 5 and 3/4-2 6 3/4 2 7 and 3/4 2 8* 3/4 2 7 and 3/4 2 7a, 3/4 2 8*

3/4 2 13 3/4 2 13 3/4 3 33.cnd 3/4 3 34* 3/4 3 33 and-3/4 3 34*

3/4 5 1 and 3/4 5 2* 3/4 5 1 and-3/4 5 2*

3/4.5 9* and 3/4 5 10 3/4 5 9* and 3/4 5 10 B 3/4 2 1 and B 3/4 2 2 B 3/4 2 1 and B 3/4 2 2-B 3/4 2 3 and B 3/4 2 4 8 3/4 2 3 and B 3/4 2 4 L B 3/4 2 5 and B 3/4~2 6 8 3/4 2 5 and B 3/4 2 6, B 3/4 2 7 and B 3/4 2 8, B 3/4 2 9 and B 3/4 2 10 B 3/4 4 1 and B 3/4 4 2* B 3/4 4 1 and B 3/4 4 2*

B.3/4 5 l* and B 3/4 5 2 B 3/4 5 l* and B 3/4 5 2 6 21 and 6 21a 6 21 and 6 21a,-6-21b

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l ENCLOSURE 3 (Cont'd)

V0GTLE ELECTRIC GENERATING PLANT REQUEST FOR TECHNICAL $PECIFICATIONS CHANGE 5 VANTAGE 5 FUEL DESIGN -

INSTRUCTIONS FOR INCORPORAT10N -

The proposed amendment to the Technical Specifications would be incorporated as follows:-

Phase 2 Attachments 2a and 2b: Effective following the Vogtle 2 Cycle 2 Shutdown (Effective as of Vogtle 2 Cycle 3 Start 90)

Remove Pace Insert pace

!!! and IV* !!! and IV*

V and Vl* V and VI*

XV and XVI* XV and XVI*

O XXIll and XXIV

2 1 and 2-2, 2 2a 2-3 and 2 4 XXIII and XXIV

2 1 and 2 2 2 3 and 2 4 2 5 and 2 6 2 5 and 2 6 2 7 and 2-8 2 7 and 2 8 2 9 and 2-10 2 9 and 2 10 2 11, 2 12 and 2 13, 2 14 and 2 15, 2 11 2 16 and 2 17, 2 18 and 2-19 B 2 1 and B2 la, B 2 2* B 2-1 and B 2-2*

3/4 1 11 and 3/4 1-12 3/4 1 11=and 3/4 1-12 3/4 1 13 and 3/4 1-14* 3/4 1-13 and 3/4 1 14*

3/4 1 19 and 3/4 1 20* 3/4 1-19 and 3/4 1 20*

3/4 2 1 and 3/4 2 la, 3/4 2-2 3/4 2-1 and 3/4 2 2 3/4 2 3* and 3/4 2 4 3/4 2 3* and 3/4 2-4 3/4 2 5 and 3/4 2 6 3/4 2 5 and 3/4 2-6

-3/4 2 7 and 3/4 2 7a, 3/4 2-8* 3/4 2 7 and 3/4 2 8*

3/4 2 13 3/4 2 13 3/4 3 33 and 3/4 3 34* 3/4 3 33 and 3/4 3 34*

3/4 5 1 and 3/4 5 2* 3/4 5 1 and 3/4 5 2*-

3/4 5-9* and 3/4 5 10 3/4 5-9* and 3/4 5-10 B 3/4 2 1 and B 3/4 2 2 B 3/4 2 1 and B 3/4 2 2 B 3/4 2 3 and B 3/4 2-4 B 3/4 2 3 and B 3/4 2 4 B 3/4 2 5 and B 3/4 2-6, B 3/4 2-7 and .

B 3/4 2 8, B 3/4 2 9 and 8 3/4 2-10 B 3/4 5 l* and B 3/4 5 2 B 3/4 5 l* and B 3/4'5 2 i

-6 21 and 6-21a, 6 21b 6-21 and 6 21a O

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Attachment la Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE 5 Fuel Design Technical Scecifications Marked Vo Paaes Effective following the Vogtle 1 Cycle 3 Shutdown (Effective as of Vcgtle 1 Cycle 4 Startup) l l (3 l '\..)

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[ RC-u RE 0.1-k & ACTOR, ccM SA FE T'/ 1.im ti (g ggit) , , , 2 2,a.

INDEX ;

i j SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I S,E, CTION PAGE

2.1 SAFETY LIMITS

, 2.1.1 REACTOR C0RE..................................... .......... 2-1 l 2.1.2 REACTOR COOLANT SYSTEM PRES $URE...... 2-1 1 NIT !

FIGURE 2.1-1 REACTORCORESAFETYLIMIT...........if............ . . . . . . .-. . . . . . . . . 2-2 1

'. 2. 2 LIMITING SAFETY SYSTEM SETTINGS, 4

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............. 2-3 wtT I

_ TABLE 2.2-l' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS. .. 2-4 ,

(mat.E 3.a-la R.E AC.ToA. TAIP WSTEM. LSTMuTAT404 Tele SE rP0:4TS -12, (vWri ] 2'.)

BASES SECTION PAGE-5 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............... B 2-3

(

4.

V0GTLE UNITS - 1 & 2 III S

e 'w tn w we- *Nw- w-+wtw=c 4-s--vv= ewe ~=ww,r"--w=w*www-+e-+w-mes e e4 er-e ew=-t--+g------- w- ~----y+-y w--w wm-waw ei-mwer e:een wy--vet--y*-h----r** +v*+--y-

i I

INDEX LIMITING CONDITIONS FOR OPERATIUN AND SURVEILLANCE REQUIREMENTS SECTION g

3/4.2 POWER DISTRIDUTION LIMITS

[ UMLfI

( 3/4.2.1 AXIAL FLUX 0!FFERENCE......... .

3/4 2 1 v y' GDR( 3.2% (DEL {TE%(3/4.21 M.lAL FLLLY. DiFFERECE (own -h3/4 l 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... 3/4 g

NGb4E ht-2NREL'Eh IW '2%J l l 3/4.2.3 3/4 2-8 NUCLEAR'ENTHALP7RISEHOTCHANNELFACTOR-Fh........... 4 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-10 3/4.2.5 DN8 PARAMETERS........................................... 3/4 2-13 j 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM O- INSTRUMENTATION........................................ 3/4 3-15 TABLE 3.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3 17 TABLE 3.3 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0!NTS..........................._

4 3/4 3-28 TABLE 4.3-2 . ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-45 f TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..................................... 3/4 3-46 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-48 Movable Incore Detectors................................. 3/4 3-49 I-Seismic Instrumentation (Common System).................. 3/4 3 50-

~

O V0GTLE UNITS - 1 & 2 V Amendment No. 32 (Unit 1)

Amendmenc Po. 12 (L* nit 2)

I INDEX n

h BASES SECTION PAGE

' 3/4.0 B 3/4 0-1 APPLICABILITY...............................................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 00 RAT 10N CONTR0L.......................................... B 3/4 1-1 3/4.1..? BORAT10N SYSTEMS.......................................... B 3/4 1-2

.1/4.1. 3 MOVABLE CONTROL ASSEMBL1ES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION L1MITS................................... B 3/4 2 1 3/4.2.1 AX1AL FLUX OlFFERENCE..... 1..... ,...................... B 3/4 3/4.2.2 and 3/4.2.3 NEAT FLUX HOT CHANNEL FACTOR an W IFAR ENTHALPYRISEHOTCHANNELFACTOR-Ffg..m.N ....

B3/43 .2 %

D IfM RE B 374s 2-1 TYPlt INDICAI U AL FLU) IFFERENCNERSUd .

g THERR R POWER. ..... ..

. . . . . . . .h.d. . [NMf 3/4.2.4 QUADRANT POWER TILT RATIO. d............. ..

B 3/4 Q '

3/4.2.5 DNB PARAMETERS. -

. . . . . . . ........................ B 3/4 3/4.3 INS 1RUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1

( 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECT 10N.............................. B 3/4 3-6

[3/42.1 WIAt n.0% DnFFEREsCE (uotT n) .. .. ... ...* 63/42-l

( 3/+.2. 2. wA 3/+.2.3 HENT Flux %7 cum 4Et FAcioA. u extegg EuTkAt.W RGE. AcT ca AN96L. FACTOR.- F$ (vast D . . . . B p+ a .2.

~

3/4.2.+ cataa r %u wt exTio (eacT h . . . . . . . . . a 3p 2-3

/4. 2. 5 w6 PARAMETER 5 (vart Q . . .. . . . . . . . ... B 3/4 # 3-) i O

V0GTLE UNITS - 1 & 2 XV l

.5 1

. INDEX l i

ADMINISTRATIVE CONTROLS l l

SECTION 6.4.2 SAFEYY REVIEW BOARD (SRB)

{ l Function.................................................. 6-9 Compositinn............................................... 6-10 Alternates................................................ 6 10 Consultants............................................... . 6-10

Meeting Frequency......................................... . 6 10 Quorum.................................................... 6-10 Review.................................................... 6-11 Audits.................................................... 6-11 Records................................................... 6-12

6. 5 REPORTABLE EVENT ACTI0N..................................... 6-13 6.6 SAFETY LIMIT V10LATION..'.................................... 6-13 6.7 PROCEDURES AND PR0 GRAMS..................................... 6-13 i

-6.8 REPORTINGRE0VIREMEN15 6.8.1 ROUTINE REP 0RTS........................................... 6-17 Startup Report............................................ 6-17 Annual Reports............................................ 6-17 Annual Radiological Environmental Surveillance Report..... 6-18 l Semiannual Radioactive Effluent Release Report............ 6-19

...... 6-21 MonthlyOperatingReports........(Wrtl)................

Core Operating Limits Report..... . . . . . _ ................ 6 21 l

6.8.2 SPECIAL REP 0RTS........................................... 6- 4.Alb 6.9 ~ RECORD RETENTION............................................ 6-22

. Core CperAb 3 LiMS Report (0*T 2.) . . . . . . . . . 0 ~Alh .

V0GTLE UNITS - 1 & 2 XXIII Amendment No. 32 (Unit 1)

Amendmeet,No. 12 (Unit 2) i

,,._,:m.,.w_.._-......,~_..r_ . . , _ . _ , , . . . . . , , , , . . . , . . . . . . . . . , . . . . . . . . . . . . _ . . . . . . . . . , . _ . , , , _ , _ _ , ~ . . . - , , , . , , , . , , _ . , _ , , . ~

. 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 SAFETY LIMITS l REACTOR CORE 4

(' 2.1.1 The combination of THERMAL POWER (NI-0041, NI-0042, NI-0043, NI-0044),

pressurizer pressure (PI-0455A, B&C, PI 0456 & PI-0456A, PI 0457 & PI-0457A, PI-0458 & PI-0458A), and the highest operating loop coolant temperature (T,yg)

(TI-0412, TI-oa22. TI-0432, TI-0442) shall not exceed the limits shown in Figure 2.1-p(hTl)orRFe. 3.1-6 (vurr 2.) . ]

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop 3

average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-mentsofSpeelfication6.6.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure (PI-0408, PI-0418, PI-0428, PI-0438)

shall not exceed 2,73_5Jiigm .

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.6.1.

, MODES 3. 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.1.

i i

Where specific instrument numbers are provided in parentheses they are for e

information only, and apply to each unit unless specifically noted (to assi't in identifying associated instrument channels or loops) and are not intended to limit the requirements to the specific instruments associated with the number.

O V0GTLE UNITS - 1 & 2 2-1

~vr , - - , - , ~ , , , , , - - - v., - - . . - - - . w-r e s, e,- ~ -- -w-.,~~ ,,w~-~--n--- - -,- m-- - ,-,--,v- w- nn,

-J_,.A,- a_A-4.4 a.%4 4ab.-4-e .m-e.de-a6 __4_.aa_&.ea, w,_&A4-a* A 10 Wa4hn- 4LA- 4_,-.4s--%p u---as=14e _m.mm,.__,_s.m_ma- e.4_6~~m.

m%- a..em..

. MO .

\ \ \ \ \ \

,O \ \ \

x s

~

g N N N. ca,1..a o,E.1ron Qx

\

N N s x x ,,,,s ~4 x x 'x x

k (2"'*\ \ x \x

~ '

j'\ '

~'

x eLi g

a s

s x% 3_x y x x N ' .

v \ 's'\ xy '

\,

O if x x x '

x '

' AECEPTABLE N opfgiries is~ \x 'N N

N ' '

[_

~

! 's 1 N 'N .

o

\\ o.:

Fm o.4

\ o.s o.s

\1.0 'Nig Tion OF AkTED THERM t POWER \ '

d 5

FIGURE 2.1-1 ,

REACTOR CORE SAFETY LIMIT

,-=r -- ..-- ---..,,,--~.-i,e-----.,r-,-.----.r--- --,.-r -- .--,.vw - + --- --e---m,-=-+-----+---.r-++"--,,--w------ev-

.T_.NSERT 7 9

670-UNACCEPTABLE 660- ,

2440 psia 650-2250 psia w

640-

>==

h 650 2000 psia E

g 628-v 1935 psia k 610 U

u

~

600<

ACCEPTABLE OPERATION 5% -

500 . . . . . . .

9. .1 .2 .5 4 .5 .6 .7 .8 9 1. 1.1 1.2 FRACTION OF RATED THERMAL POWER O

l

m .+ LA, b4 -e e= mm ---es-4e m b- ams.Gn L- m->AA'e-14LM A*A--Age- ,---Am""vaiMA k'L------m-ama ~=+04 *n"W-M PA -mama--A nmkuMMMh JmAA-x,kA,4 A- -- heA kN""4anemtwndmo (THis PAGE AP9t.tcASLE 'tb VNrf 2. c4W

-- ' 680-

~

4 UNACCEPTABLE C OPERATION '

660 --

N N 24oo p*

)

N N -

640 = m w %s, .

2%

2.oo,*

l 620 m -

E -

N m N i 6. = N \

u 8

N G -

O 5

=

5eo 3

~

ACCEPTABLE OPERATION 560 540 o 0.2 0.4 0.4 0.s 1.o 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1

~

REACTOR CORE SAFETY LIMIT 3

V0GTLE UNITS - 1 & 2 2-2 l

.s y -4, , ,- . . , - , , . , ,-- - - - , w -,r

l s

i SAFETY LIM 4TS AND LIMITING SAFFTY SYSTEM SETTINGS O 2. 2 LIMITING SAFETY SYSTEM SETTINGS . .

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

' 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2*

(u wr t) er ,

APPLICABILITY: As shown for each chtnnel in , Table 3.3-1. a k 2.24 a @ Q

ACTION:

a. With a Reactor Trip System Instrumentat on or Interlock Setpoint ,

less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1 adjust the Setooint consistent with the Trip Setpoint value. , g g ,_

b. With the Reactor Trip stem Instrumentation or Interlock Setpeint less conservative the the value shown in the Allowable Values column of -Table 2.2-1, either: f gr Table. 2.2,-Q

1. Adjust the nt consistent w n.n 1.ne irip Setpoint value of

-Table 2.2-1 nd determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or i 2. Declare the channel inoperable and apply the applicabib ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 . Z + R + 5 < TA ,

Where: gg g, Z = Tne value from Column Z of Table ?. 2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel,

. S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 %

for the affected channel, and ,gg9

(. TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.

GB r

O V0GTLE UNITS - 1 & 2 2-3

..,,,.....-_-----..,.....,,......,..._-.,...,._._._m, _ _ _ , . . . _ _ . . , . . _ . . _ - , . _ . _ , , , ,._m. . . -.. . . . . . , . . _ - - , . .--,,,m__--_,,, e.- w--

O .

O HIS PAGE APPLICABtX To udtT I oH TABLE 2.2-lhsT f k ,

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i c- -

TOTAL SENSOR 5 ALLOWANCE ERROR i j FilNf!iONAL UNIT (TA) Z (5) _ TRIP SETPOINT ALLOWABLE VALUE f b 1. Manual Reactor Trip N.A. N.A. N.A. N. A., M.A.

2. Power Rang. Neutron Flux (NI-0041b.C, NI-0042B&C,

! NI-0043B&C, MI-00448&C)

a. High Setpaint 7. 5 4.56 0 <109% af RTP# <111.3% of RTP#

b. Low Setpoint 8.3 4.56 0 525% of RTP# {27.3%ofRTPf

3. Power Range, Neutron Flux, 1. 6 0.50 0 <5% of RTPf with <6.3% of RTPf with High Positive Rate i time constant i time constant (NI-0041B&C, MI-0042B&C, >2 seconds >2 seconds DELET _

NI-0043B&C,MI-00448&C) 7 tJob 4:s cli.ne_ we MM mded W W Eut-mt.1

4. ' er RariR , Neutron' lux, 1. 6 - 0.50 0 < of RTP# wi <6.3% o RTP# with h Negatig Rate i ti- constant i time to tant

( q0041B&C,41-004

->2 sec g >2 seconds

(

, NI-0043B&C, MI- W 4B&C) q

5. Intermediate Range, 17.0 8.41 0 <25% of RTP# <31.1% of RTP#

Neutron Flux (NI-00358 NI-00368)

6. Source Range, Neutron Flux 17.0 10.01 0 <105 cps 114 x 105 cps

, . (NI-00318, MI-00328)

,3 (03 4

7. Overtemperature AT (101-411C,TDI-421C, (o. sirs) [B%j 7.

M 1.

See

+3,n te 1 See Note 2 101-4310 TDI-441C) (et i) omn)

8. Overpower AT (1D1-4118,TDI-421B, h 1.54 [D95j See Note 3 See Note 4 101-4318. 10I-441B) 4.3 IA (o.s TQ (udiT 0

# RIP = RATED IHERHAL POWER e

- w

4 kHIS PAGE AMLicAe4-E 76 unrr I cw i j- TA8LE 2.2-1 (Continued) 4 o

+

REACTOR TRIP SYSTEM INSTRIMMTATION TRIP SETPOINTS(UdtT l g. TOTAL SENSOR

q Alt 0WANCE ERROR 1 m FUNCTIONAL UNIT (TA)' .Z (5) TRIP SETPOINT ALLOWABLE VALUE 1

l g 9. Pressurizer Pressure-Low 3.1 0.71 1.67 >1960 psig** >1950 psig l q (PI-0455A,8&C, PI-0456 &

y PI-0456A, PI-0457 & PI-0457A,

l , PI-0458 & PI-0458A) l i

m 10. Pressurizer Pressure-High 3.1 0.71 1.67 $2385 psig 12395 psig

! (PI-0455A,8&C, PI-0455 f.

! PI-0456A, H -C'57 a PI-0457A, i PI-0458 & PI-0458A) 1

11. Pressurizer Water Level-High - 8.0 2.18 1.67 <92% of instrument s93.9% of instrument I (LI-0459A, LI-0461A, LI-0461) span span l

j 12. Reactor Coolant Flow-Low 2.5 1.87 0.60 >90% of loop >89.4% of loop

! (LOOP 1 LOOP 2 LOOP 3 LOOP 4 Besign flow

Besign flow

FI-0414 FI-0424 FI-043< FI-0444 l FI-04IS FI-0425 FI-0435 FI-0445 j FI-0416 FI-0426 FI-0436 FI-0446) l 13. Steam Generator Water Level 18.5 17.18 1.67 >18.5% (37.8)*** >17.8% (35.9)***

low-tow (21.8)*** (18.21)*** of narrow range of narrow range

{ pp (LOOP 1 LOOP 2 LOOP 3 LOOP 4 instrument spin instrument span l gg LI-0517 LI-0527 LI-0537 LI-0547-gg . LI-USI8 LI-0528 LI-0538 LI-0548 gg LI-0519 LI-0529 LI-0539 LI-0549

, "* LI-0551 LI-0552 LI-0553 LI-0554)

! 5,E 14. Undervoltage - Reactor 6.0 0.58 0 >9600 volts >9481 volts l %

foolant Pumps 170% bus voltage) {69% bus voltage) 1

[gg[ 15. Underfrequency - Reactor Coolaat Pumps 3.3 0.50 0 >57.3 Hz >57.1 Hz

! 33 " Loop design flow = 95,700 gpm my

- Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead and v~

l I second for lag. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

- - The value stated inside the parenthesis is for instrument that has the lower tap at elevation 333"; tie

, value stated outside the parenthesis is for instrumentation that has the lower tap at elevation 438".

O O~ O I

(THis FA66 A9PuCASLE Tb cart I o4i'f]

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUNENTATION TRIr SETPOINT

c. TOTAL SEN'A j'E ALLOWANCE EwAOR d FUNCTIONAL UNIT (TA) Z (5) TRIP SETPOINT ALLOWA8tE VALUE i-

[ 16. Turbine Trip i [ a. Low Fluid Oil Pressure M.A. N.A. M.A. ->580 psig >500 psig

~

l (PT-6161,PT-6162,PT-6163)

b. Turbine Stop Valve Closure N.A. M.A. N.A. 1%.7% open 1%.7% open l 17. Safety Injection Input from ESF N.A. M.A. M.A. M.A. N.A.

i 18. Reactor Trip System

! Interlocks i

9 l J, a. Intermediate Range M.A. M.A. M.A. 11 x 10 88 ag 16 x 10 88 amp Neutron Flux, P-6

(NI-0035B,NI-00368)

~

3

b. Low Power Reactor Trips Block, P-7

1) P-10 input N. 4. M.A. N.A. -<10% of RTPF -<12.3% of RTP#

(NI-0041B&C, NI-0042B&C, NI-0043B&C,MI-00448&C) 2)' P-13 input N.A. N.A. N.A. <10% RTP# Turbine <12.3% RTPf Turbine

(PI-0505,PI-0506) Impulse Pressure Tapulse Pressure Equivalent Equivalent t

c. Power Range Neutron M.A. N.A. N.A. $48% of RTPf 150.3% of RIPf

! Flux, P-8 i (NI-0041B&C,NI-0042B&C, NI-0043B&C,NI-00448&C)

RTP = RATED THERMAL P(TJER 1

e qquer 'W

, m - - ,. .. .

(THts PAGE AveuCABLE To ou;7 g stp_ -

l. TABLE 2.2-1 (Continued) o .

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m 4 e TOTAL. -

SENSOR i 3 ALLOWANCE ERROR i d FUNCTIONAL UNIT (IA) Z (5) TRIP SETPOINT ALLOWA8LE VALUE i s d. Power Range Neutron Flux, P-9 N.A. N.A. M.A: <50% of RTP# <52.3% of RTPS ;

{ o. (NI-00418&C, NI-0042B&C, 1 m MI-0043B&C,NI-00448&C) [

t

e. Power Range Neutron N.A. .M.A. N.A 310% of RTP# 17.7% of RTPS t

j. Flux, P-10

(NI-0041B&C, NI-0042B&C. i NI-0043B&C,MI-00448&C) j i

l f. Turbine Impulse Chamber M.A. M.A. N.A. <10% RTPf Turbine <12.3% RTP# Turbine Pressure, P-13 Impulse Pressure lopulse Pressure j

= (PI-0505,PI-0506) Equivalent Equivalent l t

u

19. Reactor Trip Breakers M.A. M.A. M.A M.A. N.A. ;

t

20. Automatic Trip and Interlock M.A. M.A. M.A. M.A. M.A. !

Logic [

i l I f

e

. I i e !

i i

4 i

RTP = RATED THERMAL POWER l t

7 l

}

r _ - -

s g 8

2

)4 t

]

)

l s 3

4 u (

a 8 s 1 ,

o 0 ,

f 3 g t

= g y g

~

T - , ,

t 3 T

)

1 T, _

r .

w 'P )t T

A T

, r o f o _

o T r a f T P a

i o r r r

o .

E

(

3 v f o o t K ( r f t a L a s B o r s n

+ t o n e A a t e p C

8 L

P P

QT t

]

'T

% a s

n.

e p

m a

s n

e p

p m

o c

m o

c g

A u t o m g a

[k

) O

) n c o a l 5 ;

c ; l -

E d g G

A P i e

u n

Q S

N I t s

T A

d e

l d

a l

g a

R E

W O

P t

Ts d

a e

d l

a e ;

9, t O 1 n r a T e u l e ,

s n I ( i u e A h L e h o (u T l

u

. (T C

(1 T

A T

O N

T

[ ks f

o a

e m

l i

n d

e r

u t

i n

A M

R E

H F

h t

b y

t i

n d

e r

u

- i n d s d T d F s 2 E n o e a e [+* d e ' a z e z D e; e h

L _

z O r 2 B E 2 i

m i t n i ,

m A o l l T 2 ao l e E T D t i n i A ' ri i r n L T a t u

o t R " et t u o B

A s

n u " na u t z e r

o t es a r T

p s s a gn s r o np e e K t t t t m n a n i b n p t

a

- T A

o c t a s n t a a' om io t a m e

s n

s; e s d t c s; t e K d g ns p n e p c ns p

[ e a o m o t nc o e m r l c3 o c a ui c4 g o

, u - c c f m a c f s d e m$ g e i a e5 r a a e

a e a m d n

en m e g i 2 i hy is v a

$ M l T 1 L T I Td Tt A L

)

5 = = = = = = 5 = = = = =

3 1

T f ^

. A 2 3, s 3

1 ' 1 , i ,

E ( 1 R , 1 . , ,

U T i 3 T i 2 4 h T

A A I t 3 1 a K K f 1 T 3 R

E P l M :

E I

g e r -

R e E T h

O V a W 1 ,

E ,

T O

N O <Pjr"'

- c ' .

e i l;j!i- !f 1!llJ' 4[: ,; ; j' } l' i, . Js' .

l m m -

r-- r ~.

hels PAGE' AfYLICABLE To UMs-TT onI.,

TABLE 2.2-1 (Continued)

T

, -e TA8tENOTATIONS(ContinuedQow NOTE 1: (Continued) i h T' <

W.*f*F(untTth

[586M(Nominal T,,,%RAMQ TMLM):

$ b er fhy

. s*

K3 = MpN O I'6/f'b(""'Ti j! p-P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);

5 = Laplace transform variable, s 8; and f (al) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such t t

, C-32.o7. (outT D II07-(u"'T'h 4 (1) For g t ~9 b *n N. 478"U ft (aI) = 0,

n g t % am m A DMM POWER in the top and bottom halves of the core respectively, and g g + qb is total THEllMAL POWER in percent of RATED THERMAL POWER; 32.o?. (ca:Tih (2) For each percent that the magnitude of q g g exceeds RQ the AT Trip 5etpoint shall

.% automatical w r=Aared by _JofitsvalueatRATEDTHE -

3.257,(owir s') +81.07. (ouiT t))

(3) For each percen at i.he magnitude of q qb exceeds b l % the AT Trip setpoint shall

, t be automatically reduced b D 3%lof its value at RATED THERMAL POWER.

1.9'37.

cair O NOTE 2: The channel's maximum Trip se nt shall not exceed its computed Trip Setpoint by more than k (4A5.17. (oo.T i)

(

1 1

~

. \ ^

E

'f.

O

.2

, si 3,2 F -

o l- . .

, I t .

-[

e nn , E 4

~

5 *

. - u e -

t n'

- k k 2 p - + 1 ,

u 6 O E 2

. u o j ._ t g

- 3 1

^i E

Y. i 8

4 t

3; o

/- ~

J g.

=

t 3,

kur v

w

v I T l

4 o 4

+

.g ^2,

,6 ,e R

g g ('

t, x'

3 e

El - +

,- T '

Djm j

u **

3 a

3

.ex, , 1, e

~ k, 1

$t -

~

g 4~

a s

X - ,E t

3 -

=

} 1

-7 8 T vi

Y '-

C y ?

f4 u .:: ce t .:: S t- 3 .::

i "5p- +0 ~

ol 3*

~E :

y E

5 e :

3

'i

TJ J ! E t E

.= - -

[ t v..

b . % o. W in . . b "2* .

2 v 3,x .

j -

- 82 s.! 3 j T )

J e ~ e

- a

j/ $ 3 % y E.

t

%\

.e. g y v. 8 v. 2 "$

-ev

8 i

., a, .

- e .8 %, .ua F. s" e l i t .:

u e

^2 n n n

n

-- -e.--v - -

n a vi as

n n n

~ + mm m m

- ON m '

- O m N

+ + . s , #

C+

q GG; %-~ ; e : %o 22 . -

+

~

+ +

00 E )

, a

, o .2

f 1

.E

\

V0GTLE UNITS - 1 & 2 2-10

L ,

O O y ,, ,_ ,_,,

O TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

E NOTE 3: (Continued) is = Time constant utilized in the measured T lag compensator, is=0s

~ .co2.o/*F (oenTS

[ Ks >

M13K for T > T" and Ks = 0 for T 1 T",

T = Average Temperature, 'F; T" = Indicated I at RATED THERMAL POWER (Calibration temperature for AT instrumenta on, 5 5B8%) Srg.f *F (voiT ({

5 = Laplace transform variable, s-I; and r,o f 2(AI) = 0 for all al.

O NOTE 4: Jhe_ channel'smaximumTrip5etpointshallnotexceeditscomputedTripSetpointbymorethan If4Klof AT span.

1.3 7. (o.s.r li\

L t

@HS PAGE APPucABLE To ucrr 2 on'j TABLE 2.2-1 - OuiT 2].

. b REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS 7,, .

c TOTAL SENSOR 5 ALLOWANCE ERROR

$ FUNCTIONAL UNIT (TA) Z (5) TRIP SETPOINT ALLOWABLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

- . 2. Power Range, Neutron Flux N

(NI-0041B&C, NI-0042B&C, -

NI-00438&C,NI-00448&C)

a. High Setpoint 7.5 4.56 0 <109% af RTP# <111.3% of RTP#

b. Low Setpoint 8.3 4.56 0 {25% of RTP# ~27.3% of RTP#

3. Power Range, Neutron Flux, 1. 6 0.50 0 <5% of RTP# with <6.3% of RTP# with High Positive Rate a time constant i time constant (NI-0041B&C, NI-00429&C, >2 seconds >2 seconds m NI-00438&C,NI-00448&C) gm] N.TE %;s %ms.p4sy op361 by Iche GLv-o t%I h 4. N r Range, D utron Flu' ~

HighNegativhRate

't. 6 OS 0 0 \<5%ofRTPhith a, time consta'nt

< 6. of RTPf1 3 time constant h_'

(NI-DQ41B&C, N1

" _ NI-004 %&C, NI- )

->2' seconds

\ \ ~>2 seconds

~\

5. Intermediate Range, 17.0 8.41 0 <25% of RTP# <31.1% of RTP#

Neutron Flux (NI-00358,NI-00368) -

6. Source Range, Neutron Flux 17.0 10.01 0 -<105 cps -<1.4 x 105 cps (NI-00318,NI-00328)

7. Overtemperature AT 6.6 _ 1.95 See Note 1 See Note 2 (IDI-411C, TDI-421C, QueT2y. (o-n

+ 0.50

, 101-4310, TDI-441C) -

@n 2_T)

8. Overpower AT 1.54 1.95 See Note 3 See Note 4 (101-4118,.TDI-421B, '"

TDI-431B, 10I-4418) ooer 2.)

RIP = RAIE0 lHERNAL POWER

- - - = . - - - - . - _ _ _ _ . _ . ,______________j

I r

O O L - -

O.

l ha S SEGE MPucAB2 To 04tT 2. 04C/

j TABLE 2.2-1 (Continued) y 8 l l p +

m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS(-tunT 2.]

g TOTAL SENSOR

y ALLOWANCE ERROR i

FUNCTIONAL UNIT ;

i .

(TA) -

Z (5) TRIP SETPOINT ALLOWABLE VALUE j g 9. Pressurizer Pressure-Low 3.1 0.71 1.67 >1960 psig** >1950 psig l- 3 (PI-0455A,B&C, PI-0456 & i i y PI-0456A, PI-0457 & PI-0457A, *

, PI-0458 & PI-0458A) m j j 10. Pressurizer Pressure-High 3.1 0.71 1.67 12385 psig 123% psig l

(PI-0455A,B&C, PI-0456 & -

i' PI-0456A, PI-0457 & PI-0457A, .

PI-0458 & PI-0458A) -

11. Pressurizer Water Level-High - 8.0 2.18 1.67 <92% of instrument <93.9% of instrument f (LI-0459A, LI-0460A, LI-0461) span span i 12. Reactor Coolant Flow-Low 2. 5 1.87 0.60 >90% of loop >89.4% of loop .

(LOOP 1 LOOP 2 LOOP 3 LOOP 4 Besign flow

3esign flow * !

I j FI-0414 FI-0424 FI-0434 FI-0444 Q- FI-0415 FI-0425 FI-0435 FI-0445 v

FI-0416 FI-0426 FI-0436 FI-0446) i i

13. Steam Generator Water Level 18.5 17.18 1.67 >18.5% (37.8)*** >17.8% (35.9)***

i Low-tow !

(21.8)*** (18.21)*** Bf narrow range 6f narrow range

, yg (LOOP 1 LOOP 2 LOOP 3 . LOOP 4 instrument span instrument span i

! ** LI-0517 LI-0577 LI-0537 LI-0547 !

yh LI-0518 LI-0528 LI-0538 LI-0548 LI-0519 LI-0529 LI-0539 LI-0549 55 LI-0551 LI-0552 LI-0553 LI-0554) i I

y[ 14. Undervoltage - Reactor 6.0 0.58 0 >9600 volts >9481 volts j Coolant Pumps {70%busvoltage) {69%busvoltage) f j

3

[

gg

15. Underfrequency - Reactor Coolant Pumps.

3.3 0.50 0 >57.3 Hz >57.1 Hz !

I ?? [

j my

Loop design flow = 95,700 gpm I L

vv

- Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead and i 1 second ' for lag. CHAl#0EL CALIBRATION shall ens e that these time constants are adjusted to these values. i l ***The value stated inside the parenthesis is for instrunct that has the lower tap at elevation 333"; the i j value stated outside the parenthesis is for instrumentation that has the lower tap at elevation 438".

4 h

~

O O (THIS FAGE APruCAecE To unrT 2. oud )

O TABLE 2.2-1 (Continued)

REACTORTRIPSYSTEMINSTRtMENTATIONTRIPSETPOINTS(ouiT

~ SENSOR TOTAL.

c- ERROR ALLOWANCE 5 (TA) (5) TRIP SETPOINT ALLOWABLE VALUE d FUNf.TIONAL UNIT Z_

b 16. Turbine Trip

- H.A. N.A. N.A. 1580 psig 3500 psig y a. Low Fluid Oil Pressure (PT-6161, PT-6162 PT-6163)

N.A. N.A. 196.7% open >96.7% open

b. Turbine Stop Valve Closure N.A.

N.A. N.A. N.A. N.A.

17. Safety Injection Input from ESF N.A.

18. Reactor Trip System Interlocks a

N.A. N.A. N.A. 31 x 10 18 amp 16 x 10 88 amp

a. Intermediate Range g3 Heutron Flux, P-6 (NI-0035B,NI-0036B) l

~

b. Low Power Reactor Trips Block, P-7

1) P-10 input N.A. N.A. N.A. 110% of RTP# $12.3% of RTP#

(NI-0041B&C, NI-0042B&C, HI-0043B&C,NI-0044B&C)

N.A. N.A. <10% RTP# Turbine <12.3% RTP# Turbine

- 2) P-13 input N.A. Impulse Pressure (PI-0505,PI-0506) Impulse Pressure Equivalent Equivalent e

i Power Range Neutron N.A. N.A. N.A. 148% of RTP# $50.3% of RTP#

l c.

' Flux, P-8 (NI-0041B&C, NI-0042B&C, NI-0043B&C,NI-0044B&C)

RIP = RATED THERMAL POWER

w

j 'm85 PAGE APPucAtkE To uurr 2 out.

TABLE 2.2-1((Continued)

, o REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS {uurr 2]

c- TOTAL SENSOR 5 ALLOWANCE ERROR i U FUNCTIONAL UNIT (IA) Z (5) TRIP SETPOINT ALLOWA8LE VALUE b d. Power Range Neutren Flux, P-9 N.A. M.A. M.A. $50% of RTP# 152.3% of RTPi o- . (NI-0041B&C, MI-0042B&C, m NI-0043B&C, NI-00448&C)

e. Power Range Neutron N.A. M.A. M.A 310% of RTPf 17.7% of RIPf Flur, P-10 (NI-0041B&C, MI-0042B&C, NI-0043B&C,NI-00448&C)

f. Turbine' Impulse Chambes M.A. M.A. M.A. <10% RTPf Turbine <12.3% RTP# Turbine lopulse Pressure i

Pressure, P-13 Impulse Pressure

.i m (PI-0505,PI-0506) Equivalent Equivalent

19. Reactor Trip Breakers N.A. M.A. M.A N.A. M.A.

j _

20. Automatic Trip and Interlock N.A. M.A. N.A. N.A. N.A.

m logic e

i j- t

RIP = RATED THERNAL POWER

, _ . . . , ~ _ . . . - . ~ . . . s ,_ .. .a - - - .~ . - . . . - - . . -

..--e._...- ..~. .~.. _.-. -.a

. i J

=

^i 43 . > @

1 i

Q * ^'

o .

2 ,, a c' I O I a > F. O j .

A h e W

. u O C 4 > 0 %

i 1 ' t <

9 u u A 2 O 6 u o p

w L u 2 3 %.

v o u , c e

=

+

wa m a

%. 3 -

W c a -

W -

o e v '

$ p : v g 5 3= g i ,

o ^ v m - m .. v .. - T e

< 3 % i

4 1

w

'a*

) * * '** '

- .".a "

s - e t &. -

.. N ?

l e - - - =

. < E a y - y

+-

e3 w 2

w

.e c

w c c 3 f ,

9 nn 2- E e

o t

k , *-

.. t 5 y*

('

N W w W. C @ . @ l W c ,

W g *

2 u .: # .: 9 ,e .: . # %

W W

- e.4 Q -

>.= .

~

'J C Q

J kcc k +5

& 3 C

~ * =

g **J

0 .-

t 8 u a % u 1 g bC bc f bC ed

b. w . b. . C

% v 3. . . G. 2. ,* g.. .-

av 3C. .. ,

8 3 W , a Ce C W w C

.- W . O O *J 0 v C & O

"" v v o b 7 w 3 .. 4-v' R v

% $ %w I V'

.- ~ .o. .-

I

'T

~

E. E ,m I"

.' b. a

. .I-vi a >- - s .- - -

o e- *-

<- a e-W 11 11 Il il it il vi 18 11 11 11 il 11 2

WM + WW W WW A

< pa4

>= >

u H

n n

W W W

W e

g, -. n

! >- M M 4 e4 c=4 W M > d M MM M > H e=4 W 5 ~~ 3 w+

I

+'

i. w--- U

- ~~ u )

ac =

w >-

-.g < f w -

- 3

' {

V0GTLE UNITS - 1 & 2 2 1

l 1

0 -

O QTwis fwGE AroucABLE To o,1 T 2 caa-y]

O

TABLE 2.2-1 (Continued) f TABLE NOTATIONS (Continued)(UntT'2.

j NOTE 1: (Continued) -

g y d y h r dv< )e-T' S 588.5'F '(Nominal T,,gkRATkTHEllMA8sk >;

K3 = 0.00056/psig; (omT 2_)

f [ P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);

5 = Laplace trarssform variable, s 1; andf(al)isafunctionoftheindicateddiffererEebetweentopandbottomdetectorsofthe power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For q t ~9 b ten -33.5% aM + 6.5[ f (al) = 0, htq and qb "" "'" "^

j } POWER in the top and bottom halves of the core respectively, and qt *9 b is total THEllMAL l

! POWER in percent of RATED THERMAL POWER; l (ud'TE (2) For each percent that the magnitude of g t 'b exceeds - 33.5%, the AT Trip Setpoint shall t RATED THERMAL POWER; and beautomaticallyreducedby1.27%gfitsv uort t (3) for each percent that the magnitude of q t g c weeds + 6.5%, the AT Trip Setpoint shall I be automatically reduced by 0.83% of its value at RATED THEllMAL POWC'.

u.nr zh (ua.Tt)

. MOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.5%.

L e

4 4

5 i

s fwc,E A#LCAed To oestT 2.043 TABLE . .-1 (Continued)

TABLE NOTATIONS (Continued m

E N010 3: ,0VERPOWER AT d O (1 + t,5) ( 1 1 15 ) ( 1 ) 1

. (1 + T25) (1 + 13 5') $ AT-, {K4 - Ks ((y ,7 77 5) gy Tsg) T - % [T ((1 + ts5)) - T9 - f2 (AI))

Q*

Where: AT = Measured AT by RTD manifold instrumentation; 1+t y ,'5S = lead-lag compensator on measured AT; ti, r2 = Time constants utilized in lead-lag compensator for AT, ty 3 8 5,12 I 3 5; 1

1+r5

= Lag c mpensator on measured AT; 13 = Time constants utilized in the lag compensator for AT, 13=0s; 2 AT, = hdicated AT at RATED THERMAL POWER; e K4 5 1.089 )

Ks 1 0.02/*F for increasing avarage temperature and > 0 for decreasing average temperature,

=

1 17 5 The function generated by L..e rate-lag compensator for T,,g dynamic compensation, ty =

Time constants utilized in the rate-lag compensator for Tyg, t7310s, a =

1+ 5 Lag compensator on measured Tag; w

~ ,

.m 'J n .x . ,

(THt51%GE Af'Puqad Tb Uort 2.och. .

TABLE 2.2-1 (Cont!nuedl TABLE NOTATIONS (Continued) m E NOTE 3: (Continued). ,

-4 7 is

= Time constant ut-

in the measured T,yg lag compensator, ;

I6 = 0 s; g i

[ -Ks > 0.0013/*F for T > T" and Ks = 0 for T $ T",

. T = Average Temperature, *F;. ,

1" = ' Indicated T avg at RATED THERMAL POWER Calibration temperature for AT !

instrumentation,5;588.5'F), (u,4 T 2) i S = Laplace transform variable, s-1; and j m f 2(AI) = 0 for all AI.

NOTE'4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.4% of AT span. :

3 3 (UN6T 2) i i

1 i

i i

i

IS PAGE hPPL.KASW To uFf l o D '

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR 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 fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operstion above the upper boundary of the nucleate boiling regime could

] result in excessive cladding temperatures because of the onset of departure

~ from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer

-A coefficient. ONB is not a directly measurable parameter during operation and

-: therefore THERMAL POWER and reactor coolant temocrature and oressure have been

-3 O Oridi cerreletien. Tim W-3 G Grid DiiB l d3 related to ONB throughitM keeMi= hn!been developed to predict the ONB flux and the location of DNB 4for ax' ally uniform and nonuniform heat flux distributions. The local DNB Mheat flux ratio (DNBR) is defined as the ratio of the heat flux that would

.3 cause DNB at a particulrr core location to the local heat flux and is indicative l 2 of the margin to DNB.

.v The minin' value of theNNBR during s'taady-state ope'retion, normal h \op ational trans nts, and anttedpated trans ts is limitechto 1.30. This s

value orresponds t 95% pronabi Rty at a 95% fidenceleve\thatDNB y 1 n ur and is t en as an apgopriate marg to ONB for 41 operating 0 IM L The curves of Figure 2.1-1 show reactor core safety limitskhic,ere!

\

rT [ . 1 o n kH bA bE b C013N 5 es-l

-sure and average temperature which satisfy the following criteria: e aN l

y DSS FNng The average enthalpy at the vessel exit is feDabtolthe enthalpy or j A.

saturated liquid (far left line segment in each curve),

t l The minim DNBR is no less than t' design limi all the other line I segments in ach curve).

s l C. Th ( hot channel xit quality not great than the uppar limit of'the l quanty range of he W-3 (R-Gr d) correlati which is 15% middle lirie s segmen es, 2400 psia nd 2250 x psia; kon Reactor Coolant SystemNoressure cutht(is s N not a l'imiting i

1

[Repaw$

O

- ( htz.

V0GTLE UNITS - 1 & 2 8 2-1

/ INSERT 1 The'DNB thermal design criterion is that the probability that DNB will not occur en the most limiting rod 11:s at least 95% (at a 95%

confidence level) for.any Condition I or II event.

In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear.and thermal parameters, fuel fabrication pecameters-and computer codes must be considered. As described-in the FSAR,.the effects of these uncertainties have been statistically combined with the correlation uncertainty.

Design limit DNBR values have been determined that satisfy the DNB design criterion.. I Additional DNBR margin is maintained by performing the safety analyses.to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR. penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

INSERT 2 B.' The minimum DNBR satisfies the DNB design criterion (all the

-other line segments in each curve). The VANTAGE 5 fuel is analyzed using the WRB-2 correlation with design limit DNBR

[)

values of 1.24 and 1.23 for the typical call and thimble

(_/ calls, respectively. The LOPAR fuel is analyzed using the WRB-1 correlation with design limit DNBR values of 1.23 and 1.22 for the typical and thimble cells, respectively,

c. The~ hot channel exit quality is not greater than the. upper limit of.the quality range (including the effect of uncertainties) of the DNB correlations. This is not a

limiting criterion for this plant.

is PAGE, APPLicA6[E To vulT 2 out.

2.1 SAFETY LIMITS m.

U BASES 2.1.1 REACTOR CORE (- upti p 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 fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime 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 measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the W-3 (R Grid) correlation. The W-3 (R Grid) DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative vi tne margin to DNB.

3 The minimum value of the DNBR during steady-state operation, normal (V operational transients, and anticipated transients is limited to 1.30.

value corresponds to a 95% probability at a 95% confidence level that DNB This will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

d The curves of Figure 2.1-1#show reactor core safety limits which are determined for a range of reactor operating conditions. The core limits represent the loci of points of THERMAL POWER, REACTOR COOLANT SYSTEM pres-sure and average temperature which satisfy the following criteria:

A. The average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid (far left line segment in each curve).

B. The minimum DNBR is not less than the design limit (all the other line segments in each curve).

C. The hot channel exit quality is not greater than the upper limit of the quality range of the W-3 (R-Grid) correlation which is 15% (middle line i segment on Reactor Coolant System pressure curves, 2400 psia and 2250 psia; this is not a limiting criterion for this plant).

em i

O V0GTLE UNITS - 1 & 2 B 2-1

SAFETY LIMITS BASES RTP REACTOR CORE (Continued) bH These curves are based on an enthalpy hot channel factor, k D 4 !and a reference cosine with a peak of 1,55 for axial power shape. "X?n llowance is luded for an increase in F N g at reduced power based on the expression:

F F1-Mi(1+ RF3il-P))

m P is the fraction of 'L^T!? 'MER"."L PA'3 These limiting heat flux conditions are higher than those calculated for the range of all contro' reds fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t(61) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the in'.e>rity of the Reactor

()

V Coolant System (RCS) from overpressurization and thereby ; events the '

tase of radionuclides contained in the reactor coolant from reaching the cot , inment atmosphere.

The reactor vessel, prassurizer, 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 entire RCS is hydrotested at 125% (3107 psig) of design pressure, to demonstrate integrity prior to initial. operation.

g g,

6 N A con cesm w<,rs eswccac' P% isk Paw % der Muthen u kr F5a spwf d in k coLR, and '

U .

V0GTLE UNITS - 1 & 2 8 2-2 -

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN ,

1 LIMITING CONDITION FOR OPERATION l 3.1. 2. 5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. 'A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of 9504 gallons (19%

of instrument span) (LI-102A, LI-104A),

2) A boron concentration between 7000 ppm and 7700 ppm, and

3) A minimum solution temperature of 65'F (TI-0103).

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 99404 gallons (9% of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A),

2) A boron concentration between 2400 ppm and 2600 ppm, and

3) A minimum solution temperature of (TI-10982).

(O/ APPLICABILITY: MODES 5 and 6.

F(vutTder 5t'F (o ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

, 1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and

3) When the boric acid storage tank is the source of borated water and the ambient temperature of the boric acid storage tank room

( (TISL-20902, TISL-20903) is <72 F, verify the boric acid storage tank solution temperature is > 65 F.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature (TI-10982) when it is the urce of borated water and the outside air temperature is less than % .y p o *F (vair I) or So'F (varr h). ,

V V0GTLE UNITS - 1 & 2 3/4 1-11 .

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPfRABLE as required by Specification 3,1,2.2:

a. A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of 36674 gallons (81%

of instrument span) (LI-102A, LI-104A),

2) A boron concentration between 7000 ppm and 7700 ppm, and

3) A minimum solution temperature of 65 F (TI-0103),

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 631478 gallons (86%

of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A),

2) Aboronconcentrah.ionbetween2400ppmand2600 ppm,

3) A minimum solution temperature nf b

@'F(umT f) or 5+'F(umT 2.I)

4) A maximum solution temperature of 116"F (TI-10982), and

5) RWST Sludge Mixing Pump Isolation Valves capable of closing on RWST low-level.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With the Boric Acid Storage Tank inoperable and being used as one of the above required borated water sources, restore the tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN as required by Figure 3.1-2 at 200'F; restore the Boric Acid Storage Tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the RWST inoperable, except for the Sludge Mixing Pump Isolation Valves, restore the tank to OPERABLE status within I hour I or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD

- SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

VCGTLE UNITS - 1 & 2 ~3 /4 1- 12 .

REACTIVITY CONTROL SYSTEMS (3

i,,) LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued)

c. With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the i

valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolate the sludge mixing system by either closing the manual isolation valves or deenergizing the OPERABLE solenoid pilot valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months cnd maintain closed. '

SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration in the water,

2) Verifying the contained borated water volume of the water source, and

3) When the boric acid storage tank is the source of borated water and the ambient temperature of the boric acid storage tank room

(] (TISL-20902, TISL-20903) is < 72*F, verify the boric acid storage v tank solution temperature is > 65'F.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST tempera (TI-10982)whentheoutsideairtemperatureislessthan@

c. At least once per 18 months by verifying that the Sludge Mixing Pump Isolation Valves automatically close upon RWST low-level test signal, F(unN I) or 50F (uul' 9

O e

k V0GTLE UNITS - 1 & 2 3/4 1-13

REACTIVITY CONTROL SYSTEMS ___

( ROD DROP TIME (votThor 2.2 (vw17 2.

QtITINGCONDITIONFOROPERATION 3.1.3.4 The individual shutdown and control red drop t me rom the physical l fully withdrawn position shall be less than or equal to ' seconds frota beginning of decay of stationary gripper coil voltage to cashpot entry with:

a. T,y (TI-0412, TI-0422, TI-0432, TI-0442) greater than or equal to 551 F, and s

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the drop time of any 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 REOUIREMENTS O

Q F 4.1.3.4 The rod drop time 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 which could affect the drop time of thosi specific rods, and

c. At least once per 18 months.

f

(

1 1

1 l

h(.s V0GTLE UNITS - 1 & i? 3/4 1*'9 Amendment No29 (Unit 1)

Amendment No.10 (Unit 2)

@ephch f. 2. I for Ud / N/

4 A nut me. ,

3/4.2 POWER DISTRIBUTION LIMIT 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION m

.. The indicated (NI-00418, NI-00428, NI-00438, NI-00448) AXIAL FLUX FFERENCE (AFD) shall be maintained within the target band (flux difference i u ts) about the target flux difference. The target band is specified in the CC3 OPERATING LIMITS REPORT (COLR).

The in icated AFD may deviate outside the required target band at greater han or equa to 50% but less than 90% of RATED THERMAL POWER provided the i icated AFD is w hin the Acceptable Operation Limits specified in the COLR an the l

, cumulative penalty deviation time does not exceed I hour during the 'evious 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated FD may deviate outside the required target band greater than 15% but less th 50% of RATED THERMAL POWER provided the cumvfative penalty l deviation time d s not exceed I hour during the previous 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

APPLICABILITY: MOD 1, above 15% of RATED THERMAL POWE #

ACTION:

a. With the indicaged AFD outside of the r utred target band and with l THERMAL POWER graater than or equal t 90% of RATED THERMAL POWER, within 15 minutes ither:

1. Restore th indicated D to within the target band limits, o!

2. Reduce THERMA POW to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD ou- de of the required target band for more l than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulati- ce Ity deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside t . Acceptable Operation Limits specified in the i COLR and with THERMA) POWER les than 90% but equal to or greater I than 50% of RATED ERMAL POWER, educe:

1. THERMAL P0 R to less than 50% f RATED THERMAL POWER within 30 minut , and i 2. The P er Range Neutron Flux * - Hig etpoints to less than or equ- to 55% of RATED THERMAL POWER w1 5in the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Specia Test Exceptions Soecification 3.10.2.

Surveillptce testing of the Power Range Neutron Flux Chann ay be performec (belowf90% of RATED THERMAL POWER) pursuant to Specification 42 .1.1 providec the jitdicated AFD is maintained within t: 1e Acceptable Operation L .its speci-f r' ear in the COLR. A total,of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulate with the A outside of the above required target band daring testing without gnalty eviation. .

N

( '

)

O V0GTLE UNITS - 1 & 2 3/4 2-1 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

THIS PAGE APPLICABLE TO UNIT 1 ONLY 3/a.2 POWER 01STRIBUT10N LIMITS A 3/4.2.1 A 1 AL FlVX DIFFERENCE - UNIT 1 b dmT 1 On1 3 llMITING CONDITION FOR OPERATION 3.2.1 The indicated (N1-0041B, N1-0042B, N1-00438, NI-00448) AxlAL FLUX OlFFERENCE (AFO) shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

APPL!CABILITY: MODE 1 ABOYE 50 PERCENT RATED THERMAL. POWER".

ACTION:

a. With the indicated AKIAL FLUX OlFFERENCE outside of the limits specified in the COLR,

1. Either restore the indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux * - High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD it within the limits specified in the COLR.

SURVEILLANCE RE001REMENTS O>

4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2) At least once per hour until the AFD Monitor Alarm is updated af ter restoration to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging,

c. The provisions of Specification 4.0.4 are not applicable.

4.2.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE escore channels are indicating the AFD to be outside its limits. -

5ee Special Test Exceptions Specification 3.10.2.

v

, V0GTLE UNITS - 1 & 2 3/42-1

fHis PAGE APfuCA6LE To vehT 2 cAG 3/4.2 POWER DICTRIBUTION LIMITS 3/4.2.1 AXIALFLUXDIFFERENCE(Ot4LTh LIMITING CONDITION FOR OPERATION 3.2.1 The indicated (NI-00418, NI-0042B, NI-00438, NI-00448) AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference. The target band is specified in the CORE OPERATING LIMITS REPORT (COLR).

The indicated AFD may deviate outside the required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is within the Acceptable Operation Limits specified in the COLR and the l cumulative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

' The indicated AFD may deviate outside the required target band at greater than l 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER * #

ACTION:

a. With the indicated AFD outside of the required target band and with l THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

p 1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD outside of the required target band for more l than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of c'nulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outsiue the Acceptable Operation Limits specified in the COLR and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:

1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and

2. The Power Range Neutron Flux * - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

See Special Test Exceptions Specification 3.10.2.

, Surveillance testing of the Power Range Neutron Flux Channel may be performed (below 90% of RATED THERMAL POWER) pursuant to Specification 4.3.1.1 provided-the indicated AFD is maintained within the Acceptable Operation Limits speci-fred in the COLR. A total.of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the above required target band during testing without penalty O viation. .

O V0GTLE UNITS - 1 & 2 3/4 2-1 P A mendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

+15 9A GE APPuCA6' " TO UdFT" 2. M 'f) g POWER DISTRIBUTION LIMITS

! i

\v/

LIMITING CONDITION FOR OPERATIONh- VMIT 2.)

ACTION (Continued)

c. With the indicated AFD outside of the required target band for more ,

than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THER M POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the required target band and the cumulative penalty devia-tion has been reduced to less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , glRVEILLANCE RE0VIREMENTS

4. 2. L 1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFO for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is'0PERABLE, and

2) At least once per hour until the AFD Monitor Alarm is updated after restoration to OPERABLE status.

/] b. Monitoring and logging the indicated AFD for each OPERABLE excore V channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging, 4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the required target band shall be l accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER 1evels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of eacn OPERABLE excore channel shall be )

determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable, 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the i most recently measured value and 0% at the end of the cycle life. The provi-OV sions of Specification 4.0.4 are not cpplicable.

V0GTLE UNITS - 1 & 2 3/4 2-2 Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

.)

i I

\

f O

/

i FIGURE 3.2-1 (DELETED)

~

f .

G b V0GTLE UNITS - 1 & 2 kM Amendment No. 32 J

(Unit 1)

Amendment No. 12 (Unit 2)

POWER DISTRIBUTION LIMITS O. i Q 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION

)

3.2.2 F q(Z) shall be limited by the following relationships:

F9 (Z) 1 F 0 [K(Z)] for P > 0.5 P

Fq (Z) 1 F 0 , [K(Z)] for P $ 0.5

0. 5 .

Where: Fq RTP= the qF limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR),

Where:

P = THERMAL POWERRATED THERMAL POWER' and K(Z) = the no,rmalized F specified in the COLR. 9(Z) as a function of core height l

APPLICABILITY: MODE 1.

ACTION:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL ~ POWER at least' 3 for each 1% qF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within-the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints ave been reduced at least 1% q for each 1% OF (Z) exceeds th limitt and b.

vo.lue.ofK4p-Identify and correct the cause of the out.-or timit condition

b. h C SPM prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be 9

within its limit.

l O A

~

V0GTLE UNITS - 1 & 2 3/4 N Amendment No. 32 (Unit 1)

Amendment No. 12 (Unit 2)

~

o i

s i

I i

! N,

\

\ -

D 1 l

\

i i

\

, , FIGURE 3.2-2 (DELETED) ,

4- -

t V0GTLE UNITS - 1 &'2 - 53 Amendment No. 32 (Unit 1)

. Amendmedt No. 12 (Unit 2)

epic.c.o bxve&nce. Segvice.md:. K POW 2R DISTRIBUTION LIMITS N 4.2.2.1 4. 2.2.1 a 4. 2.M g wM nw P"q Gurva'd\aue. Pgred Sec h s 4. 2, t . ) , 4. 2 2. 2. W 4. 2.2. 3 SURVEILLANCE REQUIREMENTS UthT I O*J L'1[fkges .2 4,2-5,2-lo 2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2 2 F xy shall be evaluated to determine if F (Z) is within its limit y:

q ah Using the movable incore detectors to obtain a power distri tion NmapatanyTHERMALPOWERgreaterthan5%ofRATEDTHERMAL WER before exceeding 75% of RATED THERMAL POWER following ea fuel 1bading. '

b.

N Increasing the measured F xy component of the power stribution map by 3% ko account for manufacturing tolerances and further increasing the valu' by 5% to account for measurement unce ainties,

c. Comparing t = F xy computed x(Ff)obtainedin pecification 4.2.2.2b. ,

above to:

1) The F li%tsforRATEDTHERMALPCWER(FRTP)fortheappropride xy x measured cora planes given in Sp i fication 4.2.2.2e, and f. ,

below, and

2) The relationshi :

F =F xRTP ppy pg l

Where F*Y is the li t forYfractional THERMAL POWER operation RTP expressed as a funct of Fxy , PF xy is the power factor multiplier for F pecified in the COLR, and P is the fraction of RATED THERMAL OWER at ch F xy was measured.

d. Remeasuring F ace dingtothefd{owingschedule:

xy

1) When F is reater than the F mit for the appropriate measured te plane but less tha7 the F relationship, additional power d stribution maps shall be take dF C compared to F x and either:

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% o RATED THERMAL )

C POWER or greater, the THERMA'. POWER at whic%F was last !

. determined, or b) At least once per 31 Effective Full Power Days-(EFR ), ~

whichever occurs first. g p

J

./ -

J

~

V0GTLE UNITS - 1 & 2 3/4 2-6 Amendment No. 32 (Unit 1) -

Amendment No. 12 (Unit 2) 1

, _ _ _ . _ . _ + , _

$tPl actmuE P age. Er Fq SveN 4e OM 1 ny y THIS PAGE APPLICABLE TO UNIT 1 ONLY l h

V POWER DISTRIBUTION LIMITS SURVEllLANCE RE0VIREMENTS - UNIT 1 l 4.2.2.1 The provisions of Specifications 4.0.4 are not applicable.

4.2.2.2 FQ (Z) shall be evaluated to determine if it is within its limit by: l

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

C

b. Determining the computed heat flux hot channel factor, FO (7), ,3 follows:

Increase the measured QF (Z) obtained from the powe- distribution map by 3% to account for manuf acturing tolerances and further

increase the value oy 5% to account for measuremer.t uncertainties,

c. Verifying that FQ (Z), obtained ir. Specification 4,2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the fellowing relationship:

C RTP Fg (Z) FO x K(2) for P > 0.5 P x W(Z)

RTP Fg (7) FO x K(Z) for P < 0.5 0.5 x W(Z)

Where F 0(Z) is obtained in Specification 4.2.2.2b above, F RTP Q is the g limit, K(Z) is the nonnalized F 0(Z) as a function of core height, P is the f raction of RATED THERMAL POWER, and W(I) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

FQ , K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT a., per Specification 6.8.1.6.

e. Measuring F Q (Z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which F0(I) was last determined *, or

. 2. At least once per 31 Effective Full Power Days, whichever occurs first.

"During power escalation af ter each f uel loading, power level may be ;

increased until equilibrium conditions at any power level greater than l

, or equal to 50% of RATED THERMAL POWER have been achieved and a power l distribution map obtained.

(

V0 GILE UNITS - 1 & 2 3/4 2-4

m _ _ - - _ . _ . _ _ _ . . _ _ . _ . _ . . . _ _ _ _ _ _ _ _ .. . _ . . _ -.

htP *'R&g3l

% sun;lknQ 74 . THIS PAGE APPLICABLE TO UNIT 1 ONLY l POWER DISTRIBUTION LIMITS i SURVEILLANCE REOUIREMENTS (Continued) - UNIT 1 l

f. With measurements indicating i maximum I C FQ (7) l over Z ( K(Z) j C

has increased since the previous determination of FQ (Z) either.

of the following actions shall be taken:

C

1) -Increase FQ (Z) by 2% and-verify that this value satisfies the relationship in Specification 4.2.2.2d, or

2) Fg C(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that ,

-maximum IFO(2)) is not increasing. _,

l l

-over Z K(Z)j J

g. With the relationships specified in Specification 4.2.2.2d above not being satisfied:' ; 1 l

1) Calculate the percent FQ (Z) exceeds its limits by the I l following expression: '

r T l / -

c l ! 1 (maximum F0 (2) x W(Z) j ,

x M0 for P > 0.5

.[ max imum C

Fo(Z)xW(2)'\ -I over Z FO RTP

,g(7) $x100forP<0.5,and

- 0.5- -

2) Thofollowingaction'shallbetaker[.
Within 15 minutes, control the AFD to within new AFD limits
. which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by 1% AFD for each percent Fn(Z) exceeds its limits as determined in Specification 4.2.2.29 1. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm selpoints
to-these modified limits. !
-p V
V0GTLE UNITS - 1 & 2 3/4 2-5
l pacemd pay. for Fq Smeillang-4,. vu a i Niy THIS PAGE APPL
CABLE TO UNIT 1 ONLY l POWER DISTRIBUTION LIMITS SURVEft. LANCE RE0VIREMENTS (Contir.ded) - UNIT 1 l
h. The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e. 0 - 100%, inclusive.
1. The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:
1) Lower core region f rom 0 to 15%, inclusive.
2) Upper core region f rom 85 to 100%, inclusive.
4.2.2.3 When Fg(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(Z) shall be obtained from a power distribution map and increased by 3%
to account for manufacturing tolerances and further increased by 5%
to account for measurement uncertainty.
O 1
)
I O V0GTLE UNITS - 1 & 2 3/4 2-6 ~-
I
_ _ - - -~ ----
P GE A99LicA8LE, To oraiT 1 om y j POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS b dlT D _
n_
4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 F xy shall be evaluated to determine if qF (Z) is within its limit by:
a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER before exceeding 75% of RATED THERMAL POWER following each fuel loading.
b. Increasing the measured F xy component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, C
c. Comparing the F.Y ^ computed (F*Y) obtained in Specification 4.2.2.2b.,
above to:
1) The F limits for RATED THERMAL POWER (FRTP) for the appropriate xy x measured core planes given in Specification 4.2.2.2e and f.,
below, and
2) The relationship:
F =F RTP x
gy pp 1,p)
Where F*Y' is the limit for fractional THERMAL POWER operation RTP expressed as a function of Fxy , PFxy is the power factor multiplier for F xy specified in the COLR, and P is the fraction of RATED THERMAL POWER at which Fxy 'i/as measured,
d. Remeasuring F xy according to the following schedule:
1) When F x
is greater than the F RTP limit for the appropriate measured core plane but less than the F ' relationship, additional power distribution maps shall be taken a d F compared to F P and F eitHr:
a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL )
C POWER er greater, the THERMAL POWER at which F*Y was last
- determined, or b) At least once per 31 Effective Full Power Days (EFPD), ~
whichever occurs first. , g o g (Unit 1)
V0GTLE UNITS - 1 & 2 3/4 K Amendment No.
Amendment No.
32 12 (Unit 2)
- gts PAGE APPUC.ASLE To 04tT 1 ou'Q POWER DISTRIBUTION LIMITS
-D- SURVEILLANCE REQUIREMENTS (Continued (-- O4IT 2).
2) When the F,C is less than or equal to the F P limit for the appropriate measured core plane, additional power distribution C
4-maps shall be taken andyF , compared to F,RTP andF,hatleast once per 31 EFPD.
e. The F limits used in the Constant Axial Offset Control analysis forRhEDTHERMALPOWER(F,RTP)shallbespecifiedforallcoreplanes containing Bank "0" control rods and all unrodded core planes in the COLR per Specification 6.8.1.6; f.
The F,y limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
1) Lower core region from 0 to 15%, inclusive,
2) Upper core region from 85 to 100%, inclusive, -
3) Grid plane regions at 17.8 2 2%, 32.1 2%, 46.4 2 2%, 60.6 1 2%,
and 74.9 1 2%, inclusive, and p 4) Core plane regions within t 2% of core height [1 2.88 inches) about the bank demand position ef *"o Bark "0" control rods,
g. - With F x exceedingF,ftheeffectsofF o F9 (Z) shall be evaluated to determine if F q (Z) is within its limits.
4.2.2.3 When Fq (Z) is measured for other than Fxy determinations, an overall measured F (Z) shall be obtained from a power distribution map and increased 9 _
by 3% to account for manufacturing tolerances and further increased by 5% to account for-measurement. uncertainty, i
1:
i L .
1 O O C/ V0GTLE UNITS - 1 & 2 3/4 2-7 ' Amendment No. 32 (Unit 1)
Amendment No. 12 (Unit 2) 4 -- ,
(s\
POWER DI5iRIBUTION LIMITS
" P3 W # +f3 "'I U 3/4.2.5 DNB PARAMETERS LIMITINO CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits:
a. R (TI-0412 TI-0422. TI-0432, TI-0442),
5D or TCoolant System (vo T,I_T E) w SWF (vun 2.
592.5*R
b. Pressurizer Pressure (PI-U495A,5&C, Pi-(T4 6__& PI-0456A, PI-0457 &
PI-0457A, PI-0458 & PI-0458A), ((2M ( pst @
c. Reactor Coolant System Flow (FI-0414, FI-0415, FI-0416, FI-0414, FI-0425, F1-0426, FI-0434, FI-0435, FI-0436, FI-0444, FI-0445, FI-0446) 1p (,198 p m K e APPLICABILITY: MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
O V SURVEILLANCE REQUIREM_ENTS l
l 4.2.5.1 Reactor Coolant System T,yg and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . RCS flow rate shall be monitored for degradation at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In the event of flow degradation, RCS flow rate shall be
, determined by precision heat balance within 7 days of detection of i flow degradation, l
l 4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL l CALIBRATION at each fuel loading and at least once per 18 months.
I l 4.2.5.3 After each fuel loading, the RCS flow rate shall be determined by l precision heat balance prior to operation above 75% RA1ED THERMAL POWER. The RCS flow rate shall also be determined by precision heat balance at least once per 18 months. Within 7 days prior _to_per.
l forming the precision heat balance floFineasurement, the instrument-l I ation used for performing the precision heat balance shall be I calibrated. The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
" Limit not applicat,le during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
~
" Includes a f* flow measurement us.certain_ty. _ _
2.1% (od T 0- 3.5% (voiT 2.) )
V0GTLE UNITS - 1 & 2 3/4 e n
O -
O O TABLE 3.3-3(CopMmed)
E$
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS g ,
TOTAL SENSOR E ERROR y ALLOWANCE ALLOWABLE VALUE
FUNCTIONAL UNIT (TA) Z (S) TRIP SETPOINT 9
~ 7. Semi-Autom tic Switchover to Containment Emergency Sump (Continued)
b. RWST Level--Low-Low 3.5 0.71 1.67 >275.3 in. from >264.9 in. from l Coincident With Safety Iank base tank base Injection (>39.1% of (>37.4% of (LI-0990A&B, LI-0991A&B, iEstrument iEstrument LI-0992A,LI-0993A) span) span)
8. Loss of Power to 4.16 kV ESF Bus t
N. A. N.A. N.A. >2975 volts >2912 volts
a. 4.16 kV ESF Bus
';' Undervoltage-Loss of Voltage Uith a < 0.8 Gith a < 0.8 second time second time g delay. delay.
b. 4.16 kV ESF Bus N.A. H.A. N.A. >3746 volts >3683 volts Undervoltage-Degraded Uith a <20 sith a <20 Voltage second time second time L- 2000 delay. __ delay.
9. Engineered Safety Features (odt [-)
Actuation System Interlocks N.A. N.A. 0 sig < 1980 psia
a. Pressurizer Pressure, P-11 N.A. <
(PI-0455A,8&C, PI-0456 & - (omT z KumT 2)]
PI-0456A, PI-0457 & PI-0457A, PI-0458 & PI-0458A)
b. Reactor Trip. P-4 H.A. H.A. N.A. N.A. N.A.
1
. 3/4.5 EMERGENCY CORE COOLING SYSTEMS M1: A Maind boded d vak d b s..,6ss5 3/4.5.1 ACCUMULATORS , M 09 # an> o.7 7.@ of .'as shM
s. useenX er aL1-0950,
~~a Aa O ( LI-09 56, LI-09 5 1-om, L1-o% ,
L1-o%3 1.1 0954, t.1 o%,
3 LIMITING COND_ ON FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
a. solation valve open,
@ pwr 2. O A coritained borated water volume of between 6616 (36% of instrument span) and 6854 gallons (64% of instrument span) (LI-0950, LI 0951, LI-0952, LI-0953, LI-0951, LI-0955, LI-0956, LI-0957),
c. A boron concentration of between 1900 ppm and 2600 ppm, and
d. A nitrogen cover pressure of between 617 and 678 psig. (PI-0960A&B.
PI-0961A&B, PI-0962A&B, PI-0963A&B, PI-0964A&B, PI-0965A&B, PI-0966A&B, PI-0967A&B).
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next
{'
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,
b. With one accumulator inoperable due to the isolation valve being closed, either fmmediately open the isolation valve or be in a' least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressve to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1) Verifying t'he contained borated water volume and nitrogen cover pressure in the tanks, and'
2) Verifying that each accumulator isolation valve is open (HV-8808A, B, C, 0).
Pressurizer pressure above 1000 psig. -
t V0GTLE UNIT. - 1 & 2 3/4 6-1 1
M 6
BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION-FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
)
a. A minimum contained bo sted water volume of 631,478 gallons (86% of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A).
b. A boron concentration of between 2400 ppm and 2600 ppm of boren,
c. A minimum solution temperature of, n 4*F(vmT her 54*F (ve D
d. A maximum solutico temperature of 116 F (TI-109B2)
e. RWST Sludge Mixing Pump Isolation valves capable of closing on RWST low-level.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a. With the RWST inoperable except for the Sludge Mixing Pump Isolation Valves, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
L O
NO b. With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolate the sludge mixing system by either closing the manual isolation valves or deenergizing the OPERABLE solencid pilot valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and maintain closed.
SURVEILLANCE REQUIREMENTS
~
4.5.4 The 'RWST shall be demonstrated OPERABLE:
a. At least once per 7 days by:
1)~ Verify ng the contained borated water volume in the tank, and
2) Verify ng the boron concentration of the water.
b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying e ST temperature when
'the outside air temperature is'less than * ^
. c. At least once per 18 months by verifying that the sludge mixing pump isolation valves automatically close upon'an RWST low-level test )
signal.
F (va rt h or So'F (udiT 6
e9 d
+*
sPAG4 A#PucA6LE To um7 l o 3/4.2 POWERDISTRIBUTIONLIMITShu9ffh eetshg h D96 du;y u%
BASES The specifications of this section provide assurance of fuel integrity
' during Condition I (Normal Operation) and II (Incidents of Moderate-Frecuenev) ,/
events by: (1)[m'autainino thNnimum DNBlMD the corabater than bhecual _ \
Entsejouring normai operation and in snort-term transients, and (2) lim 1 ting
~
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 criteria limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
Fg (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevatier, 7, divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Ri.se Hot Channel Factor, is defined as the ratio of the integral of linear power alon the red with the highest integrated power to the average rod power .
(Z), itadial Peaking Factor, is detined as ths\ rat 4 of pea \pcwer deri ty !
~
to ' average power daqsity in t% horizontal'Alene at coreglevati t
3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper n
bound envelope of the F limit q specified in the CORE OPERATING LIMITS REPORT j (COLR) times K(Z) ib not exceeded during either normal operation or in the i event of xenon redistribution following power changes.
- Target flu \ difference is eterminedathu111briumxen(ncondition\
' rods may be pqsitioned vf 5i the core in Accordance with\their respective in tion limits ahd should u stat peration at .Kghpowerlevel ins rtedThe near the'tp(normal value d the target posiden for steady-flux difference f obtai under these Ronditions divi .d by the fraction of RATED THERMAL POWER \ ,
is the t'ir Ncore burnup conditions.
\get flux diffqrence Target at RATEDfor flux differences \ THERMAL POWER other TtiERMAL POWER \for the associhted leNels are i obtained by' multiplying the RATED THERMAL \ POWER value by the appropriate fractional THE MAL POWER level.
dif ference val (ue is necessar'ys to reflect core. burnup considerations.The periodic upd
'\ \ N \ . A
\
l l
V0GTLE UNITS - 1 & 2 B 3/4 2-1 Anendment No. 32 (Unit 1)
Amendment _No. 12 (Unit 2)
i his PA66 MNcA%E ~4 @T ion POWERO!STRIBUTIONLIMITSb O eA$a AXIAL FLUX O!FFERENCE (Continued)
Although it is ihtended that thelplant will be %perated uithke AFD !
within th.e target band Nequired by 'pecification 3.2.1 about the target flux I differencet during rapid' plant THERMAL POWER reductions, contrni rod thation i will cause the AFD to devtete .9.tside of'the target bands at reduced THERMAL j YOWER levels.NThis deviation will not affect the xenon redistribution suf -
clotttly to change the envelobe of peaking fh tors t which may be reached on a subsa>quent return \to RATED THER$AL POWER (witA the AFD witniq the target band g provided the time dbration of the deviation is Nimited. Accordingly, a 1-hour '
s
\ penalty eration' deviation
'outside oflim'it the,cumulstiv'e target bandduring the previous but within the limits 24 hours is specified in provided
he for s
COLRwhileatsTHERMALPOWERlevelsbetween50%and90%ofR For 7 'ERMAL POWER levels be' tween 15% a'd n 50% of RATEDsTHERMAL POWER, deviations of the FD'outsidkof the ta'eget band aresless sdgnificant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months aptual timesreflects this reduced Sjgnificance. '
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and Sot provides an alarm message immed(atelv if the,AFD for two or more OPERABLE ex ors Q annels are outside tP.e(tit d t h qdD nd the THERMAL POWER is creater_ s thar(cQLipf RATED THERMAL POWER. 'Ing operation 'at THERMAL \ POWER levels 4 TeTwTeiN0% andN(0% and tetween 16% a 50% RATED THERMAL POWER, the computer O -
ou'tputs a}h41 arm mqssage khen the penaltAdeviaRon merDmulates beyond the \
limil k of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months anh2 hours \ respect'ivelyq
^
gg & gg,. gg l
x se m s e % c. s
@gureWL4 PkthewsNypicQonthlyTaget b&nQpmk gu kvehm wp 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE H0TCHANNELFACTOR-Fh ('O The limits on heat flux hot channel factor auf nuclear enthalpy rise hot channel factor ersure that: (1) the d n limithon peak local power density thniillust DNRjats not exceeded and .in the event of a LOCA the peak fuel 11ad temperature will not e_xcemb e he 2200'F ECCS acceptance criteria limit, b eh of these is measurable but will normally only be determined i pe*iodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
a. Control rods in a single group move together with no individual red I insertion differing by mo than i 12 steps, indicated, from the group demand position;
b. Control rod banks are sequenced with a co_nstant tip-to-tip distance between banks as M e3 V FT h e 3,w 4.p .
, (2.Sk 04 deyn I eH en is meh "b b 9N ^M' O N "o V0GTLE UNITS - 1 & 2 B 3/4 h Amendment No. 32 (Unit 1)
Amendment No.12 (Unit 2) ,
4 e----.--,,v y..m...-.
N D GO fl QNLT n.
I l
I 0.90 N \ / '
0.80 x I
/ <
f l -
g 0,70
\ !
t
/ -
E \ /
.a \ l 0, E ENCE 0.60 ,
p N Y i
O 0.50
/ \
s
u. l' e 2 0.40 x
e /i
$* 0.30
/ \ \ I
[ t
'\
/. I
\
0.20 I
/ 1
\
0.10 ,
/ \ ,
(
0'
30
/ -20 -10 0 l
+10 +20 +30 INDICATED AXtAL FLUX DIFFERENCE (percent)
FIGURE B 3/4 2-1 j r TUICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER
(
V0GTLE UNITS - 1 & 2 B 3/4 M
@is PAGE APPLICABLE fM VNIT 1 ouLh 1
p0VER0!STRIBUTIONLIM!TCvmT] '
BASES
' HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (continued)
c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
d. The axial power distribution, expressed in terms of AXIAL FLUX OlFFERENCE, is maintained within the limits.
Ffg will be maintained within its limits provided Conditions a. through
d. above are maintained. The relaxation of F N as a function of THERMAL POWER l allows :hanges in the radial power shape for all permissible rod insertion limits.
J When an F g easurement m is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance. E 3 WhenFfg is measured, (i.e , inferred), measurement uncertainty (i.e., .
the appiopriate uncertainty on the incore inferred hot rod peaking factor) must be allowed for and 4%-is th6 appropriate allowance for a full core map taken with the incore detection system.
Fuel rod bowin reducesthevaluebDNBratio. C\r i t is availabl to (fset this reducti n the generic marg . .The
. 9.1% ONBR completely o fset any rod bow pen ties. generic This ma'harnins, totalin includes the follbwing:
a esign limit DNBR ofQ.30 vs 1.28, l \ b. Gr pacing (K,) of OD046 vs 0.059,
c. Therma f fusion Coef fibent of_0.038 vs 0. 9, '
d. DNBRMultipierof0.86vsht8,and i
.{ f l e. Pitch reductio l
The applica .e values of rod w penalties are eferenced in the SER. i t
V0GTLE UNITS - 1 &*2 B 3/4 % ,
.c- --,,-..,.v. -a v.en+,w. , , .,,,.,.,,,,-.a.,,,a,-_na,.,,,,n,.,,m,,,.. -,-,,,wa- a a .w,.,e , _ , - . u._ , ,-.,,+,,e n ,,,-,-emav
INSERT The heat flux het channel factor Fg(Z) is sessured periodically and increased by a cycle and height dependent power factor appropriate to KAOC operation.
W (I) . to provide assurance that the limit on the heat flux not channel factor.
W(Z) accounts for the effects of normal operation transients Fo(I) is set.
within the AFD band and was determined from expected power control manuevers over the full range of burnup conditions in the core. The W(Z) function for normal operation and the AFD band are provided in the CORE OPERATIhG 1,IMITS REPORT per Specification 6.8.1.6.
O I
O.
6 P A GE $49u cA8tg my p r i ON L-POWER DISTRIBUTION LIMIT grT v BASES
~
HEAT FLUX H0T CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 4
(Continuto)
The R%al Peaking hector, F,y(Z) is measuredhriodically to provide 1
Z), remains w hin its limit. he a urancs tha the Hot Chargel Factor, FRTPq
. F imit for RMED THERMAL PQWER (F )a pecified in te COLR per Spec i- l I ca ion .8.1.6wabetermined me ected po r control ma rs over the !
full ra ge of burnup\ conditions it the core.
3/4.2.4 OVADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during $TARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowan:e for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greate'r 4 O than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fqis reinstated by reducing the maximum allowed power by 3% fr.,r each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incere detectors are used to confirm that the normali:ed symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, C-11, H-3, H-13, L-5, L-11, N-6.
3/4.2.5 DNB PARAMETERS f 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 ara consistent with the I
initial FSAR assumetions and have been analytically demonstrated adequate to
'TmTtntain '1NtinimumMNBR of B0 )throughout each analy:eo transient. The indicated T value ofi?SEJand the indicated pressuri:er pressure value of M sig c r aspond to nalytical limits of $3172SEEland[2M5(psig respec-tively, with allowance or measurement uncertainty. -
g zu.s r cawe .
V 1-3 V0GTLE UNITS 1&2 B 3/4 % Amendment No. 32 (Unit 1)
Amendment No.12 (Unit 2)
, , . - - - . . , - - ~ . , ~ ,
5 pact, Apit.icABt.E "TC) VMIT'MLC PO\'ER DISTRIBUTION LIMITS O BASES 3/4.2.5 DNB PARAMETERS (Continued)
The 12-hour periodic surveillance of these parameters through 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 the flow rate degradation on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis. A change in indicated percent flow which is greater than the instrument channel inaccuracies and parallax errors is an appropriate indication of RCS flow degradation.
O I
D V0GTLE UNITS - 1 & 2 B 3/4
7_______.__._____________
("THis PAGE APrucABLE To untT 2. com]
3/4.2 POWER O!STR!dVT10N LIMIT $ Q m T g 4
BASES l
The specifications of this section provide assurance of fuel integrity
, during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting j- 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 criteria limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
Fg (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; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z). Radial Peaking Factnr, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.
3/4.2.1 AXIAL FLUX O!FFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F g
(Z) upper bound envelope of the F limit q specified in the CORE OPERATING LIMITS REPORT (COLR) times K(Z) is not exceeded during either_ normal operation or in the event of xenon redistribution-following power changes.
Target flux difference is determined at equilibrium' xenon conditions.
The rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for'staady-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 >
l is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux dif ferences for other THC.'ti'.',L POWER levels are
'( obtained by multiplying the RATED THERMAL = POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
1 i
! 2.-5 V0GTLE UNITS - 1 & 2 8 3/4 Amendment Ao. 32 (Unit 1)
Amenament No. 12 (Unit 2)
his PAGE MPLr AME To vciT P. ewL.D POWER Of STRIBUTION LIM!T5Qum T 2].
BASES AXfAL FLUX O!FFERENCE (Continued)
Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux I difference, during rapid plant THERMAL POWER reductions, control rod motion 4
t will cause the AFD to deviate outside of the target band at reduced THERMAL POWER 1evels. This deviation will not affect the xenon redistribution suffi-
, factors which may be reached on a ciently toreturn subsequent change theTHERMAL to RATED envelope.of POWERpeaking (with the AFD within the target band provided the time duration of the deviation is limited. Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits specified in the COLR while at THERMAL POWEE levels between 50% and 90% of RATED THERMAL POWER.
For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD'outside of'the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL' POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the O
limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.
Figure B 3/4 2-1 shows a typical monthly target band.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOTCHANNELFACTOR-Fh The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.
Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
a. Control rods in a single group move together with no individual rod e insertion differing by more than i 12 steps, indicated, from the group demand position;
'b. Control rod banks are secuenced with a constant tip-to tip distance between banks asI f m b . - . ... - g g g.g -
p 3. t.L Q
.2 - 4 V0GTLE UNITS - 1 & 2 B 3/4 Amendment No. 32 (Unit 1)
Amendment No. 12 (Unit 2) n.- . . , . . - . . . _ _ . _ _ _ _ _ _ . . _ _ . _ _ _ - . _ _ - - . _ _ . _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - -
r QHis PACE A0PucAOLE "To vasi 2 out.7)
(
1.00 I
l ,
I I
I 0.90 I !
l i l
i 0.80 i l
i i '
I 0,70 ,
g l
! ! l g
TARGET FLUX p , f-- DIFF ERENCE 0.80 3
5
/
5 Y i f O 0.50 I
E I I
o -
0.40 I g
i i c I !
E
"* 0.30 l l l
1 l
1 0.20 1 ;
I l
0,10 -
I L l
.L o' I
- 30 -20 -10 0 +10 +20 +30 .
, INDICATED AXlAL FLUX DIFFERENCE (percent)
FIGURE B 3/4 2-1 TYAltAL lilDICATED AXIAL FLUX DIFFERENCE VER$35 THERMAL POWE mt ums . n 2 e 2ngn 4
bl5 PAGE APPLICA6m *mgg i ^
POWER DISTRIBUTION LIMIT
VMl"T?].
BASE $
HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 4
(Continued)
I
c. The control rod insertion lielts of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
i. d. The axial power distributios, expressed in terms of AXIAL FLUX i
DIFFERENCE, is maintained within the limits.
IhwillbemaintainedwithinitslimitsprovidedConditionsa.through
. d. above are maintained. The relaxation of F g as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
l When an Fg measurement is taken, an allowance for both experimental error
and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.
WhenFfg is measured, (i.e'., inferred), measurement uncertainty (i.e.,
the appropriate uncertainty on the incore inferred hot rod peaking factor) must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system.
Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. .The
. 9.1% DNBR completely offset any. rod bow penalties. generic margins, totalingThis marg following:
a.. Design limit DNBR cf 1.30 vs 1.28,
b. Grid Spacing (K,) of 0.046 vs 0.059,
c. Thermal Dif fusion Coef ficient of 0.038 vs 0.059,
p. ONBR Multiplier of 0.86 vs 0.88, and e, Pitch reduction.
, TheapplicaolevaluesofrodbowpenaltiesarereferencedintheFSER.
V0GTLE UNIT $ - 1 & 2 B 3/4 e -e
+"-gr---er -
g i hl5 9 AGE AP91.lC'0610 UNIT 104 POWER O!STRIBUTIO!i LIMITS 7v nTV J
. BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY R!$E HOT CHANNEL FACTOR
~
(Continued)
The Radial Peaking Factor, Fxy(Z), is measured periodically to provide ,
assurance that the Hot Channel Factor, F (Z), remains within its limit. The RTPg F,y limit for RATED THERMAL POWER (F ) as specified in the COLR per Specifi- l cation 6.8.1.6 was determined from expected power control manuevers over the
( full range of burnup conditi us in the core.
3/4.2.4 0VADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.
Radial }ower distribution measurements are made during STARTUP testing and periodica?ly during power operation.
The limit of 1.02, at which cor . action is required, provides ONB and linear heat generation rate r on with x y plane power tilts. A limit _of 1.02 was selected to pr, an allowance for the uncertainty associated with the indicated po' silt.
The 2-hour time allowance fe' operation with a tilt condition greater
, O than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F isq reinstated by reducing the maximum allowed power by 3% fer each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the mouable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT
' POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two rats of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are
.C-8. E-5, E-11, H-3, H-13, L-5, L 11, N-8.
\
3/4.2.5 DNB PARAMETERS The limits on the ONB-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 I initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The indicated T avg value of 591'F and the indicated pressurizer pressure value of 2224 psig correspond to analytical limits of 592.S'F and 2205 psig respec-tively, with allowance for measurer.ent uncertainty. -
i 2 't V0GTLE UNITS 1&2 B 3/4 2-*
Amendment No. 32 (Unit 1)
Amendment No.12 (Unit 2)
"n+15 PAGE APPLicA6LE Tb umi 2. odLY]
POWERDISTRIBUTIONtIMITS(--vqty BASES 3/4.2.5 DNB PARAMETERS (Continued)
The 12-hour periodic surveillance of these parameters through 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 the flow rate degradation on 4 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis. A change in indicated p'ercent flow which is greater than the instrument channel inaccuracies and parallax errors is an appropriato indication of RCS flow degradation.
r"%
4 l
l t
1 .
(
l
() V0GTLF UNITS - 1 & 2 8 3/4 2-2-10 .
. - . . . .- . - _ ._ _ ~ . - -__ -- -
3/4.4 REACTOR COOLANT SYSTEM N U
B_ASES dtQri Crkreon
(
/
3/4.4.1 REACTOR COOLANT LOOPS AND C00LANT JfRCULAT10N _
The plant is desianed to ocerata_ h all reactor coolant loops in operation and % M in M -vu 1.30iduring all normal operations and antici-
'. pated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however a single reactor coolant loop provides sufficient heat removalcapacityIfabankwithdrawalaccidentcanbeprevented,i.e.,by opening the Reactor Trip System breakers.
In HODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR train provides sufficient heat removal capability for removing decay heat; but single failure cons'derations require that at least two trains / loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but single failu*e considerations, and the unavailability of the steam generators as a O heat removing component, require that at least two RHR trains be OPERABLE, The d locking closed of the required valves in Mode 5 (with the loops not filled) precludes the possibility of mcontrolled boron dilution of the filled portion of the Reactor Coolant System. This action prevents flow to the RCS of unborated water by closing flowpaths from sources of unborated water. These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.
The operation of one reactor coolant pump (RCP) or one RHR pump provides acequate 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 an RCP with one or more RCS cold legs less than or equal to 350'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant 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 Aependix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the-RCS cold leg temprature s. h 1
eTc 1 "fMs ebe. m utAusly rausbl W (der GW-piqo; 2n MODE
4 h, 54a,6q ef an RCP den no eb RcP 4 epm 3 , d ]
-.n,a % .n # .m .,. mpremte.
% ,:hl .
anQdo md 4. Jem.arirati. M A MA. desip Is ut ucec QH A rel;d yaho art Wd f>r Ac5 onrpeswo profeeb.
V0 GILE UNITS - 1 & 2 B 3/4 4-1
j i
(~ EMERGENCY CORE COOLING SYSTEMS
. ( i BASES 4
f ECCS SUBSYSTEMS (Continued)
I The limitation for all safety injection pumps to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. ,
l The Surveillance Requirements provided to ensure OPERABILITY of each ;
component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Reauirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.
Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance I configuration,-(2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS LOCA analyses, (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that as.umeds in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow which is directed through the seal injection path is less than or equal to the amount assumed in the safety analysis. The surveillance requirements for leakage testing of.ECCS check valves ensure a
,O failure of one valve will not cause an intersystem LOCA. In MODE 3, with either HV-8809A or B clossd for ECCS check valve. leak testing, adequate ECCS flow for core cooling in the event of a LOCA is assured.
3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that sufficient negative reactivity 15 injected into the core to counteract any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss of-coolant accident, or a steam line rupture.
The limits on RWST minimum volume and boron concentration ensure that
1) suf ficient water is available within conta.inment to permit recirculation cooling flow to the core, 2) the reactor will remain suberitical in the cold condition following a small LOCA or steamline break, assuming complete mixing of the RWST, RCS, and ECCS water volumes with all control rods inserted except the most reactive control assembly (ARI-1), and 3) the reactor _will remain whe *" W in the cold condition following a large break LOCAlOtr uk (hnt 1 M.0M assuming complete mixing of the PMST, RCS. ECCS water ano other -
sources of water that may eventually reside in tne suup post-LOCA with all control rods assumed to be outg '
h The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
(von 2) w d w4rol reds insot.d utept & A.h usi reame. coaty.) uwe s (veer i) .
V0GTLE UNITS - 1 & 2 B 3/4 5-2 <
ADMINISTRATIVE CONTROLS SEMIANNUAL RADICACTIVE EFFLUENT RELEASE REPORT (Continued)
The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gAstous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.
MONTHLY OPERATING REPORTS
6. 8.1. 5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT N 6.8.1.6 Core coerating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:
a. SHUTOOWN MARGIN LIMIT FOR MODES 1 and 2 for Specification 3/4.1.1.1,
b. SHUT 00WN MARGIN LIMITS FOR MODES 3, 4 and 5 for Specification 3/4.1.1.2,
c. Moderator temperature coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,
d. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,
e. Control Rod Insertion Limits for Specification 3/4.1.3.6,
. f. Axial Flux Dif ference LimitsMd Car 4et itaqdl for Specification 3/4.2.1, g
g. H , Flux Hot Channel Factor, K(Z) ,Ntg Powa(FacttnQultiph(rj and x
for Specification 3/4.2.2,
h. Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.
The analytical methods used to determine the core operating limits shall be those previously approved by the NRC in:
O V0GTLE UNITS - 1 & 2 6-21 Amendment No. 32 (Unit 1)
Amencment do. 12 (Unit 2)
1 l
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued V46
a. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", l July 1985 (W Proprietary). 1 (Methodology for Specifications 3.1.1.3 - Moderator Temperature !
Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit. 2.t 1.s - )
__ Control B&n Insertion Limits J 3N.1 kAxial' Mux 01fhrence N.2.N IMeabflux ht Chhpal FactorNand 3.2.3 - Nuclear Enthalpy Risle -
~ Hot Channel Factor.)
b. WCAP 355,"POWFR0 TRIBUTION CON) OL ANO LOAD
- TOP AL REPORT", 5 gtember 1974 ( Prep ,etary).
LLOWINGPROMDURES]
(Method logy for Speci ication 3.2.I Axial Flux fference (Constan Axial Offset ntrol).)
c. T. M. Ander .n to K. Kniel (Chief of Cor Performance anch, NRC) '
enuary 31 --Attachmen ' Operation d Safety Ana ysis Aspects o an Improved ad Follow P kage.
(Me odology fo pecification 3.2.1 - Axia Flux Differen (Cons nt Axial Of set Control).
NUREG 08 , Stan'ard' d view Plan, S. Nuclear agulatory Commission, Section 4. , Nuclear De gn, July 198 h Branch echnical Po tion CPB 3-1, Westing ouse Constant Axial Offset trol (CAOC) Rev. 2, ly 1981.
O (
(M (Con odology fo Specifica ion 3 2.1 - ial Flux Dif rence nt Axial fset Cent 1). )
% WCAP-9220-P-A, Rev. 1. " WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION", February 1982 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g. , fuel thermal-mechanical limits,- core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMIls REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and i Resident Inspector. [
N PECIAL REPOR 1 Administrator of 6 Special repo shall be su ted to the Regi e
)
Regi ffice of the C within the period speci for each report.
1
@ SERT b O
V0GTLE UNITS - 1 & 2 6-21a Amendment No. 32 (Unit 1) l Amendment No.12 (Unit 2)
- _ ~-. _ _ _.._ _ _ _ _ _ _ _ __
. - - . - - . - ~ _ .
d i
i 1
1
, INSERT)
- b. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET '
i- CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June d
1983 (H Proprietary).
s (Methodology for specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot channel Factor (W(Z) surveillance requirements for Fq Methodology).)
4 e
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4 4
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.-;.-,. .--._,.._..,-.2.--._.-._..,_..__.._...__..-_,...._......_._...._.__._.__...-____m._. . _ . _ . . . . . . , . . . . . - . - . - , , ~ ~ . -
ADMINISTRATIVE CONTROLS g
O SEMIANNUAL RA0bCTIVE EFFLUENT \ REL.ASE REPORT (Continae \ I '
q x x TheSemiannualNadioactiveEffluchgC '1ase Reports sh also includ t following: an ex anation as to whv ' inoperability o iquid or gase eff ent monitoring ins umentation we orrected ,?lthin t time specifie in Sp ification 3.3.3.9 3.3.3.10, n pec vely; and descript n of the events ading to liquid ho up tanks or gas s rage tanks exceed g the limits of cification 3.11. 4 or 3.11.2.6, re ectively.
MONTHLY OPERAt% G REPORTS 8.1.5 Routine r orts of ope, rating tatistics and sh down experience, in luding documentat of all challeng to the PORVs or afety valves, sha be submitted on a onthly basis to e Director, Offi of Resource Manage nt U.S. Nuclear gulatory Commiss n, Washington, O. 20555, with
! copy to e Regional Admin trator of the Reg 4nal Office of th NRC, no late than the h of each month lowing the calenMar mont5 covered y the report.
4 CORE OPERATING LIMITS REPORTb UWT E w -
6. 8.1. 6 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part
of a reload cycle for the following:
SHUTDOWN MARGIN LIMIT FOR MODES I and 2 for Specification 3/4.1.1.1,-
6 a.
b. SHUTDOWN MARGIN LIMITS FOR MODES 3, 4 and 5 for Specification 3/4.1.1.2
c. Moderator temperature coefficient BOL ano EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 4
d. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,
e. Control Rod Insertion Limits for Specification 3/4.1.3.6,
. f. Axial Flux Difference Limits, and target band for Specification 3/4.2.1,
g. Heat Flux Hot Channel Factor, K(Z), the Power Factor Multiplier and F for Specification 3/4.2.2,
h. Nuclear Enthalpy Rise Hot Channel Factor .imit and the Power Factor-Multiplier for Specification 3/4.2.3.
The analytical methods used to determine the core operating limits shall be those previously e nroved by the NRC in:
L 2
V0GTLE UNITS - 1 & 2 Amendment No. 32 (Unit 1)
Amendment No. 12 (Unit 2) e
.-.m------,,e---+-e, ,rwm,=,,%w- es-wc awww-.------,-r-- - - , . , - ---ev_m-m
_ ...-.._-.w-- - - - - .--__-.e- -_.- _ t
ADMINISTRATIVE CONTROLS l
( g}
1 v
j CORE OPERATING LIMITS REPORT (Continued) Q utarT {
a. WCAP-9272-P A, " WESTINGHOUSE RELOAC SAFETY EVALUATION METHODOLOGY",
July 1985 (W Proprietary). ,
(Methodology for Specifications 3.1.1.3 - Moderator Temperature !
Coefficient 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 - ) ,
Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 '
Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise !
Hot Channel Factor.) )
b. WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES
- TOPICAL REPORT", Septe.aber 1974 (W Proprietary). ,
(Methodology for Specification 3.2.1 Axial Flux Difference (Constant Axial Offset Control].)
c. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC)
January 31, 1980--
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Of fset Control].)
d. NUREG-0800, Standard Review Plan, U. S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2. July 1981.
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control].)
e. WCAP-9220-P-A, Rev.1, " WESTINGHOUSE ECCS EVALUATION M00EL-1981 VERSION", February 1982 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g. , fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPCRT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. l SPECIAL REPORTS i
6.8.2 Special reports shall be submitted to the Regional Administrator of the #
Regional Of fice of the NRC within the time period specified for each report.
i
(,- 21 b
i. )J V0GTLE UNITS - 1 & 2 IP1!Q . Amendment No. 32 (Unit 1) l Amendment No.12 (Unit 2) 1 j
1
.k Atig.r,binent 1b Vogtle Flectric Osnerating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE-5 Fuel Design Technical Soecifications Tyned Paget Effective following the Vogtle 1 Cycle 3 Shutdown (Effective as of Vogtle 1 Cycle 4 Startup)
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(O) k' 5(CTION 5,ffd
}/4.2 DDW[R_ Pl$f#lBUTION LIMITS 3/4.2.1 u !AL Flux O!FFERENCE (Ukit )). . . 3/4 l-1 3/4.2.1 AXIAL FLU 1 DIFF(RINC[ (UNIT 2) . . . .. 3/4 2-la )
3/4.2.2 H[lil FLUA N01 CHANNEL FACTOR Eg(Z). 3/4 /.3 !
3/4,2,3 NUCLEARINTHALPYRI$tHOTCHANNil.FAC10R-F$H. . .. . 3/4 2-0 3/4.2.4 OVADRANT POWER T!Li RAi!0. . .. ..... . . .. .. 3/4 2-10 3/4.2.! LNB PARAMETER 5... ......... ... . . . . . . 3/4 2-13 3/4.3 lqiRUWINTAil0N 3/4.3.1 ar,4c;04 TRIP $YSTLM INURUMLNIAi!ON.. . .. . . .. . . 3/4 3 1 TABL[ 3.3 1 REACf0R fRIP $YSTEM IN51 RUM [NIA110N. . . 3/4 3 2 TABLE 4.3 1 R(ACTOR TRIP $YSTlM IN51 RUM [NTA110N $URVillLANCE REQUIRLMENTS.... . ..... ....... .. ..... .... , .. . ... 3/4 3-9 3/4.3 ? ENGIN((R(0 $AFITY F(ATURi$ ACTUAi!ON $YST[M INSTRUMENTAil0N..... .... ,. . ..... . ...... . . 3/4 3-15 1ABLC 3.3-2 [NGINElRED SAFLTY FEATURI$ ACTUATION SY$ FEM IN$1RUMENIAfl0N. ... .. ................ . . .. .. .... 3/4 3 li i TABLt 3.3-3 [NGINt(R(D $AFETY FtATUR[$ ACTUATION $YSTEM V IN$1 RUM [NT Ai!ON TRIP MIPOINi$. . . . . .., . ... .. .. 3/4 3-28 TA6LE 4,3-2 [NGIN([RIO $AF[TY F(efURI$ ACTUA1!0N SY$1tM
[N57RUMINIAT10N $URVi!LLANCE RLOUIR(MENTS. . .. . 3/4 3-36 3/4.3.3 MONITORING INSTRUM[NTAi!0N Radiation Monitoring For Plant Operations. ... .... 3/4 3-45 TABLE 3.3-4 RADIATION MON!10 RING INSTRUMENTAi!DN FOR PLANT OPERAi!ON$ ...... . .......... .. .., ,. ... 3/4 3 46 TABLE 4.J-3 RA0!Atl0N MON!TORING IN$iRUM[NTATION FOR plani OPIRATIONS $URV(!LLANCC REQUIREMENT $ .... ... .. ,. . 3/4 3-48 Movable incore Detectors....... ....... .. .... .. . .. 3/4 3-49 Seismit Instrumentation (COMMON SYST[M) . ..... ........ 3/4 3-$0
/O.
V0 GILE UN!15 - 1 & 2 V l
l
_.. _ _ . - . _ . , _ . -- ~- _ _ _ ._. - -
1 i
Mil l !!J5t $
O
. I
! (
1
' [A,9.[
[Ci10N
?/4,0 APPL'CA91Lliv. F 3/4 O J 3/4.1 8tAttlYtty CONikOL SY5t[W5 3/4.1.1 60RAll0N CONfROL. . . . B 3/4 1-i 3/4.1.2 BORA110N SY$1[MS. . B 3/4 1*?
i 3/4.1.3 MOVABLE CON 1ROL A15LMBLits.. . . . . . B 3/4 1 3 F 3/42 POWER 0!$1R1BU110N llMIT$. . , , . . . B 3/4 2-1 1
3/4.2.1 AA1 AL F LUA O!FFERtNCI (UN!I !). . B 3/4 2*l 3/4.2.2 and 3/4.2.3 wtA1 (LUX HOT CHANNil FACTOR and NUCLIAR (N1PfEl *t$t NOT (HANNil F AC10R - flH (UNil 1). . . B 3/4 2 2 3/4.2.4 QUADRANT POWER TILi RA110 (UNIT 1). . . . . . .... B 3/4 2-3 1
! 3/4.2.5 DNB PARAM[ttR$ (UNIT 1)...., . . . . .. , ,, B 3/4 2-3 3/4.2.1 AXI AL FLUX OlFFtRtNC[ (UNil 2). ... . . . ,, . .., , . B 3/4 2-$
3/4.2.2 and 3/4.2.3 Ht A1 FLUX HOT CHANNEL F ACT,0R and NUCl( AR [
[NTHALPY Rl5[ HOT CHANNil FACTOR - F$H (UN!12). .... B 3/4 2-6 j 3/4.2.4 OUADRANT POWER flLT RA110 (UN11 2). . . . .. B 3/4 2-9
! 3/4.2,5 DNB PARAM[IlR$ (UN]f 2). , ... , , B 3/4 2-9 3/4.3 INSTRUMENTAtl0N 1
3/4.3.1 and 3/4.3.2 REACTOR 1 RIP SYSTEM and INGINEERED $AF(1Y l FIATURES ACTUA110N SYSTEM IN$1 RUM [NTAT10N... .. ... . B 3/4 3 1 3/4.3.3 MONITORING INSTRUMENTA110N.. . ... ... ... .. ...... ,, B 3/4 3 3 3/4.3.4 TURBINE OVER$ Pit 0 PR0tlCTION.. ... . ... . . . ... .. B 3/4 3-b l
l V0G1LE UNITS - 1 & 2 xV
,...,--.-,,#w-..*-----r-...
i l
INDI A ADMINI$fRAflyt CONTROL $ i f-^ !
( _ _ . -
I
\s.
M.L.19N fAGt 6.4.2 $ Aft 1T R[vitw BOARD ISFB)
Function.. . . . . . . 6-9 Composttton. . . . .. . . . . . . 6-10 Altertates.. . . . . ... 6-10 Consultants... .. . .. . .. . . . . . 6-10 Meeting F recuenc y. . .. . .. . . .. , . ... 6-10 Ouorum.... . ..... . ...... .. .. . ....... 6-10 Revtew.... .. . . ... ... . .. . , . 6-11 Audits... ........ . .. .. .. . .. . . .. . . . b-11 Recores.... .. . .. . ..... 6 12 6.5 REPORT ABtt (YtNT ACTION. . ........ ..... . ......... .. 6-13 6.6 $AftTY LIMIT VIOLATION... ................ ... . ............. 6-13 6.7 PR0tt0VRi$ AND PROGRAMS. ... ............ .... . ..... ...... 6-13 6.8 REPORilNGRIOUIRIMM 6.0.1 ROU11Nt REPORTS..... ..... .. ... 6-17
. gg ..... ... .. .. .. .....
e
\, Startup Report........ . .......... ...... .. . ........ 6 Annual Report........ ............ ............. .. ........ 6-17 Annual Radiological Environmental Surveillance Report., . 6-1B Semiannual Radipactive [ffluent Release Report.... .... ... . 6-19 Monthly Operating Repor15.. .... ....... .... . . ....... . 6-21 F
Core Operating Limits Report (UNIT 1)...... ...... ...... ... 6-21 Core Operating Limits Report (UNIT 2)... .. ....... ......... 6-21a 6.8.2 $PICIAL REPORi$ .. .. . ... ... ...... .. . . .. .. . 6-21b i 6.9 RECORD RiftNT!0N........... ...... ....... . .. ... ........... 6-22 i
e
\
v0 GILE UNtf 5 - 1 & 2 xxill
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETi!NGS* _
2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER (NI-0041 N!-0042, NI-0043, N1-0044),
pressurizer pressure (PI-0455A, B&C, PI-0456 L PI-0456A, P!-0457 L P!-0457A, PI-0458 & P!-0458A), and the nighest operating loop coolant temperature (T ..)
(TI-0412, T1-0422. 11-0432. T1-2442) shall not esteea the limits shown in Figure 2,1-1 (Unit 1) or Figure 2.1-la (Unit 2). l APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressuricer pres 5Ure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the reautre-ments of Specification 6.6.1 REACTOR COOLAL'T SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure (PI-0408, PI-0418, PI-0428, PI-0438)
O shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES I and 2:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6,6,1.
MODES 3, 4 and 5:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6,6,1,
Where specific instrument nutbers are provided in parentheses they are for information only, and apply to each unit unless specifically noted (to assist O in identifying associated instrument channels or loops) and are not intenced to limit the requirements to the sDecific instruments associated with the numDer.
V0GTLE UNITS - 1 & 2 2-1
TH!b PAGE APPLICABLE TO UNIT 1 Of1LY I 670 l
UNACCEP"ABLE OPERATlfsN
% 2440 psia 650
%- s, 2250 psia 640 N ,
N m 3
2 5 m A N -
7 % 2000 psia T l o
N A N s, \3 620 g 3 e N \ N 8
< 610 N m A '
$ 1935 rxia 600 m ACCEPTABLE OPERATION 590 1 580
0. .1 .2 .3 .4 .5 ,6 .7 .8 .9 1.0 1.1 1.2 l FRACTION OF RATED THERMAL POWER O FIGURE 2.1-1 REACTOR CORE SAFETY Ll?ilT - UNIT 1 l V0GTLE UNITb 1&2 22
i THIS PAGE APPLICABLE TO UNIT 2 ONLY l 680
- UNACCEPTABLE OPERATION 600 2400 psia
~
640
"-- 2250 psia
~
2000 psia
[
620
- 1775 psia O - - - - - -
o 600 T
8 C 580 3 ACCEPTABLE OPERATION 560 540 1.0 1.2 0 0.2 0,4 0.6 0.8 FR ACTION OF R ATED THERMAL POWER i
l l
l O FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - UNIT 2 V0CTLE UNITS 1 '. 2 2-2a
. ~__- - _ - - - - - - . - - - _ _ . - - - - . - . - - . - . . - -
O l.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O
2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 4
2.2.1 The Reactor Trio System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1 (Unit 1) or Table 2.2-la (Unit 2).
APPLICABILfiY: As shown for each channel in Table 3.3-1.
ACTION: ,
a. With a Reactor Trio System instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the 9hiue shown in the Allowable Value
]- column of Table 2.2-1 or Table 1.2-la, adjust the Setpoint consistent l with the Trip Setpoint value.
b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1 or Table 2.2-la, either: l i
1. Adjust the Setpoint consistent with the Trio Setpoint value of Table 2.2-1 or Table 2.2-la and determine witran 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that l Ecuation 2.2-1 was satisfied for the affected enannel, or Declare the channel inoperable and apply the applicable ACTION 2.
statement 7equirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted Consistent with the Trip Setpoint value.
Equation 2.2-1 Z+R+5 i TA Where:
Z The value from Column Z of Table 2.2-1 or Table 2.2-la for the l affected channel, R The "as measured" value (in percent span) of rack e ror for the affected channel, l'
l S . Elther the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 [
or Table 2.2-la for the affected channel, ano i TA The value from Column TA (Total Allowance) of Table 2.2-1 or Table 2.2-la for the affected channel. l l
O I
VCGTLE UNITS - 1 & 2 2-3
O V THIS PAGE APPLICActt 10 UNil 1 ONLY V
ES TABLE 2.2 UNii 1
?
{} REh" TOR TRIP SYSTEM I'NSTRUMENTATION TRIP SE1POINIS z
El 101 Al. SENSOR ALLOWANCE ERROR FUNCTIONAL UNIT (TA) Z (S) , TRIP SEIPOINI ALLOWABl E VAllit
1. Manual Reactor Trip N.A. N.A. N.A. N.A. H.A.
}
2. Power Range, Neutron Flux (N1-00418&C, NI-0042B&C, N!-0043B&C, NI-0044B&C)
a. liigh Setpoint 7.5 4.56 0 <109% of RIP # $111.3% of RIP #
b. Low Setpoint 8.3 4.56 0 25% of RlP# $27.3% of RIP #
3. Power Range, Neutron Flux, 1.6 0.50 0 $$% of RIP # with 56.3% of RIP # with High Positive Rate a time constant a time constant n, (NI-0041H&C,-NI-0042B&C, 22 seconds >2 seconds j, NI-0043BLC, N1-00448&C)
4. Deleted.
5. Intermediate Range, 17.0 B.41 0 $25% of RIP # $31.1% of RIP #
Neutron Flux (NI-0(358, NI-00368) l
6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.4 x 105 cps (NI-00318, NI-00328)
7. Overtemperature al 10.7 7.04 1.96 See Note 1 See Note 2 (101-411C IDI-421C, (Unit 1) (Unit i) + 1.17 101 -4310, 101 -441 C )' (Unit 1) I
8. Overpowar al 4.3 11 . 5 4 1.96 See Note 3 See Note (
( 1 D 1 -411 B , 10 I-4 23 8, (Unit 1) (Unit 1) 101-4318, T 01-441B) l l#1P = RAlED THERIiXL POWER 6
m 11115 PAGE ) CABLE 10UNII 1 ONLY {
i TABLE 2.2-1 (Continuedi l
RE AC1;'R IRIP SYSTEM INSTRUMtNTATION TRIP SElPOIN15 - UNI 1 1 101AL SENSOR m
ALLOWANCE ERROR e
[S)___ IRIP SETPOIN( Alt 0WA8tt VAIUE ftJNCIIONAL UNIT (TA) 1 1.61 21960 psig** >1950 psig Pressurizer Pressure-tow 3.1 0.71 ra 9.
(PI-0455A,8&C, PI-0456 &
PI 0456A, PI-0451 & PI-0457A.
PI 0458 & PI-0458A) 3.1 0.71 1.67 $2385 asig $2395 psig
10. Pressurizer Pressure-tilgh (PI-0455A.8&C. PI-0456 &
PI 0456A, PI-0451 & PI-0451A.
PI 0458 & PI 0458A)
Pressurizer Water Level -fligh 8.0 2.18 1.61 $92% of instrument $93.9% of instrument II. span span (I I 0459A. I I 0460A,11 046l)
>90% of loop >89.4% of loop Reac tor Coolant F low-Low 2.5 1.87 0.60
12. design flow
design flow *
'. (LOOPl LOOP 2 LOOP 3 t00P4 fl 0414 FI 0424 F I -0434 II-0444 FI 0415 I I -0425 I I -0435 fi-0445 I I 0116 II 0426 II 0136 *I-0446) 18.5 17.18 1.67 >18.5% (37.8)*** 217.8% (35.9)***
13. Steam Generator Water tevel of narrow range of narrow range t ow l ow (21.8)*** (18.21)*** instrument span instrument span LOOP 2 LOOP 3 LOOP 4 (LOOPI II 0511 LI 0521 11-0537 (1-0541 1I 0518 11 0528 LI 0538 LI-0548 il 0519 11 0529 11-0539 LI-0549 11 0551 II 0552 II 0553 II-0554) 6.0 0.58 0 >9600 volts >9481 volts
14. Ilndervoltage - Reactor (70% bus voltage) (69% bus vo l t age)
Contarit Pumps 0 >5 7.3 sti >st. tir tinderfrequency - Reactor 3.3 0.50 15.
Coolant Pumps
Loop des ign f low e 95.700 gpm
lime constants ut ilized in the lead-lag controller f or Pressurizer Pressure low ara 10 seconds for lead anil I second for lag. CitANNil CALIBRAIION shall ensure that these time constants are ad*usted to these values.
'the value stated inside the parenthesis is for instrumentat ion that has the lower tap at elevat ion 333"; the value stated outside the parenthesis is for instrumentation that has the lovar tap at elevat ion 4:ut"
5 I'
O O!
1HIS : PAGE ' APPLICAtsil,10 ' UNIT .. I 'ONLY - O 7j'
. 5' , . ,
S. 1ABlE 2.211 (Continuedl- i l E
,g REACTOR TRIP. SYSTEM'INSTRl1 MENTATION TRIP SETPOINTS 1 UNil 1 -
s o
-i
_ Ut "10TAL SENSOR-j , .
' ALLOWANCE ERROR .,
,_, IUNCTIONAL UNIl- (TA) .Z (S) IRIP SLIPOINI At 10WABI L val HL 3 i . o. .
j m
~16. lurbine Irip. l l- a. Low FluidI0il Pressu're N.A. .N.A. N.A. 2580 psig 2500 psig~ -lr l
(Pl -6161, PI-6162, PT-6163)
. J
, b. Turbine.Stop Valve Closure -~ N . A . N.A.. N.A. 296.7% open .196.7% open. ] !
11. Safety Injection Input from ESF
'h.A. N.A.- N.A. N.A. N.A. [
18. Reactor 1 rip-System _[ '
Interlocks N t b a. Intermediate Range N.A. N.A. N.A. 21 x 10 ~ m amp 26 x 10 at-amp Neutron Flux. P-6 ;
i (NI--00358, NI-00368)
l D. Low Power Reactor Irips Bl oc k , P --7.
l 1) - P-10 input L _ .
N.A. N .' A . N.A. 510% of R1Pt $12.3% of RIP #
(NI-0041B&C,'NI-0042B&C,, {'
NI-0043B&C, NI-00448&C)
2) P-13 input. .N.A. N.A. N.A. $10% RIP # lurbine $12.3% RIP # lurbine (PI't)S05;- P1 -0506) Impulse Pressure Impulse Pressure- j
- Equivalent Equivalent l f
c. Power-Range Neutron 'N.A. N.A. N.A. $48% of RIP # $50.3% of RIP #
b0 &C, NI-0042B&C, i NI-0043B&C, NI-0044B&C)
.F
-# RIP ' RAILD lHERMAL POWER' 1
i t
-- .w ., n u , ,
^
lHIS PAGE 'APPLICABLL' 10 UNIl 1 ONt1 rf!.
I
~
! Lc5
' El 1 ABLE 2.2-1 (Continuedl - ~j E$- .
- 4 eg ' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINIS"- UNIl 1 l.- !
! EN "101AL' SENSOR- _;
ALLOWANCE . ERROR !
l. . - . ~ FUNCTIONAL UNIT .(TA) Z .{S) TRIP SEIPOINI ALLOWABtl VAlHL.
j' c. -i n3 d. L Power Range Nuetron Flux, P-9 N.A. N.A. N.A. $50% of RIP # 552.3% of RIP #
(NI-0041B&C, NI-0042B&C.
NI-0043B&C,=NI-00448&C) ,
l f e. Power. Range Neutron! N.A. N.A. 'N.A. >10% of RIP # .
>7.7% of RIP #- ;
Flux. P-10 ;
! (NI-0041B&C, NI-0042B&C, i
'NI-0043B&C, NI-00448&C)
f. luEbine Impulse Chan6er N.A. 'N.A. N.A. $10% RIP # lurbine $12.3% RIP # lurbine -f
Pressure, P-13 Impulse Pressure Impulse Pressure -
p.. Q3 (PI-0505 PI-0506) Ecuivalent Equivalent l
'd L !
L 19. Reactor Irip Breakers N.A. N.A. N.A. N.A. H.A.
! I
20. Automatic Trip and Interlock _' N.A. N.A. N.A. N.A, N.A. '
logic I
t I
'I
EIP = ~ RAILD lHERMAL' POWER 1
1
/. s
_O L) 1HIS PAGE APPLILnot.E 10 UNil 1 ONLY Lf l
4 h 1 ABL E 2.2-1 (Continued)_
TABLE NOTATIONS - UNil 1 5_ !
] Nalt 1: OVERILMPERA1URE AT b al
(
S)f' I o (K 2 -K 2 II !_ - I'I
K 2(P - P') - f ,(at))
o.
ro (1 +1 5) \1 + 2 2 3 5)l 5 al (1 + tsS) il - .Sj Where: AT = Measur ed AT. (Unit 1);
l
'2S = lead-lag compensator on measured al; 1 + t,S 1,, t, = . Time constants utilized 'in lead-lag compensator for al, r, ? 8 s, t, 5 3 s; 1
= Lag. compensator on measured al; r;a 1 + 1,5 13 = Time constants utilized in the lag compensator for al, r, = 0 s; al o = Indicated al at RAlED lilERMAL POWER; K1 5 1.12 (Unit 1);
K, = 0.0224/*F (Unit 1); !
I * ** = Th'e function generated by the lead-lag compensator for l ayg 1 + 1,5 dynamic compensation; 1., i, =
Time constants utilized in the lead-lag compensator for layg, r, 2 28 s, ts 5 4 s; i
1 = Average temperature, 'F; 1
=
Lag compensator on measured Tavg; 1 + t.S 1 =
lime constant utilized in the measured lavg lag compensator, s. = 0 s; l
O O O l THIS PAGE APPLIC*ott 10 UNil 1 ONLY 1
' TABLE 2.2-1 (Continuedl-r-
" TABLE N01 ATIONS (Continuedl - UN!1 I l l
E
] N0lE 1: (Continued) l l' 5 588.4*F (Unit 1) (Nominal l ayg operat ng temperature;)
i l
K, = 0.00ll5/psig (Unit 1);
P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);
5 = Laplace transf orm variable,. s~2 ;
the and f,(AI) is a function of the indicated difference between top and bottom detectors of power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
between -32.0% (Uniit 1) and + 11.0% (Unit 1), f ,(at) = 0, where qt and 4b are l (1) l-or qt - 4b Ub percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and at &
is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - Ab exceeds - 32.0% (Unit 1), the al Irip Setpoint shall be automatically reduced by 3.25% (Unit 1) of its value at RATED THERHAL POWER; and (3) For each percent that the magnitude of qt - gb exceeds + 11.0% (Unit 1), the al 1 rip Setpoint shall be automatically reduced by 1.97% (Unit 1) of its value at RATED lHERHAL POWER.
NUlE 2: the channel's maximum 1 rip Setpoint shall not exceed its computed Irip Setpoint by more than 3.17 (Unit I) of al span.
.s
, -THIS PAGE APPLIC ~tE 10 UNIlL1 ONLY -l. . _ .
8-y- -
__ TABLE 2.2-l'(Continued)
j
i. r- . >
' m ^
TABLE NOT ATIONS '(fontinued). - 11 Nil - 1 l _'
-F .
3 Z
vi N0ll 3: OVERPOWERLaT' 33.(I 61 5) -( 3 I
) $ alf'{K ,~-.K , I } l'- K. [1 I - I"J' if,('al)) j j: 3 ._(1 + 1 25) (1 + TsS). -
(1 + T 25) (1 +: TsS) (IL* 2.5) '
m
. -+
Where: . al -
= Measured al (Unit 1);- :
_l I'*'15,
-- Lead -lag. compensator on measured al; _
, 1 + 1,5 -
t 1,. ,
s- lime constants' utilized in . lead-leg compensator.
for al, 11 2 8 S, 12 $-3 s; ,
.1 1
= Lag compensator on measured:al; 1+-1 35 l 75 #
E .
13 = lime constants utilized in the lag compensator for al, j 1 3=0s;- i i
i al o = Indicated AT at RAlED lHERMAL POWER; i
) i
5 1.08 (Unit 1),
K. l '
l-
-K, 2 0.02/*F for increasing average temperature and 2 0 for decreasing aveiage_ .
temperature -
}
= The function generated by the rate-lag compensator for l ayg dynamic 1 + .1,5 compensation, i 1, = - Time constants utilized in the rate-lag compensator for layg, r , 2 10 s ,
1
, ;.= Lag compensator on measured Tavg; j 1_ + r,S i
f i
?
, . , . _ ~ . . . . e' m --- . , , . - . _ . .-
g"" jg'T % .M#*b.
t 4 ( } \ $
'w/ \_,/ V' g IHIS PAGL APPilCAbtE 10 UNIl 1 DNLY I SS T ABL E 2.2-1 (Continued)~
pd TABLE NOIA110NS (Continuedl - UNil I l E
(( N0lt 3: (Continued) ,
[, i.
= lime constant utilized in the measured l ayg lag compensator,
,, t. = 0 s; K. > b.9020/*F (Unit 1) for T > T" and K. = 0 for I 5 1". l 1 = Average Temperature, 'F; 1" = Indicated Tavg at RAILD THERMAL POWER (Calibration temperature for al instrumentation, 5 588.4*F (Unit 1)), l S = Laplace transform variable, s 2; and f,(al) = 0 for all al.
,Y Nolt 4: The channel's maximum Trip Setpoint shall not exceed its computed trip Setpoint by more than 1.9% (Unit 1) of AT span.
l
t'h ,"s r ~%
THIS PAGE APPLICAum. TO UNil 2 ONLY 8 TABL E_ 2.2-la - UNil 2 r-REACTOR TRIP SYSTEM INSTRUMENTATION 1 RIP SCIPOINIS z
El 101AL SENSOR
'" ERROR ALLOWANCE IUNCTIONAL UNil (TA) Z_ (S) TRIP SE1POINI ALLOWABLE VAlut.
1. Manual Reactor Trip N.A. N.A. N.A. N.A. H.A.
2. Power Range, Neutron Flux (NI-0041B&C, NI-0042B&C, NI-0043B&C, N1-00448&C)
a. liigh Setpoint 7.5 4.56 0- $109% of RIP # $111.3% of RIP #
b. Low Setpoint- 8.3 4.56 0 $25% of RlP# $27.3% of RIP #
3. Power Range, Neutron Flux, 1.6 0.50 0 55% of RIP # with 56.3% of HIP # wille liigh Positive Rate a time constant a time constant m (N1-0041B&C, NI-0042B&C, 22 seconds 22 seconds
,'. N1-0043B&C, N1-00448&C)
4. Deleted.
5. Intermediate Range, 17.0 B.41 0 $25% of rip # $31.1% of RIP!
Neutron Flux (NI-0035B, NI-00368)
6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps 14x 1 los (ps (NI-00318, N1-00328)
1. Overtemperature al 6.6 3.37 1.95 See Note 1 See Note 2 (101 -411C, 101 -4 210, (Unit 2) (Unit 2) & 0.50 101-431C. 101-441C) (Unit 2)
8. Overpower AT 4.9 1.54 1.95 See Note 3 See Note 4 (101 -4118. 101 -421 B , (Unit 2) (Unit 2) 10I-4318, 101-4418)
FRTP T RAIED iHERMAE POWER
k \ )
IHIS PAGE erPLICABLE 10 UNIT 2 ONLY h TABLE 2.2-la (Continued)
N" REACIOR TRIP SYSTEM INSTRUMENI ATIG:3 TRIP SETPOINTS - U.1II 2 E
G 10TAL SENSOR
ALLOWANCE ERROR FUNCTIONAL UNIT (TA) Z (S) TRIP SETPOINT Att0WACLL VAIUE
9. Pressurizer Pressure-Low 3.1 0.71 1.67 21960 psig** 21950 psig
" (PI-0455A,2&C PI-0456 &
PI-0456A, PI-0451 & PI-0457A, PI-045R & PI-0458A)
10. Pressurizer Pressure-High- 3.1 0.71 1.67 52385 psig $2395 psig (PI-0455A.B&C, PI-0456 &
PI-0456A, PI-0457 & PI-0457A.
t PI-0458 & PI-0458A)
II. Pressurizer Water Level-High 8.0 2.18 1.6T $92% of instrument $93.9% of instrument (LI-0459A, LI-0460A, LI-0461) span span
'? 12. Reactor Coolant Flow-Low 2.5 1.87 0.60 290% of loop 289.4% of loop design flow
design flow
C (LOOPl LOOP 2 LOOP 3 LOOP 4 F1-0414 FI-0424 FI-0434 FI-0444 F I -0415 FI-0425 F1-0435 FI-0445 FI-0116 FI-0426 FI-0436 FI-0446)
13. Steam Generator Water Level 18.5 17.18 1.67 218.5% (37.8)*** 217.8% (35.9)***
Low-Low (21.8)*** (18.21)*** of narrow range of narrow range instrument span instrument span (LOOPl LOOP 2 100P3 LOOP 4 LI-0517 LI-0527 L1-0537 Lt.-0547 11-0518 LI-0528 LI-0538 L I -0548 L1-0519 LI-0529 LI-0539 LI-0549 LI-0551 11-0552 LI-0553 LI-0554) 14 Undervoltage - Reactor 6.0 0.58 0 29600 volts 29481 volts Coolant Pumps (70% bus voltage) (69% bus voltage)
15. Underfrequency - Reactor 3.3 0.50 0 257.3 HZ 257.1 Ht Coolant Pumps
Loop design flow - 95.700 gpm
1ime constants . utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds f or lead and I second for lag. CHANNEL CALIBRA110N shall ensure that these time constants are adjusted to these values.
1he value stated '.aside the parenthesis is f 6r instrumen*ation that has the lower tap at elevation 333"; the value stated outside the parenthesis is for instrumentation that has the lower tap at elevation 438".
M, THIS'PAGL APPLICABLE.10 UNIl 2 ONLY-8
$3 TABLE'2.2.1a (Continuedl., '
Ei
. e REACTOR TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINIS - UNIT 2 't l '- . 55 -
SENSOR' ;
Ut ,
.101AL
' ALLOWANCE .
-ERROR-j (TA) Z. (S) ALLOWABLE val.4L.
- IUNCTIONAL UNIT l1 RIP SEIPOINI ';
I o.
n, 16. lurbine 1 rip .
, r
a. Low Fluid Oil Pressure N.A. N.A. N.A. 2580 psig 2500 psig- :
(Pl -6161, PT-6162, PT-6163)
b. lurbine-Stop Valve Closure ~ N.A. N.A. M.A. 296.7% open 296.7% open i
II. Safety injection input _from ESF N.A. N.A. N.A. N.A. N.A.
t
18. Reactor Irip' System z i Interlocks' -[
! 7' 7: a. Intermediate Range N.A. N.A. N.A. 21 x 10 20 amp 26 x 10 ta amp ;
i Neutron Flux, P-6 .
(NI-00358,:NI-00368) 4 I l b. Low Power Reactor Trips j
Block, P-7 ;
1) P-10 inputL 'N.A. N.A. N.A. 510% of RIP # $12.3% of RIP #
! (N1-0041B&C, NI-0042B&C, NI-0043B&C, NI-0044B&C) i
2) P 13 input . .
N.A. N.A. N.A. 510% RIP # lurbine $12.3% RIP # lurbine (PI-0505, PI-0506) Impulse Pressure Impulse Pressure.
Equivalent Equivalent
'c. Power Range. Neutron N.A. N.A. N.A. 548% of R8P#. 550.3% of RIP # l Flux; P-8' (NI 00418&C,!NI-0042BP.C, NI-0043B&C, NI-00448&C)
- i '
. - -# RIP
RATED 1HERMAL POWER-- j r c. -
?
1HIS PAGE APPIICABLE 10 UNil'2 ONLY 8
$ 1ABLE 2.2 'a-(Continued}_
i E' . .
.;i j c REACIOR TRIP SYSTEM INSTRUMENTATION lRIP SETPOINIS - UNil '2 -
' -5 . .
d- 101AL ..
SENSOR
=- ALLOWANCE ERROR I f_UNC110NAL UNIT _ (TA)__ Z 15) TPIP_SEIP0lN1_ Al LOWABLL_ VAlllll l
QuD .
, y d. Power Range Nuetron F lux, P-9 N.A. N.A. ..N.A. 550% of-RTP#. 552.3% of' RIP # ,
(N1-0041B&C, NI-0042B&C, N1-0043B&C, NI-00448&C)~
i e. Power Range Neutron N.A. N.A. N.A. >10% of RIP # >I.7%:of RIP #
F l ux , : P -10 . .
(N1-0041B&C. NI-0042B&C, N1-0043B&C, N1-0044B&C)-
i. Turbine. Impulse Chamber- N.A. N.A. N.A. $10% RIP # lurbine 512.3% RIP # lurbine.
i Pressure, P-13. Impulse Pressure impulse Pressure
(P1 -0505, P1-0506) Equivalent Fquivalent m
, 19. Reactor Irip Breakers .N.A. N.A. ~N4A. N.A. N.A.
20. Automatic 1 rip and. Interlock N.A. N.A. N.A. N.A. N.A.
Logic I
4
~ -
#R I'P e RATED 11tERMAL POWER-
^ " - ' - - '
+
-_ __ _ _ _ _ _ __ __________.___________________L
( I (,-- )
,v ~
v 1HIS PAGE APPLICABLE 10 UNil 2 ONLY 8 T ABL E 2.2-la (Continuedl A
TABLE NOTATIONS - UNil 2 E
h N0lt I: OVERILMPERAIURE al 31 (I * *tS) I 5 ATo (K1 - K 2 ( +
} [I - I' ]
K3 (P - P') - f 1 (al)]
e- (1 +1 2 5) I & 12 5 - (1 + TsS) 1+1 56 m
Where: a1 = Measured al by RID Manifold Instrumentation (Unit 2);
I*** = Lead-lag compensator on measured al;
_1+T 2 5 1x, 12 = lime constants utilized in lead-lag compensator for al,1 1 > 8 s.
13 5 3 s; m r lag compensator on measured al;
,'. _1+v 35
\
13 = lime constants utilized in the lag compensator for al.1 3 = 0 s; alo = Indicated AT at RAlED 1HERMAL POWER; Ki $ 1.10 (Unit 2);
K2 = 0.012/*F (Unit 2);
I * **3 = The function generated by the lead-lag compensator for lavg 1 & tsS dynamic compensation; i., is = lime constants utilized in the lead-lag compensator fo. la vg. '. 2 28 s, Ys 5 4 5; '
1 = Average temperature, *F; I =
lag compensator on measured Tavg; 1 & T.S 1 =
Time constant utilized in the measured lavg lag compensator, 1 = 0 s;
. O O THIS PAGE APPLICautE 10 UNil -2 ONLY O i
~
. TABLE 2.2-la (Continuedl -
i .
TABLE NOTATIONS (Continued) UNil 2 ;
a g1 NOIE.1: (Continued) !
I' '< 588.5 F (Unit 2)7(Nominal T avg operating temperature); l-K, = 0.00056/psig (Unit 2); !
{
' Pressurizer. pressure, psig; P =
i P' = . 2235 psig (Nominal RCS operating pressure);
5 = Laplace transform variable, s-1;
~
i and f,(al) is a function of the indicated difference between top and bottom detectors of_the l power-range neutron ion chambers; with gains to be selected based on measured instrument !
response during plant startup tests such that: '
. ro h (1) For qt 4b between -33.5% (Unit 2) and + 6.5% (Unit 2), f,(al) - 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt-* 4b j' .['
l is total THERMAL' POWER in percent of RATED THERMAL POWER-1 i (2) For each percent that the magnitude of 'It - 4b exceeds - 33.5%-(Unit 2), the al .1 rip Setpoint shall l be automatically reduced by 1.27% (Unit 2) of its value at RAILD 1HERMAL POWER;:and ;
(3) For each percent that the magnitude of qt .- gb exceeds
6.5% (Unit 2), the' al Irip Setpoint shall be automatically reduced by 0.83% (Unit 2) of its value at RATED THERMAL POWER.
l t
N01E 2: The channel's maximum Trip Setpoint shall.not exceed its computed Irip Setpoint by more than 2.5% (Unit 2). l t
i
'[
l t
f
.l t
- l
"^-
r~
\ /
! ) 0 UNil 2 ONLY
'^~~'
li!IS PAGE APPLIL _t hk T ABLE 2.2-la (Continuedl A
l
"' TABLE NOTA 110NS (Continuedl - UNil 2 E
_4 unu 1- "'I'nPOWER al m
I gj (I * ' S) ( I ) < gi g 1,5 )( 1
) j _ g gj ( l __} _ y . j _ g,43g)y (1 +t 5) (1 +1 3 5)
[K, _ K,(1(_ + i,5) (1 + 1.5) (1 + 1.5) 2 3 t
ro I Where: AT = Measured aT by RID manifold instrumentation (Unit 2);
l * '25 = lead-lag compensator on measured al; 1 + 1,5 2 1 12 = Time constants utilized in lead-leg compensator for AT, 11 2 8 S. 22 5 3 s; 1
~
= Lag compensator on measured al; 7 1 +v 35
~
co r = lime constants utilized in the lag compensator for al, 3
1 3=0s; al o = Indicated AT at RAlED THERMAL POWER; K. 5 1.089 (Unit 2),
K, 1 0.02/*F for increasing average temperature and > 0 f or decreasing average temperature,
= The function generated by the rate-lag compensator for layg ilynami(
1 + t,5 compensation, i, = lime constants utilized in the rate -lag compensator f or layg, r, 2 10 s.
1
= Lag compensator on measured lavg; I + 1.5
w THIS PAGE APPilC. \r'10 UNil 2 ONLY
-=
g TABLE 2.2-la (Continuedl EN TABLE NOIAI10NS (Continued) - UNil 2 c-z
q NOlt 3: (Continued) m i, = Time constant utilized in the measured l avg lag compensator,
r. = 0 s;.
n.
^3 K. > 0.0013/*F (Unit 2) f or T > T" and K. = 0 for T < 1",
l 1 = Average Temperature, *l ;
1" =
Indicated Tavg at RAILD TifERMAL POWER (Calibration temperature for al instrumentation, < 588.5*F (Unit 2)), l S = Laplace transform variable, s-1; and f,(al) = 0 for all al.
Y 0$ NOIL 4: The channel's maximum Trip Setpoint shall not exceed its computed Irip Setpoint by more than 2.4% (Unit 2) of AT span.
l u
i l
i l .
1 I
I l
l l
D 'HIS PAGE APPLICABLE TO UNil 1 DNLY
- l )
i 1
-2.1 SAFETY I1MITS BASES 2.1.1 REACTOR CORE - UNIT 1 I The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforatian which would result in the release of fission
. products to the reactor coolant. Overneattnq of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transf er coef ficient is large and the cladding surf ace temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime 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 dir?ctly measurable parameter during operation and therefore THERMAL. POWER and reactor coolant temperature and pressure have been related to ONB through correlations which have been developed to predict the l
-DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The-local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local. heat flux and is indicative of the margin to DNB.
, The DNB-thermal design criterion is that the probability that DNB will
\
not occur on the most limiting rod is at least 95% (at a 95% confidence level) for-any Condition I or II-event, In meeting the DNB design criterion,-uncertainties in plant operating parameters, nuclear and thermal parameters, fuel f abrication parameters, and computer codes must be considered. As described in-the FSAR, the effects.of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the-DNB design criterion.
Additional DNBR margin is maintained by performing the safety-analyses to a-higher DNBR limit. This margin between the design and safety analysis limit ONBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and .to provide DNBR margin for operating and design flexibility.
The curves of Figure 2.1-1 show reactor core :cf et 3 limits for a range of THERMAL POWER, REACTOR COOLANT SYSTEM pressure, and anrage temperature which satisfy the following criteria:
A. The-average enthalpy at'the vessel exit is less than the enthalpy of saturated liquid (far-left line segment in each curve).
B. The minimum DNBR satisfies the ONB design criterion (all the other line segments in each curve). The VANTAGE 5 fuel is analyzed using the WRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical cell and thimble cells, respectively. The LOPAR fuel is analyzed using the WRB-1 correlation with design limit DNBR values of 1.23 and l . t(.~ 1.22 for the typical and thimble cells, respectively.
I C. The hot channel exit ouality is not greater than the upper limit of the quality range (including the effect of uncertainties) at the UNB correlations. This is not a limiting criterion for this plant.
! V0GTLE UNITS - 1 & 2 8 2-1
f
-THIS PAGE APPLICABLE TO UNIT 2 ONLY l 7 .
1 L 2.1 SAFETY LIMITS BASES-2.1.1 RE ACTOR CORE - UNIT 2 l
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-fuel cladding is prevented by restricting-fuel operation to within the nucleate' boiling regime where the heat transfer coef ficient is large and the cladding surf ace temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures. because of the onset of departure from' nucleate boiling (DND) and the resultant sharp reduction in heat transfer coef ficient 10NB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related_to DNB through the W-3 (R Grid) correlation. The W-3 (R Grid)' DNB correlation has been developed to predict the=DNB flux and the location of DNB for axially uniform and nonuriform heat flux distributions. The local DNB heat flux ratio (DNBR) is-defined as the ratio of the heat flux that would (3 cause.DNB at a particular core location ~ to the local heat flux and is.
_() indicative of the margin to-DNB.
The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability-at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all _ operating conditions.
The curves of Figure' 2.1-la show reactor core ~ safety limits which are l determined for a range of' reactor operating conditions., The, core limits represent _the. loci of points of THERMAL POWER, REACTOR COOLANT SYSTEM pres-sure and average temperature which satisfy the following. criteria:
A. The average- enthalpy at the vessel exit is equal -to the enthalpy of saturated-liquid (f ar lef t line segment -in each curve).
B. .The minimum DNBR is not less than the design limit (all the.Other .line segments in each curve). !
C. :The hot channel exit. quality is-not greater than the upper limit of the q'ality u range of the W-3'(R-Grid) correlation which is 15% (middle line segment on Reactor Coolant System pressure curves, 2400 psia and 2250 psia; this-is not a limiting criterion for this plant).
O V0GTLE UNITS -.1 & 2 B 2-la
. - . . _ ~ - _ _ _ _ _ - . --
l
. SAFETY LIMITS.
BASES- ,
REACTOR CORE (Continued)
These' curves are based on an enthalpy hot channel f actor, F$(( and a reference cosine with- a peak of 1.55 for axial power shape. An allowance is ,
includedforanincreaseinF$gatreducedpowerbasedcntheexpres: ion:
F$H=FgkPg pp (),p))
Where: pgPis the limit at RATED THERMAL POWER (RTP) specified in.the CORE OPERATING LIMITS REPORT (COLR),
PFaH is the Power Factor Multiplier for F$t:
tied in the COLR, and P is the fraction of RTP,
~ These limiting heat flux conditions are higher than those calculated for the range of-all controlirods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits;of the fc (al) function of the Overtemperature-trip, When the_ axial power imbalance
- - 'is not within the tolerance, the axial power imbalance ef f ect on the Over-temperature AT trips will reduce the Setpoints-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 Reacto_r
. Coolant' System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant.f rom reaching the _ containment
atmosphere.
The reactor vessel, pressurizer, and the RCS piping, valves and fittings are-designed to-Section 111 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 entire-RCS is hydrotested at 125% (3107 psig) of design pressure, to
demonstrate integrity prior to initial operation.
LO V0GTLE UNIls - 1 & 2 8 2-2 l'
.,...--.-..--.,..----..-----...-..~.-.-.~.--_1
-)
O REACTIVITY' CONTROL SYSTEMS
- BORATED WATER SOURCE - SHUTDOWN l.!MITING CONDITION FOR OPERATION L3~.1.2.5- As a minimum, one ~ of the following borated water sources shall be OPERABLE:
a. A Boric Acid Storage Tank withi
1) A minimum contained borated water volume of 9504 gallons (19%
of instrument span) (L1-102A, L1-104A),
2) A boron concentration between 7000 ppm and 7700 ppm, and
3) - A minimum solution temperature of 65'F (TI-0103).
b. The refueling water storage tank (RWST) with:
1) A minimum contained borated water volume of 99404 gallons (9% of instrument span) (L1-0990A&B, LI-0991A&B, LI-0992A, LI-0993A)',
2) A boron concentration between 2400 ppm and 2600 ppm, and.
3) 'A' minimum solution temperature of 44'F (Unit 1) or 54*F (Unit 2) - l'
- ( TI-10982) .
APPLICABILITY: MODES 5'and 6.
ACTION:
With no borated water. source-0PERABLE, suspend all operations involving. CORE 4 ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.1.2.5, The above required b' orated water source shall be demonstrated OPERABLE:
a. At least once'per 7 days by:
, 1) Verifying the boron concentration of the water,
2) Verifying the contained borated water volume. and-
- 3) When the' boric acid storage tank is the source of borated water and the ambient temperature of _the boric acid storage tank' room
- (TISL'-20902, TISL-20903) is 172*F, verify the boric acid storage O b.
tank solution temperature b >65'F.
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by . verifying the RWS1 temperature (T!-10982)-
when ,it is the source of borated water and the outside air temperature
is ~less than 40*F (Unit 1) or 50'F (Unit 2). l V0GTLE UNITS - 1 & '2 3/4 1-11 W v '
r-m- h ya v'-V- " - ' ' -
m.m m. - .._. .___.__-._-._...__._._______._.__...._.__...,.m t
-i
)
t 1s ; REACTIVITY CONTROL SYSTEMS
BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION ._ _ ___,
~
3.1.2.6 As a minimum: the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
-a. A Boric Acid Storage Tank with: .
1) A minimum contained borated water volume of 36674 gallons (81%
of instrument span) (LI-102A, LI-104A),
2) A boron concentration between 7000 ppm and 7700 ppm. and 3 )' A minimum solution temperatore of 65'F (TI-0103). .
~b.- The refueling water storage tank (RWST) with:
.' ) A minimum contained borated water volume of 63147B gallons (86%
of instrument span)-(LI-0990A&B, L1-0991A&B, L1-0992A, LI-0993A),
() 2) A boron. concentration between 2400 ppm and 2600 ppm,
3) A minimum solution temperature of 44'F -(Unit 1)--or 54'F-(Unit' 2), ,
4) A maximum solution temperature of 116'F (TI-10982), and
5) "RWST Sludge. Mixing Pump Isolation _ Valves capable cf closing on RWST low-level.
- APPLICABillTY: MODES 1, 2,'3,-and 4.
ACTION:
a. 'With-the Boric: Acid-Storage Tank' inoperable and being used as one'of the above required borated water sources, restore the
~
+
tank to OPERABLE' status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be -in at least HOT STANDBY within-the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to-a-SHUTDOWN MARGIN as required by Figure 3.1-2 at 200*F; restore the Boric Acid .
Storage Tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the nextE30 hours. .
! b. With the RWST-inoperable, except for-the Sludge Mixing Puma
- j isolation Valves,. restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ll or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD-SHUTDOWN within the following 30 hotrs. .
O .
' V0GTLE UNITS - 1 & 2 3/4 1-12
. _ _ _ _ __ . _ _ _ _ . __ _ _ . ~ _ _ - .
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATIO_N (Continued) _
ACTION (Continued)
c. With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolate the sludge mixing system by either closing the manual isolation valves er deenergizing the OPERABLE solenoid pilot valve within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and maintain closed.
d SURVEllLANCE REQUIREMENTS 4.1.2.6 Each borated water source shai' be demonstrated OPERABLE:
a .~ At least once per 7 days by:
1) Verifying tne boron concentration in the water.
2) . Verifying the contained borated water volume of the water source, and
3) When the boric acid Mor39e tank is the source of borated water and the ambient temperature of the boric acid storage tank room (TISL-20902, TISL-20903) is 5 72'F, verify the boric acid storage tank solution temperature is 165'F.
b. At least once per ?4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by verifying the RWST temperature (TI-10982) when the outside air temperature is less than 40*F (Unit 1) or 50'F (Unit 2).
c. At least once per 18 months by verifying that the Sludge Mixing Pump Isolation Valves automatically close upon RWST low-level test signal.
i L
U V0GTLE UhlTS - 1 & 2 3/o 1-13
'O V REACTIVITY CONTROL SYSTEMS ROD OROP TIME I.!MITING CONDITION FOR OPERATION 3.1.3.4 The individual shutdown and control rod drop time from the physical fully withdrawn position shall be less than or equal to 2.7 (Unit 1) or 2.2 (Unit 2) seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
a. Tavg (TI-0412, TI-0422,11-0432, TI-0442) greater than or equal to 551 F, and
b. All reactor coolant pumps operating.
APPLICABillTY: MODES I and 2.
ACTION:
With the drop time of any 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.
< p SURVEILLANCE REQUIREMENTS
'4.1.3.4 The rod drop time shall be demonstrated through measurement prior to reactor criticality:
a. For all rods following each removal of the reactor vessel head,
b. For specifically affetted individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and
c. At least once per 18 months.
O l i
V0GTLE UNITS - 1 & 2 3/4 1-19 l
1
THIS PAGE APPLICABLE TO UNIT 1 ONLY l 3/4,2^ DOWER DISTRIBUTION LIMl?S !
3/4.2l1 AXIAL FLUX O!FFERENCE - UNIT 1 LIMITING CONDITION FOR OPERATION _ ,
r ,
3.2.1 The indicated (NI-0041B, NI-00428, NI-0043B, NI-00448) AXIAL FL'UX OlFFERENCE-(AFD) shall be maintained within the limits specified in the CORE I OPERATING LIMITS REPORT (COLR).
i APPLICABILITY: MODE 1 ABOVE 50 PERCENT RATED THERMAL POWER *.
ACTION:-
g a.
With the' indicated AX1AL FLUX O!FFERENCE outside of the limits specified in the COLR,
1. Either restore the indicated AFD to within the limits within 15 minutes, or 2; Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux *:- High Trip setpoints to less than or equal to 55 percent of RATED ' '
THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months b.
. THERMAL POWER shall not be increased above 50% of RATED THERMAL ~ POWER unless the indicated-AFD is within the limits specified in the COLR.
v SURVElllANCE REOUIREMENTS 4 . 2 .1,1
'The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED. THERMAL POWER by:
a. Monitoring-the indicated AFD for each OPERABLE excore channel: I
'l) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and'
2) At least once' per hour until the AFD Monitor Alarm is updated af ter restoration to OPERABLE status.
b. Monitoring and logging the indicated AFD for each OPERABLE excore channel- at least once per hour' for the .first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoper-able.- The logged values of the indicated AFD shall be assumed to exist during the interval r. receding each loggir.g.
c. The provisions- of Specification 4.0.4-are not applicable.
'4,2.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD :o be outside its limits.
"See Special Test Exceptions Specification 3.10.2.
, V0GTLE UNITS - 1 & 2 3/4 2-1
\
3
/
L)L THIS PAGE APPLICABLE 10 UNIT-2 ONLY 3/4.2 DOWER 0157RIBUTION LIMITS 2/4i2.1~ AX1 AL FlVX OlFFERENCE 'JNif 2
~ l LIMITING CONDITION FOR OPERATION 3.2.1 The-indicated (N!-00418. NI-00428, NI-00438, NI-00448) AXIAL FLUX.
'OlFFERENCE (AFO) shall be maintained within the target band (flux difference
' units) about the target flux difference. The target band. is specified.in the CORE OPERATING' LIMITS REPORT (COLR).
The indicated AFD'may deviate outside the reouired target band at greater than or. eaual to' 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is vdthin the Acceptable Operation Limits specified in the LOLR and the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
'The indicated AFD may deviate outside the required target band at greater than IST. but less than 50% of RATED THERMAL' POWER provided the cumulative penalty deviationitime does not exceed'I hour curing the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
APPLICABILITY:'. MODE 1, above 15% of RATED THERMAL-POWER * #
Q v
' ACTION:-
a.- With the indicated AFD outside of the required target band and with
-THERMAL POWER greater than or eaual to 90% of RATED THERMAL POWER, within 15 minutes either:
.l .
Restore the indicated AF0~to within-the target band limits, or 2.
Reduce-THERMAL. POWER to less than 90% of' RATED THERMAL POWER.
b; With .the indicated AFD outside of the required target. band for-more than ILhour of. cumulative penalty deviation- time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside' the Acceotable Operation Limit; specified in the
~ COLR.and with THERMAL POWER 11ess than190% but equal to or greater
'than 50%-of RATEDLTHFPMAL POWER, reduce:-
1.
THERMAL 30' minutes,POWER and :o less than 50% of- RATED THERMAL ~ POWER within
2. The Power Range Neutron Flux * - High Setpoints to.less than or equal- to' 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
"See Special Test Exceptions Specification 3.10.2.
/'. #Surveillance testing of the Power-Range Neutron Flux Channel may be performed (celow 90% of RATED THERMAL POWER) pursuant to Specification 4.3.1.1 provided the indicated AFD is . maintained within the Acceptable Operation Limits speci-fied in the COLR. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ooeration may be accumulated with the
- a AFD outside of the above requirea target bana during testing without penalty
'wiation.
'!OGTLE UNITS - 1 & 2 - 3/4 2-la '
THIS PAGE APPLICABLE TO UN?T t ONLY
.- l
-Q [ POWER DISTRIBUTION LIMITS t!MITING CONDITION FOR OPERATION - UNIT 2 l ACTION (Continueo)
c. With the indicated AFD outside of the required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with IHERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER snali not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is withir. the required target band and the cumulative penalty devia-tion has been reduced to less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
SURVEILLANCE REOUIREMENTS 4 . 2 . I '.1 The indicated AFD shall be determined to be within its' limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
a. . Monitoring the indicated'AFD for each OPERABLE.excore channel:
1) At least once per 7 days when.the AFD Monitor Alarm is OPERABLE, and.
i
2) At least once per hour until the AFD Monitor Alarm is updated after restoration to OPERABLE status,
b. . Monitoring and logging the indicated AFD for each OPERABLE excore char.3el at least Ece per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least
--once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. .The logged values of the indicated AFD shall be assumed to exist during the interval preceding each le 1ing.
-4.2.1.2 The indicated AFO shall'be considered outside of its target band when-
, two.or more'0PERABLE excore channels are indicating-the AFD to'be outside the
-target band. Penalty deviation outside of the required target band shall be Laccumulated on a time basis of:
a. One minute penalty deviation for each 1 minute.of POWER.0PERATION outside of-the -target band at THERMAL POWER levels equal -to or;above 50% of RATED THERMAL POWER.-and
b. One-half minute penalty. deviation for each 1 minute of POWER OPERATION outside of the target band at THEitMAL POWER levels between 15% and 50% of RATED THERMAL-POWER.
4.2.1.3 -The target. flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.
The provisions of Specification 4.0.4 are not appl ~icable.
4.2.1.4 The' target flux dif ference shall be updated at least once per O
._31 pursuant Effective Full Power Days to Specification 4.2.1.3 by above either determining or by linear the target luxbetween interpolation difference the most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable.
V0GTLE UNITS - 1 & 2 3/4 2-2
Y k
POWER DISTRIBUTION' LIMITS 3/4.2.2 HEAT Fi UX' HOT' CHANNf'. FACTOR - Fg{JJ LIMITING' CONDITION FOR r/ERATION
,- 3.2.2 FQ (Z) shall be limited by the following relationships:
-FQ (Z) 5-FQ" [K(Z)] for P > 0.5 P
FQ(Z) $ FO RTP [K(Z)] for P 5 0.5.
0.b RTP Where: Fg = the FQ limit at RATED THERMAL POWER (RTP)_
specified in the CORE.0PERATING LIMITS REPORT (COLR),
Where:_ P_= THERMAL POWER
, and
[.. RATED THERMAL POWER K(Z) = the normalized Fn(Z) as a function-of core height specified in the COLR.
APPLICABILITY: ' MODE-1.
ACTION:'
~With F Q (Z) erceeding.its limit:
a. - Reduce THERMAL ~ POWER at least 1% for ecch 1% Fg(Z) exceeds the' limit within 15 minutes-and'similarly. reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4
hours;.. POWER OPERATION may proceed for up-to a total of-72.
hours;_ subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints ~(value of K ) have' been- reduced .
at least ~1%:(in ai span)lfor each 1% Fg(Z) exceeds .the -
limit;-and a
-b.- Identify and. correct the cause of the out-of-limit condition prior'to increasing. THERMAL POWER above.'the reduced limit re-quired by ACTION a., above;. THERMAL POWER may then be increased ~
provided Fn(Z) is demonstrated through incore mapping to be within its limit.
,0 h _V0GTLE UNITS - 1 & 2 3/4 2-3 o
THIS PAGE APPLICABLE TO UNIT 1 ONLY POWER DISTRIBUTION LIMITS SURVEllLANCE REQUIREMENTS - UNIT 1 4.2.2.1 The provisions of Specifications 4.0.4 are not applicable.
4.2.2.2 FQ (Z) shall be evaluated to determine if it is within its limit by:
a. Using the movable incore detectors to obtain a power distribution map at any THERMA'. POWER greater than 5% of RATED THERMAL POWER.
b. Determining the computed heat flux hot channel f actor, FQ (Z), as follows:
Increase the measured QF (Z) obtained from the power distribution map by 3% to account f or manuf acturing tolerances and f urther increase the value by 5% to account for measurement uncertainties,
c. Vr.rifying that FQ (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.
d. Satisfying the following relationship:
RTP FQ (Z) FO x K(Z) for P > 0.5 P x W(Z)
RTP FQ (7) Fg x K(Z) for P 5 0.5 0.5 x W(Z)
Where FQ (Z) is obtained in Specification 4.2.2.2b e ove, FQ is the FQ li. nit, K(Z) is the normalized F0(2) as a a unction of core height, P is the fraction of RATED THERMAL POWER, and W(Z) is the tw ?' dependent function that accounts for power distributio' e _ients encountered during normal operation.
FQ , K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specificction 6.8.1.6.
e. Measuring F Q (Z) according to the following schedule:
1. Upon achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined *, or
2. At least once per 31 Effective Full Power Days, whichever occurs first.
"During power escalation af ter each f uel loading, power level may be O- increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.
V0 GILE UNITS - 1 & 2 3/4 2-4
W~
e 1
\
/
p U THl5 PAGE APPLICABLE TO UNIT 1 ONLY i POKER _ DISTRIBUTION i IMITS SVRVilllANCE REOUIREMENTS (Continued) UNIT 1 l
f.
With measurements indicating niaximum
[Fo(Z)h over Z
( K(Z))
C has of the increased followingsince theshall actions D.?vious be taken:determination of FQ (Z) either
1) Increase F Q (Z) by 2% and verify that this valus satisfies the relationship in SDecification 4.2.2.2d, or E
2) Fg (Z) shall be measured at least once per 7 Ef f ective Full Power Days until two successive maps indicate that p:imum C
[Fg(Z)) is not increasing, over 2 (K(Z)j g.
With the relationships specified in Specification 4.2.2.2d above not being tatisfied:
1) Calculate the percent Fn(Z) exceeds its limits by the following expression:
[ maximum IF0(Z) x W(Z)
< RTP ~1 over Z FO X MU x 100 for P > 0.5
(
- P -
l
) ~ $
6 E
[ maximum Fo(Z)xW(Z)'\ I I
over Z RIP '~
Fg x K(2) > x 100 for P 1 0.5, and
- 0.5 -
?) The following action shall be taken.
Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specifiad in the CORE OPERATING LIMITS REPORT by 1% AFD for ::ach percent CN FQ (2) exceeds its limits as determined in Specification 4.2,2.29 1. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to V tnese modified limits.
V0G1LE UNITS - 1 & 2 3/4 2-5
- . . . . . . _ _ _ . - . . ~ . _ . . _ . _ _ . . . _ _ _ . _ _ _ _ . _ _ _ _ . _ - - _ . . _ _ _ . _ . _ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ .
THIS PAGE APPLICABLE TO UNIT 1 ONLY l POWER DISTRIBUTION LIM 115 SURVE!LLANfC REOUIREMENTS (Continued) - UNIT 1 _., _ l
h. The limits specified in Specification 4.'2.2.2c are applicable in all core plane regions, i.e. 0 - 100%, inclusive,
i. The limits specified in Spec'.fications 4.2.2.2d, 4.2.2.2f, and i 4.2.2.29 above are not applicable in the following core plane regions:
1) Lower core region f rom 0 to 15%, inclusive.
2) Upper core region f rom 85 to 100%, inclusive.
4.2.2.3 When fg(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.* an overall measured FQ(Z) shall be obtained from a power distribution map and increased by 3%
to account far manufacturing tolerances and further increased by 5%
to account for measurement uncertainty.
O .
O V0GTLE UNITS - 1 & 2 3/4 2-6
I l
THIS PAGE APPLICABLE TO UNIi 2 ONLY l POWER DISTRIBUTION LIMITS SURVE!LLANCE REQUIREMENTS - UNIT 2 'l 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 Fxy shall be evaluated to determine ifQF (Z) is within its limit by:
a. Using the movable intore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER before exceeding 75% of RATED THERMAL POWER following each fuel loading.
b. Increasing the measured Fxy component of the power distribution map by 3% to account for manuf acturing tolerances and f urther increasing the value by 5% to account for measurement uncertainties,
c. Comparing the Fry computed (Fxy) obtained in Specification 4.2.2.2b.
dbove to:
1) measuredcoreplanesgiveninSpecificatio[n4.2.2.2e.andf.The below, and .
2) The relationship:
Fxy = Fh (l &PFxy(1-P) },
Where Fxy is the limit for f ractional THERMAL POWER operation expressed as a function of F f PFx is the power factor multiplier for Fxy specified 5n,the bOLR, and P is the f raction
, of RATED THERMAL POWER at which Fxy was measured.
d. Remeasuring Fxy according to the following schedule:
1) WhenFahisgreaterthantheF!flimitfortheappropriate measured core plane but less than the Fxy relationship, a(ditional powerdistributionmapsshallbetakenandFxhcomparedtoF!hP and Fxy either:
a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWERorgreater,theTHERMALPOWERatwhichFfwaslast determined, or b) At least once per 31 Ef fective Full Pwer Days (EFPD),
whichever occurs first.
V0GTLE UNITS - 1 & 2 3/4 2-7
- _ = _ _ _ _ _ . - - , -
m THIS PAGE APPLICABLE TO UNIT 2 ONLY l POWER DISTRIBUTION LIMITS SURVE!LLINCE RE0VIREMENTS (Contirued) - UNIT 2 l l
2) When the fxy is less th6n or equal to the F limit for the appropriate measured co*e plane, additional power distribution i maps thall be taken and F and F xy at least xy coinpared to F once per 31 EFPD.
e. The Fxy limits used in the Constant Axial OfIset Control analysis l
for RATED THERMAL POWER (F ) shall be specified for all core planes containing Bank "0" control rods and all unrodded core planes in the COLR per Specification 6.8.1.6;
f. The Fxy limits of Specification 4.2.2.2e. above are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
1) LWer core region f rom 0 to 15%, inclusive.
2) Upper core region f rom 85 to 100%, inclusive, O 3) Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%,
and 74.9 1 2%, inclusive, and
4) Core plane regions within i 2% of core heignt [1 2.88 inches]
about the bank demand position of the Bank "0" control rods.
g. WithFxhexceedingFxytheeffectsofFxyonFQ(Z)shallbeevaluated to determine if FQ(Z) is within its limits.
4.? 2.3 When QF (Z) is measured for other than Fxy determinations, an overall measured FQ(2) shall be obtained f rom a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
!' O d
V0GTLE UNITS - 1 & 2 3/4 2-7a l
. ~ - . . _ . - - - - - - . - . - - . - . - . - . -
4 i
DOWER 01STRIBUT10N LIMITS ,
3/4.2.5 DNB PARAMETERS LIMITING 00N0!T10N FOR OPERATION 3.2.5 The following DNB-related parameters $ ball be maintained within the limits: .
a. Reactor Coolant System Tav (T1-0412, TI-0422, 11-0432, TI-0442),
5592.5'F(Unit 1)or559$'F(Unit 2). l
b. Pressurizer Pressure (PI-0455A,B&C, PI-0436 & PI-0456A, PI-0457 &
PI-0457A, PI-0458 & PI-0458A),12199 psig* (Unit 1) or 2224 psig*
(Unit 2).
c. Reactor Coolant System Flow (F1-0414, F1-041',, F1-0416, F1-0424, F1-0425, F1-0426, F1-0434, F1-0435 F1-0436. F1-0444, F1-0445, .
F1-0446) 1391,225 gpm** (Unit 1) or 396,19e gpm" (Unit 2). l APPllCABILITY: MODE 1.
ACTION:
With any of the'above parameters exceeding its limit, restore the parameter to O within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . .
SURVElllANCE RE0VIREMENTS 4.2.5.1 Reactor Coolant System Tavg and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . RCS flow rate shall be monitored for degradation at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In the event of -flow degradation, RCS flow rate vall be determined by precision heat balance within 7 days of detection of-flow degradation.
4.2.5.2 The RC$ flow rate indicators shall be subjected to CHANNEL-CAllBRATION at each fuel loading and at least once per 18 months.
4.2.5.3 After each fuel loading, the RCS flow rate shall be determined by '
precision heat balance prior to operation above 75% RATED THERMAL POWER. -The RCS flow rate shall also be determined by precision heat balance at least once per 18 months. Within 7 days prior-to per-forming the precision heat balance flow measurement, the instrument-ation used for performing the precision heat balance shall be calibrated. The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
4 Limit not applicable during either a THERMAL POWER ramp in excess of 5% of O- RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
" Includes a 2.2% (Unit 1) or 3.5% (Unit 2) flow measurement uncertainty. l V0GTLE UNITS 1&2 3/4 2-13
_ . _ ~ . _ ~ _ _ _ . _ . _ _ - _ . . _ _ _ _ . . - _ _
O O
,(Continuedl O
< TA_BtE 3..
ES
d ENGINEERED SAFETY FEA10RES ACTUATION SYSTEM INSlRilMLNI AIlON TRIP Sell'OINIS m
101AL SEN50R 3 ALLOWANCE ERROR q IRIP SETFOINI All0WABIE VAIUt
FUNCil0NAL UNIT (TA) Z _{Sl_
~7. Semi-Automatic Switchover to I '* Contain.nent Emergency Sump (Continued) ru
b. RWSI Level- -Low -tow 3.5 0.71 1.67 2 275.3 in. from 2 264.9 in. from Coincident With Safety tank base tank base Injection (2 39.1% of (2 31.4% of (LI-0990A&B, LI-0991A&B, irstrument instrument span) span)
Ll-0992A, L1-0993A) 1
8. loss of Power to 4.16 kV ESF Bus J. 4.16 kV ESF Bus N.A. N.A. N.A. 2 2975 volts ? 2912 volts Undervoltage-Loss of Voltage with a 5 0.8 with a 5 0.8
, second time second time g delay. delay.
Y U b. 4.16 kV LSF Bus N.A. N.A. N.A. 2 3146 volts _> 30H3 volts Undervo ltage -Deg raded with a 5 20 with a $ 20 second time second time Voltage 'c la y .
delay.
9. Engineered Safety Features Actuation System Interlocks N.A. N.A. $ 2000 psig $ 2010 psig
a. Pr essurizer Pressure P-11 N.A.
(Unit I)
(PI 0455A B&C, PI-0456 & (Unit 1)
PI-0456A, PI-0457 & PI-0451A, $ 1970 psig $ 1980 psig P1 04W L "i -0458 A), (finit 2) (llnit 2)
N.A. N.A. N.A. N.A N.A.
': Reactor Irip, P-4 L
v 3/4.5 EufRGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATCRS l
LIMITING CONDITION FCR OPERATICN l
3.5.1 Eacn Reactor Coolant System (RCS) accumulator snail te OPERABLE with: I
a. The isolatich valve ocen, b, Unit 1: A containec boratea water volume Of tetween 6555 (29.2% of instrument scan) and 6909 gallons (70.7% cf instrument scan > (LI-0950.
LI-0951, LI-0952 L!-0953, LI-0954, LI-0955, I-0956. LI-0957),
Unit 2: - A containea corated water volume Of terween 6616 ( 36% of instrument scan) and 6854 gallons (64% of instrument scan) ( LI ')950.
LI-0951, LI-0952, LI-0953, LI-0954, LI-0955, LI-0956. LI-0957).
c. A boron concentration of between 1900 com ana 2600 Opm, s.v,
d. A nitrogen cover-cressure of between 617 and 678 Osig, (P'-0960A&B, PI-0961 ALB, PI-0962A&B, PI-0963ALB, PI-0964A&B, P!-096 !A&8,',
PI-0966ALB, PI-0967A&B)
O Q APPLICABillTY: . MODES 1, 2, and 3*
ACTION:
a. With one accumulator inoperable, e.: cept as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b, With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
a, At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1)
Verifying the contained corated water volume and nitrogen cover-pressure in the tanr,s, and
2) verifying that eacn accumulator isolation valve is coen (HV-8808A, S, C, 0).
' Pressurizer pressure above 1000 psig.
V0GTI.E UNITS - 1 & 2 3/4 5-1
l BORON INJECT 10N SYSTEM 3/4.5.4 REFUELING WATER STORAGE 'ANr LIMITING CONDITION FOR OPERATION 4
i 3.5.4 The refueling water storage tant (RHST) snall be OPERABLE sith:
a. A minimum contained borated water volume of 631,478 gallens (867. of i instrument span) (LI-0990ALB, LI-0991A&B, LI-0992A, LI 0993A). !
l
b. A boron concentration of between 2400 ppm and 2600 com of boron, j i
c. A minimum solution temDerature of 44*F (Unit 1) or $4*F (Unit 2), and J
d. A maximum solution temDerature of Il6*F (TI-10982).
9. RHST SludQe Mixing Fum0 Isolation valves capable of closing on RHST low-level.
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:
i a. With the RHST inoperable except for the Sludge Mixing Pume Isolation Valves, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -
b. With a Sludge Mixing Pump Isolation V61ve(s)-inoperable, restore the valve (s) to OPERABLE status within 24 . tours or isolate the sludge mixing
-system by either closing the manual 1sclation valves or deenergizina the OPERABLE solenold pilc: valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and maintain closed.
SURVE!LLANCE REQUIREMENTS 4.5.4 The RHST shall be demonstrated OPERABLE:
a. At least once per-7 days by:
1) Verifying the contained borated water volume in the tank, and
2) Verifying the Doron concentration of the nater.
b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RHST temperature wnen the outside air temperature is less than 40*F (Unit 1) or 50*F (Unit 2).
c. At least once per 18 months by verifying that the sludge mixing pumo isolation valves automatically close upon an RHST low-level test signal.
V0GTLE UNITS - 1 & 2 3/4 S 10
THIS PAGE APPLICABLE 10 UNIT 1 ONLY 1/4.? POWER OlSTRIBUTION LIMITS - UNIT I BASES The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) meeting the DNB design criterion during normal operation and in short-term transients, and (2) 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 1 events provices assurance that the ir.itial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking facters as used in these specifications are as follows:
FQ(Z) Heat Flux Hot Channel Factor is defined as the maximum local heat flux on the surface of a fuel rod at core elevation 2 divided by the average fuel rod heat flux, allowing for manuf acturing tolerances on fuel pellets and rods; and l
O' F$H Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power, 3/4,2,1 AX1AL FlVX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFO) assure that the Fg(Z) upper bound envelope of the Fg limit specified in the CORE OPERATING LIM 11S REPORT event of xenon (redistribution following power changes,(COLR) times K Z) is Provisions for monitoring the AFD on an automatic basis are derived f rom '
the plant process computer throagh the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs W provides an alarm message immediately if the AFD for two or more OFEnABLE excore channels are outside the allowed of power operating space for RAOC operation specified within the COLR and the THERMAL POWER is greater than
, 50% of RATED THERMAL POWER, EQ V0G1LE UN115 - 1 & 2 83/42-1 l
- . - _. .- - -- -_- .----..- - -. . - ~ _ - - . - _ . - . _ - _ . . _ - - -
I M
11115 PAGE APPLICABLE 10 UNIT 1 ONLY E0WER DISTolBUTION LIMITS UNIT 1 BASES
~
AX1AL Ft.UX OIFFERENCE (Continued) 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALP tiOTCHANNELFACTOR-F$H The limits on heat flux hot channel f actor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limit on peu Iccal power density is not exceeded, (2) the DNB design criterion is met, and (3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS accephnce criteria limit.
Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
a. Control rods in a single group move together with no inidivdual rod insertion differing by more than i 12 steps, indicated, from the group demand position;
b. Control rod groups are sequenced with a constant tip-to-tip distance between banks as described in Specification 3.1.3,6;
c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F$HwillbemaintainedwithinitslimitsprovidedConditionsa,thr.ough
d. above are maintained. TherelaxationofF$HasafunctionofTHERMALPOWER allows changes in the r6Jial power shape for all permissible rod insertion limits.
When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a f ull core map taken v :th the incore. detector flux mapping system and a ,
3% allowance is appropriate for manufacturing tolerance.
The heat flux hot channel factor F Q(2) is measured periodically and increased by a cycle and height dependent power factor appropriate to RA00 operation, W(2), to provide assurance that the limit on the heat flux hot channel factor, Fn (2), is met. W(2) accounts for the effects of normal operation transients within the AFD band and was determined f rom expected
'.. power control maneuvers over the full range of burnup conditions in the core. The W(2) f unction for normal operation and the AFD band are provided in the CORE OPERATING LIM 115 REPORT per Specification 5.8.1.6.
V0GTLE UNITS - 1 & 2 8 3/4 2-2
-a.%gse,e ,-:----m -
a- 3 N 971e w'm--- + d--'--
THl5 PAGE APPLICABLE 10 UNii 1 ONLY PtNER 015TRIBUT10N LIMITS - UNIT 1
, BASES HEAT Fl.UX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
WhenF$Humeasured,(i.e., inferred),measurementuncertainty(i.e..,
the appropriate uncertainty on the incore-inf erred hot rod peaking f actor) must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system.
-F4.2.4 OVADRANT PORTILT RATIO ,
The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
.gQ The limit of 1.02, at which corrective action is required, provides ONB
(/ and linear heat generation rate protection with x-y plane power tilts. A limit r,f 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hoor 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 control rod. In the event such action does not
c. rect the tilt, the mergin for uncertainty on Fn is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER Tit.T RATIO wnen one excore detector is inoperable, the moveable ir. core detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANI POWER TILT RATIO. The incore detector n.onitoring it done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C -B , E-5, E > 1 l , H-3, 4-13, L-5, L-11, N-8.
3/4.2.5 DNB PARAMETE_R.S 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 i- meet the DNB design criterion throughout each analyzed transient. The indicated lavg value of 592.5'F and the indicated pressurizer pressure value of 2199 psig correspond to analytical limits of 594.4'F and 2185 psig respec-t hely, with allowance for measurement uncertainty.
V0GTLE UNil5 - 1 & 2 8 3/4 2 3
( THIS PAGE APPLICABLE 10 UNIT 1 DNLY ,
POWER DISTRIBUTION 1.lMITS - UN11 1 BASES 1
3/4.2.5 ONB PARAMETERS (Continued)
The 12-hour periodic surveillance of these parameters through instrument readout is suf ficient 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 suf ficitnt verification of the flow rate degradation on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis. A change in indicated percent flow which is greater than the instrument channel inaccuracies and parallax errors is an appropriete indication of RCS flow degradation.
O t
EO ff0GTLE UNITS - ) &2 [t 3/4 p.4
. _ _ _ _ . _ . , . . _ . , , - . . _ . - . . _ . . . _ . . _ - . . _ . _ _ _ . _ . . - _ _ . _ . ~ . . _ _ _ . . _ _ . . . _ _ _ _ _ _ _ _
l l
THIS PAGE APPLICABLE TO UNIT 2 ONLY
_3 /4 . 2 POWER OISTRIBUTION LIMITS - UNIT 2 BASES The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and 11 (Incidents of Moderate Frequency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) 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 Coridition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
FQ (Z) :leat 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 f uel rod heat flux, allowing for manuf acturing *olerances on O fuel pellets and rods; V N FaH Nuclear Enthalpy Rise Hot Channel Factor. 4: defined as the ratio of the integral of linear pov:r along the rod with the highest integrated power to the average rod power; and Fxy(Z) Radial Peaking Factor, is defined as the ratio cf peak power density to average power density'in the horizontal plane at core elevation Z.
3/4.2.1 AX1AL FLUX 01FFERENCE The limits on AXIAL FLUX O!FFERENCE (AFD) assure that the Fg (Z) upper bound envelope of the FQ limit specified in the CORE OPERATING LIMITS REPORT (COLR) times K(2 event of xenon re) distribution following power changes.is not exceeded during eith Target flux difference is determined at equilibrium xenon conditons.
The rods may be positioned within the core in accordance with their retpective 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 l' obtained under these conditions divided by the f raction of RATED THERMAL POWER j is the target flux difference at RATED THERMAL POWER for the associated core
i. burnup conditions. Target flux dif ferences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate f ractional THF.RMAL POWER level. The periodic updating of the target flux dif ference value is necessary to reflect core burnup considerations.
1 u.)
V0GTLE UNITS - 1 & 2 8 3/4 2-5
3 (Q THIS PAGE APPLICABLE TO UNIT 2 ONLY POWER DISTRIBUTION LIMITS - UNIT 2 BASES AXIAL FLUX O!FFERENCE (Continued)
Although it is intended that the plant will be operated with the AFD within .he target band required by Specification 3.2.1 about the target flux dif f er .nce, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of sne target band at reduced THERMAL POWER levels. This deviction will no', af fect the xenon redistribution suf fi-ciently to change the envelope of peaking f actors which may be reached on a subsequent return to RATED THERMAL P0,'ER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits specified in the COLR while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.
For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of
? hours actual time reflects this reduced significance.
Provisions for monitoring the AF0 on an automatic basis are derived from A the plant process computer through the AFD Monitor Alarm. The computer deter-Q mines the 1-minut average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.
Figure B 3/4 2-1 shows a typical monthly target band.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOTCHANNELFACTOR-F$H The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel f actor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.
l Each of these is measurable but will normally only be determined J periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic l
surveillance is suf ficient to ensure that the lim;n are maintainea provided:
a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from the
~
group demand position;
b. Control rod groups are sequenced with a constant tip-to-tip distance be . een banks as described in Specifica tior 3.1.3.6;
{
l- V0 GILE UNITS - 1 & 2 8 3/4 2-6
n THIS PAGE APPLICABLE TO UNIT 2 ONLY l 1.00 , y 1
0.90 l I
i 1
0.80 j i
1 I
0.70 e I
) Target Flux 0 g r Difference j 0.60 l ,/
! le w I 0.50 I
(~']N
<- y 4 i
I O 0.40 l 5 I '
1 5
< l E 0.30 p-I I
0.20 - --
-# I
\
l '
O.10 =
j l
1
' I t 0
30 20 10 0 +10 +20 +30 INDICATED AXI AL FLUX DIFFtiRENCE (percent) i g
(sl FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THER.'1AL POWER l
V0GTLE UNITS 1 & ' B 3/4 2-7
THl$ PAGE APPLICABLE TO UNIT 2 ONLY POWER DISTRIBUTION LIMITS - UN!12 BASES HEAT FEUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c, The control rod insertion limits of Specifications 3.1.3.5 and
, 3.1.3.6 are maintainea; and
d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F!H will be maintained within its limits provided Conditions a through
d. above are maintained. TherelaxationofF$HasafunctionofTHERMALPOWER allows Changes in the radial power shape for all permissible rod insertion limits.
' . When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerante must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.
WhenF$Hismeasured.(i.e., inferred),measurementuncertainty(i.e.,
the copropriate uncertainty on the incore inferred hot rod peaking factor) must be allowed for and 4% is the appropriate allowance for a f ull core map taken with the incore detection system.
Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties. This margin includes the following:
a. Design limit DNBR of 1.30 vs 1.28,
b. Grid Spacing (Ks) of 0.046 vs 0.059,
c. Thermal Diffusion Coefficient of 0.038 vs 0.059,
d. DNBR Multiplier of 0.86 vs 0.88, and
e. Pitch reduction.
The applicable values of rod bow penalties are ref erenced in the FSAR.
O V0GTLE UNITS - 1 & 2 B 3/4 2-8
i THIS PAGE APPLICABLE TO UNIT 2 ONLY
$ POWIR OISTRIBUTION tIMITS - UNIT 2 j BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR (NTHALPY RISE HOT CHANNEL FACTOR
(Continued) ,
The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor,Q F (Z), remains within its limit. The Fxy limitforRATEDTHERMALPOWER(Fxf)asspecifiedintheCOLRperSpecifi-cation 6.8.1.6 was determined from expected power control maneuvers over the f ull range of burnup conditions in the core.
3/a 2.4 OVADRANT POWER TILT RATIO The OVADRANT POWER I!LT RAl!0 limit assures that the radial power distribu-tion satisfies the_ design values used in the power capability analysis.
4 Radial power distribution measurements are made during SfARTUP testing and periodically during power operation.
The limit of.l.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x-y plane power tilts, A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indi:ated power tilt.
The 2-hour time allowance for operation with a tilt condition greater -
than-l.02 but less than 1.09 is provided to allow identification and correction -
of a dropped or misaligned control rod. in the event such action does not correct the tilt, the margin for uncertainty on F0 is reinstated by reducing i the maximum allowed power by 3% for each percent of tilt in excess of 1.
'For purposes of monitoring QUADRAN1 POWER i!LT RATIO when one excore detecter is.. inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four synenetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll , H-3, H-13. L-$ . L-ll , N-8.
3/4.2.5 DNB PARAMETERS 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 O maintain a minimum DNBR of 1.30 throughout each analyzed transient. The indicated Tavg value of $91'F and the indicated pressurizer-pressure value of 2224 psig correspond to analytical limits of $92.5'F and 220$ psig respec-tively, with allowance for measurement uncertainty, V0 GILE UNITS - 1 & 2 B 3/4 2-9
THIS PAGE APPLICABLE 10 UNIT 2 ONLY POWER O!STRIBUTION LIMITS - UNIT 2 BASE $
i 3/4.2.5 DNB PARAMETERS (Continued)
The 12-hour periodic surveillance of these parameters through 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 suf ficient verification of the flow rate degradation on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis. A change in indicated percent flow which is greater than the instrument channel inaccuracies and parallax errors is an appropriate indication of RCS flow degradation, O
V0GTLE UNITS - 1 & 2 83/42-10
. _ _ _ _ _ _ _ _ . . - . ._ .m _ .__ _ _ _ . ._ _ . _ _ _ _ _ _ _ _ _ _ .
4 3/4.4 REACTOR CC3 TANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation ar.d meet the ONB design criterion during all normal operations and l 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
In MODE 3, two reactor coolant loups provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers, in MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR train provides suf ficient neat remeval capability for removing decay heat; but single failure considerations require that at least two trains / loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR trains be OPERABLE. The locking closed of the required valves in Mnde 5 (with the loops not filled)
\ precludes the possibility of uncontrolltJ boron dilution of the filled portion of the Reactor Coolant System. These actions prevent flow to the RCS of unborated water in excess of that analyzed. These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.
The operation of one reactor coolant pump (RCP) 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 an RCP with one or more RCS cold legs less than or equal to 350'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant 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 RCP5 to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures, in MODE 4 the' starting of an RCP, when no other RCP is operating, is restricted to a range of temperatures that are consistent with analysis assumptions used to demonstrate that the RHR design pressure is not exceeded when RHR relief valves are used for RCS overpressure protection.
f.
!U l
l V0GTLE UNITS - 1 & 2 B 3/4 4-1
EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSNSTEMS (Continuec)
The limitation for all safety injection pumps to be inocerable below 350'F orovides assurance that a mass 80dition pressure transient can De relievec ey the operation of" a single FORV.
The Surveillance Requirements provided to ensure OPERABILITY of eaCn
omponent ensure that at a minimum, tne assumotions usea in the iafety analyses are met and that suosystem OPERABILITY is maintained. Surveillance Requirements for tnrottle valve cosition steos and flow balance testing orovide assurance that crocer ECCS flows will ce maintained in the event ;f a LOCA.
Maintenance of crocer flow resistance ano cressure droo in tne cloing system to each injection Dolnt is necessary to; (I) prevent total pumo flow from exceeding runcut conditions when the system is in its minimum resistance configuration, (2) provide the proper flow spilt between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, (3) provide an acceptaDie level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow which is directed through the seal injection path is less than or equal to the amount assumed in-the safety analysis. The surveillance requirements for leakage testing of ECCS check valves ensure a failure of one valve will not cause an intersystem LOCA, In MODE 3, with either HV-8809A or B closed for ECCS check valve leak testing, acequate ECCS flow for core cooling in the event of a LOCA is assured.
3/4.5,4 REFUELING WATER. STORAGE TANK The OPERABILITY cf 'Be Refueling Water Storage Tank (RHST) as part of the ECCS ensures that suff ant negative reactivity is injected into the core to counteract ey positive increase in reactivity caused by RCS cooldown, RCS cooldown can be causeo by inadvertent depressurization, a loss-of-coolant
. accident, or a steam line rupture.
The limits on RHST minimum volume and boron concentration ensure that
1) sufficient water is available within containment to permit recirculation cooling flow to the core, 2) the reactor will remain succritical in the cold condition following a small LOCA or steamline break, assuming comolete mixing of the RHST, RCS, and ECCS water volumes with all control roos inserted except the most reactive control assembly (ARI-1), and 3) the reactor will remain subcritical in the cold condition following a large break LOCA assuming complete mixing of the RHST, RCS, ECCS water and other sources of water that may eventually reside in the sumo, post-LOCA with all control rods assumed to be out (Unit 2) or all control rods inserted except for the two p most reactive control assemolies (Unit 1).
The contained water volume limit includes an allowance for water not usable cecause of tank 31scharge lire location or other physical characteristics.
YOGTLE UNITS - 1 S2 3 3/4 5-2
I ADMINISTRAT!vE CONTROLS O , g NNUAL PA010 ACTIVE EFFLUENT RELEASE REe0RT (Continued)
The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous ef fluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3 3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Soecification 3.11.1.4 or 3.11.2.6, respectively.
MONTHLY OPERATING REPORTS 6.8.1.$ Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Of fice of Resource Management, U.S. Nuclear Regulatory Commission. Washington 0.C. 20$56, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the l$th of each month following the calendar month covered by the report.
I CORE OPERATING llMITS REPORT - UNIT 1
6. 8.1. 6 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:
a. SHUTDOWN MARGIN LIMll FOR MODES I and 2 f or Specification 3/4.1.1.1,
b. SHU100WN MARGIN LIMITS FOR MODES 3, 4, and $ for Specification 3/4.1.1.2,
c. Moderator temperature coef ficient BOL and EOL limits and the 300-ppm surveillance limit f or Specification 3/4.1.1.3,
d. Shutdown Rod insertion Limit for Specification 3/4.1.3.5,
e. Control Rod Insertion Limits for Specification 3/4.1.3.6,
f. Axial Flux Difference Limits for Specification 3/4.2.1, l
g. Heat Flux Hot Channel Factor, K(Z) and W(2) for Specification 3/4.2.2,
h. Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.
The analytical methods used to determine the core operating limits shal; be those previously approved by the NRC in:
O
-V0GTLE UNITS - 1 & 2 6-21
fi '
AQMl_NI$7RATIVE CONBOL -
-.=
CORE OPf RATING LEESJfPP.O.R.I (Continued) - UNIT I l
a. WCAP-9272-P-A, #WESitNGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY."
1 July 1985 (W Proprietary).
(Methodo?ogy for Specification 3.1.1.3 - Moderator Tempa' 'ure Coeff$t*ent 3.1.3.$ - Shutdown Bank Insertion Limit. 3. .6 -
Control Bank insertion Limits, and 3.2.3 - Nuclear Entha. Rise l ,
Hot Chafine) Factor.) '
i A WCAP-10216 + A. " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F Q SURVEILLANCl TECHNICAL SPECIFICATION," June 1983 (W Proprietary).
, (Methodology f or Specifications 3.2.1 - Axial Flux Of f ference (Relaxed Asial Offset Cohtrol) and 3.2.2 - Heat Flun Hot Channel Factor (W(1) surveillance reovirements for FQ Methodology).)
c. WCAP-9220 P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION M00EL-1981 VERSION," February 1962 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g., f uel therinal-mechanical limits, core thernal-hydraulic limits, ECCS liraits, nuclear limits such as shutdown margin, and transient and accident anolysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS RE.90RT, including any aid-cycle revisions or supplements thereto, shall be provided upon issuance, for_each reload cycle, to the NRC Document Control Oesk with copies to the Regional Administrator and Resident inspector.
CORE OPERATING LIMITS REPORT - UNIT 2 6.8.1.6 -Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part
-of a reload-cycle for the following:
a. SHUTOOWN MARGIN limit for MODES 1 and 2 for Specification 3/4.1.1.1,
b. SHUTOOWN MARGIN limits for MODES 3, 4, and 5 for Specification 3/4.1.1.2,
c. Moderator temperature coef ficient BOL and EOL limits and the 300-ppm surveillance limit for Specification 3/4.1.1.3,
d. Shutdown Rod Insertion Limits for Specification 3/4.1.3.5,
e. Control Rod Insertion Limits for Specification.3/4.1.3.6,
f. A'xial. Flux. Difference Limits, and target band for Specification-3/4.2.1,
~
g. Heat-Flux Hot Channel Factor, K(Z), the_Fower Factor Multiplier and F[forSpecification3/4.2.2,
h. Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.
The analytical methods used to determine the core operating limits shall be those previously approved by the NRC in:
'l
,_ V0GTLE UNITS - 1 & 2' 6-21a lr c______.-___--
l AJMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) - UNIT 2 l
a. WCAP-9272-P-A,
WESTINGHOUSE RELOAD SAFETY EVALUAl!ON MElH000 LOGY," .
July 1985 (W Proprietary). !
(Methodology for Specification 3.1.1.3 - Moderator T- .erature Coef ficient 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 -
Control Bank ;rasertion Limits, 3.2.1 - Axial Flux Dit f erence, 3.2.2
- Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot C^annel Factor.)
b. WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES
- TOPICAL REPORT " September 1974 (W Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference
[ Constant Axial Offset Control).)
l c. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC)
January 31, 1980 --
Attachment:
Operation and Safety Analysis Aspects .
of an improved Load Follow Package.
(Methodology f or Specification 3.2.1 - Axial Flux Dif f erence
[ Constant Axial Offset Control].)
d. NUREG-0800 Standard Review Plan, U. S. Nuclear Regulatory Conynission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouw Constant Axial Of f set Control (CAOC), Rev.2, July 1981.
( -
(Methodology for Specification 3.2.1 - Axial Flu:: Difference
-V- (Constant Axial Offset Control).)
.i e. WCAP-9220-P-A, Rev. 1. " WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION, February 1982 (W Proprietary).
(Methodology f or Specification 3.2.2 - Heat Flux Hot Channel Factor.)
1 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met, lhe CORE OPERATING LIM 115 REPORI, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector. ,
i SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Of fice of the NRC within the time period specified for each report.
I l-LA V0GTLE UNITS. - 1 & 2 6-21b l
l l
n v
1 Attachment 2g Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE 5 Fuel Design Technical Specifications Marked-Vo Paoes i Effective following the Vogtle 2 Cycle 2 Shutdown !
(Effective as of Vogtle 2 Cycle 3 Startup) '
!O .
O
!N0tt tartty Llutt5 AND t!uf ttNG SaptTv tysttu ![tt1NGs b\
V una m -
)
2,1 !artty ilutt$
1 2.1.1 REAC10e CORI. . .. . . .. 21 1 2.1.2 #EACTOR COOLANT SYSitM PRI55URL. . .. ... 2 -1 I FIGURL 2.1 1 REAC10R CORI 5 Aft 1Y limit:M d. . . 22 URL 14RIACTORNORE5 Aft 1YIT (Uhli 2). .. . . 2 2a 2.2 tiuftfNGSaittvSv57tMStTitNG5 2.2.1 dtACTOR TRIP $YS1tM INSTRUMENTATION Sitt0lN?$.... ... .. 2-3 TABLE 2M-1 "T T In.V 5tfPOINi$
. l( W .'OR TRIP .
... SY5itM
.... . . .IN$1RUMENTA 10N
. . 24 l
. Lt 2.. a REACTO R!P $YS IN51RUML Atl0N TRIP . TP0lNTS 2-12 (UN 2 ... .... .... .
g ... . .
l nast$
5tCTION O 2.1 5AttTv tIMITS 2.1.1 RtACTOR CORE., .. ......... .. ............... ... ........ B 2-1 2.1.2 RCACTOR COOLANT 5751tM PRt55URL... ... .... . ....... .. B22 limit 1NG 5AftTV SY$ttM SITTING $
2.2.1 RE ACTOR 1 RIP . titM IN5tRUMENT Afl0N SLTPOINT5. . . .. . . . . . B23 v0GtLI UNITS - 1 & 2 !!!
v .
IN.211 t!=1tf NG CONDI?!?N5 80R crf # ATION HD RevittuNtt t!Dy.LatutNTS
ttCTitN J C hl
}/4 2 00wto Oi. 3 1eutt0N !!alTJ 3/4.2.1 AtlAL FLUA DIFFERENCE . . . 3/4 2 1
%.1 4A k tut 0175 M Ntt (uni k .
.A. .
.A3/4hak 3/4.2.2 MEAT Flux Mot CMANNEL F ACTOR - Fn(Z). ... 3/4 2 3 3/4.2 3 NUCL(AR [NIMALPY RI$t MOT (MANNtt FACTOR F$11.. . 3/4 2-0 3/4.2.4 OVA 0 RANT P0wtR TILT RATIO. . .. . . . .... .. 3/4 2 10 3/4.2.5 ONB PARAMtitR$. ... ... 3/4 2-13 3/4.3 INsf oUME NT AflQ3 3/4.3.1 REACTOR 1 RIP $Y$ftM IN$1RUMINIATION. . .. . 2/4 3 1 TABLE 3.3 1 # TACTOR TRIP $v5 TIM INSTRUMENTATION. . 3/4 3-2 TABLI 4.3-1 REACTOR TRIP SY$1tM IN$TRUMENTATION $URVi!LLANCE Rt0VIREMENTS. . . .. . . .. .. 3/4 3-9 3/4.3.2 (NGlNttRt0 $AftTY FIATURt$ ACTUAT!0N $YSTEM IN$TRUMENTATION, .. .......... .. . . .. ... 3/4 3 15 TABLE 3.3-2 (NGINttRED $AftTY FtATUR[$ ACTUATION SYSTIM INSTRUMENTATION.................................... ..... 3/4 3-1T TABLE 3.3-3 [NGINitRED $AFETY FI ATURL$ ACTUATION $YST[M g INSTRUMENTAtl0N TRIP $lTP0lNTS...... ... ............ ... 3/4 3 2B TABLE 4.3 2 [NGINt!Rt0 $AFETY FIATURt$ ACTUATION $Y$ TEM INSTRUMENTATION $URVEILLANCE Rt0UIREMENT1.. . . ..... 3/4 3 36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations. . .. 3/4 3-45 TABLt 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERAtl0N1.................... .. . . . 3/4 3 46 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR P(ANT OPERATIONS $URVilLLANCE RIOUIREMENTS.... .. .. ... . . 3/4 3 4B Movable Incore Detectors......... .................. .. 3/4 3 49 Seismic Instrumentation (COMMON SYST[M) ... .. . .... .. 3/4 3 $0 1
0 V0GTLt UNil$ - 1 & 2 V .
. . _ _ _ . _ . _ . - _ . . m__ _ _ _ . _ . _ _ - . _ _ _ _ . _ _ _ _ - . . . . _
Mtt n
As.!$
I V
StCt:0N g 3/4.0 A8pt1CAtttfiv. , , . . . i 3/4 0.'
3/4.1 8tattivity CONtt0L systtus 3/4.1.1 60RAfl0N CONTROL. . . i 3/4 1-*
3/4.1.2 BORA110N SYSitF5.. . . B 3/4 12 3/4.1.3 MOVABLt CONTROL A55tMBLil5.. . B 3/4 1 3 3/4.2 P M R 015ftleuff0N t1utts. . . . .. . t 3/4 2 1 3/4.2.1 AxlAl. FLUX O!FFERtNCE h .. j',.. ... . .. . . . B 3/4 21 3/4.2.2 and 3/4.2.3 HEAT Flux Hot CHANNEL FACTOR and NU* LIAR IN1HALPY A!5[ MOT CHANNEL FACTOR F$g}: '] . B 3/4 2 2 B 3/4 23 3/4.2.4 OUADRANIPowtRTILTRAflo{i.irj . . . . . ....... .
3/4.2.5 DN8 PARAMti!R5 ,ii M. . . ... . . . ... . 8 3/4 '3
/4.2.1 .. .. . ..... ... 8 3/4 2 5 AA1 AL F%Dif f tRtNCE (UNil 2) .. . ..
3/4. 2 and 3/4.2.3 Flux Hof CHAhNEL FACTOR NUCLLAR
[]
(j ' NIHALPY R!st NOT $ NNtLFACTOR-F$H(UNIT . .. .. , 8 3/4 -6 .
I 3/4.2.4 OUAD NT P0 Win TILT RAll ' Nil 2) . . . .. ...
. $ 3/4 2-9 l 3/4.2.5 DN8 PARAM R$ (UNIT 2),. . .... ... . .. . . .. 6 3/4 2 9 I
- s a 3/4.3 TNSTRuutNTAft0N 3/4.3.1 and 3/4.3.2 REACTOR TRIP SY5itM and INGINEERED SAFtTY FLATURES ACTUA110N SY5ftM INSTRUMENTAi!0N...... ... . .. 8 3/4 3-1 3/4.3.3 MONITORING IN51RUMENTAT10N......... ....... .... . ... . B 3/4 3 3 3/4.3.4 TURBINE OvtR5 Pit 0 PR0ftti!ON., ....... ... ..... ... , . B 3/4 3 6 O
! v0G1'I UNlts - 1 52 Av
~ ' ' * ,wn-e , ., , ,
INDE!
atutNtstantivt 00Niects
,~~s 6
i
' [(TION gag 6.a.2 SAFt1Y Rivi[W BOARD (188)
Function. . . . ... . . . . ... .. . . .... . 6-9 Compolition. . .. . . ... .. .. . . . . . . 6-10 Alterratel... .... . . .... . ...... . . . . . . 6 10 Conlultantl........ . . ... ...... . . .. .. . .. 6-10 Meeting Frecuency....... . ...... .. ... .. ... . . 6-10 Overum. .. .. ... .. .. . .. . . . . .. . 6-10 Review, ..... . .. ,, . ..... .. . .. . . . . 6511 Audit 1... . . . . . .. .. .. .. . 6-11 Records. . ..., ...... .... .. ... ..... . ...... ... 6-12 6.6 8tPORTAett tytNT ACTION.,......., , . ......, . .... , , . 6-13 6.6 S Af TTY ' t u t t vt0t AT104 . . . .. ... . ... ... .. 6-13 6.7 ##0Ct0V#ts AND PROGRAMS. ............. ......, . .. ...... 6-13 6.8 RtPORTING pt00tRtutNT5 6.8.1 ROUflNL REPORT 5.......... ...... . ...... . . . ............ 6-17 5ta rtup Report . . . . . .. .......... ..... ............... ... 6-17 s
v,
-~s) .
Annual Report............................................... 6-17 Annual Radiological Invironmental Surveillance Report........ 6-18
$emiannual Radioactive [ffluent Release Report............... 6-19 Monthly Operating Reports.. .......... ............ ......... 6 21 Core Operating Limits Report igNi; iib.................... 6-21
(:;r: c::::ta;.; .:;::. .a.,t testi r,.
___ :J 6.8.2 SPECIAL REP 0RTS..................... .... ................... 6- *l 6 22 69 #ECORD #tTENT10N....................... ....................... i i l l 6-21 l l (O) v0GTLC UNITS - 1 & 2 xx!!! v
e 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER (NI-0041, NI-0042, NI-0043, NI-0044),
pressurizer pressure (PI-0455A, BLC, PI-0456 L PI-0456A, PI-0457 L PI-0457A, PI-0458 L PI-0458A), and the highest operating loop coolant temperature (T,,,) T T .na a M ehall M* axceed the limits shown in (TI-0412,TI-y',TI-04U Figure 2.1-1 eUC .> riwurej.1-;a (Uoii 23 l APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating 1000 average temperature and THERMAL POWER has exceeded the appropriate pressurtzer pressure line, be in HOT STANOBY within I hour, and comply with the require-ments of Specification 6.6.1. REACTOR COOLANT SYSTEM PRESSURE 2.I ? The Reactor Coolant System oressure (PI-0408, PI-0418, PI-0428, PI-0438) Im shall act eaceed P 35 psig. . APPLICAt!!LITY: M00E$ 1, 2, 5, ,, and 5. (CTION: MODES I and 2: Whenever the Reactor Coolant System p* essure has exceeded 2735 Mig, be in HOT STANDEY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.6.1. MODES 3, 4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.1.
'Hnt'*e spectftc instrument numbers are provided in parentheses they are for information only, and apply to each unit unless specifically noted (to assist In ldertifying associated instrument channels or looos) and.are not intenced to limit the requirements to the specific instruments associated with the numoer.
VCGTL. UNITS - 1 & 2 2-1 l 1
APPL;;ASLE TO-UNIT 4 OkY lh5 e 670 - l
~j UNACCEPTABLE OPERATION 660 A
% 2440 psia N A '
650 2250 psia u. o 340 A, C 3 630 2000 psia T I
* % i ! 620 N. N - ' 'm N3 }
i 8-
\ - \ N .
S 610 - g 3
$ 1935 psi 600 A
7 ' k l ACCEPTABLE i OPERATION. l ; 590 1 ! l 580
.5 .6 .7 .8 .9 1.0 1.1 1.2
0. .1 .2 .3 .4 FRACTION OF R ATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIttlTI '2':' ;) l
.~
V0GTLE UNITS - 1 & 2 2-2
p THIS PAGE APPLICABLE TO UNIT 2 ONLY l
> 680 - UNACCEPTABLE OPERATION 60 2400 psia 640 N'N X % /
3
$ 2250 psia ~
2000 psia j "' s s N N ^ g - 1775 psia /
;o 600 X N / \ 7 \/ T ~
O i C 580 , ACCEPTABL OPERATIO/ 560 540 I 0.8 1.0 1.2 0 0 0.4 0.6 FRACTION OF R ATED THERMAL WER
y
\
FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - UNIT 2 I l V0rTLE UN:TS - 1 i 2 ]
i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS O 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trio Systam Instrumentation ar.d Interlock Setpoints shall he set _ consistent with tne Trio Setootnt values shown in Table 2.2-1 )Ud t '4 Gi au; :.:-u m 2t. APPLICABILITY: As shown "w each channel in Table 3.3 1. ACTION:
a. With a Reactor Trio System mstrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value snown in the Allowable Value column of Taol: 2.2-lGx 6 &e . . - - c ;, adjust the Setpoint consistent I dith the Trip Setootnt value,

b. With the Reactor Trio System Instrumentation or Interlock Setootnt less conservative than the value snown in the Allowable Values column of Table 2.2-1 br 'E ; 2. - j, ei tner: l

1. Adjust the Setoolat_ consistent with the Trio Setcoint value of Taole 2.2-1 W ~ W a ::: '9,and determine within 12 hours that l
\ Ecuation 2.2-1 was satisfied for the affected channel, or
2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z+R+51 TA Where: Z - The value from Column 2 of Table 2.2-lhbr 4b4 2.2-jd for the i affected channel, R . The "as measured" value tin percent span) of rack error for the affected channel, S - Either the "as measured" value (in percent span) of the sensor _ error, or the value from Column S (Sensor Error) of Tacle 2.2-1 W 2
, ...- ;jfor the affected channel, ano l TA ._Ihe value from Column TA (Total Allowance) of Tacle 2.2-1 @
pa;.: . :- W for the affected channel. l 0 - V0GTLE UNITS - 1 S 2 2-3 O _ . . _ _ _ _ _
3 O C f!!'!S "?.Ci *: Tilt ,[ TO L%i; I 0%LY
..e . _
lS T ABLE 2.2-1 jt"!!?} _4
r. . .
i REACTOR TRIP'SYSitM INSTRUMENTATION IRIP SEIPOINIS c-2 3 _ 101AL SENSOR
"' ERROR AtLOWANCL
. FUNC l l0NAt - ilNI T (TA) Z (S) IRIP SE1POINI ALLOWABl.L val 0L I
1. Manual Reactor.irip N.A. N.A. .N.A. N.A. N.A.

ns

2. Power Range, Neutron Flux (N1 00418&C, NI-0042B&C.
4 -HI-0043B&C NI-00448&C) t
a. High Setpoint: 7.5 4.56 ,0 1109% of R1Pt $111.3% on RIP #

b. tow-Setpoint 8.3 4.56 f- $25% of.R1P# $27.3% of RTP#

3. Power Range, Neutron Flux, 1.6 0.50 0 $$% of RIP # with 16.3% of RIPJ with liigh Positive Rate a time constant a time constant n, (NI-0041B&C, NI-0042B&C.- 12 seconds' 22 seconds k NI-0043B&C, NI-0044B&C) ,
4
4. . Deleted. ;

5. Intertnediate Range, 17.0 8.41 0 $25% of RIP # $31.1% of HIP #
Neutron Flux ; (NI-0035B NI-00368)
6. Source Range, Neutron Flux '7.0 10.01 0 $105 cps $1.4 x 105 cps (RI-00318, NI-00328)
J. Overtemperature al . 10.7 7.04 1.96 See Note i See Note 2 . (101-4110, 101-421C, ]("e!' ') (r*4
+ 1.17 i'
101 -4 310, 101 -441 C)' ! 113 t H. Overpower-al 4.3 , 1.54 _1.96 See Note 3 See Note 4 [ ( 101 -4118, 101 -4 21 B, jf#a46-44 ((iin n i }I, , 101-4318, 101-4418)
.I #R I P - 'RAllD lill"RMA[ POWER ;
t . . _
ENf , f3 -- (3
%.) (~)/
q,
'--1H15 PAGr PPLICABLE-10-ilN!+ + 0Ht+
b TABLE 2.2-1 (Continued 1 REACTORTRIPSYSTEMINSTRUMENTATipNTRIPSETP0lNIS[--UNFi-4] c: TOTAL SENSOR UI AttDWANCE ERROR e TRIP SETPOINT ALLOWABLE VALUE
- IUNCIl0NAI UNIT (TA) Z [51___
3.1 0.71 1.67 11960 psig** 11950 psig ro 9. Pressur zer Pressure-Lew i (PI-0455A,8&C, PI-0456 & PI 0456A, PI-0457 & PI-045FA, PI 0458 & PI-0458A) 0.71 1.67 12385 psig 12395 psig
10. Pressurizer Pressure-High 3.1 (PI-0455A,8&C PI-0456 &
PI 0456A, PI-0451 & PI-0451A, P1 0458 & PI 0458A) Pressurlier Water Level-High 8.0 2.18 1.67 192% of instrument $93.9% of instrument
.l. span span (11 0459A,1.1 -0460A, L I 9161) 1.87 0.60 190% of loop 189.4% of loop Reactor Coulant F low-Low 2.5 design f low *
12. design flow *
"' LOOP 3 100P4 . (t00PI LOOP 2 fi-0414 fi-0424 F 1 -0134 II-0444 11 0415 (1 0425 11-0435 II-0445 11 0416 II 0426 11-0436 Il-0446) 18.5 17.18 1.61 118.5% (37.8)*** 117.8% (35.9)***
13. Steam Generator Water Level of narrow range of narrow range tow-Low (21.8)*** (10.21)*** instrument span instrument span LOOP 2 LOOP 3 LOOP 4 (LOOPI 11 0517 Li-0521 L1-0537 11-0547 11 0518 LI-0528 LI-0538 11-0548 11 0519 Li-0529 LI-0539 L1-0549 11 0551 11-0552 11-0553 tl 0554) 6.0 0.58 0 19600 volts 19481 volts

14. IJndervoltage - Reactor (70% bus voltage) (69% bus voltage)
Coolant Pumps 0.50 0 157.3 Itz >57.1 ter tinder f requency - Peactor 3.3 15. Coolant Pumps
*l oop des ign i Iow = 95,160 gpm ^^ t ime (onstants utilized in the lead-lag controller for Pressurlier Pressure-tow are 10 seconds f or lead and i second f or lag. CHANNEL CAllBRAll0N shall ensure that these time constants are adjusted to these values. **^lhe value stated inside the parenthesis is for instrumentation that has the lower tap at elevation 333*; the value staled outside the parenthesis f or instrumentation that has the lower tap at elevation 438"
' O
. (' ' ' (a) v . -4H!S onCE ^= !C. - 10 U^' ! ' ' '?"' v l i
_1 i o 1ABl.E 2.2-1 (Continued}_ ;t Q 5 . 1 c REACIOR 1 RIP SYSTEN INSTRUNENIA710N TRIP SEIPOINIS _ 07:!l 4l.- l , 2, - o' 101AL SENSOR- , , Alt 0WANCE ERROR , _ IUNCildNAL UNil ( T A)_ Z [5) 1 RIP SE1PolNI AILOWABl.E VAIUL' ') i o. . .
-m 16. Turbine Irlp {
a. I ow f luid Oil Pre.ssure N.A. N.A. N.A. 25B0 psig 3500 psig !
(P1 -6161, PI:6162, PI-6163) t i ! 13 Turbine Stop Valve Closure N.A. N . A .' N.A. ?96.7% open ?96.7% open
17. Safety injection input from ESF N.A. N.A. N.A. N.A. N.A.
I
18. Reactor-Irip System I
. interlocks 1 m
i S a. Intermediate Range N.A. M.A. N.A. 21 x 10-to amp >6 x 10 - *
arap Neutron Flux, P-6 (NI -00358, NI-00368) _ ;
> b. I osa Power Reactor 1 rips' l Block, P-7 f
1) P-10 input N.A. N.A. N.A. $10% of RIP # $12.3% of RIP #
(Ni-0041B&C, N1-0042B&C, NI-0043B&C, N1-0044B&C)
-. 2) P-13 input N.A. N.A. N.A. $10% RIP # lurbine $12.'l% RIP # lurbine (PI-0505, P1-0506) litepulse Pressure lapulse Pressure f Equivalent Equivalent i
c. Power Range Neutron N.A. N.A. N.A. $48% of RIP # $50.3% of HIP # i flux, P-8 (NI -004186C, NI -0042B&C, NI-0043B&C, NI-00448&C)
# RIP- RAll. D lillHMAl_ POWLR j F % - . . _ _ _ . . I
O O O
- m -11f15 l'AGL APPt-WBel !U U"!!
CNLY l 8 I!LBLE_2. 2_;[_[ Con t inuedl l
= REACTOR TRIP SYSl[M INSTRUMENTA110N IRIP SEIPOINIS = j 101 Al S[NSOR d AttOWANCE ERROR l Alt 0WABIE VALUE, j
_ IR_lP SETPOM I i
- FUNCIIONAL UNIT (IA L Z 15) I N.A. N.A. N.A. $50% of RIP # $52.3% of RIPg m d. Power Range Nuetron Flux, P-9 (NI-00418&C, NI-0042B&C, Ni 0043B&C, N1-00448&C) >10% of RIP # >7.1% of RIPg Power Range Neutron N.A. N.A. N.A.
e. i lux, P-10 (N1-0041B&C, NI-0042B&C, NI-0043B&C, NI-00448&C) N.A. N.A. N.A. 510% RIP # lurbine $12.3% RIPg lurbine
f. lurbine impulse Chamber Impulse Pressure impulse Pressure Pressure, P-13 Equivalent Iquivalent
','2 (PI-0505, PI-0506)
N.A. H.A. N.A. N.A.
19. Reactor 1 rip Breakers N.A.
N.A. N.A. N.A. N.A.
20. Automatic Irip and Interlock N.A.
togic 4 l
# RIP r RAltu IHfRMAL POWER u_ ..
p .r-O . C
-1HIS PAGE APPtionBtfl0-ilNil-1-1DNLP . l.
JABtE 2.2-1 (Continuedl n l TABLE NOTATIONS}~ -4 ! c - l - z. Z v> NOIL 1: OVERIEMPERATURE al I [ -(I *s3) 1 ATo [K, - K,( ** .T" 31 '[ - T * ] + K,(P - P') - f ,( AI) }
o. (1 + 1,5) (1 + t,Sj (1 + t,5) ; + t.S f +
ro Where: ai = Measured AT (Unit I) ; j I***S = Lead-lag compensator on measured AT: 1 + 1,S t,, 2, = Time constants utilized in lead-lag compensator for al, .t, 28s, 1, 5 3 s; .i l 1
= Lag compensator on measured AT;
, e,o 1 + t,S t a2 , ; Time constants utilized in the' lag compensator for al, i, = 0 s;
~
t, =
}
aT o = Indicated AT at RAlED. THERMAL' POWER; ! K,- .$ 1.12 [fUnit-+j; i
K, = . 0.0224/*F lfunit-+j; .!
'I*** = The function generated by the lead-lag compensator for T avg 1 & t,$' dynamic compensatlon; t i
1., 1, = Time constants utilized in the lead-lag compensator for Tavg. T. 2 28 s, j
. Y, 5 4 s; l \
i = -Average temperature. *F; i 1 I
=
Lag compensator on measured Tavg; l 1 1 + 1.5 , i- 1 = Time constant utilired in the measured lavg lag compensator, s. = 0 s; ,
, ~ . . - -, - ,s , ..
^ ' +gis._sy r anos 2.c. s it gg! > ggty j u -l S$ IABIE 2.2-1 (Continued} .A' I -Ij TABIENOIAIIONS(Continued).hU:1! .l h N0lt 1: (Continued)-
, la :5 588.4*F (UnitM(Nominal _.lavg opera {iing temperature;) I c-m x, .= 0.00n s/psig (ve t t '-t: i i P = Pressurizer pressure, psig; 1 P' = 2235 psig (Nominal RCS operating pressure); . 5 = Laplace trani. form variable, s-2;
.and f3 (at) is a function of the indicated difference between top and bottom detectors of the ,
power-range neutro:1 ion chambers: with gains to be selected based on measured instrument
~
response during plant startup tests such that: n 43 (1) for qt - gb between -32.0%!(Un!! 'N and + 11.0% [Unii If,f,(al)=0,wheregtandqbare l percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt
gb-is total THERMAL POWER in percent of RA1ED THERMAL POWER:
For each percent that the magnitude of qt _- gb exceeds - 32.0%({unii ih',thealIripSetpointshall
~
(?) ; beautomaticallyreducedby3.25%l(Unit?)-{of~ltsvalueatRAltDIHLRMALPOWER;and i t t (3) f or each percent that the magnitude of ot - gb exceeds
11.0% h"-it -M{ the al Irip Setpoint shall ;
be automatically reduced by 1.97% p !' 1)[of its value at RATLD 1HERMAL POWLR.
~
N0lt 2: 1he channel's maximum 1 rip Setpoint shall not exceed its computed Irip Setpoint by more than 3.1%[('h:i of AT span. b t
3
?;;;5 PAGi AFPL . 'O UN!' ' 9H -
l IABIE 2.2-1 (Continued 1 8
"' T ABil NOI Ail 0NS (Continued] UN!' '( l E
Z N0lt 3: OVERPOWER al u, aj (I * ' S) ( I ) $ at,(g, _g, I 's5 )( I _)1 - g, [i (_ 3 _j - 1* ] - f ,(at )) 3 ( 1 + r ,5) ( 1 + t ,5) (1 + t ,5) (1 + 5.5) (I & r.5) m Where: AI = Measured al. (Unit ?) ; i I * 'd = lead-lag compensator on measured al; I + i,5 r,, r, = lime constants utlitzed in lead-leg compensator for al, is 2 8 5. *2 5 3 s; l
= lag compensator on measured al; " _l + i,5.
5 = Time constants utilized in the lag compensator f or al. 1, t,=0s; al o = Indicated al at RAILD 1HERMAL POWlR; l r, 5 1.08 (Un!! ? , K, 1 0.02/*l- f or increasing average temperature and 2 0 f or decreasing average temperature,
'1 = The f unction generated by the rate-lag compensator f or 1,9 3 dynamic 4+ i,5 compensatlon, r, = lime constants utilized in the rate -lag compensator f or layg,1, 2 10 s, 1 - tag compen;ator on measured Tavg; ~1 + r,$
x . . .
\ ,c,,_.\
f 4 w \j C') . T
-!!!! S- PA OL Ai' . YO UNM ' UM!Y l 1
3 S a T Af3! E 2.2-1 (Continuedl o r-TABil NOIA110NS (Continuedli l'!!! ! l l s C nn N0ii 3: (Continued) r, = Time constant utilized in the measured l avg lag compensator, o.
t. = 0 s; K. >_ 0.0020/*F h .". j for i > I" and K. = 0 for I $ I". l T = Average Temperature. *F; 1 I" = Indicated T ayg at RAILD 1HERMAL POWiR (Calibration temperature for al instrumentation,5588.4*F](Unii1f), i l 5 = Laplace transform va iable, s' ; and f,(al) = 0 for all al.
Y ~ Nolt 4: maximum Trip Setpoint shall not exceed its computed Irip 5etpoint by more than The 1.9%gcp( U % nitof;)fAT span. l
~
o o :g;
- Tills PAGE APPLICnoL4~ 10 UNil 2 ONLY .c -
8 TABtfu2.2-la - UNII 2
'" REACTOR TRIP SYSTEM INSTRUMENTAT10N 1 RIP SLTPOINIS E
Q - 101AL SENSOR Al.10WANCE ERROR. Ib,w Ilut4AL UNil (TA) - Z_ (5) TRIP SLIPOINI LOWABtE VAIDE
1. Monual- ctor Irlp N.A. N.A. N.A. N.A. N.A.

2. Power Range, N ron Flux i (NI-0041B&C, N1- 42B&C, HI-0043B&C, NI-0044 )

a. til t h Setpoint 7.5 4.56 0 $109% of R1Pt $111.3% of RIP # -

b. Low Setpoint 83 4.56 0 $25%'of RIP # $27.3% of RTP#'

3. Power Range, Neutron Flux, 1. 0.50 0 $5% of RTP# with $6.3% of RIP # with l
.High Positive Rate a time constant a time constant + -0 C N O 40 C & 4. Deleted.
i t
5. ' Intermediate Range, 17.0 8.41 0 q5%ofRIP# $31.1% of RIP #
Neutron. Flux- ; (NI-00358, NI-00368) l
-6. Sourc'e Range, Neutron F x 17.0 10.01 0 <105 cps \ <1.4 x 105 cps :
(NI-00318. N1-0032 ;
7. ,0vertemperatur al 6.6 3.31 1.95 See Hote 1 See Note 2. [
(101-4110 01-4210 (Unit 2) (U31t 2) s- G,5G i 101-431 , 101-441C) (tinit ,2) . r
8. Over wer AT .4.9 1.54 1.95 See Note 3 See Note l 401 -411 B, ' 101-4218, {t:ntt 2) I (11 nit 2) 101 -43 B, IDI-441B)
#RTP =~ PATED Ti1ERRAE POWER. . e -- -----------.__.__---_-_-_._____..-_._.-_.-.1- . - . _ _ _ _ ~
3- , e O O O . glis PAL ,. A. ice 8tt 10 Utti T 2 ONLY l TABLE. f.2-la _(Cont,inued) ,
.?
- r-R(A[TM.13 SYSTEN INSTRtMNTA(ION TRIP SEIP0lNIS J UNil 2 -{. f : E l G TOTAL SENSOR
Att0MANCE EnRDR
-(TA) ' Z L51 ' TRIP SEiP0lNT Att0WA81[ vat _tg .'
N IUNC110NAL ONi! '-- -
.q-g
9. Pressurt2er Pressure-Low 3.1' O.11 1.67 11960 psig** 11950 psig-t
" (PI-04554.B&C, PI-0456 & ;
PI-0456A, PI-0457 & PI-0457A,
\ 1-0458 & PI-OiS8A) i Pressur s r Pressure-High 3.1 0.71 1.67 _2365 psig <2395 psig -l 8- 10. ;
(PI-0455 , &C, PI-0456 & I PI-0456A, P 457 & PI-0457A, PI-0458'& PI-04 ) .] y
<92% of instrument <93.9% of instrument i 11.
Pressurizer Water Level-H gh 8.0 2.18 1.67 Qi span span (LI-0459A.'LI-0460A, L1- 1) -1 2.5 1.87 0.60 1905 of loep : 289.4% of . loop i 12. Reactor Coolant Flow-Low design i10w* design iIow* , (LOOPI t00P2 LOOP 3- _l . .'4 - . F1-0414 F1-0424 F1-0434 ~F1-0444 F1-0415 F1-0425 FI-0435 FI-0445 - FI-0416 FI-0426 FI-0436 FI-0446) I
/ 18.5 17.18 1.67 118.55 (37.8)*** 117.8% (35.9)*** j 13 .- Steam Generator Water Level of narrow range of narrow range.
tow-Low (21.8)*** (18.21)* instrument span instrument span
- (LOOPl LOOP 2 L LOOP 4 e L1-0517 LI-0527 L 53F L1-0547 '
. . LI-0518 11-0528 1-0538 11-0548 LI-0519 L1-05 Ll-0539 .LI-0549 -* L I -0551 LI- 52 LI-0553 LI-0554) .
14. Underv tage'- Reactor 6.0 0.58' 0 19600 volts 29481 voits C ant Pumps (70% bus voltage (69% bus voltage)
] 3.3 0.50 0 157.3 Hz 25 Hz
15. nderfrequentf - Reactor Coolant Puelis .
*toop <8esign f les .- 95.700 gpm x / . ' ** lime condan, utilized in.the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead:and N f I second for . lag. CHANNEL.C't.lBRA*lDN shall ensure.that these time constants are adjusted to these values. ' , ***lhe value stated inside the parenthesis is. for. lostrumentation that has the Irn.er tap at e!evation 333"- the s value stated'outside the pt renthesis is for instrumentation that has the lower tap at elevation 438".
i t
o P oL 1HIS PAGE.APPLICAstt. 10 UNIl 2 DNLY Q
~ .S- TABLE 2.2-la (Continued). q .g e REACIOR TRIP SYSTEN INSIRUMENTATION TRIP SETPOINIS - UNil 2 ~
15
d 101AL . SENSOR i- ,- . ALLOWANCE ERROR
. FUNCTIONAL. UNIT (TA) Z {S) 1 RIP SEIP0 INT. AlLOWABLELVALUL o. m 16. Turbine 1 rip i i a '. L Fluid 011 Pressure N.A. N.A. N.A. 3580 psig 2500 psig j , (P1 - 161, PT-6162 PT -6163) ; i <
b. Turbine Sto alve Closure N.A. N.A. .N.A. 296.7% en ?96.7% open

17. Safety injection input om ESF N.A. N.A. N.A. . N.A. -

18. Reactor 1 rip System 1 Interlocks-
, a. Intermediate Range N. . - . N.A. 21 x 10 20 emp 26 x 10 in amp ; J]E - Neutron Flux, P-6 : (N1-00358. NI-00368) ,
b. Low Power Reactor 1 rips y Bicck, P-7 ,
t
1) P 'l input N.A. N.A. N.A. 510 RIP # $12.3% of RIPJ' (N1 '941B&C, NI- 2B&C, .
NI-Ob.3B&C, NI 0448&C) ! l
2) P-13 inpy N.A. N.A. N.A. $10% RIPJ lurbine <12.3% RIPJ lurbine .;
(Pi-05 5. PI-0506) Impulse Pressure 1 1se Pressure- [ Equivalent Equi ent
c. P er Range Neutron N.A. M.A. N.A. $4B% of RIP # $50.3% of R F -
(NI-0041B&C. NI-0042B&C, ,.
/ flux'P-8 NI 0043B&C, NI-00448&C)
I
-6 _.
L 4 O
^
4 O O : ^ 1HIS PAGE APPLICABLE 10 UNIl 2 ONLY ! o 3 --TABIE 2.2-la (Continuedl 3 "m c .REACIOP TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOIN15 - UNil 2 5 5 10T AL - SENSOR g_ , t ALLOWANCE ERROR '
' IUNC1 ONAt UNil (TA) .Z L(S) 1 RIP SETPOINT AILOWABitA ALUE. ,
o- - / i
'd. P ange Nuetron Flux, P-9 N.A. N.A. N.A. 550% of RTP# $52 % of RTP# '[
m *
(NI- 1 &C, NI-00428&C, NI-0043B& , NI-00448&C) j

e. Power Range Neu n N.A. N.A. N.A 210% of RTP# 37.7% of RIP # ;
Flux, P-10 (N1-0041B&C..NI-00428 NI-0043B&C, NI-00448&C). .
f. . Turbine Impulse Chamber .A. N.A. N . A .- < % RIP # lurbine $12.3% RIP # lurbine Pressure.-P-13 mpulse Pressure Impulse Pressure
. . ( PI -0505, PI-0506) Equivalent' Equivalent i - 19. Reactor Trip Breakers. N.A. N. N.A. N.A. N.A.
1 I
20. Automatic 1 rip and Interlock N.A. .A. . . N.A. N.A.
Logic - , l s t ' j l
# RIP = RA1ED ll*ERMAL POWER
O O O THIS PAGE APPLICABLE TO UNil 2 ONLY-
-/ .-/
O 1 ABL E 2.2-la (Continued) y 9.-
" /
TABLE NOTATIONS - UNIT 2
\ ,. ' .f- -4 NOIL 1: DVERIEMPERATURE AT I - aT I N '25) 1 5 ATo [K, - 3K II * **3I [T - T'] + K 3 (P - P*),- f 3(AI))~
o* (1 + t'25) - 1+2 53 (1 + tsS)- 1'+ t.S , to Where: .ai = Measured AT by RID Manifold Instrumentatioh{UM _ - M (s' ; 3 I * ** = lea -lag compensator on measured AT;- 1 + v2S l ta, t2 = Time const n(s utilized in lead-lag compensator for AT, is 2 8 s, , 22 53s* .; T . Lag compensator on h asured aT: ! = ! g i + 7,5.
= Time constants' utilized in the lag compensator for AT, 1, = 0 s; 1, , ,N ATo - Indicated AT at RATED THERMAL'POWCR;, .
K2 5 1.10 (Unit 2); NN N , K, = id12/*F.(Unit 2); I * *$b
= The function generated by the lead-lag compensator for Tavg i 1,/1,5 dynamic compensation ,' ., 1, = Time constants utilized in the lead-lag compensator for lavg. T. 2 28 s, v s 5 4's; ;
T = Average temperature. *F; I
= . Lag compensator on measured Tavg; 1 + t.S 1 a . Time constant utilized in the measured lavg lag compensator, s. = 0 s; - - o - - - + _ -T 1
A y/ . N-~. THIS PAGE APPla..8tE TO UNIT 2 ONLY -
/
TABLE 2.2-la (Continued) TABLE NOTATIONS (Continued) - UNil 2. E
] NOIE 1: (Continued) .- * ? ,
588.5 F (Unit 2) (Nominal Tavg operating temperature); I'
.,/ l ;
K, ,= ' O.00056/psig=(Unit 2); ,/ 'l
/ !. .
P [' Pressurizer pressure, psig; j/ P'
=- 2235 psig (Nominal RCS operating pressure); .S - Laplace transform variable, s-2; '
and f3 (aI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with' gains to be s 1ected based on measured instrument ; response during plant startup tests such'that: ' r.>
* \ r ; (1) f or qt - 4b between -33.5% (Unit 2) andNf>.5% (Unit 2), f,(al) = 0, where qt and qb are l percent RATED THERMAL POWER in the top'and'bptten halves of. the core respectively, and gt & gb ~
is total THERMAL POWER in. percent of RATED THERMAL POWER-
/ a (2) for each percent that the magnitude of qt'- Ab **C Ed5 - 33.5% (Unit 2)..the AT 1 rip Setpoint shall '
l be automatically reduced byl.27% (Unit 2) of its value at RATED THERMAL POWER; and y . (3) For each percent that y the arqnitude of qt - Ab exceeds + y (Unit 2), the al Irip Setpoint shall be automatically r duced by U.83% (Unit 2) of its value at RA'T 0 THERMAL POWER. ; N01E 2: The channel's maximum
/ Trip Seipoint shall not exceed its coe%.1 Trip Setpoint N
by more than 2.5% (Unit 2). l .
/ 1 N :
x
\ j r / s f
,y ., . 'w- )
11115 PAGE APPL..a _. 10 UNil 2 DNLY h 1 ABLE 2.2-la (Continuedl TABli NOTAT10NS (Continuedl - UNil 2 s N - 3 Ndit 3: OVERPOWER AT in
\ \ol\ 1I + r1 }I ,5) ( 1 + s ,5)
I $ alo (K - K, ( ( 1 - K. [1 I }- "] - f,(al)) 3 (1 + s ,5) (1 + 1.5) (1 + 1.5 ro a Where: AT = Heasured al by RTO manifold instrumentation (Unit 2 , I Y = lead-lag compensator on measured al; 1 . ,5 i,t, = y constants utilized in lead-leg c ensator for E , si 28S.22535;
= Lag compens oronmeasuredAd 1+v 35 / /
b t, = lime constants uti edinthelagcompensatorforal, t,=0s; al o = Indicated at RAILD THER A POWER; K, $ 1.089 dit2), ! K, t 0 02/*F for increasing average temperat e and 2 0 for decreasing average temperature,
= The function generated by the rate-lag compensato or l ayg dynamic 1 t' t ,5 compensatlon, \
i, = Ilme constants utilized in the rate-lag compensator for lav i,2 10 s. 1
/ =
Lag coopensator on mer.sured Tavg;
/ l + 1,5
\ s' lHIS PAGE APPLit.*dts. TO UNIT 2 UNLY TABLE 2.2-la(Continuedl n
b TABLE NOTATIONS (Continued) - UNil 2 E g NOTE : (Continued) nn i.
= Time constant utilized in the measured T ayg lag compensj or, 1 = 0 s; ]
N ' K. . 0.0013/*F (Unit 2) for T > I" and K. = 0 for T ". l I - Average Temperature. *F; l' = Ind at RAILO Tiler. MAL POWEV(Calibrat ion temperature f or al instrum(edTavg en etion, s 588.s r (Unit d, I 5 = Laplace trans . variable, 1; and f,(al) = 0 for all al. NOIE 4: The channel's maximum Trip Setpoint f Il not exceed its computed Trip Setpoint by more than 2.4% (Unit 2) of ai span. l l
,/
. . _ - . _. . _ _ _. _ - . - _ __ _.~ _ _ _. - ._ _. . _ __ _.
g 3 % N;; 'S uC? nE m m " ' Day y l
./ 2.1 SAFETY LIMITS 4
BASE $ 2.1.1 RE ACTOR CORE - GM 1 f I The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission prooucts to the reactor. coolant. Overneating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transf er coef ficient~ is large and the claddint surf ace temoerature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime 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 measurable parameter during-Cperation and theref ore THERMAL POWER and reactor coolant temperature and pressure have been related tc ONB through correlations which have been developed to predict the l DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB. Tr.e DNB thermal design criterion is. that the probability .that DNB will not occur on the most limiting rod is at least 955 (at a 955 confidence level) for any Condition I or 11 event. b meeting the DNB design criterion, uncertainties in plant oDeratint rarameters, nuclear and thermal parameters, fuel f abrication parameters, ai t t.mouter codes must be considered. As described in the FSAR, 'he effects of , taese uncertainties have been statistically combined with L.e correlation
- uncertainty. Design limit DNBR values have been determined that satisfy the ONB design criterion.
Additional DNBR margin is maintained by performing the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and trans* on core) and to provide DNBR margin for operating and design
< flexib .ity.
The curves of Figure 2.1-1 show reactor core safety limits for a range of THERMAL POWER, REACTOR COOLANT SYSTEM pressure, and average temperature which satisfy the following criteria: e A. The #7erage enthalpy at the vessel exit is less than the entaalpy of satur ted liquid (f ar lef t line segment in each curve), f S. The minimum DNBR satisfies the ONB design criterion-(all the other line segments in each curve). The VANTAGE 5 fuel is analyzed using the WRB-2
. correlation with design limit DNBR values'of 1.24 and 1.23 for the typical cell and thimble cells, respectively. The LOPAR fuel is analyzed using the WRB-1~ correlation with design limit DNBR values of 1 23 and 1.22 for the typical and thimble cells, respectively.
Lp C. The hot channel exit quality is not greater than the upner limit of the ! t
, quality range (inc;uding the effect of uncertainties) or the DNB correlations. This is not a limiting criterion for this plant.
V0GTLE IJNITS - 1 & 2 B 2-1 i m._ _ _ _ _ _ _ _ ___ _ _ _ . _ ___ ___ . _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ __ _ _ _ . _ , _ . . . _ -.
J THIS PAGE APPLICABLE TO UNIT 2 ONLY
2,1 SAFETY LIMITS BASES-
\ /
2.1.1 ACTOR CORE - UNIT 2 l The strictions of this Safety Limit prevent overheating f the fuel and possib cladding perforation which would result in the r ease of fission products to he reactor coolant. Overheating of the fuel c dding is prevented by restrictin fuel operation to within the nucleate boil 9g regime where the heat transfer c ef ficient is large and the cladding surf t.te temperature is slightly above t coolant saturation temperature. Operation abov the upper boundary of the nucle te boiling regime could result in excessive adding temperatures because the omet of departure from nucleate boiling DNB) and the resultant sha p reduction in heat transfer coefficient. DNB is no. a directly measurable rameter during operation and therefore THERMAL POWER a d reactor coolant t perature and pressure have been related to DNB-through the -3 (R Grid) corr ation. The W-3 (R Grid) DNB correlation has been develo d to predict e DNB flux and the locat%n of DNB-for axially uniform and nonun form heat f ux distributions. The 1 a ' heat flux ratic (DNBR) is defi d as th ratio of the heat flux t5 cause DNB at a particular core catio to the local heat flux at. indicative of the margin to DNB. The minimum value of the DN du ng steady-state operation, normal operational transients and ant cipated transients is limited to 1.30. This value corresponds to a 95% pr ability a a 95% confidence level that DNB
-will not occur and is chosen s an appropr te margin to DNB for all operating conditions.
The curas of Figur 2.1-la show reactor re safety limits which are l
- determined for a range f reactor operating cond. ions. The core limits represent the loci of oints-of THERMAL POWER, REA TOR COOLANT SYSTEM pres-sure and average te erature which satisfy the foll ing criteria:
A.- The average e thalpy at the vessel exit is equal t the enthalpy of , saturated li uid (f ar lef t line segment in each cury . B. The minim DNBR is not less than the design. limit (all he other line segment in each curve). C. The hpt channel exit quality is not greater than the upper 1 it of the qual /ty range of the W-3 (R-Grid) correlation which is 15% (m (die line se ent on Reactor Coolant System pressure curves, 2400 psia an 2250 p ia; this is not a limiting criterion for-this plant). V0GTLE UNITS - 1 & 2
,,L-, ~ ,. . ., . _ - . . _ - . - - . . . . - , , - . - , . , . . . . . . . _ - , - -
~ _.~.__.._....._.___.____._.._.___.-__m . _ . . . . . - . . _ _ . _ . _ _ _ . . _ _ _ . -
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE'- SHUT 00WN k LIMITING CONDITION FOR OPERATION 3.1. 2. 5 As a minimum, one of the following borated water sources shall be - . OPERABLE:
a. A Boric Acid Storage Tank with:
1
1) A minimum' contained borated' water volume of 9504 gallons (19%
of instrument span) (LI-102A, LI-104A), 2). A boron concentration between 7000 ppm and 1700 ppm, and
3) A minimum solution temperature of 65'F (TI-0103).

b. The refueling water storage tank (RWST) with:-

1) A minimum contained borated water volume of 99404 gallons (9% of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, t.1-0993A),

2) A boron concentration between 2400 ppm and-_2600 ppm, and -
3)- A minimum solution temperature of 44'F Lbt'Inesaerg';;;y l O' -( T I -109 B2 ) . APPLICABILITY: MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEllLANCE REQUIREMENTS
.4.1.2.5 The above required borated water source 'shall be demonstrated OPERABLE:
a, 'At least once per 7 days by:
1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and 3)- When the be-ic acid storage tank is-the source of borated water-and the ambient temperature of the boric acid storage. tank rr n (TISL-20902,.TISL-20903) is 172'F, verify the boric. acid ste -: 4 tank-solution temperature is >65?F.

b. At least once per 24 hours by verifying the RWS1 temperature (TI-10982)
O when it is the soy'm* nf barated water and_the outside-air temperature is less than 40'F "ai+ 'i nr s n.g, , , g, V0GTLE UN115 -1 & 2 - 3/4 1-11
. u . > .. -.- .-- . . . . - . , -_- .- - - . - . _
p REACTIVITY CO.NTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1. 2. 6 As a minimum: the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
a. A Boric Acid Storage Tank with:

1) A minimo ntained borated water volume of 36674 gallons (81% '
of instrument span) (LI-102A, L1-104A),
2) A boron concentration between 7000 ppm and 7700 ppm, and

3) A minimum solution temperature of 65'F (T!-0103).

b. The refueling water storage tank (RWST) with; i) A minimum contained borated water volume of 631478 gallons (86%
of instrument span) (LI-0990A&B, LI-0991A&B, L1-0992A, LI-0993A),
2) A boron concentration between 2400 ppm and 2600 ppm, LC 3) Aminimumsolutiontemperatureof44'Fks,'t'.)orWI'Unii 2) l i
d .
4) A maximum solution temperature of ll6'F (TI-10982), and l
l
5) RWST Sludge Mixing Pump Isolation Valves capable of closing on RWST low-level.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: l a. With the Boric Acid Storage Tank inoperable and being used as one of the above required borated water sources. restore the tank to OPERABLE status withiv 72 hours or be in at Itast HOT l STANDBY within the next 6 hours and borated to a SHUTLOWN MARGIN as required by Figure 3.1-2 at 200*F; restore the Boric Acid Storage Tank to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours.
b. With the RWS1 inoperable, except for the Sludge Mixing Pump l
Isolation Valves, restore the tank to OPERABLE status within I hour or be in at least HOT STANDB" within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. O v V0GTLE UNITS - 1 & 2 3/4 1-12
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued)
c. With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the valve (s) to OPERABLE status within 24 hours or isolate the sludge mixing-system by either closing the manual isolation valves or deenergizing the OPERABLE solenoid pilot valve within 6 aours and maintain closed.
SURVEILLANCE REOUIREMENTS 4.1.2.6 Each borated water source t ,all be demonstrated OPERABLE:
a. At least once per 7 days by:

1) Verifying the boron concentration in the water,

2) Verifying the contained borated water volume of the water source, and

3) When the boric acid storage tank is the source of borated water and the ambient temperature of the boric acid storage tank room f
( (TISL-20902, TISL-20903) is 5 72'F, verify the boric acid storage tank solution temperature is > 65'F.
b. At least once per 24 hours by verifying the RWST temperature gI-10982 gen the outside air temperature is less than 40*F hu .ii. ii ed
, - - . ~ ~..s o,
c. At least once per 16 months by verifying that the Sludge Mixing Pump Isolation Valves automatit. ally close upon RWST low-level test signal.
O V0GTLE UNITS - 1 & 2 3/4 1-13
REACTIVITY CONTROL SYSTEMS ROD OROP TIME LlhiTING CONDITION FOR OPERATION 3.1. 3 . 4 The individual shutdown and control rod drop time f rom the ohvsica fully oithdrawn position shall be less than or equal to 2.7 p0M ') er 2.2 ['l;r. ' t 23 seconds from beginning of decay of stationary gripper coil voltage To dashpot entry with:
a. Tavo (TI-0412, TI-0422, TI-0432 TI-0442) greater than or equal to
$51'i, and
b. All reactor coolant pumps operating.
APPi1CABillTY: MODES 1 and 2. ACTION: With the drop time of any rod determined to exceed the above limit, restore the red drop time to within the 'bove limit prior to proceeding to MODE 1 or 2. S_URVEILLANCE REOUTREMENTS 4.1.3.4 The rod drop time .. .. i 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 which could affect the drop time of those specific rods, and

c. At least once per 18 months.
V0GTLE UNITS - 1 & 2 3/4 1-19
"i; FAGE AF^u;AOLE 70 6:T ' 0'! i .3/a.2 DOWER OfSTRIBUTION LIMITS > , 3/4.2.1AxtAL FLUX'OlFFERENCE IT i ,
LIMITING CONDITION FOR OPERATION-1.2.1 The indicated (N!-06418. N!-00428, NI-00438, N!-00448) AX1AL FLUX f DIFFERENCE (AFD) shall be maintained within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). APPLICABILITY: MODE 1 AB0VE 50 PERCENT RATED THERMAL POWER" ACTION:
a. With the indicated AXIAL FLUX OlFFERENCE 69tside of the limits ; L specified in the COLR, i-

1. Either restore the indicated AFD to within the limits !
within 15 minutes, or }'
2. RedJee THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux * - Hign' Trip setpoints to less than or equal to 55 percent of RATED-
. THERMAL POWER within the next 4 hours.
I ;
b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.- I' SURVEILLANCEREOUiREMENTS 4.2.1.1- The indicated AFD shall be detemined to be within its limits during POWER' OPERATION above 50% of RATED. THERMAL: POWER by: l' a.. Monitoring the indicated- AFD for each 0PERABLE excore channel:
.1) At least once per 7 days when the AFD Monitor Alarm is OPERABl.E, od
2) At least once per hour-until2 the AFD Monitor Alarm is updated

af ter restoration to OPERABLE status.
'b. Monitoring and' logging the indicated AFD for each OPERABLE excore .
channel at least once per hour for the-first 24 hours and at least once' per 30 minutes thereaf ter, when the. AFD Monitcr Alarm is inoper - able. !The logged values of the indicated AFD shall be assumed to exist during the interval preceding each' logging.
c. The provisions of Specification 4.0.4 are not applicable.
4.2.1.2 The-indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside-its 'j-
' limits. ! *See Special Test Exceptions Specification 3.10.2. , V0GTLE UNITS - 1 & 2 3/42-1
THl$ PAGE APPLICABLE 10 UNil 2 ONLY l (Q g 3/a.2 DOWER DISTRIBUTION LIMITS 2 .2.1 AX1AL FlVX DIFFERENCE UNIT 2 l LIM NG CONDIT10N FOR OPERATION
\ /
3.2.1 T indicated (N1-0041B, N!-0042B, NI-00438. NI-00448) AXI FLUX DIFFERENCE difference units) about AFD)shallbemaintainedwithinthetargetband(flu he target flux difference. The target band is 5 e /cifiedinthe CORE OPERATIN LIMITS REPORT (COLR). The indicated AF may deviate outside the recuired target d at greater than or anual to 50% bu less than 90% of RATED THERMAL POWER ovioed the indicated AFD is within the A eptable Operation Limits specified the LOLR and the cumulative penalty de iation time does not exceed I hou during the previous 24 hours. The indicated AFD may dev te outside the required arget band at greater than 15% but less than 50% of R ED THERMAL POWER pro ea the cumulative penalty deviation time does not exce a 1 hour curing th previous 24 hours. APPllCABILITY: MODE 1, above ' % of RATED TH MAL POWER.* # ACTION: [] V
d. With the indicated AFD outs .e THERMAt. POWER greater than o f the reauired target band and with cual to 90% of RATED THERMAL POWER, within 15 minutes either:
1 Restore the indicated AFD to ithin the target band limits, or
2. Reduce THERMAL POW to less theo 90% of RATED THERMAL POWER,

b. With the indicated A- outsida of the quired target band for more than I hour of cumu ative penalty deviat on time during the previous 24 hours or outsi the AccGotable Operat n Limits specified in the COLR and with TH MAL POWER less than 90% b t equal to or greater than 50% of RAT THERMAL POWER. reduce:

1. THERMAL P WF.R to less than 50% of RATED THE AL POWER within 30 minu s, and

2. The P wer Range Neutron Flux * - High Setpoints t less than or eau to 55% of RATED THERMAL POWER within the nex 4 hours.
'See Specia Test Exceptions Specification 3.10.2. *$urveill nce testing of the Power Range Neutron Flux-Channel may be pgrformed iceiow 0% of RATED THERMAL POWER) pursuant to Specification 4.3.1.1 provided the 1 dicated AFD is maintained within the Acceptable Operation Limits sgeci-p fie in the COLR. A total of 16 hours operation may te accumulated with the l AF outside of the above requireo target band during testing without penalty
a ution.
'!0GTLE UNITS - 1 &2 _
~~
IHIS PAGE APPLICABLE 'O UN11 2 ONLY 7l
WER DISTRIBUTION LIMITS i141 'NG CJNDITION FOR OPERATION - tlNIT '
S ACTION ntinued) ,'
c. W i the indicated AFD outsiae of the required target bad for more tha 1 hour of cumulative penalty deviation time during the previous 24 ho rs and with THERMAL POWER less than 50% but greater than 15% of RATED ERMAL POWER, the THERMAL POWER snail not be /ncreased equal to or greater than 50% of RATED THERMAL POWER untiVthe indicated AFD is within\the required target band and the cumulat'ive penalty devia- ,
tion has been reduced to less than 1 hour in the' previous 24 hours. 1 N '
/ l i SURVE llL ANCE REOUIREME'NTS s ,
4.2.1.1 The indicated AFD\shall be oetermined to be within its limits during POWER OPERATION above 15% ofs RATED THERMAL POWER,-by: i '\ /
a. Monitoring the indicated AFD f or eacV 0PERABLE excore channel:
l
1) At least once per 7' days whenfthe AFD Monitor Alarm is OPERABLL.
and ' x / j 2) AtleastonceperhourundttheAFDMonitorAlarmisupdated i af ter restoration to OPENABLE status. I b. Monitoring and logging the' indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes the'reafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist dering the/ interval p:* ceding eacn logging. i / 4.2.1.2 The indicated AFD'shall be considered outside of its target band when two or more OPERABLE exc4re channels are indicating the AFD to be outside the target band. Penalty Oeviation outside of the required target band shall be accumulated on a time basis of: 'q a. x One minut penalty deviation for each 1 minute o,f POWER OPERATION outside f the target band at THERMAL POWER levels equal to or above 50% of ATED THERMAL POWER, and i
\
b. One, half minute penalty deviation for each 1 minute f, POWER OPERATION oupide of the target band at THERMAL POWER levels between 15% and 5.0% of RATED THERMAL POWER.
4.2.1.3 he target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.
/
4.9.1.4 The target flux difference shall be updated at least once per 3)' Ef f ective Full Power Days by eithe' determining the target t lux uif f erence pursuant to Specification 4.2.1.3 above or by linear interpolation between the p}.
\--
most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification. 4._0.4
. = .
are nut appli. cable _ ,.. _ -
'N,M NG6 3MT0dT'c4 At.td LEPr B{}. ' ~ '!0GTLE UNITS - 1 &? ' 7 /4 2-?
__ . . _ - ._ __ ~._ m . .. _ . . - IHi3 EAd AFFLiCA6Lt. 40 UNIT 1 y l O POWER DISTRIBUTION LIMITS , SURVEILLANCE REOUIREMENTS - uE , - t l 4.2.2.1 The provisions of Specifications 4.0.4 are not applicable. 4.2.2.2 Fn(Z) shall be evaluated to determine if it is within its limit by: l 4
a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Determining the computed heat flux hot channel factor, FgC (Z), as follows:
Increase the measured F 0 (Z) obtained f rom the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties. - C
c. Verifying :..at FQ (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the following relationship:
C RTP Fg g7) pQ x K(Z) for P > 0.5 P x W(Z). C RU FQ g7) pQ x K(2) for P 5,0.'5
~
i 0.5 x W(Z) C RD-Where FO (7).is obtained in Specification 4.2.2.2b above FO is the FQ limit, K(Z) is the normalized FQ(Z) as a function of core height, P is the fraction of RATED THERMAL POWER, and W(2) is the cycle dependent function that accounts for power l' distribution transients encountered during norinal operation. RTP FO , K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6,8.1.6.
e. Measuring.F Q (Z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which F0(Z)'was last determined *, or

2. At least once per 31 Ef fective Full Power Days, whichever occurs first.
"During power escalation af ter each f uel loading, power level may be l increased until equilibrium conditions at any power level greater than
, or equa! to 50% of' RATED THERMAL POWER have'been achieved and a power distributio,n map obtained.
V0GTLE UNITS - 1 & 2 3/4 2-4
- -. ~
.-. - - . - . - . . - - - - - - . . . - . . - . - , _ - . . . - - . . . - . . - - _ ~ - . _ . . . - - . . - +
_T M10 PAGE APPLICAhi. TU Unii iUnu[1 l
\ POWER DISTRIBUTION LIMITS- -l ~
SURVEitLANCE RE0VIREMENTSr (Continued)T UC T 'l _ l
f. With' measurements indicating U
maumum FQ (2) over 2 K(2)j has increased since the previous determination of FOC (2) either of the_following actions shall be taken: Increase FQ C (2) by 2% and verify that this value satisfies 'l 1) the relationship in Specification 4.2.2.2d, or C
2) Fg (2) shall be measured at least once per 7 Effective Fu'll Power Days until two successive c.,aps indicate that t t maximum- IF Q(2)h C is not increasing.
1 1 over 2.. -K(Z)j r
g. With the relationships specified in Specific' ation 4.2.2.2d above gO .
.not.being satisfied: i 1
1) Calculate the percent FQ (2) exceeds its limits by the '
following expression: , e- 3
~
[ maximum- F0 (2) x W(2) 0.5 c
<l over 2 -
FQ P x K(2) 1 [xl l
~
I
)[ maximum F0 (2) - x W(2)~\'-I $x100forP_<0.5,and < over Z F0 .x K(2) - 0.5 -
l
2) ~ The following action shall be taker [.

Within 15 minutes, control the AFD.to within new AFO limits which=are. determined by reducing the.AFD limits specified-in the CORE OPERATING LIMITS REPORT by l% AFD for each percent FQ (2) exceeds its limits as determined in Specification.
Within 8 hours, reset the AFD alarm setpoint; to l- 4.2.2.29 1, 1 these modified limits. V0GTLE UNITS - 1 & '2 3/4 2-5
b :'S "%E ^oPL-!Cn0LE TO UEi ' ONL") l POWER DISTRIBUTION llHITS SURVEILLANCEREOUTREMENTS(Continued)hUNii d l
h. The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e. 0 - 100%, inclusive,

i. The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.29 above ara not applicable in the following core plane regions:

1) Lower core region from 0 to 15%, inclusive.

2) Upper core region f rom 85 to 100%, inclusive.
4.2.2.3 When QF (2) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured FQ(Z) shall be obtained f rom a power distribution map and increased by 3% to account for manuf acturing tolerances and further increased by 5% to account for measurement uncertainty. O Lo V0GTLE UNITS 'l & 2 3/4 2-6
T s PAC-E INTENTM ALLN GT BLW l r THIS PAGE APPLICABLE TO UNil 2 ONLY l o \ Q POWER DISTRIBUT10N IIMITS l
. \ ; \
l SURVEtLLANCE RE0VIREMENTS - UNIT 2 l l l ~\ l \ l i 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 Fxy'shall be evaluated to determine ifQ F (Z) is within its limit by: Il i a. Using the movable incore detectors to obtain a power distribution
! map at any THERMAL POWER greater than 5% of RATED THERMAL POWER l before exceeding 75% of RATED THERMAL POWER following each fuel loading,
b. Increasing the measured Fxy component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties,

c. Comparing the Fxy computed (Fxy) obtained in Specification 4.2.2.2b.
above to:
1) The Fxy limits for RATED THERMAL POWER (Fh ) for the appropriate fm measured core planes given in Specification 4.2.2.2e. and f.
below, and ( )
2) The relationship:
Fxy = F [1&PFxy(1-P)), Where Fxy is the limit for fractional THERMAL POWER operation expressedasafunctionofFhfP, PFx is the power factor multiplier for Fxy specified in the OLR, and P is the f raction
; of RATED THERMAL POWER at which Fxy was measured, i
I
d. Remeasuring j fxy accoiding to the following schedule:

1) When.F'xhisgreaterthantheFhy limit for the appropriate
, mgasuredcoreplanebutlessthantheFxyrelationship, additional powerdistributionmapsshallbetakenandFxhcomparedtoFhTP and Fxy either: I a) Within 24 hours after exceeding by 20% of RATED THERMAL POWER or greati , the THERMAL POWEP at which Fxy was last I / determined, or l l
/ b) At least once per 31 Ef Uctive Cull Power Day; (EFPD), / whichever occurs first. .
V0GTLE UNITS - 1 & 2 3/4 2-7
rg THIS PAGE APPLICABLE TO UNIT 2 ONLY l WER DISTRIBUTION LIMIT.i SU LLANCE REOUIREMENTS (Continued) - UNIT 2 l
\ _ /
2) hen the Fxy is less than or equal to the F 1 it for the a repriate measured core plane, additional po r distribution RTP maps hallbetakenandFhcomparedtoF x and F xy at least once p 31 EFPD.

e. The F xy limi sed in the Constant Axial ffset Control analysis for RATED THERMAL OWER(FhP)shallb specified for all core planes containing Bank "0" ontrol rods and I unrodded core planes in the COLR per Specific ion 6.8.1.6; f, The Fxy limits of Specifi tion .2.2.2e. above are nct applicable in the following core plane r gions as measured in percent of core height from the bottom of th uel:

1) Lower core region f r 0 to 1 ", inclusive,

2) Upper core region rom 85 to 100 inclusive,
+ 2%, 46.4 1 2%, 60.6 1 2%,
i p) v
3) Grid plane reg ns at 17.8 1 2%, 32.
and 74.9 1 2 , inclusive, and
4) Core plan regions within i 2% of core hei t [1 2.88 inches) about t bank demand position of the Bank ' " control rods.

g. With Fw exceeding Fxy the effects of Fxy on FQ(Z) s 11 be evaluated to detfrmine if FQ (2) is within its limits.
4.2.2.3 Wh_ FQ (Z) is measured for other than Fxy determinations, a overall measured (Z) shall be obtained from a power distribution map and inc eased by 3% t account for manuf acturing tolerances and further increased by (to I accou for measurement uncertainty. l V0GTLE UNITS - 1 & 2 l
- -- . - . . ---. ~ . . . .- .~ - ..
l POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDIT10r /OR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits:
a. Reactor Cociant System T_. (Tf-0412, TI-0422, TI-0432, TI-0442),
5 592.5'F t ' ' L ; ;; f " ( Uni ;, ;j. l
b. Pressurizer Pressure (P!-0455A,BLC, PI-0456 & PI-0456A. PI nao L PI-045]A, PI-0458 & PI-045BA), 3 2199 psig*\' Ti; ;) v, ::::
n d-g""f M.
c. Reactor Coolant System Flow (FI-0414, F1-0415, FI-0416, FI-0424, FI-0425. FI-0426, F1-0434_. F1-0435. FT-ndjA FT-0444 FI-0445, FI-0446) 2391,225 gpm**IUn't '; a 300. MS ;;r * (Un;. Q . l APPLICABILITY: MODE 1.
ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. .f SURVEILLANCE REOUTREMENTS . 4.2.5.1 Reactor Coolant System Tavg and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours. RCS flow rate shall be monitored for degradation at least once per 12 hours. In the event of flow degradation, RCS flow rate shall be determined by precision heat balance within 7 days of detection of flow degradation. 4.2.5.2 The RCS flow rate indicators shall be subjected to-CHANNEL CAllBRATION at each fuel loading and'at least once per 18 months. 4.2.5.3 After each fuel loading, the RCS flow rate shall be determined by. precision 1 heat' balance prior to operation above 75% RATED THERMAL POWER. The RCS flow rate shall also be determined by precision heat balance at least once per 18 months. Within 7 days prior to per-forming the precision heat balance flow measurement, the instrument-ation used for performing the precision heat balance shall be calibrated. The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
*L1mit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED . THERMAL POWER per minute or a THET4AL POWER steo in. excess of .10% of RATED THERMAL POWER. ** Includes a 2.2% h t 'l ^- ).Z wait @ flow measurement uncertainty. l 4
V0GTLE UNITS - 1 & 2 3/6 2-13
l O O O i
< : TABLE 3. ^.,(Continued)_ >
8
'l ;d ENGINEEREb SAFETY FEATURES ACTUATION SYSTEM INS 1RUMENIA110N TRIP SE1PolN15 . l t,
m -j 3 101Al SLHSOR ] Q ALLOWANCE ERROR :
", FUNCTIONAL dNil ~(TA) Z _(S_}_, IRIP SEIPOINT Alt 0WABIE VAIUL -
e
~
7. Semi -Automat ic Switchover. to ,
j 4
** Containment Emergency Sump (Continued) - 'b. RWSI Level---Low-Low 3.5 0.71 1.67 > 275.3 in. from > 264.9 in. from- !
Coincident With Safety tank base tank base
injection . (> 39.1% of (> 31.4% of ;
(1.1-0990A&B, LI-0991A&B, instrument instrument- ! 1I-0992A,Tt.1-0993A) span) span). f i
8. loss of Power to 4.16 kV ESF Bus I

a. 4.16 kV ESF-Bus N.A. 'N.A. N.A. > 2975 volts >-2912 volts Undervoltage-Loss of Voltage Uith a $ 0.8 Uith a'$ 0.8 -l g second time second time l delay. delay. j

l., I
's b. 4.16 kV ESF Bus N.A. N .A. N.A. > 3146 volts > 3683 volts Undervoltage-Degraded Uith a $ 20 Uith a'$ 20 Voltage. :second time- second time i delay. delay. l t
9. Engineered Safety features ;
Actuation-System Interlocks t
~
a. Pressurizer Pressure, P-11 N.A. N.A. N.A. 5 2000 psig $ 2010 psig (i (PI-0455A,B&C,JPl-0456 & n ij r
] nit-a; PI-0456A, PI-0457 & PI-0457A, $1 sig 5.1 psig ;
PI '1458 & PI -0458A), (Unit 2 Junit2 -; ' b. Reac tor Irip, P-4 H.A. N.A. N.A. M.A N.A. I r 2 4
- . , ~ . - _ , , ,m . .- - -. , ,
3/4.5 EMERGENCY CORE COOLING SYSTEMS
3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPEcATION 3.5.1 Each Reactor Coolant System (RCS) accumulator snall be OPERABLE with

a. The isolation valve coen,

b. (OiiTt k] A contained borated water volume of between 6555 (29.27. of instrufent scan) ana 6909 gallons (70.7% of instrument scan) (L!-0950.
LI-0951, LI-0952, L!-0953. LI-0954, LI-0955, LI-0956. LI-3957), t 2: - A co ined Dorate ater volume o't tween 6616 ' 7. o f k ins ment scan) a 6854 gallons 4*'. of instrum scan) (LI-v 0, lLI-095 , '-0952, LI-v.' LI-0954. t 0955. LI-0956. .1-0957),
c. A boron concentration of between 1900 :pm ano 2600 com, Ana

d. A nitrogen cover-pressure of between 617 and 678 psig. (PI-0960A&B, PI-0961ALB, PI-0962A&B, PI-0963A&B PI-0964A&B, PI-0965A&B, PI-0966A&B, PI-0967A&B)
APPLICABILITY: MODES 1, 2, and 3' ACTION:
a. With one accumulator inoperable, except as a result of a closed Isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANOBY within the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours.

b. Hith one accumulator inocerable due to the isolation valve being closed, either immediately open the isolation valve or ce in at least HOT STANDBY within 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours.
SURVEILLANCE REOUIREMENTS 4.5.1.1 Each accumulator snall be demonstrated OPERABLE:
a. At least once per 12 hcurs ty:

1) Verifying the contained borated water volume and nitr0 gen cover-pressure in tae tanks. and

2) Verifying that eacn accumulator isolation valve is oren (HV-8808A, S C, 0).
D) (,
Pressurizer pressure above 1000 psig.
V0GTLE UNITS - 1 & 2 3/4 5
BORON IN3ECT10N SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION I 3.5.4 The refueling water storage tank (RWST) snall be OPERABLE with;
a. A minimum contained borated water volume of 631,478 gallons (86% of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A),

b. A boron concentration of between 2400 ppm and 2600 ppm of boron,

c. A minimum solution temperature of 44'F bun ' L - U" .Jni; 2h and ,
i
d. A maximum solution temperature of Il6'F (TI-10982).
e, RHST Sludge Mixing Pump Isolation valves capaole of closing on RWSi low-level. ; APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: a, With the RHST inoperable except for the Sludge Mixing Pump Isolation Valves, restore the tank to OPERABLE status'within I hour or be In at A least HOT STAN0BY within 6 hours and in COLD SHUTDOWN within the V following 30 hours. D. With a Sludge Mixing Pump Isolation Valve (s) inoperacle, restore the " valve (s) to OPERABLE status within 24 hours or isolate the sludge mixing system by either closing the manual isolation valves or deenergizing the
.0PERABLE solenoid pilot valve within 6 hours and maintain closed, SURVEILLANCE REOUIREMENTS 4.5.4 The RHST shall be demonstrated OPERABLE:
a. At least once per 7 days by:

1) Verifying-the contained berated water volume-in the tank, and

2) Verifying the boron concentration.of tne water.

b. At least once per 24 hours by verifying the RWST tamaerature unen tne curside air temperature is less than 40'F O ~
. ::^O
[#
~
23
c. - At least once per 18 months by verifying that the sludge mixing pumo isolation valves automatically close upon an RHST low-level test signal. -
V0GTLE UNITS - 1 & 2 3/4 5 10
.m _ .. m_ _ _ _ _ _ _ _ _ . . _ . _ . - . . . _ _ . . . _ . _ . _ . - _ . _ _ _ . _ _ _ _ . _ _ . _ -
t k.O[ AFVG LAtiLL iU Uh1I l ONL 3/a .' ? POWERDISTRIBUTIONLIMITShU$i BASES-The specifications of this section provide assurance of fuel _ integrity during Condition I (Normal Operation) and_ Il (Incidents of _ Moderate Frequency) events by: -(l) meeting the DNB design criterion during normal operation and l in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design c ri te ria .- In addition, limiting the peak linear power density during Condition I eventsLprovides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria. limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these' specifications are as follows: L FQ (Z) Heat Flux Hot Channel Factor is defined as the maximum local heat
flux on-the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on-fuel pellets and rods; and l-F$H Nuclear-Enthalpy Rise Hot Channel Factor is defined as the ratio of the integral of linear power along the rod with the highest integrated l -. 4 . power to the average rod power.
3/a;2.1 AXIAL FlVX DIFFERENCE The limits on AX1AL FLUX DIFFERENCE (AFD) assure that the FQ (Z). upper bound envelope of.the FQ limit specified in the CORE OPERATING LIMITS REPORT (COLR) timescK(2) is not exceeded during either normal operation'or in-the event of xenon redistribution following power changes, l Provisions for monitoring the AFD on an automatic basis are derived from' the plant process computer through the AFD'Honitor. Alarm. The= computer deter-
' mines the l_-minute: average;of each of the OPERABLE excore detector outputs _and provides an alarm message'imediately if the AFD for two or more OPERABLE excore channels are outside the allowed. Al power. operating space for.RAOC operation specified within the COLR and the _ THERMAL POWER is greater _.than 50% of. RATED THERMAL POWER, i
O
~
V0GTLE-UNITS - 1 & 2 B 3/4 2-1 l=. ._ a _= - - - --
-- ..~ . .
buisPAGEAPP1;CA0l:.10UM1.UnQ POWER DISTRIBUTION f.lMITd UM" ] BASES' AXIAL FlVX DIFFERENCE (Continued) 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR-AND NUCLEAR ENTHALPY RISE HOTCHANNELFACTOR-F$H The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limit on peak local power density is not exceeded, (2) the DNB design criterion is met, and (3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS~ acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:
a. Control rods in a single group move together with no inidivdual rod insertion differing by more than i 12 steps, indicateu, from the group demand position;
.b. Control rod groups are sequenced with a constant tip-to-tip distance between banks as described;in Specification'3.1.3.6;
c. The- control rod insertion limits of Specifications. 3.1. 3.5 and 3.1.3.6 are maintained; and
~d. The axial power distribution, expressed in terms of AX1AL FLUX.
DIFFERENCE,-.is maintained within the limits. F$Hwillbemaintainedwithin'its'limitsprovidedConditionsa,through _
d. above are maintained. TherelaxationofF!Has'afunctionofTHERMALPOWER~
allows changes in the' radial power shape for all permissible rod insertion U limits. i When an QF measurement is taken, an al.lowance for both experimental error and manufacturing tolerance must;be made. An allowance of.5% is. appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance:iscappropriate for manuf acturing tolerance; The heat flux-hot channel factor F Q(Z).is. measured periodically and increased by a cycle and height dependent. power factor appropriate to:RAOC operation, W(2), to provide assurance that the' limit on the'. heat flux l hot channel factor, Fn(Z),.-is' met, W(2) accounts'.for the effects of normal operation transients within the AFD band ~and was determined from expected power control maneuvers -over th.: full range of burnup conditions in the core, TheW(Z) function for normal. operation'and the AFD band are provided in the CORE OPERATING LIMITS REPORT per Specification 6.8.1.6. V0GTLE UNITS - 1 & 2 B 3/4 2-2
l L., m . ,. . _ _ . . . . -
,,m,, ,2 O m - . .
u-vow om e rt m u. mi POWEROISTRIBUTIONLIMITS{-UN!T-O BASES l HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) WhenF$Hismeasured,(i.e., inferred),measurementuncertainty(i.e., the appropriate uncertainty on the incore inferred hot rod peaking factor) must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. 3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x-y plane power tilts. A A limit of 1.02 was selected to provide an allowance for the uncertainty Q associated with the indicated power tilt. The 2-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 control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reduc %'; the. maximum allowed power by 3% for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-S , E-11, H-3, H-13, L-5, L-11, N-8. 3 /4 . 2. 5 DNB PARAMETERS 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. meet the ONB design criterion throughout each analyzed transient. The indicated Tavg value of 592.5'F and the indicated pressurizer pressure value of 2199 psig correspond to analytical limits of 594.4*F and 2185 psig respec-
. tively, with allowance for measurement uncertainty, b
V0GTLE UNI 15 - 1 & 2 B 3/4 2-3
h5 PAGE UFi.ic ABLE ;Q uisii i U:iLD POWER DISTRIBUTION I,1MITS_ iiii7 BASES 3/a.2.5 DNB PARAMETERS (Continued) The 12-hour periodic surveillance of these parameters through 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 f-low indication channels with measured flow such that the indicated percent flow will provide sufficient verification of the flow rate degradation on a 12 hour basis, A change in indicated percent flow which is greater thOn the instrument channel inaccuracies and parallax errors is an appropriate indication of RCS flow degradation, I, g V0GTLE UNITS - 1 & 2 8 3/4 2-4 -
I l THIS PAGE APPLICABLE TO UNii 2 ONLY [] 0/4.2 POWER 01STRIBUT10N LIMITS - UNIT 2 BASE
\ / _
The 5 cifications of this section provide assurance of fuel integrity during Cond ion I (Normal Operation) and 11 (Incidents of oderate Frequency) events by: ( maintaining the minimum ONBR in the core reater than or equal to 1.30 during ormal operation and in short-term trans nts, and=(2) limiting the fission gas lease, fuel pellet temperature, and ladding mechanical properties to with n assumed design criteria. In ad tion, limiting the peak linear power densit during Condition I events prov ies assurance that the initial conditions as umed for the LOCA analyses a e met and the ECCS acceptance
, criteria limit of 2200 is not exceeded.
The definitions of c tain hot channel an peaking factors as used in these specifications are as follows: Fg(2) Heat Flux Hot Channel actor, is d fined as the maximum local heat flux on the surface of fuel roi at core elevation 2 divided by the average fuel rod heat f.1 , all Wing for sanuf acturing tolerances on fuel pellets and rods;
.O F$H Nuclear Enthalpy Rise Hot Ch nnel Factor, is defined as the ratio of long the rod with the highest integrated
(/ the integral of linear pow power to the average rod wer; and Fxy(2) RadialPeakingFactor,/sdefined s the ratio of peak power density to average power dens ty in the hor gontal plane at core elevation 2. 3/4.2.1 AXIAL FLUX DIFFER NCE The limits on AX1AL LUX DIFFERENCE (AFD) assur 2) upper bound envelope of the 0 limit specified in the CORE that the Fg(IMITS PERATING L REPORT (COLR) times K(2) is ot exceeded during either normal operation or in the event of xenon redi ribution following power changes. Target flux f f erence is determined at equilibrium xe n conditons. The rods may be ositioned within the core in accordance wit their respective insertion lim s and should be inserted near their normal pos ion for steady-state operat n at high power levels. The value of the target lux difference obtained u er these conditions divided by the fraction of RATED HERMAL POWER is the tay'get flux dif f erence at RATED THERMAL POWER for the assoc ated core burnup cpnditions. Target flux differences for other THERMAL POWER evels are obtained by multiplying the RATED THERMAL POWER value by the appropri te f ract The periodic updating of the target f ux dif'e/onalTHERMALPOWERlevel. rence value is necessary to reflect core burnup considerations. O) L V0GTLE UNITS - 1 & 2 , l
-l .\-
p ,
\
THIS PAGE APPLICABLE TO UNIT 2 ONLY POWER'0!STRIBUT10N LIMITS - UNIT 2 BASES-
~
AXI Ats FLUX DIFFERENCE (Continued) A ough it is intended that the plant will be operated with the AFD within t target band required by Specification 3.2.1 about the target flux difference during rapid plant THERMAL POWER reductions, control. rod motion. will cause fe AFD to deviate outside of the target band at reduced THERMAL POWER levels \ This deviation will not affect the xenon redistribution suffi-ciently to chahpe the envelope of peaking factors which may be reached on a subsequent returg to RATED THERMAL POWER (with the AFD within the target band) providedthetime\durationofthedeviation-islimited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operationoutsideohthetargetbandbutwithinthelimitsspecifiedinthe COLR while at THERMAL \ POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outsido of the target band are les's significant, The penalty of 2 hours actual time reflects this reduced significance. ,
\ /
Provisionsformonitorhg.theAFDonanautomaticbasisarederivedfrom-the plant process computer thzough the/AF0 Monitor Alarm. The computer-deter-A mines the 1-minute average of each of the OPERABLE excore detectc* outputs 'and g provides an. alarm message immed k tely if the AFD for two-or more OPERABLE ' , excore channels are outside the t'a,rget band and the THERMAL POWER is greater l than 90%-of RATED THERMAL POWER / Iluring operation at THERMAL POWER levels
.between 50% and 90% and between'15% and \ 50% RATED THERMAL POWER, the computer outputs'an alarm message when the pedalty deviation accumulates beyond the limitsof'lhourand2 hours",respectihely. e Figare _B 3/4 2-1 shews a typical monthly target band. '3/4.2.2'ano 3/4.2.3 /EHA FLUX \ AND NUCLEAR ENTHALPY RISE T HOT CHANNEL FAITOR HOTCHANNEL-FACTOR'-F!H The limits /on heat flux hot channel factor and nuclear enthalpy rise hot:
channel f actor' ensure that

(1) the design limits \on peak local power density l -and minimum f 0NBR are not-exceeded and (2) in the eveqt of a LOCA:the pu k-fuel
- clad temperature will not exceed the 2200*F ECCS acce tance criteria limit.
L [ Each of these is measurable but will normally only be determined periodically as-specified in Specifications 4.2.2 and 4.213. This periodic surveillance is sufficient to ensure that the limits ar_e ma.intained provided:
, a. Control rods in a single group move together with.ndsindividual rod insertion dif fering by more than i 12 steps, indicated, f rom the
_ group demand position; j e
b. Control rod groups are sequenced with a constant tip-to-t'ip distance l , -between banks as described in Specification 3.1.3.6; l V0GTLE UNITS - 1 & 2 / 5 M 2 'l ,
q THis PAGE APPLICABLE TO UNIT 2 ONLY l G 1.00 1 i' I l 0.90 i / 0.8 i l i / 0.70
\ \ /
{ Targ/ Flux 0 p 0.60 -- s f l i f [ Di rence g \ lA l g 0.50 --
\
N. i / 9 n <
'V i \l,,-
0 0.40 8 /gI' ig 0.30 7
/ ! \ g 1
0.20 , 1 N 0.10 -
- \i \ g i l 'l 0 .$' 20 10 0 +10 +20 30 l
INDICATED AXIAL FLUX DIFFERENCE (percent) f)
'd FIGURE B 3/4 2-1 -
TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THER.'1AL POWER V0GTLE UNITS 1 & 2 /4 : 7
.n . . . -- - . _ _ . ~ _ - - - - - . - - . - - . - - . _ _ . - _ . - . - . - - ,
.THIS PAGE' APPLICABLE TO' UNIT 2 ONLY O PJWER Di$TRIBUT10N LIMITS - UNil 2 /
7 BASES
\ /
HEAT FLUX H0 CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHAN L FACTOR (Continued)
c. :The cont pl. rod insertion limits of Specifications .l.3.5 and 3.1.3.6 af maintained; and <

d. The axial po r distribution, expressed in te s of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F$Hwillbemaintaine within its' limits prov ed. Conditions a. through.
-d. above are maintained.. Th relaxationofF$H a function of THERMAL POWER allows changes in the radial wer shape for al permissible rod insertion limits.
When an Fg measurement is ta en, an al owance for both experimental error and manuf acturing tolerance must b made. An allowance of 5% is appropriate for a full core map taken with the cor detector flux ~ mapping system and a 3%-allowanc6 is appropriate for manuf uring tolerance.
. WhenF$Hismeasured,(1.e.,i err )',. measurement uncertainty'(i.e.,
the appropriate. uncertainty on th incore nferred hot rod peaking factor) must. be allowed for and '4% is th appropria e allowance for a full core map
~taken with.the incore detection system.- -
Fuel rod, bowing reduces the value of DNB r tio. Credit'is.available to
-offset-this reduction in t generic margin. Th generic margins, totaling -
9.1%'ONBR. complete'y offs any rod bow penalties. .
This margin includes the-following:-
a. ' Design limit NBR of 1.30'vs 1.28,
~
b. Grid Spac g;(K5) .of.0.046 vs 0.059,

c. . Therma Diffusion Coefficient of 0.038 vs 0.059,

d. DNB Multiplier of-.0.86 vs 0.88, and

e. tch reduction.
The a p icable values of rod bow' penalties are referenced in the FSA. rO ~ V0GTLE UNITS - 1 & 2 c 0 3/4 2 S1
)
i G THIS PAGE APPLICABLE TO UNIT 2 ONLY V) , i N OWER DISTRIBUTION llMITS - UNIT 2 BAS 5
/ \
HEAT F HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continue The Rad 1 Peaking Factor, Fxy(2), is measured periodically to pr vide assurance that e Hot Channel Factor, QF (Z), remains within its 1 it. The Fxy limit for RATE THERMAL POWER (Fx ) as specified in the C R ,)er Specifi-cation 6.8.1.6 was de rmined f rom expected power control maover uvers e/ the full range of burnup to ditions in the core. 3/4.2.4 00ADRANT POWER T T RATIO N The QUADRANT POWER TILTIATIO limit assures that the radial Dower distribu-tion satisfies the design valuby used in the powerf'apability analysis. Radial power distribution measu ments are made d ring STARTUP testing and periodically during power operati . O The limit of 1.02, at which cor etive a ion is required, provides ONB' and linear heat generation rate prete ion with x-y plane power tilts. A () limit of 1.02 was selected to provide a a owance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for opefatio with a tilt condition greater than1.02butlessthan1.09isprovifedto low identification and correction of a dropped or misaligned control yod. In th event such action does not correct the tilt, the margin for ujicertainty on g is reinstated by reducing the maximum allowed power by 3% or each percent f tilt in excess of 1. For purposes of monitori QUADRANT POWER TILT ATIO when one excore detector is inoperable, the . Veable incore detectors re used to confirm that the normali-M symetric po er distribution is consiste t with the OVADRANT POWER TILT RATIO. The inc re detector monitoring is don with a full incore flux map or two sets of fdur symetric thimbles. The two ets of four symmetric thimbles is a unique se of eight detector locations. The locations are C-8, E-5. E-11. H 13, L-5, L-il, N-8. 3/4.2.5 DNB PARAME ERS The limits on the DNB-related parameters assure that each of the parameters are maintaine within the normal steady-state envelope of operation'\ assumed in the transier}( and accident analyses. The limits are consistent with\che initial FSAs assumotions and have been analytically demonstrated adeahate to maintainjd minimum DNBR of 1.30 throughout each analyzeo transient. Tfte indicat,ed T avg value of 591*F and the indicated pressurizer pressure valNe of 2224 ,d'sig correspond to analytical limits of 592.5'F and 2205 psig respec l) V tiv y, with allowance for measurement uncertainty. V0GTLE UNITS - 1 & 2 2/E
THIS PAGE APPLICABLE TO UNIT 2 ONLY (~} \ POWER JSTRIBUTION ..lMITS - UNil 2 BASES
\ / _
3/4.2.5 ONB PARAME RS (Continued) The 12-hour perio surve ance of these parameters through instrument readout is sufficient to ens hat the parameters are restored within their limits following load ch and other expected transient operation. The 18 month periodic meas ment the RCS total flow rate is adequate to detect flow degradation a nsure corr tion of the . low indication channels with measured flow s c that the '..ijicat percer. flow w'll provide sufficient verificatio f the flow rate degrada on on a 12 hour basis. A change in indicat ercent flow which is greater an the instrument channel inaccuracies andf p rallax errors is an appropriate indi tion of RCS flow degradation. O o) L -
~
V0GTLE UNITS - 1 & 2 2M
j -
-)
EMERGENCY CORE COOLING SYSTEMS BASES
'ECCS SUBSYSTEMS (Continuec)
The limitation for all safety injection pumps to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
.The Surveillance Requirements provided to ensure OPERABILITY of each
component ensure that at a minimum, the assumptions used in the safety-analyses-are met and that subsystem OPERABILITY is. maintained. Surveillance Requirements for throttle valve position stoos and flow balance testing provide assurance-that oroper ECCS flows will be maintained in the event of a LOCA.
Maintenance of procer flow resistance and oressure droo in the piping system to each injection point is-necessary to: (1) prevent total pump flow from-exceeding runout conditions when the system is in its minimum resistance - configuration, (2) provide the proper flow split between injection points t
'in accordance with the assumptions used in the ECCS .0CA analyses,.(3) provide an' acceptable level of total ECCS flow to all injection points equal to or above-that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pumo injection flow'which _is directed through the seal injection path is less than.or equal to the amount assumed in the safety e.nalysis. The LL . surveillance requirements for leakage testing;of ECCS check valves' ensure _a
!' failure of one valve willLnot cause an intersystem LOCA. In MODE 3, with Leither HV-8809A'or B closed for ECCS check valve leak testing, adequate ECCS flow for core cooling _in the event of a LOCA is' assured. 3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY.of the' Refueling Water Storage Tank (RWST) as part of the. ECCS ensures-that sufficient negative reactivity is. injected into the~ core to counteract any positive increase in reactivity caused by RCS cooldown. RCS
-.cooldown can De caused by inadvertent depressurization, a loss-of-coolant' accident, or a steam:line rupture.
The limits on RHST minimum volume and boron concentration ensure that
1) sufficient water is available within containment to permit recirculation cooling flow to the' core,-2) the reactor will remain subtritical.In the cold condition following a small LOCA or steamilne break, assuming complete mixing of the-RHST,..RCS, and ECCS-water volumes with'all control rods inserted except-the most reactive. control assembly-(ARI-1), and 3) the-reactor will remain, subcritical in the cold condition following a large break LOCA assuming L ' complete mixing of-the RHST, RCS, ECCS water and other scur m of water-Ethat'may eventually reside 'in the sump, post-LCCA with fA;' scd r0c: 7 In:= R h w w ; :. M all_ control rods inserted except for the two "most reactive control assemoltes Fun't '4 O :The contained water volume limit includes an allowance for water not usable because of' tank discharge line location or other physical characteristics, V0GTLE UNITS - 1 & 2 S 3/4 S-2 n
1 . .- - . - . . , . - . -
+ . - . - . -.- . . . - - - . . - _ . ~ . . = - - - . - _ ~ - . _ _ ,
t ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) Thel Semiannual Radioactive Ef fluent Release Reports shall also include tN following: an explanation as to why the inoperability of liquid or gaseous
~ 'f fluentimonitoring instrumentation was not corrected within the time specified i
in Specification 3.3.3.9 or- 3.3.3.10, respectively; and description of the events -leading to liquid holdup tanks or gas storage tanks exceeding the
= limits of _ Specification 3.11.1.4 or 3.11.2.6, respectively. . MONTHLY OPERATING REPORTS 6.8.i.5 Routine reports of operating statistics and shutdown experience. . including documentation of- all challenges to the PORVs or. safety valves, .
shall- be submitted on a monthly basis to the Director, Of fice of Resource Management, U.S. . Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Of fice' of the NRC, no later than the:15th of 'each month f ollowing the calendar month covered by the report. CORE OPERATING' LIMITS REPORT - Uu!T ' 6.8.1.6- Core operating limits shail be established and documented in the CORE OPERATING LIMITS _ REPORT (COLR) before each reMad cycle or any remaining part. of a reload. cycle for the following:
a. SHUT 00WN MARGIN' LIMIT FOR MODES l_ar.d 2 for Specification 3/4.1.1.1,

b. -SHUT 00WN MARGIN LIMITS FOR MODES 3, 4, and 5 for_ Specification 3/4.1.1.2,

c. Moderator temperature coef ficient- BOL and EOL limits and the 300-ppm surveillance limit for Specification 3/4.1.1.3,

d. Shutdown Rod Insertion. Limit for Specification 3/4.1.3.5,

e. ' Control Rod Insertion -Limits f or Specification.3/4.1.3.6,
.f. Axial' Flux Difference Limits for Specification ~3/4.2.1, l
g. Heat Flux Hot Channel Factor, K(Z) and W(Z) for Specification 3/4.2.2, l' -h. Nuclear Enthalpy Rise Hot Channel- Factor Limit and the Power Factor. Multiplier for Specification 3/4.2.3..
The analytical methods used;to_ determine the core operating-limits shall-be
.those previously' approved by the NRC in:
O V0GTLE UNITS - 1 & 2 6-21 l
, _ _ _ . _. __. - _ _ _ _m .- . _._..-__m__ ._ ' ADMINISTRATIVE CONTROLS /~N .COREOPERATINGLIMITSREPORT(Continued)[UN! ) l 4
a,-lWCAP-9272-P-A " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (M Proprietary)' . (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient,'3.1.3.'S - Shutdown Bank Insertion Limit,=3.1.3.6 - Control Bank Insertion limits, and 3.2.3 - Nuclear Enthalpy Rise l
' Hot Channel Factor.)
b. WCAP-10216-P-A. " RELAXATION OF CONSTANT AXI AL OFFSET CONTRCL Fg SURVEILLANCE TECHNICAL SPECIFICATION "' June 1983 (W Proprietary).
(Methodology for Specific;tions 3.2.1 - Axial Flux Dif ference (Relaxed Axial Offset Con,"l) and 3.2.2 - Heat Flux Hot Channei Factor (W(Z) surveillance requirements for FQ Methodology).).
c. WCAP-9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION M00EL-1981 VERSION " February 1982 (W Proprietary).
-(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits ECCS limits, nuclear-limits such as shutdown margin, and transient and accident analysis' limits) of the safety analysis'are met, t The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto,' shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk-with copies to the Regional Administrator and Resident Inspector. . RE OPERATING LIMITS REP 0tti - UNIT 2-6.8 the CORE OPERATI Coreoperatinglimitsshallbeestablishedanodocumentedi"ingpart LIMITS. REPORT (COLR) before each reload. cycle or any remain of.a relo le for the following:
a. SHUT 00 RGIN limit for MODES l'and 2 for Specif ation 3/4.1.1.1,

b. SHUTDOWN MAlHilN limits for MODES 3, 4,' and 5 r Specification 3/4.1.1.2,~

c. Moderator temperatu -coefficient 80L y d EOL limits and the 300-ppm surveillance it f or Spec rication 3/4.1.1.3,

d. -Shutdown Rod: Insertion Lim r Specification 3/4.1;3.5,

e. Control Rod Insertion Lim ts forN[pecification 3/4.1.3,6, I

f. : Axial Flux Different imits, and tar t band for Specification 3/.4.~ 2.1, 9 Heat Flux HotAhannel Factor, K(Z), the Power' actor Multiplier and Ffy)fo DIcification3/4.2.2,

h. Nucle.ar Enthalpy Rise Hot Channel Factor Limit and th'e Power.
Fac<t'or Multiplier for Specification 3/4.2.3. b V The ana(ytical methods used to determine the core operating limits sht 1 be thove'previously approved by the NRC in:
.1 '
V0 GILE UNITS - 1 & 2 6-21a
l
. ADMINISTRATIVE CONTROLS - q RE OPERATING LIMITS REPORT (Continued) - UNIT 2
a. . WCAP-9272-P-A.._" WESTINGHOUSE RELOAD SAFETY EVALUATION METHODO GY,"
July 1985-(W Proprietary). ~ y (Methodology for Specification 3.1.1.3 - Moderator Tempe ture l oefficient 3.l'.3.5 - Shutdown Bank Insertion Limit,- ,1.3.6 - l C trol' Bank Insertion Limits, 3.2.1 - Axial Flux -01 erence','3.2.2 3
-H t Flux Hot Channel Factor, and 3.2.3 - Nucle Enthalpy Rise i Hot C. nnel Factor.) . ,
b. WCAP-8385, " POWER DISTRIBUTION CONTROL AND AD-FOLLOWING PROCEDURES
- TOPICAL R PORT," September 1974 (W Prop etary).
(Methodology r Specification 3.2.1 - . ial Flux Difference (Constant Axial ffset Control).) I i c, .T. M. Anderson to . Kniel (Chief f Core Performance Branch, NRC) i January 31, 1980 -- A tachment: Operation and Safety Analysis Aspects l { of an Improved Load Fo ow Pa age. ' (Methodology for Specifi t n 3.2.1 - Axial Flux .Dif f erence j (Constant Axial Offset Co 01).) {
d. NUREG-0800 Standard view Pihn, s
U. S. Nuclear Regulatory l Consnission, Section .3, NuclearNDesign, July 1981. Branch i Technical Position P8-4.3-1, Westinghouse Constant Axial Offset ! Control-(CAOC), v.2, July 1981. t (Methodology f -Specification 3.2.1 - Axial Flux Difference !' (Constant Ax l-Offset Control).)
e. WCAP-9220 -A,-Rev.- 1 " WESTINGHOUSE CCCS ALUATION MODEL-1981 VERSION February 1982 '(W Proprietary). . !
(Metho ology for Specification 3.2.2 - Heat F x Hot Channel Factor.) l The core op ating limits shall be-determined so that all plicable limits
.(e.g , fue thermal-mechanical limits, core thermal-hydraul limits, ECCS j' limits, clear limits such as shutdown margin, and transient d accident l- j analysi limits) of the. safety analysis are met. l l The- RE OPERATING LIMITS REPORT, including any mid-cycle revisions f' su lements thereto,'shall be provided upon issuance, for each reload cle, 3 Et .the NRC-Document Control Desk with copies to the Regional.Administrat.
nd Resident Inspector. i l. SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional 0ffice of the'NRC within the time period specified for each report. t l: [ { V0GTLE UNITS'- 1 & 2
-.tv?)
Attachment 2b Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE-5 Fuel Design Technical Soecifications Tvoed Paaec t Effective following the Vogtle 2 Cycle 2 Shutdown p (Effective as of Vogtle 2 Cycle 3 Startup) ( I l l l l l-i
IN0tl ,m $AFtTV timit$ AND LIMITING $AffTY $Y$T[M_f(TTING$ _ [ v
$tCT10N g 2.1 $AFtTY L1=lVS .................... .... ... 2-1 2.1.1 REACTOR C0Rt................
2.1.2 REACTOR COOLANT $YSTEM PRI$$URE................... . ... 2-1 flGURt 2.1-1 REACTOR COR! $AFETY LIMIT................ .... ... 22 2.2 LIMll,1NG SAFtTY SYSTEM stTTING} 2.2.1 REACTOR TRIP $Y$ftM INSTRUMENTAT!DN $tTP0!NT$............... 23 2 -4 TABLE 2.2-1 REACTOR TRIP $YSTEM INSTRUMENTAT!DN TRIP $tTPOINT$...
!!tts $[CTION 2.1 SArtTY LIMITS J.1.1 REACTOR C0RE................................................
82-1 o B 2-1 2.1.2 REACTOR COOLANT $Y$ TEM PRt$$URE............................. tlM1 TING SAftTY SYSTtM stTTING$ 2.2.1 Rf ACTOR TRIP $Y$f tM IN$f RUMENT Afl0N $lTP0lNT$, . . . . . . . . . . . . . . 123 l THl$ FAGE BECOMi$ APPLICA8.t FOLLOWING $ HUT 00WN F't0M UNIT 2 CYCLE 2 OPERATIO l n V0GTLE UNII$ - I & 2 !!!
'1 s
5 ihtLE - LIMITING CONDITIONS FOR OPfRATION AND $URV[lLLANCf Rf0tlIR[M[NT$ , (
$tCTION fill 1/4.2 POW [R DISTRIBUTION LIMIT $
3/4.2.1 A11AL FLUX 0lFFERtNCE..................................... 3/4 2 1 { 3/4,2.2 HEAT FLUX HOT CHANNEL FACTOR - F 0 (!)................... .. 3/4 2-3 3/4.2.3 NUCLEARENTHALPYR!$tHOTCHANNELFACTOR-F$N............ 3/4 2-8 3/4.2.4 OUADRANT POWER TILT RAT 10................................. 3/4 2-10 l 4 ' 3/4.2.5 DNS PARAMEftk$............................................ 3/4 2-13 t 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP $Y$ttM INSTRUM(NTAT10N....................... 3/4 3 1 TABLt 3.3-1 REACTOR TRIP $Y$ttM INSTRUMENTAfl0N.................... 3s*32 T ABLt 4.3-1 REACTOR TRIP $YSTEM INSTRUMENTAT10N $URVE!LLANCE Rt0UIRtMtNT$.............................................. 3/4 3-9 3/4.3.2 ENGINitRED $AftTY FIATURt3 ACTUATION $YSTEM INSTRUMENTATION......................................... 3/4 3-15 TABLt 3.3 2 ENGINttRLD SAftTY FLATUtts ACTUATION $YSTEM INSTRUMENTATION........................................... 3/4 3-1T-TABLE 3.3 3- (NGINEERt0 $AFETY FLATURE$ ACTUAT!DN $YSTEM INSTRUMENT AflDN T RIP $t TPOIN T3. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-28
~/ ' \'~ TABLE 4.3 2 (NGINitRt0 $AFETY FIATutt$ ACTUATION $YSTEM a-
INSTRUMENTATIDN $UkvtlLLANCE REQUIREMENTS................. 3/4 3-36 i 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................. 3/4 3 45 TABLE 3.3-4 RADIATION MONITORING INSTRUMENTAT!DN FOR PLANT OPERATIONS....................................... 3/4 3-46 .l r
TABLE 4.3 3 RADIAT!br 40NITORING INSTRUMENTAfl0N FOR PLANT OPERATIONS $URVEILLANCE REQUIREMENTS...................'... 3/4 3-48 Movable Incore Detectors.................................. 3/4 3-49
Seismic Instrumentation (CDMMON SYSTEM) . . . . . . . . . . . . . . . . . . 3/4'3-50 P
t THl3 PAGE BECOMES APPLICABLE FOLLDWING $HUTODWN FROM UNIT 2 CYCLE 2 CPERATION. l , r ( V0GTI.E UNIT $ - 1 & 2 V i e
-,-,_._..--..._..#._ .._x _ , - - - _ _ . . ~ . _ - - . ._ . - - - __..m --
1!$ll LAlts L A
$[CTION PAGt 3/4.0 APPLICABillTY........... ......... ... ,, . . . . . . . B 3/4 0-1 3/4.1 REACTIVITY CONTROL 3YSTEMS 3/4.1.1 BORAT10N CONTROL.................. ..... . . . . . . . B 3/41-1 3/4.1.2 BORAT10N $Y$7tM$ ........ ....... ..................... . B 3/4 1-2 3/4.1.3 MOYABLE CONTROL A$$tMBLit$......... ...................... B 3/4 1 3 3/4.2 POWE R 01 ST9 t BUT 10N L I MIT $ . . . . . , , . . . . . . . . . . , . . . . . . . . . . .. B 3/4 2 1 3/4.2.1 AX1AL FLUX OlFFERENCE........... .......... . . . . . . . . . . . . B 3/4 2-1 3/4 2.2 and 3/4.2.3 HEAT FLUX HOT CHANNtt FACTOR AND NUCLEAR INTHALPYR!$tHOTCHANNELFACTOR-F!H................. . . B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO. .................... ... ....... B 3/4 2-3 3/4.2.$ DNB PARAMLitR$...................................... ...... B 3/4 2-3 3/4.3 INSTRUMENTAtl0N
,F] \ 3/4.3.1 and 3/4.2.2 REACTOR TRIP $YSTIM and (NGINEERED SAFETY FEATURES ACTUAY10N $YST[M INSTRUMENTATION.................. B 3/4 3-1
%J 3/4.3.3 MONITORING IN$fRUMENTAT10N.......... . .................... B 3/4 3 3 3/4.3.4 TURBlHE OVER$PilD PR0ftCT!0N............................... B 3/4 3 6 I
THis PAGI BECOMES AFPLICABLE f0LLOWING $ HUT 00WN FOR UNIT ; CYCLE 2 OPERATION. I m /' \' YOGTLE UNITS - 1 & 2 XV
, --.m y --- _
l!ilil AtutN15TRAftvt C0kTR0ts
!ICT10N f,,Q(
6.4.2 $AFETY ktvl[W BOARD ($RB)
. ... . .... 6-9 Funttion..... ............... .. ... .
Composition....................... .. . .. . . . ... 6-10 Alternates.... ...... .. ................. . .. .. .. ... . 6-10 Consultants................................... .. .... .. . 6-10 Meeting Frecuency......................... . . .... ....... . 6-10
.......... 6 10 Quorum........................................... ... . . ... . . .. .. 6-11 Review................................
6-11 Auditt.................................. ..... .. . ..... ...
... .... . ... 6-12 Records.......... ............... ...... . ..
6-13 6.6 # t POR T ABLE [*/t NT ACT10N . . . . . . . . . . . . . . . . . . . .. ............. .. SAFETY LIMIT V10LAT10N.............. ................... ...... 6-13 6.6 6.1 P R0t t LUR E S AND P e 0GR AMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13
.kij REPORTING Rt0VIPtMENT5 6-17 6.8.1 ROUTINE RtP0RTS.............................................. $ t a rt up R epo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 Annu a l R e po rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 Annual Radiological invironmental Surveillance Report.. ..... 6-10 Report............... 6-19 Semiannual Radioactive [ff!uent Release Month l y Ope r a t i ng R e po rt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 Core Operating Limits Repo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 6.8.2 $PECIAL REP 0RTS.............................................. 6-21al 6-22 6.9 RECORD RETINT10N...............................................
THis PAGE BECOMt$ APPLICABLt FOLLOWING $HUTDOWN FROM UNIT 2 CYCLE 2 OPE V0GTLt UNITS - 1 & 2 xx!!! 1 . 1
2.0 SAFETY LIMITS AND LIMITING SAFETY SY5 TEM SETTINGS
2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER (NI-0041, NI-0042, NI-0043 NI-0044),
pressurizer pressure (PI-0455A, B&C PI-0456 & PI-0456A, PI-0457 & PI-0457A, PI-0458 & P!-0458A), and the highest operating loop coolant temperature (T. .) (TI-0412, 71-0422. TI-0432. 71-0442) thall not exceed the limits shown in Figure 2.1-1. APPLICABILITY: MODES I and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour, and comply with the require-ments of Specification 6.6.1. R_EACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure (PI-0408, PI-0418, PI-0428, PI-0438) O*
shall not exceed 2735 :sig.
APPL!CABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES I and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.6.1. MODES 3, 4 and 5: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.1.
*Where specific instrument numbers are provided in parentheses they are for information only, and apply to each unit unless specifically noted (to assist in Identifying associated instrument channels or loops) and are not intended to . limit the requirements to the specific instruments associated with the rumber.
O LJ THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNITS - I & 2 2-1 _ . , - - . . _ . . . - . _ , _ ~ __
f'y} v i THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 nPERATION, i 670 g UNACCEPTABLE ' OPERATION g 660 A
% 2440 asia w N s, 650 2250 psia t % w % --
640 , 3 5 % N 630 T I
$ T % 2000 osin D
N h 3 g 620 A N s, \3 3 8 t3 N N m A
\ 'N 8'
m 1935 psin } 000 m ! ACCEPTABLE OPERATION 500 3
~'-
580 1.2 4 .5 .6 .7 .8 .9 1.0 1.1
.0 .1 .2 .3 FRACTION OF RATED THERMAL POWER O FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT V0GTLE UNITS 1 & 2 22
l
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ! 2.2.1 The Reactor Trio System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION:
a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trio Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1. adjust the Setpoint consistent with the Trio Setpoint value,

b. With the Reactor Trip System Instrumentattor, or Interlock Setpoint less conservative than the value f"m., in the Allowable Values column of Table 2.2-1, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of r Table 2.2-1 and determine within 12 hours that Ecuation 2.2 1 was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 2 + R + S f, TA Where: Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel, THIS PAGE BECCMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 2-3
~
! O O O
-TABLE 2.2-1 I <
I 8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
?
TOTAL SENSOR
E ALLOWANCE .ERRoa
!- ~Q FUNCTIONAL UNIT (TA) - Z (5) TRIP SETPOINT Att0WT.8tE VALUE. j .m
l. Manual Reactor Trip N.A. N.A. M.A. N.A. N.A.
f-
" Power Range, Neutron Flux
. 2.
" (Ni-0041B&C NI-0042B&C, l ..
i' NI-0043B&C,'NI-00448&C)
a. High Setpoint. 7.5 4.56 0 $109% of RTPg $111.3% of RIPg i $25% of RTPg.

b. Low Setpoint 8.3 4.56 0 127.3% of Rit?

3. Power Range, Neutron Flux, 1.6 0.50 0 $5% of RTPg with $6.3% of RIPg with
. High Positive Rate a time constant a time constant
l. (NI-0041B&C, NI-0042B&C. 22 seconds 12 seconds
! NI-0043B&C NI-0044B&C)
m.
l E 4. Deleted.
5. ' Intermediate Range, 17.0 B.41 0 $25% of RIPg $31.1% of RIPg
! Neutron Flux. l- (NI-0035B NI-00368)
6. Source Ringe, Neutron Flux ~17.0 10.01 0 $105 cps $1.4 x 105 cps j (NI M 18 NI-00328)

7. Overtemperature AT . 10.7 7.04 1.96 See ?%te 1 See Note 2 {'
(101-4110. T01-421C, + 1.17 I 101-4310, 101-4410) -
8. Overpower a'; 4.3 1.54 1.96 See' Note 3 See Note 4 l

(101 -4118 TDI-4218, l 101-4318. 10I-4418) 4 F
# RIP = IIAllif THERNAL POWER
} THIS PAGE BECONL5 APPLIC.\BLE'FOLLOWING SHUIDOWN FROM UNIT'.2 CYCLE 2 OPERATION. l m - c._ -_ - . . - _ , . , - _ _ _ _
_ y ,r y r\ TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS TOTAL SENSOR E ALLOWANCE ERROR
] [UNCTIONAL UNIT (TA) Z [5) TRIP SETPOINT ALLOWA8tt VALUL
9. Pressurizer Pressure-t.ow 3.1 0.71 1.67 21960 psig** 21950 psig
~
(PI-0455A,B&C, PI-0456 & PI-0456A, PI-0457 & PI-0457A, PI-0458 & PI-0458A)
10. Pressurizer Pressure-High 3.1 0.71 1.67 12385 psig $2395 psig (PI-0455A,BLC, PI-0456 &
PI-0456A, PT-0457 & PI-0457A, PI-0458 & PI-0458A)
11. Pressurizer Water level-High 8.0 2.18 1.67 592% of instrument $93.9% of instrument (LI-0459A, LI-0460A, LI-0461) span span

12. Reactor Coolant Flow-Low 2.5 1.87 0.60 190% of loop 289.4% of loop (LOOPI LOOP 2 LOOP 3 LOOP 4 design flow
design flow

m FI-0414 FI-0424 FI-0434 FI-0444

fI-0415 FI-0425 FI-0435 FI-0445 FI-0416 FI-0426 FI-0436 f i-0446)
13. Steam Generator Water Level 18.5 17.18 1.67 218.5% (37.8)*** 217.8% (35.9)***
Low-Low (21.8)*** (18.21)*** of narrow range of narrew range instrument span Instrument span (LOOP 3 LOOP 2 LOOP 3 LOOP 4 LI-0517 LI-0527 LI-0537 LI-0547 LI-0518 LI-0528 LI-0538 LI-0548 LI-0519 LI-0529 LI-0539 LI-0549 L I--0551 LI-0552 LI-0553 LI-0554)
14. Undervoltage - Reactor 6.0 0.58 0 29600 volts 19481 volts l Coolant Pumps (.0% bus voltage) (691 bus voltage,4

15. Underfrequency - Reactor 33 0.50 0 257.3 Hr 157.1 Hz Coolant Pumps
*toop design flow - 95.700 gpm **1lme constants utilized in the lead-lag controller for Pressurizer Pressure-l.ow are 10 . econds for lead and 1 second for lag. CHANNEL CALIBRATION shall ensure that these time constants are adjust.*d to these values. ***Ihe value st. Ped inside the parenthesis is for instrumentation that has the lower tap at Tevation 333*: the value stated wtside the parenthesis is for instrueentation that has the lower tap at eleva tion 438".
IHIS PAGE 8tCOMES APPLICABLE FOLLOWING SHUIDOWN FROM UNIf 7 CYCtt 2 OPERATION. l
s i ' i TABLE 2.2-1 (Continued) !. 5 I' REACTOR' TRIP S UTEN INSTRUM NTATION TRIP SETPOINTS g TOTAL SENS02- ; i~ ' i ~ Al_LOWANCE ERROR N FUNCTIONat UNIl (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE ! -e t 1-
~ 16. lurbine Trip b
f

ro .a. Low f luid Oil Pressure ~ N.A. M.A. N.A. 1580 psig 1500 psig. j (PT-6161 PT-6162, PT-6163)
. b. Turbine Stop Valve Closure N.A. N.A. N.A. 196.7% open 196.7% open l q. N.A. N.A. N.A. N.A. j 1T. Safety Injection Input from ESF
.N.A.
I
18. Reactor Trip System i i Interlocks'

a. Intermediate Range M.A.. M.A. N.A. 11 x 10 n0 amp 16 x 10-s* a v i l! y Neutron Flux, P-6 j
m - (NI-00358, NI-00368) h b. Low Power Reactor Trips l Block, P-7 j 1) P-10 input N.A. M.A. M.A. 110% of RIPJ $12.3% of RTPJ I (NI-00418&C, NI-0042B&C, [ NI-0043B&C,'NI-0044B&C)-
2) P-13 input N.A. M.A. N.A. $10% RTP# Turbine 112.3% RTPg Turbine j (PI-0505, PI-05%) Impulse Pressure Impulse Pressure i
Equivalent- Equivalent l t
c. Power Range Neutron N.A. N.A. N.A. 148% of RTPJ 150.3% of RTPJ f Flux, P-8 i

(NI-0041B&C, NI-0042B&C, [
f NI-0043B&C. NI-0044B&C) l !' # RIP - Rn1LD THERNAL POWER j i i j THIS PAGE BEC0HLS APPLICABLE FOLLOWING SHUIDOWN FRON UNIT 2 CYCLE 2 OPERA 110N. l [ 1' ; I" '
!. ' ~ T 1
.a TABLE'2.2-1 (Continued) f I' ~8 - .
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
'E 101AL SENSOR l ERROR *'* ALLOWANCE 1- .-e '! " FUNCTIONAL UNIT' (TA) I (S) TRIP SETPOINT ALLOWABLE VALUE
! d. Power Range Noetron Flux, P-9 N.A. N.A. N.A. 550% of RTPg 152.3% of RIPf j
** (NI-0041BE, NI-0042BE, j N NI-0043BE, NI-0044BE)
e. Power Range Neutron M.A. N.A. M.A. 110% of RTPi 17.7% of RTP#
{-
_ Flux - P-10 i (NI-0041BK, NI-0042BE, NI-0043BE, NI-0044BE) s j f. Turbine Impulse Chamber N.A. -N.A. N.A. $10% RIP # Turbine $12.3% RTP# Turbine Pressure, P-13 Impulse Pressure Impulse Pressure I (PI-0505, PI-0506) Equivalent Equivalent
)~ N.A. N.A. N.A. h.A. N.A. j y 19. Reactor 1 rip Breakers i
20. Automatic Trip and Interlock N.A. N.A. M.A. N.A. N.A.
i j Logic j- . I i j 1 i
#RIPT lih1LO 1HLRMAL POWER l
A 2 1HIS PAGL BICOHLS APPLICABLE FOLLOWING~ SHUIDOWN FROM UNI 1 2 CYCLE 2 OPERA 110N. l t
. . . . - ~ .- - . ~ . . . .
n ,-~ i' -
,]
TABLC 2.2-1 (Continued)
'8 S TABLE NOTATIONS I
G c- NOIL 1: OVERIEMPERATURE AT
" I al I * * *SI {I(1 + t ,5 4 (1 + t,5) $ AT o(K,-K,U***I[T[1+1.5i (1 + t,5) 'L - T'] + K,(P - P') - f,(AI)) " Where: AT = Measured aT l ro 4 I * **S = Lead-lag compensator on measured aT; I + r,5 1,, x, = Time constants utilized in lead-lag compensator for AT,1, > 0 s.
22 1 3 s; 1
= Lag compensator on measured AT; 1 + r,5 r, = Time constants utilized in the lag compensator for ai, r, = 0 s; aTo =
Indicated AT at RATED THERMAL POWER: K, 1 1.12; l K, = 0.0224/*F; I I*** = The function generated by lead-lag compensator for Tayg 1 + 1,5 dynamic compensation; r., 1, = Time constar.ts utilized in the lead-lag compensator for Tavg. '. t 28 s, t, 545; 1 = Average temperature 'F; 1
= Lag compensator on measured Tavg; I + v.S 1 = Time constant utilized in the measured T avg Id9 CO*De"5dl0F. '. = 0 s; Tills PAGt IllC0f1LS APPLICABLE FOLLOWING SHUTDOWN FROM UNil 2 CYCLE 2 OPERAll0N. l
O O O TABLE 2.2-1 (Continued) TABLE NOTATIONS (Continued)
] NOTE 1: (Continued) x l T* < 588.4*F (Mominal l avg operating temperature); ] = 0.00115/psig; l K,
P = Pressurizer pressure, psig; y P' = 2235 psig (Nominal RCS operating pre',sure); 5 = Laplace transform variable, s-2; and f,(al) is a function of the indicated difference between top and bottom detectors of the l power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: between -32.0% and + 11.0%, f x(al) = 0, where at and qb are percent RAIED THERMAL l (1) For qt - 4b m POWER in the top and bottom halves of the core respectively, and qt + 4b is total THERMAL 6 POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - Ob exceeds - 32.0%, the AT Trip Setpoint shall be automatically reduced by 3.25% of its valde at RATED THERMAL POWER; and (3) For each percent that the magnitude of qt - Ub exceeds + 11.0%, the AT Trip Setpoint shall be automatit.dly reduced by T.97% of its value at RATED THERMAL POWER. N01L 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.1% of al cpan. l THIS PAGL BECOMES APPLICABLE F0tt0 WING SHUIDOWN FROM UNIT 2 CYCLE 2 OPERATION.
< O O O - j ' TABLE 2.2-1 (Continued 1 ;
l. h. TABLE' NOTATIONS (Continued)L i- g NOTE 3: ' OVERPOWER aT E-

j. h AT .I * ** I I I $ ATo [K. - K, - ** ' I T - K. [T I - T*] - f,(at))
} . (1 + r,5) (1.+ t,5) (1 + 1,S) (1 + 1.5) (1 + 1.5) Where:. aT = Measured AT; l
- 'o l- I * '2$ = lead-lag compensator on meascred AT

j 1 + 1,5 . 1 3, 1, = Time constants utilized in lead-leg compensator j for AT, si 1 8 5, 12 5 3 s; I 1 l = Lag compensator on measured AT; j 1 + t,5 [ ? t, = Time constants utilized in the lag compensator ter AT, l E v3=0s; i AT o - Indicated AT 'at RATED THERMAL POWER;
K, $ 1.08, K, 1 0.02/*F for increasing average temperature and 2 0 for decreasing average j - temperature, 4
= The function generated by the rate-lag compensator for Tay, dynamic l 1 + t ,5 compensation,
[ 1, = Time constants utilized in the rate-lag campensator for Tavg. Y, 2 10 s, I i 1 Lag compensator on measured Tavg;' i 1 + 1.5 !~ THIS PAGL BECOMES APPLICA8LE FOLLOWING SHU1DOWN FROM UNIT 2 CYCLE 2 OPERA 110N. l
= . " - -_ % y,, g e ry-----.y..-e-- .- , , , - + . . . . . . . _ ,
_.,,.c.
n ,- TABLE 2.2-1 (Continued) h TABLE NOTATI9NS (Continued) A NOTE 3: (Continued) g.
q 1, =
Time constant utilized in the measured Tayg lag compensator, 4
= 0 s; ". 1 4 ~
K. > 0.0020/*F f or i > I* and K. = 0 f or T < T*, ! I " T = Average Temperature, 'F; T* = Indicated Tavg at RATED THERMAL POWER (Calibration temperature for a1 instrumentation, < 588.4*F), l 1' S = laplace transfona variable, s-*; and i i f,(AI) = 0 for all al. 4 l ru NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed frip Setpoint Dy more than [ i / 1.9% of AT span. I
~
i i i , j j i THIS PAGE BECOMES APPLICABLE F0lt0 WING SHUIDOWN FROM UNII 2 CYCLE 2 OPERA 110N. l
N 2.1 SAFETY LIMITS BA$f5
, 2.1.1 RE ACTOR CORE The restrictions of this Saf ety Limit prevent overheating of the f uel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented >
by restricting fuel operation to within the nucleate boiling regime where the heat-transfer coef ficient is large and the cladding surf ace temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime 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 measurable parameter during operation and theref ore THERMAL POWER and reactor coolant temperature and pressure have been-related to DNB through correlations which have been developed to predict the l DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to ONB. The DNB thermal design criterion is that the probability that ONB will not occur on the most limiting rod is at least 95% (at 5 95% confidence level) f or any Condition I or !! event. (D in n'esting-the DNB design criterion, uncertainties in plant operating V parameters, nuclear and thermal parameters, f uel f abricat. ion parameters, and computer codes must be considered. ' As described in the FSAR, the ef fects of these uncertainties have been statistically combined with the correlation uncertainty.- Design -limit DNBR values have been determined that satisfy the ONB design criterion.
' Additional DNBR margin is maintained by perf orming the saf ety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.
The curves of Figure 2.1-1 show reactor core safety limits for a range of THERMAL POWER, REACTOR COOLANT SYSTEM pressure, and average temperature which
. satisfy the following criteria:
A. The average enthalpy at the vessel exit is less than-the enthalpy of l saturated liquid (f ar lef t line segment in each curve). B. The minimum DNBR satisfies the DNB design criterion (all the other line segments in each curve).- The VAhiAGE 5 fuel is analyzed using the WRB-2 correlation with design limit ONBR values of 1.24 and 1.23 for the typical cell'and thimble cells, respectively. The LOPAR fuel is analyzed using the WRB-1 correlation with design limit DNBR values of 1.23 and ' 1.22 for the typical and thimble cells, respectively. , C. The hot channel exit Quality is not greater than the upper limit of the I quality range (including the effect of uncertainties) of the DNB correlations. This is not a limiting criterion for this plant. THIS PAGE DEC0l4ES APPLICABLE FOLLOWING SHU100WN FROM UN112 CYCLE 2 OPERATION. [ V0GTLE UNITS - 1 & 2 B 2-1 (
i REACTIVirY CONTROL SYSTEMS ( n BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION l 3.1.2.5 As a minimum, one of the following borated water sources shall be . OPERABLE:
8. A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of 9504 gallons (19%
of instrument span) (LI-102A, LI-104A),
l' 2) A boron concentration between 7000 ppm and 7700 pom, and
3) A minimum solution temperature of 65'F (T!-0103). ,

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 99404 gallons-(9% of instrument span) (L1-0990A&B, LI-0991A&B .LI-0992A, LI-0993A),
2)l A boron concentration between 2400 ppm and 2600 ppm, and () 3) A minimum solution temperature of 44'F (T!-10982). l t
-APPLICABILITY: MODES 5 and 6. .
ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ; ALTERATIONS or positive reactivity changes. SURVEILLANCE REOUIREMENTS , 4.1.2.5 . The above required borated water source shall be demonstrated OPERABLE:
a. At least once per 7 days by:

1) Verifying the boron concentration of the water, .

2) Verifying the contained borated water volume, and

3) When the boric acid storage tank is the source of borated water and the ambient temperature of the boric acid storage tank room (TISL-20902. TISL-20903) is $72'F, verify the boric acid storage
tank solution temperature i: 165'F.
b, At least- once per 24 hours by verif ying the RWST temperature (TI-10982) O when it is the source of borated water and the outside air temperature is-less than 40'F. l THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2- 3/4 1-11 < l= i
. _ _ . . . _ - . _ _ . _ . _ _ _ . . _ , _ - , _ . . . . . . _ . _ _ _,_.___. - . . . _ _ . . . . . _ , - _ . . - _ . , , . _ . ~ _ . , , .
s k REACTIVITY CONTROJ SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1. 2. 6 As a minimum: the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: 1
a. A Boric Acid. Storage Tank with: ]
l
1) A minimum contained borated water volume of 36674 gallons (81%
of instrument span) (L1-102A, L1-104A), i 2)- A boron concentration between 7000 ppm and 1100 ppm, and
3) A minimum solution temperature of 65'F (TI-0103).

b. The refueling water storage tank (RWST) with:
s
1) A minimum contained borated' water volume of 631478 gallons'(86%
of instrument span) (L1-0990A&B L1-0991A&B, L1-0992A, L1-0993A), < CJ 3) A minimum solution temperature of.44*F,
4) A maximum solution temperature of 116'F (T1-10982), and I

5) RWST-Sludge' Mixing Pump Isolation Valves capcble of closing on l l
RWST. low-level. L APPLICABILITY:. MODES 1, 2, 3, and 4. ACTION: i
a. ' With the Boric Acid Storage Tank inoperable and being used as one of the above required borated water sources, restore the tank-to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN as required by Figure 3.1-2 at 200'F; restore the Boric Acid Storage Tank to OPERABLE status within the next 7 days or'be in ;'
COLD SHUTDOWN within the next 30 hours.
b. With the RWST. inoperable, except for the Sludge Mixing Pump ,
Isolation Valves, restore the tank to 0PERABLE status within 1 hour- l or be in at least HOT STAN0BY-within the next 6 hours ard in COLD SHUTDOWN within the following 30 hours.
=
. THIS 'PAGE BECOMES APPLICABLE FOLLOWING SHU100WN FROM-UN11 ? CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 3/4 1-12 1
4 t REACTIVITY CONTROL SYSTEMS ! l LIMITING CONDITION FOR OPERATION (Continued) ! ACTION (Continued) j c. With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the , valve (s) to OPERABLE status hithin 24 hours or isolate the sludge ; mixing system by either closing the manual isolation valves or , , deenergizing the OPERABLE solenoid pil 'alve within 6 hours and - maintain closed. ; SURVEILLANCE RE0VIREMENTS 4.1.2.6 Etch borated water source shall be demonstrated OPERABLE: a.. At-least once per 7 days by:
1) Verifying the boron concentration in the water, 4

2) Verifying the contained borated water volume of the water source, and
\s,, 3) When the boric, acid storage tank is the source of borated water ',
o and the ambient temperature of the boric acid storage tank room (TISL-20902, TISL-20903) is 5 72'F,-verify the bor.ic acid storage - tank solution temperature is >-65'F.
b. At least once per 24 hours by verifying the RWST temperature (T!-10982) when the outside air temperature is less than 40'F. l
-c. At least once per 18 months by verifying that the Sludge Mixing Pump Isolation Yalves automatically close upon RWST low-level test signal, i
L + l l: V THIS PAGL BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 3/4 1-13
1-l
O REACTivlTY CONTROL SYSTEMS i
ROD OROP TIME
i. LIMITING CONDITION FOR OPERATION 3.1. 3. 4 The individual shutdown and control rod drop time from the physical fully withdrawn position shall be less than or equal to 2.7 seconds from l beginning of decay of stationary gripper coil voltage to dashpot entry wi,th:

a. TavU (T1-0412, TI-0422 TI-0432. TI-0442) greater than or equal to 1- $$1 F. and

b. All reactor coolant pumps operating.
APPLICABit.ITY: MODES 1 and 2. h.U.19.8! With the drop time of any 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. rO
i. r SURVEILLANCE RE0VIREMENTS t
4.13.4 . The rod drop t'ime 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 3=
on or modification to the Control Rod Drive System which could af fect the drop time of those specific rods, and
c. - At least.once per 18 months.
1. O -THIS PAGE BECOMES APPLICABLE FOLLOWING SHUT 00WN FROM UNIT 2 CYCt.E 2 OPERATION. l V0GTLE UNITS - 1 & 2 3/4 1-19 1:
.-. . . - - . . - - .. . - . - . . ~ - - . - ~ . - - . - . ~ . . . . . ~ . . . _ - -
f
t 3/4,2 POWER 015TRIBUT10N LIMITS 3/4,2,1 AXfAL FLUX DIFFERENCE ,
LIMITING CONDITION FOR OPERATION , 3.2.1 The indicated (NI-00418, N!-0042B, N!-00438, N!-00448) AXIAL FLUX , DIFFER *NCE (AFD) shall be maintained within the limits specified in the C(,RE OPERATING LIMITS REPORT (COLR). APPLICABILITY: MODE 1 ABOVE 50 PERCENT RATED THERMAL POWER *.
' ACTION!
a. With the indicated AX1AL FLUX DIFFERENCE outside of tre limits specified.in the COLR, [

1. Elther restore the indicated AFD to w' thin the litiits l w thin 1$ minutes, or

2. Retuce THERMAL POWEF. to less than 50% of RATED THIRMAL POWER within 30 minutes and reduce the Power Range Neutton Fhx* - High Trip setpoints to less than or equal to 55 percent of RAT 2D
. THERMAL POWER within the next 4 hours,
b. THERMAL POWER shall not be increased above 505 of P.ATED THERMAt POWER unless the indicated AFD is within the limits specified in the 50LR.
.z o },yRVEILLANCE REOUIREMENTS g 4.2.1.1 The indicated AFD shall be determined to be within its limits during i POWER OPERATION above 50% of RATED THERMAL POWER by
a. Monitoring the indicated AFD for each OPERABLE excore channel:
1). At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and ;
2) At least once per hour until the AFD Monitor Alara is updated after restoration to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoper-
'ble. The logged values of the indicated AFD shall be assumed to l
tist during the interval preceding each logging, r c e provisions of. Specification 4.0.4 are not Lpplicable. 4.2.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside its limits.
*Sce Special Test Exceptions Specification 3.10.2.
THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. l
. V0GTLE UNITS - 1 & 2 3/4 2-1 . . . _ _ . . _ , _ _ . . - _ _ _ . . ., - _ . _ - , _ . . . _ _ ~ . . _- _ --
- ..~... . .. . - - - . . , - - . . - . - - - . . _ - - . - - - . - . . . - . . - . - . - _ .
1 l l i l l i i i l This page intentionally left blank. 1 I l l 1 THIS PAGE BEC0".ES APPLICABLE FOLLOWING SHUTOOWN FROM UNIT 2 CYCLE 2 OPERATION. , V0 GILE UNITS - 1 & 2 3/4 2-2
. - . _ _ . _ . _ _ _ _ _ . _ . _ _ _ . _ _ _ = _ _ _ _ _ . _ . _ ____.__ _ _ +
6 d POWER DISTRIBUTION L1MtTS SURVEILLANCE RE0VIREMENTS 4.2.2.1 The provisions of Specifications 4.0.4 are not applicable. 4.2.2.2 FQ (Z) shall be evaluated to determine if it is within its limit by: l
a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
C
b. Determining the computed heat flux hot channel factor FQ gy), ,,
follows: Increase the measured QF (Z) obtained from the power distribution map by 35 to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties. <
c. Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d. Satisfying the following relationship:
' C RTP Fg gg) pQ x K(Z) for P > 0.5 i P x W(Z) C RTP Fg (Z) FQ , gg7).for P 5,0.5
~
0.5 x W(Z) C RD Where Fg (Z) is obtained in Specification 4.2.2.2b above, Fg is the F0 limit, K(Z) is the normalized F0(Z) as a function of core height, P is the fraction of RATED THERMAL POWER, and r W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. RTP F'g , K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.8.1.6.
e. Measuring FQ (Z) according to the following schedules

1. Upon achieving equilibrium conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined *, or

2. At least once per 31 Effective Full Power Days, whichever occurs first.
"During power escalation af ter each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power O distribution map obtained.
THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTOOWN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNITS - 1 & 2 3/4 2-4
_ _.._._.m. . _ _ _ _ _ _ ..___...___.-___.m._________._ _ . _ _ . . _ . _ _ .
-t I
1' I POWER DISTRIBUTION LIMITS j, l $URVEILLANCE RE0V!REMENTS (Continued) ' 4
f. With measurements indicating 2
maximum I FoC (g) over Z ( K(Z) j has increased since the previous determination of FQ (Z) either of the following actions shall be taken: U
1) ' Increase FQ (2) by 2% and verify that this value satisfies the relationship in Specification 4.2.2.2d, or 4

2) .FgC(Z) shall be measured at least once per_7 Effective Full Power Days until two successive maps indicate that f l C is not increasing.
maximum qFg(Z)
-over Z ( K(Z) /
L O g. With'the relationships spegified in Specification 4.2.2.2d above not being satisfied
1) Calculate the percent FQ (Z) exceeds'its limits by the l following expression:
C-
/ maximum FQ g7) x W(Z) x 100 for P > 0_.5 RTP ,) -4 1 >
x K(2) (overZ
?
l [ maximum FQU(Z) x W(Z) l x 100 for P $ 0.5, and h RTP I
over Z x K(Z)
~I[ \
I 2)' The following action shall be ta# ken: Within:15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by 1% AFD for each percent ' FQ (Z) exceeds its limits as determined in Specification 4.2.2.29 1. . Within 8 hours, reset the AFD alarm setpoints -to these modified limits. 1 O THIS PAGE BECOMES APPLICABLE FOLLOWING 3HUTDOWN FROM UNIT 2 CYCLE 2 OPERATION,
. V0GTLE UNITS & 2 3/4 2-5 4
POWER OIS1RIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued)
h. The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e., 0 - 100%, inclusive.

i. The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1) Lower core region f rom 0 to 15%, inclusive.

2) Upper core region f rom 85 to 100%, inclusive.
4.2.2.3 When QF (Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured Fe(Z) shall be obtained from a power distribution map and increasod by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. 3 (V THIS PAGE BECOMES APPLICABLE FOLLOWING SHUT 00WN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNIls - 1 & 2 3/4 2-6 l
i i i i 1 i I o - 1 l l 1 1 l
. l 4
.I 1 This page intentionally left blank, t I I THIS PAGE BECOMES APPLICABLE FOLLOWING SHUT 00WN FROM UNIT 2 CYCLE 2 OPERATICN. V0GTLE UNITS - 1 & 2 3/4 2-7 i l 1..______...___ _ _ ____.._ ___._ __ _. __ _ _ , _ _ , _ _ _ _ , _ _ _ . _ , _ , _ _ , _ _ _ _ _ _ _ ,_,_ ,,
.. _ - - . . - ~ . . . . - . - ...- - . -- . - ..~ - ..- - . - -
POWER DISTRIBUTION LIMITS 3/a.?.5 DNB PARAMETERS . LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB elated parameters shall be maintained within the limits:
a. Reactor Coolant System Tavg (TI-0412. T1-0427. TI-0432. TI-0442),
5 $92.5'F l t
b. Dressuriter Pressure (PI-0455A.B&C. PI-0456 & P!-0456A, PI-0457 &
PI-0457A, PI-0458 & P!-0458A) > 2199 psig* l
c. Reactor Coolant System Flow (Ft-0414 F1-0415. F1-0416. F1-0424, r'
F1-0425, F1-0426, F1-0434. F1-0435. F1-0436, FI-0444. F1-0445, F1-0446) >. 391.225 gpm** l APPLICABILITY: #400E 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. . SURVEILLANCE RE0VIREMENTS , -4.2.5.1 Reactor Coolant System Tavg and Pressuriter Pressure shall be verified to be within their limits at least once per 12 hours. RCS flow rate shall be monitored for degradation at least once per 12 hours. In the event of flow degradation, RCS flow rate shall be determined by precision heat balance within 7 days of detection of flow degradation. - 4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at each fuel loading and at least once per 18 months. 4.2.5.3 -After each fuel loading, the RCS flow rate shall be determined by precision heat balance prior to coeration above 75% RATED THERMAL POWER. The RCS flow rate snall 6 be detemined by precision heat I balance at least once per 18 months. Within 7 days prior to per-forming the precision heat balance flow measurement, the instrument-ation used for performing the precision heat balance shall be calibrated. The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
Limit not applicable during either a THERMAL POWER ramp in excess of.5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
**Iacludes a 2.2% flow measurement uncertainty. l THIS PAGE BECOMES APPLICABLE FOLLOWING SHUT 00WN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 3/4 2-13
_ _- _ _ .. __ _ . _ . . _~ _ _.__ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ _ - _ _ . . _ _ . . _
o p O !- 1A8LE 3.3-3 (Continued) t !i - <:
-- 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS !
i- a t t r- t !- [ TOTAL SENSOR j E _ ALLOWANCE- . ERROR ; I? Q FUNCTIONAL UNIT (TA) 7 (S) TRIP SETFOINT ALLOWABLE VALUE- i ! " i ) '.7. Semi-Automatic Switchover to ' !
". Containment Emergency Sump (Continued) !
o -
b. RWS1 Level-Low-Low . 3.5 -0.71 1.67 > 275.3 in. from > 264.9 in. from Coincident With Safety tank base tank base. .!

Injection (1 39.1% of (t 37.4% of
! (LI-0990A&B, LI-0991A&B, instrument- instrument l- LI-0992A, LI-0993A)~ span) span) i ! 8. Loss of Power to 4.16 kV ESF Bus
-a.
4.16 kV ESF Bus M.A. N.A. N.A. 2 2975. volts 2 2912 volts
, Undervoltage-Loss of Voltage with a 1 0.8 with a 5 0.8 i g second time second time j delay. delay. i 4 U b. 4.16 kV ESF Bus :M.A. N.A. N.A. > 3746 volts > 3683 volts
{ Undervoltage-Degraded Uith a $ 20 Eith a 5 20 ; i Voltage - seconJ time second time : i delay. delay. I 4 9. Engineered Safety features i l Actuation System Interlocks i I I a. Pressurizer Pressure, P-ll ,M.A. N.A. N.A. 5 2000 psig $ 2010 psig i ? (PI-0455A,8&C, PI-0456 & ! I PI-0456A, PI-0457 & PI-0457A. '[
PI-0458 & PI-0458A),

t l b .~ Reactor Trip, P-4 N.A. N.A. N.A. M.A N.A. !
t 3 THIS PAGE BECONLS APPLICABLE FOLLOWING SHUIDOWN FRON UNIT 2 CYCLE 2 OPERATION. l [ i j .. . L
, ,,, . . , . , . _ . , . . - ~ - . -~ _ _ ._ . - - - - -
4 3/4.5: EMERGENCY CORE COOLING' SYSTEMS 3/4.5.'l -ACCUMULATORS
-LIMITING CONDITION FOR OPERATION
3.5.1 EadhReactorCoolantSystem(RCS)accumulatorshallbeOPERABliwith
a. The isolation valve open, D. A contained berated water volume of between 6555 (29.2% of instrument span) and 6909' gallons (70.7% of instrument span) (LI-0950, LI-0951,- 1 LI-0952, LI-0953, LI-0954, LI-0955, LI-0956, LI-0957), q

c. A boron' concentration of between 1900 ppm'and 2600 ppm, and .
.w
d. A~ nitrogen cover-pressure-of between 617'and 678 psig. (PI-0960A&B, PI-0961A&B,' PI-0962A&B,-PI-0963A&B, PI-0964A&B, PI-0965A&B,.
-PI-0966A&B, P!-0967A&B) -APPLICABILITY: ' MODES 1. 2, and.3*. 1 ACTION:. .,
Ja,' With one accumulator inoperable. Ucept as a result of a closed isolation valve restore the^ inoperable accumulator to OPERABLE status within 1-hour or.be in at least. HOT STANDBY'within the next 6 hours and reduce pressurizer. pressure to less than 1000 psig within the-following 6 hours.
-b. With one accumulator inoperable due to the isolation valve being closed, either-Immediately open the isolation valve'or be in at least HOT STANOBY within-6 hours and. reduce pressurizer. pressure to-less than-1000 psig within the following 6. hours.
, -SURVEILLANCE _RE0VIREMENTS' 3 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
.a. At least once'per 12' hours by 1)~ Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and
2) Verifying that eacn accumulator isolation valve is open (HV-8808A,-B, C, 0).
Pressurizer pressure above 1000 psig.
THIS' PAGE BECOMES APPLICABLE FOLLOHING SHUT 00HN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNITS - 1 & ' 3/4 5-1
l. BORON' INJECTION SYSTEM !
D; 3/4.5.4 REFUELING WATER STORAGE TANK d - : LIMITING CONDITION FOR OPERATION ; i 3.5.4- The refueling water-s.torage tank (RHST) shall be OPERABLE with: a.' A minimum contained borated water volume lof 631.478' gallons (867.~ of
; instrument = span) (LI-0990A&B,'LI-0991A&B, LI-0992A, LI-0993A),
b. A-boron concentration of between 2400 ppm and 2600 ppm of. boron, ;

c. A minimum solution temperature of 44*F, and j-
=d. A maximum solution temperature of Il6'F (TI-10982). .e. .RNST Sludge Mixing Pump. Isolation valves capable of closing on RWST low-level.
6PPLICABILITY: ' MODES 1, 2', 3, and 4. ACTION:
'a. H1'ththeRNSTLinoperable.except.fortheSludge'MixingPump-Isola' tion .
Valves, rettore the tanKO ) 0PERABLE status within .1 hour or be in at t least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the- i O ' following 30 hours. .
.b. With a Sludge Mixing Pump Isolation Valve (s) Inoperable, restore the "
valve (s) to-OPERABLE statustwithin 24 hours or isolate the sludge mixing 1
, system by either closing the manual ~1 solation valves or deenergizing the g <0PERABLE:solenold'pliot valve Within 6 hours and maintain closed. -SURVEILLANCE REGUIREMENTS- -4.5.4h The RHST;shall-be demonstrated OPERABLE:
a. At~least once.per 7 days by:
.g 11 )- Verifying the: contained borated water-volume in the tank,-and-1"-
2) Verifying the boron concentration of the water. e
-b; At_least once per 24 hours by' verifying.the RHST temperature when ';
the outside' air temperature is less.than 40'F. .l
c. At least once per-18 months:by verifying that the sludge mixing pump:
isolation valves automatically close upon an RHST low-level test'
. signal-.
THIS PAGE.BECOMES APPLICABLE FOLLOWING SHUT 00HN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNITS .l & 2 3/4 5-10
. I'. .c . . . - . . . . - - ,. 4 -,m
,a (") , 3 /4,2 POWER DISTRibOTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) meeting the DNB design criterion during normal operation and in short-term transients, and (2) 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 criteria limit of 2200'i is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: FQ (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 manuf acturing tolerances on fuel pellets and rods; and l p F$H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the _ ratio of
j the integral of linear power along the rod with the highest integrated
' power to the average rod power l 3 /4. 2.1 AXIAL FLUX OIFFERENCE , The limitu on AXIAL FLUX DIFFERENCE (AFD) assure that the FQ(Z) upper bound envelope of the FO limit specified in the CORE OPERATING LIMITS REPORT (COLR) times K(Z) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. l Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immedistely if the AFD for two'or more OPERABLE excore channels are outside the allowed al power operating space for RAOC operation specified within the COLR and the THERMAL POWER is greater than 50% of RATED THERMAL POWER. j% b THIS PAiiE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 8 3/4 2-1
. . . . . - . . . _ . . _ _. - . _ . . _ . _ _ . . _ _ m . _ . . _ _ _.
POWER-DISTRIBUTION LIMITS
! BASES -
t AXIAL FLUX OlFFERENCE (Continued) 3/4.2.2-and 3/4.2.3 -HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE
~HOTCHANNELFACTOR-F$H The limits on heat ~ flux hot channel factor and nuclear enthalpy' rise hot .l channel factor ensure that: ~(1) the design limit on peak local power density .is not exceeded, (2) the-DNB design criterion is met, and (3) in the-event of ;
a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance i c riteria - limi t.- Each of these is measurable but will normally only be determined periodically'as specified in Specifications'4.2.2 and 4.2.3._ This periodic surveillance is sufficient to ensure that the limits are maintained provided: a.- Control rods in a single group move together with no inidivdual. rod
insertion dif fering.by more than i 12 steps, indicated, .f rom the - .
group demand position; l
b. Control rod groups are ' sequenced with a. constant tip-to-tip distance
'between banks as described in Specification-3.1.3.6; l-l- c. The control rod insertion limits of Specifications 3.1.3.5 and-l, 3.1.3.6 are maintained; and d.. The axial power distribution, expressed in terms of. AXIAL FLUX -DIFFERENCE,-is maintained-within-the limits.
F!H'wil1Jbemaintainedwithin_itslimitsprovidedConditionsa.~through _
d. above are maintained. TherelaxationofF$HasafunctionofTHERMALPOWER allows changes-in-the radial _ power shape for.all' permissible-rod insertion-limits.
~ ' When an Fg measurement _ is taken, an- allowance-for.both experimental error .and manufacturing tolerance must be made. -An. allowance of-L% is appropriate for a full core map;taken with the incore ~ detector flux mapring system and a 3% _ allowance is appropriate for.manuf acturing . tolerance, l :The: heat: flux hot channel factor FQ(Z)'is measured periodically and increased by a cycle and height dependent power. factor appropriate to RAOC operation W(Z), to provide assurance that the limit on the heat flux hot channel factor,'FQ(Z). is met. W(Z) accounts-for the effects of normal operation transients-within-the AFD band and was determined from expected power-control maneuvers over the full range of burnup conditions in the core._ The'W(Z) function for normal operation and tht,-AFD band are provided in the CORE OPERATING' LIMITS REPORT per Specification 6.8.1.6. -THIS PAGE BECOMES' APPLICABLE-FOLLOWING-SHUT 00WN FROM UNIT 2 CYCLE 2 OPERATION.
V0GTLE UNITS - 1 & 2 8 3/4 2-2
i f Q'{ OWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
-(Continued)
WhenF$Hismeasured.(i.e., inferred),measurementuncertainty(i.e., the appropriate uncertainty on the incore inferred hot rod peaking factor) must be allowed for and 44 is the appropriate allowance for a full core map taken with the incore detection system. j t 3/4.2,4 09ADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and ' periodically during power operation. The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protect. ion with x-y plane power tilts.- A-limit of 1.02 was selected to provide an allowance for the. uncertainty i associated with the indicated power tilt. The 2-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 control rod. In the event such_ action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing- 1 the maximum:al. lowed power by'3% for each percent of tilt in excess of 1. .
r~

For purposes- of monitoring QUADRANT POWER TILT RATIO when one excore ' .
detector is inoperable, the moveable incore detectors are used to confirm that
- the normalized symetric power distribution is consistent with the QUADRANT POWER. TILT _ RATIO. The incore detector monitoring is done with a full incore flux map or two sets of. four symetric thimbles. The two sets of four- symmetric.
thimbles is a unique set of eight detector locations. These locations are-C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. 3/4.2.5 DNB PARAMETERS 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 meet the DNB design criterion throughout each analyzed transient, The indicated Tavg value of 592.5'F and the indicated pressurizer pressure value of 2199-psig correspond to analytical limits of 594.4*F and 2185 psig respec-tively, with allowance for measurement uncertainty.
i- V O i THIS PAGE BECOMES-APPLICABLE FOLLOWING SHUTOOWN FROM UNIT 2 CYCLE 2 OPERATION. V0GTLE UNITS - 1 & 2 B 3/4 2-3 I i- , .
I s;
4 6 ._ POWER O!STRIBUTION LIMITS 4
BASES j 3/4.2.5 DNB PARAMETERS'(Continued)- t The 12-hour periodic . surveillance of these parameters through 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 the ~ flow rate degradation on a 12 hour basis. A~cnange in: -; indicated percent. flow which is greater than the, instrument channel inaccuracies and' parallax errors.is an appropriate indication of RCS flow degradation. z i t r J f r L KO THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTDOWN FROM UNIT 2 CYCLE 2 OPERATION. p V0GTLE UNITS - 1 & 2 B 3/4 2-4
EMERGENCY CORE COOLING SYSTEMS U BASES ECCS SUBSYSTEMS (Continued) The limitation for all safety injection pumps to be inoperable below 350*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system. to each injection point is necessary to: (1) prevent total f,UE fivw from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provice the proper flow spilt between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow which is directed through the seal Injection path is less than or equal to the amount assumed in the safety analysis. The
,o surveillance requirements for leakage testing of ECCS check valves ensure a (j failure of one valve will not cause an intersystem LOCA. In MODE 3, with either HV-8809A or B closed for ECCS check valve leak testing, adequate ECCS flow for core cooling in the event of a LOCA is assured.
3/4.5.4 REFUELING HATER STORAGE TANK The OPERABILITY of the Refueling Water Storage Tank (RHST) as part of the ECCS ensures-that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident, or a steam line rupture. The limits on RHST minimum volume and boron concentration ensure that
1) sufficient water is available within containment to permit recirculation cooling flow to the core, 2) the reactor will remain subcritical in the cold condition following a small LOCA or steamline break, assuming complete mixing of the RHST, RCS, and ECCS water volumes with all control rods inserted except the most reactive control assemoly (ARI-1), and 3) the reactor will remain subcritical-in the cold condition following a large break LOCA assuming complete mixing of the RHST. RCS, ECCS water anc other sources of water that may eventually reside in tne sump, post-LOCA with all control rods inserted except for the two most reactive control assemblies.
The contained water volume limit includes an allowance for water not usable because of tank discherge line location or other physical characteristics.
O THIS PAGE BECOMES APPLICABLE FOLLOWING SHUT 00HN FROM UNIT 2 CYCLE 2 OPERATION. l V0GTLE UNITS - 1 & 2 B 3/4 5-2
s J
' ADMINISTRATIVE CONTROLS ~
f% _ (> q
' SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT lContinur.d)
The Semiannua1' Radioactive Effluent Release Repr,rts shall also include the following: -an explanation as to why the inoperability of liquid or gaseous ef fluent monitoring instrumentation was not corrected within the time spt cified in' Specification 3.3.3.9 or 3.3 3.10, respectively; and description of ine-events leading to liquid holdup tanks or gas. storage tanks exceeding tre slimits Lof Spcification 3.11.I'.4 or 3.11.2.6~, respectively. MONTHLV OPERATING REPORTS 6.8.1.5 Routine reports of' operating statistics and shutdown experience, includi1g documentation.of all challenges to the PORVs or safety valves, shall b submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar. month covered by the report. CORE OPERATING' LIMITS REPORT 6.8~1.6 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT-(COLR) before each reload cycle or any remaining part of a' reload cycle for the following:
~ 'a. SHUTDOWN MARGIN LIMIT FOR MODES 1 and 2 for Specification 3/4.1.1.',
b. SHUTDOWN MARGIN LIMITS FOR MODES 3, 4, and 5 for Specificatim 3 /4.1. l . 2,

c. Moderator temperature coefficient BOL and EOL limits and the 300-ppm surveillance limit for Specification 3/4.1.1.3,,
'd. Shutdown Rod Insertion Limits for Specification 3/4.1.3.5,
e. Control' Rod Insertion Limits for Spe 'ification 3/4.1.3.6

f. Axial Flux Difference Limits for Specification 3/4.2.1,

g. Heat Flux Hot Channel Factor,.K(Z) and W(Z), for Specification. l 3/4.2.2,

h. Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification =3/4.2.3.
The analytical methods used to determine the core operating limits shall be those previously approved by the NRC.in: THIS PAGE BECOMES APPLICABLE FOLLOWING SHUTOOWN FROM UNIT 2 CYCLE l 2 UP V0GTLE UNITS - 1 & 2 6-21
~~ + - - - - , , , ~ - , . , , , . _ , , , , , ,
. ., . . - _ .- -_ .-.- - - .~ . . . _ - . . . . - t RMINISTRATIV" CONTROLS CORE OPERATING LIMITS REPORT (' Continued)
a. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985-(W Proprietary). (Methodology for Specifications 3.1.1.3 - Moderator Temperature
-Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 '
control Bank Insertion Limits, and 3.2.3 - Nuclear Enthalpy Rise l Hot Channel Factor.)
b. WCAP-10216-P-A, " RELAXATION OFJune CONSTANT AXI'AL OFF5ET CO SURVEILLANCE: TECHNICAL SPECIFICATION," 1983 (W Proprietary (Methodology for Specifications 3.2.1 - Axial Flux Dif ference (Relaxed Axial Offset Control) and 3.2.2 . Heat Flux Hot-Channel ,
Factor.(W(Z) surveillance requirements for FQ Methodology).)
c. WCAP-9220-P-A, Rev.1, " WESTINGHOUSE ECCS EVALUATION M00EL-1981 VERSION, " February 1982 (8 Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) The core operating limits'shall be determined so that all applicable limits
-(e.g., f uel thermal-mechanical' limits, core thermal-hydraulic limits, ECCS limits.pnuclear limits such as shutdown margin, and transient and accident-analysis limits).of the safety analysis are met. .The CGNE-OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements'thereto, shall be provided upon issuance, for each reload cycle, 'to the NRC Document Control Desk with copies to the Regional Administrator -and Resident Inspector.
SPECIAL REPORTS 6.8.2 ' Special reports shall-be submitted to the Regional Administrator of the
. Regional Office of the NRC within the time period specified for each report.
THIS-PAGE BECOMES APPLICABLE FOLLOWING SHUT 00WN FROM UNIT 2 CYCLE 2 OPERATION. ] V0GTLE UNITS - 1 & 2 6-21a
Enclosure 4 Vogtle Electric Generating Plant Units 1 and 2 Request for Technical Specifications Changes VANTAGE-5 Fuel Design Safety Evaluation Recort O 1 l-A v l
b TABLE OF CONTENTS Section 1111g gitgg
1.0 INTRODUCTION
AND CONCLUSIONS 2' , 2.0 MECHANICAL EVALVATION 7 3.0 NUCLEAR EVALVATION 17 4.0. THERMAL AND HYDRA 0LIC EVALUATION .19
-5.0' . ACCIDENT EVALVATION 26 6.0
SUMMARY
OF TECHNICAL SPECIFICATIONS CHANGES 48
7.0 REFERENCES
. 50 LIST OF TABLES d Table No. M -Eggg 21 ' Comparison of 17x17 LOPAR and 17xl? VANTAGE 5 Fuel Assembly .9 Design Parameters 4 Thermal and Hydraulic Design Parameters for VEGP Units 1 and 2 '21 6-1 LSummary and Justification for V0GTLE Units 1 and 2 Technical -49 Specifications Changes for VANTAGE 5 Fuel L LIST OF FIGURES Fiaure No. Title Ejtqg-2.1 17x17-' VANTAGE 5/LOPAR Fuel- Assembly Comparison 11 l' 4'494F:6 900926- 1
~
1.0 INTRODUCTION
AND CONCLUSIONS The Vogtle Electric Generating' Plant (VEGP) Units 1 and 2 are currently operating with a Westinghouse 17x17 low parasitic (LOPAR) fueled core. -For subseouent cycles,'it is planned to refuel and operate the VEGP Units 1 and 2 with the Westinghouse VANTAGE 5 improved fuel design. -As a result, future core loadings would range from approximately 55% LOPAR, and 45% VANTAGE 5 i transition cores to eventually an all VANTAGE 5 fueled core. The VANTAGE 5 ; fuel . assembly is designed as a modification to the current 17x17 LOPAR (standard fuel) and the Optimized fuel Assembly (OFA) designs, Reference 1. The VANTAGE 5 design features were conceptually pcckaged to be licensed as a single entity. This was accomplished via the NRC review and approval of the
" VANTAGE 5 Fuel Assembly Reference Core Report," WCAP-10444-P-A, Reference 2.
The initial irradiation of a fuel region containing all the VANTAGE 5 design features occurred in the Callaway Plant in November 1987. The Callaway VANTAGE 5 licensing submittal was made to the NRC on March 31, 1987 (ULNRC1470,DocketNo.50-483). NRC approval was received in October 1987. Several of the VANTAGE 5 dasign features, such as_. axial blankets, reconstitutable top nozzles, _ extended burnup modified fuel assemblies and integral fuel burnable absorbers have been successfully licensed as individual design features and are currently operating'in Westinghouse = plants. The VEGP Units 1 and 2 will be operating in reload Cycles 4 and 3, respectively, with M fuel containing the following features:' ' Integral Fuel Burnable Absorbers (IFBAs),' Intermediate Flow Hixers -(IFM) grids,-Reconstitutable Top Nozzles (RTN), (currently operating in both VEGP Units 1 and 2), and. fuel assemblies modified for extended burnup (currently operating in both VEGP Units 1 and 2). In' addition, both the VEGP Units 1 and 2 fuel assemblies are currently-operating with the Debris Filter Bottom Nozzle (DFBN). _The axial blankets are- _ optional'on the first transition _ Cycle, i.e., Vogtle Unit 1, Cycle 4 and Vogtle Unit 2, Cycle 3.~ A brief summary of the VANTAGE 5 design features and the major advant%3s of the improved fuel design are given below. These features and /igures illustrating the VANTAGE 5 design are presented in more deta'.1 in Section 2.0. 4494F:6 900926 2
O t
V'i Intearal Fuel Burnable Absorber (IFBA) - The IFBA features a thin boride coating on the fuel pellet surface in the central portion of the enriched UO2 pellets. In a typical reload core, approximately forty percent of the fuel rods in the feed region are expected to include IFBAs. IFBAs provide power peaking and moderator temperature coefficient control. Intermediate Flow Mixer (IFM) Grid - Three IFM grids located between the four upper most zircaloy grids provide increased DNB margin. Increased margin permits an increase in the desigt basis F6H and Fg . Reconstitutable Too Nozzle (RTN) A mechanical disconnect feature facilitates the top nozzle removal. Changes in the design of both the top and bottom nozzles increase burnup margins by providing additional plenum space and room for fuel rod growth. Extended Burnuo - The VANTAGE 5 fuel design will be capable of achieving extended burnups. The basis for designing to extended burnup is contained in
) the approved Westinghouse topical WCAP 10125 P A, Reference 3.
Blankets - The axial blanket consists of a nominal six inches of natural U02 pellets at each end of the fuel stack to reduce neutron leakage and to improve uranium utilization. For VANTAGE 5 reload cores, low leakage loading patterns (burned blankets) are shown to further improve uranium utilization and provide additional pressurized thermal shock margin. This submittal is to serve as a reference safety evaluation / analysis report for the region by-region reload transition from the present VEGP LOPAR fueled core to an all VANTAGE 5 fueled core. The submittal examines the differences between the VANTAGE 5 and LOPfR fuel assembly designs and evaluates the effect of these differences on the core performance during the transition to an all VANTAGE 5 core. The VANTAGE 5 core evaluation / analyses were performed at a core thermal power level of 3565 MWt for Units 1 and 2 with the following conservative assumptions made in the safety evaluations: a full power F6H of 1.65 for the VANTAGE 5 fuel and 1.57 for the LOPAR fuel, and 1.70 both for the VANTAGE 5 fuel non-LOCA analyses and the small bret.k LOCA
\
4494F:6-900926 3
lh(D analysis, an increase in the maximum Fg to 2.50,'107. plant total steam generator tube plugging for both Units 1 and 2, and a core bypass flow of 8.4Y. with thimble plugs removed. The analysis assumption of core bypass flow with thimble plugs removed is conservative for operation with thimble plugs and/or Wet Annular Burnable Absorber (WABA) rods. The axial offset strategy will be the licensed Relaxed Axial Offset Control (RAOC) with Fg surveillance. RAOC uses a +10/ 20% Axial Flux Difference (AFO) band at 100% RTP for the safety evaluations. The standard reload design methods described in Reference 4 will be used as a basic reference document in support of future VEGP Units 1 and 2 Reload Safety Evaluations (RSE) with VANTAGE 5 fuel reloads. Sections 2.0 through 5.0 summarize the Mechanical, Nuclear, Thermal and Hydraulic, and Accident Evaluations, respectively. Section 6.0 gives a summary of the Technical Specifications changes needed. Consistent with the Westinghouse standard reload methodology, Reference 4, parameters are chosen to maximize the applicability of the safety evaluations v for. future cycles. The objective of subsequent cycle specific RSEs will be to . verify that: applicable safety limits are satisfied based on the reference evaluation / analyses established in this RTSR. In order to demonstrate early performance of the VANTAGE 5 design product features i.n a commercial reactor, four VANTAGE 5 demonstration assemblies (17x17) were loaded into the V. C. Summer Cycle 2 core and began power prodiaction in December of 1984. These assemblies completed one cycle of irradiation in October of 1985 with an average burnup of 11,357 MWD /MTV. Post-irradiation examinations showed all 4 demonstration assemblies were of good mechanical integrity. No mechanical damage or wear was evident on any of the VANTAGE 5 components. Likewise, the IFM grids on the VANTAGE 5 demonstration assemblies had no effect on the adjacent fuel 1 assemblies. All four demonstration assemblies were reinserted into V. C. Summer for a second cycle of irradiation. This cycle was campleted in M sch of 1987, at which time the demonstration assemblies achieved ar. average burnup of about 30,000 MWD /MTV. The observed behavior of the four assemblies at the end of 2 cycles of irradiation was as good as that observed at the end of the first cycle of irradiation. The four assemblies were reinserted for a third cycle of irradiation 4494Ft6 900926 4
O which was completed in November 1988 (EOG burnup 46,000 MWD /MTU). The observed behavior. of the four assemblies was -as good as that observed at the end' of. the first and second cycles of irradiation. In addition to V. C. Summer, individual VANTAGE 5 product features have been
demonstrated at other nuclear plants. IFBA demonstration fuel rods have been
-irradiated in Turkey Point Units 3 and 4 for two reactor cycles. Unit 4 contained 112 fuel rods equally distributed in four demonstration assemblies.
The IFBA coating performed well with no loss of coating integrity or adherence. The IFM grid feature has been demonstrated at McGuire Unit 1. The demonstration assembly at McGuire was irradiated for three reactor cycles and showed good mechanical integrity-.
'The. following plants are currently operating with regions of VANTAGE 5 fuel . assemblies: Callaway, V. C. Summer, Shearon Harris, Diablo Canyon and Byron /Braidwood.
The results of_ the evaluation / analysis described herein lead to the following conclusions:
' l .. The Westinghouse VANTAGE 5 reload fuel assemblies for the VEGP Units 1 and '
2 are mechanically compatible with the current LOPAR fuel assemblies, control rods, and reactor internals interfaces. The VANTAGE 5/LOPAR fuel assemblies satisfy the current design BASES for the VEGP Units 1 and 2.
2. . The structural integrity of 'the 17x17 VANTAGE 5 fuel assembly design for seismic /LOCA loadings has' been evaluated for both VEGP Units 1 and 2.
Evaluation of the 17x17 VANTAGE 5 fuel assembly component stresses and grid impact forces due to postulated faulted condition accidents verified that the VANTAGE 5 fuel assembly design is structurally acceptable.
3. Taking credit for. Leak Before Break technology, it has been demonstrated that RCCA insertion will occur during a LOCA with a limited displacement break. This allows for consideration of RCCA insertion when performing the evaluation to demonstrate that the core will remain subcritical during V the post-LOCA long term core cooling phase of the event.
4494F:6-900926 5
i O V 4. Change,: _in the nuclear characteristics due to the transition from LOPAR to , VANTAGE 5 fuel will be typical of the normal cycle to-cycle variations experienced as loading patterns change.
5. The reload VANTAGE 5 fuel assemblies are hydraulically compatible with the LOPAR fuel assemblies. from previous cycles of operation.

6. The core design and safety analyses results documented in this report show the core's capability for operating safely for the rated VEGP Units _1 and 2 design thermal power with.FAH of 1.65 for, the VANTAGE 5 fuel and 1.57 for the LOPAR fuel, Fg of 2.50 and steam generator tube plugging
' levels up to 10L The analysis was performed at 3565 MWt which is conservative compared to the current rated thermal power, 3411 MWt.
7. The steam generator tube _ rupture analysis to support the transition to VANTAGE 5 fuel shows that offsite doses for the VEGP Units 1 and 2 are well within the allowsle gt'delines specified in the SRP (NUREG-0800),
O._ v FSAR Chapter 15.6.3 and 10CFR100.
8. The projected increase in the fuel burnup levels with the use of VANTAGE 5 fuel _.has a negligible effect on' the radiological consequences of accidents
.duelto the very small changes in the core inventory of fission products.
9. The-previously reviewed licensing basis continues to-be met when the VEGP Units 1 and 2 are reloaded with VANTAGE 5 fuel. Plant operating limitations given in the Technical Specifications will be satisfied with the proposed changes noted in Enclosure 3 of this submittal. A reference is established upon which to base Westinghouse reload safety evaluations for filture reloads with VANTAGE 5 fuel.
( 4494F:6 900926 6
d 1
~
2.0 MCCHANICAL (VALUATION Introduction'and Summary This section evaluates the mechanical design and the compatibility of the 17x17 VANTAGE 5 fuel assembly with the current low parasitic (LOPAR) fuel assemblies during the transition through mixed fueled cores to an all VANTAGE 5 core. The VANTAGE 5 fuel assembly has been designed to be compatible with , the LOPAR fuel assemblies, reactor internals interfaces, the fuel handling > equipment, and refueling equipment. The VANTAGE 5 design is intended to replace and be compatible with cores containing fuel of the LOPAR design. The VANTAGE 5 design dimensions are essentially equivalent to the current VEGP Units 1 and 2 LOPAR fuel assembly design from an exterior assembly envelope and reactor internals interface standpoint.- References in this section are 1 made to WCAP-10444 P-A, " VANTAGE 5 Fuel Assembly Reference Core Report," Reference 2, and.to WCAP-9500-A, " Reference Core Report 17x17 Optimized Fuel Assembly," Reference 1. ( The significant new mechanical features of th'e VANTAGE 5 design relative to the initial core / Cycle 1 LOPAR fuel design for both VEGP Units 1 and 2 include the-following: o Int agral Fuel Burnable Absorber (IFBA) i o Intermediate Flow Mixer (IFM) Grids o Reconstitutable Top Nozzle (RTN) o Extended burnup capability including slightly longer fuel rods o Axial blankets-o Replacement of six intermediate inconel grids with zircaloy grids o- Reduction in fuel rod, guide thimble and instrumentation tube diameter o Redesigned fuel rod bottom.end plug to facilitate reconstitution capability o Snag resistant inconel grids (top and bottom) o Debris Filter Bcttom Nozzle (DFBN) l-l O l' 4494F:6 900926 7
The _RTN, OFBN, redesigned fuel rod bottom.end plug, snag resistant grid, and the fuel assembly extended burnup modification have been introduced previously in both VEGP Units 1 and 2. These features will continue to be utilized in the VANTAGE 5 design. Table 2-1 provides a comparison of the LOPAR and VANTAGE 5 fuel assembly design parameters.
- f,yel Rod Performance fuel rod design evaluations for VEGP Units 1 and 2 were performed using the NRC' approved models, References 5 and 6, and the extended burnup design methods in Reference 3. Fuel rod performance for'all VEGP fuel is shown to satisfy the NRC Standard Review Plan (SRP) fuel rod design basis on a region by region basis. These same BASES are applicable to all fuel rod designs, including the LOPAR and VANTAGE 5 fuel designs, with the only difference being that the VANTAGE 5 fuel'is designed to_ operate with a higher FAH limit.
The design BASES for Westinghouse VANTAGE 5 fuel are discussed in Reference 2. There _is no effect from a fuel rod design standpoint having fuel with more than one type of geometry simultaneously residing in the core during the transition cycles. _The mechanical fuel rod design evaluation for each region incorporates all appropriate design features of the region, including any-changes to the fuel rod or pellet geometry from that of previous fuel regions (such as the presence of axial blankets or changes-in the fuel rod and plenum length, for example). Analysis of IFBA rods includes any geometry changes necessary to model the presence of the burnable absorber, and conservatively
models the gas ~ release from the zirconium diboride pellet coating.
Fuel performance evaluations are completed for each fuel region to demonstrate that the design criteria will be satisfied for all fuel rod types in the core under the-planned operating conditions. Any changes from the plant operating conditions originally evaluated for the mechanical design of a fuel region (tar example an increase in the peaking factors) are addressed for all effet.ted fuel regions as part of the reload safety. evaluation process when the plant t.5ange is to be implemented. O 4494F:6 900926 8
'I l
j' TABLE 2 1 1 Comparison of 17x17 LOPAR and 17x17 VANTAGE 5 Fuel Assembly Design Parameters-17x17 17x17-PARAMETER LOPAR DESIGN VANTAGE 5 DESIGN s Fuel Assy length, in. 159.765 159.975. Fuel Rod Length, in.- 151.560 152.285 Assembly Envelope, in. 8.426 8.426~ Compatible with Core Internals'- Yes Yes Fuel Rod Pitch', in. 0.496 0.496 Number _of. Fuel Rods /Assy. 264 264 Number / Guide Thimble' Tubes /Assy. 24 24 >
. Number / Instrumentation Tube /Assy. I 1 ~
Fuel Tube Material Zircaloy 4 Zircaloy 4 Fuel Rod Clad 00 , in. 0.374 0.360-Fuel Rod Clad Thickness, in. 0.0225- -0.0225 Fuel / Clad Gap. mil. 6.5 6.2
. Fuel: Pellet 01ameter, in. 0.3225 0.3088 Fuel Pellet: Length, in. 0.387 :0.370:
i 4494F:6-900926 9 L
i-
~
Grid Assemblies The. top and bottom inconel (non-mixing vane) grids of the VANTAGE 5 fuel assemblies. are similar in design to the inconel grids of the Cycle 1, LOPAR fuel assemblies for VEGP Units 1 and 2. The differences are:' 1) the spring and dimple heights have been modified to accommodate the reduced diameter fuel rod,
2) the top grid spring force has been reduced to minimize rod bow, 3) the VANTAGE 5 top grid uses type 304L stainless steel sleeves instead of 304 stainless steel sleeves used for the LOPAR top grid,' 4) the top and bottom grids have a snag resistant design which minimizes assembly interactions during core loading / unloading, 5) the top and bottom grids-have dimples which are rotated 90* to minimize fuel rod fretting and dimple cocking, and 6) the top and bottom grid heights have been increased to 1.522 inches. The snag resistant design and the 304L stainless steel sleeves were introduced in Vogtle Unit 1, Cycle 2/ Region 4 and in the Vogtle Unit 2 initial core / Cycle 1. Rotated dimples and 1.522 inch grid heights were introduced in Vogtle Unit 1, Cycle 3/ Region 5 and Vogtle Unit 2, Cycle 2/ Region 4. These features will continue to be j utilized. The six interme'diate (mixing vane) grids are made of zircaloy ,
material rather than inconel which is currently used in the LOPAR design. The IFM grids shown in Figure 2.1 are located in the three uppermost spans between the zircaloy mixir.g vane structural grids and incorporate a similar mixing vane array. Their prime function is mid-span flow mixing in the hottest fuel' assembly spans. Each IFH grid cell contains four dimples which are designed.to prevent mid-span channel closure in the spans containing IFMs and fuel rod contact with the mixing vanes. This simplified cell arrangement allows short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop. The IFM grids are not intended to be structural members. The outer strap configuration was designed to be similar to current fuel designs to preclude grid hang-up and damage during fuel handling. Additionally, the grid envelope is smaller which further minimizes the potential for damage and reduces l calculated forces 'during seismic /LOCA events. Impact loads during these L Lo . 4494F:6 900926 10 l
- .-- . , . . . , , . . . . . . . . . ~ . - . ~ . . - . . . . . - ~ ~ ~ . . . . - - . . . . . , . . ~ . - . ,
d k 1H 975 3 479 152.265 3,333 I p t I _
- 122.3 f == 101.74 - 41.21 -=
g3 2, 4 to-113 40 === 133.47 - t11.32 - $1.37 . - 71.32 31Jt- = m72 g.gsg 17X17 VANTACC $ NEL ASSEMBLY
'Se /1 71 oNel (2X3Ms) O xakes taxu - - 1sto / ten oxo e own 2.73a / 2.asJ -= --
18144 / t$1.741 / 164.20 5 4426 c - n 5
. f. . .. . . . . . . . I .
1 l.i'tT2" - ln exa) aus) . pus) owei pus) ausi ows
==1&L90(att' = t3101 === 19 2.44 == 9 9.91 == 71Je ==96A1 == m24 === $.436-e inst it2 a2 . e2 27 n.n st it me e ieo tueeone) . (1X2X4) -
13173(3) (tX2X4) (1K2X4) (1X2 Net (1XIN4) (1X2X4) (tX2X4) 17X17 RECONS 717UTA8LE.LOPAA NEL ASSEMBLY t (t) estint cont / cruz e imer 1 (2) cent 2/etcupe 4 yast t (3) cycle 3/Mcich 6 usef 1 (4) esmak colut/CTCn.E t ves? 2 (S) CYCIA 2/Mcaosee LMT 2 O ' yogya. 17%17 YANTACE S / LOPAA Georgia Power ELactitic ctetAf.aso
=i - Mast?
FUEL ASSIMBLJ COMPARISON TICURE 2.1 11
m & 4 1-e L. ..<ss& 4, +n _ _ . _ , ,
?
QI events have been calculated for the IFM grids for typical Westinghouse reactors. A coolable geometry is, therefore, assured.at the IFM grid elevation, as well as at the structural grid elevation. Reconstitutable Too Nozzle and Debris Filter Bottom Nozzle The RTN for the VANTAGE 5 fuel assembly differs from the welded top nozzle design in two ways: a groove is provided in each thimble thru-hole in the nozzle plate to-facilitate removal, and the nozzle plate thickness is reduced to provide additional axial space for fuel rod growth. The RTH feature was previously introduced in Vogtle Unit 1, Cycle 2 and Vogtle Unit 2, Cycle 2 and will continue to be utilized. In the . VANTAGE 5 RTN design, a stainless steel nozzle insert is mechanically connected to the top nozzle adapter plate by means of a pre-formed circumferential bulge near the top of the insert. The insert engages a mating groove in the wall of the adapter plate thimble tube thru hole. The-insert has 4 equally spaced axial slots which allow the insert to deflect inwardly at the elevation of the bulge, thus permitting the installation or removal of the nozzle. The insert bulge is positively held in the adapter plate mating groove by placing a lock tube with a uniform ID identical to that of the thimble tube into the -insert. The lock tube is secured in place by two means. First, a top flare creates a tight fit. - Second, six non-yielding projections ~on the 00 which interface with the concave side of the insert
-preclude escape during core component transfer..
The full' complement of these joints comprises the structural connection (reconstitutable design feature) between the top nozzle and the remainder of the VANTAGE 5 fuel assembly. The nozzle insert to-adapter plate bulge joints-replace the uppermost grid sleeve to-adapter plate welded joints found in 4 current fuel assemblies. The nozzle insert-to thimble tube multiple 4-lobe bulge joint located in the lower portion of the -insert represents the
- structural connection between the insert and the ren ainder of the fuel assembly below the elevation of the insert. The uppirmost grid sleeve is connected to the thimble tube by similar 4-lobe bulgt joints, l-4494F:6-900926 12 F
O To remove the top nozzle, a tool is first inserted through the lock tube and' expanded radially to engage the bottom edge of t' O ie . An axial force is then exerted on the tool which overrides the Ok tube deformations and withdraws.the lock-tube from the insert. Aft. J., lock tubes have been , withdrawn, the nozzle is removed by raising it off the upper slotted ends.of the nozzle inserts which deflect inwardly under the axial lift load. With the top nozzle removed, direct access is provided for fuel rod examination or replacement. Reconstitution is completed by the-remounting of the nozzle and the insertion of new locked tubes. The design BASES and evaluation of the RTN are given in Section 2.3.2 in Reference 2. The VANTAGE 5 design will include the use of the DFBN to reduce the possibility of fuel rod damage due to debris induced fretting. For the DFBN design the relatively large flow holes in the LOPAR bottom nozzle are replaced with a new pattera of smaller flow holes. The holes are sized to minimize passage of debris particles large enough to cause damage while providing sufficient flow area, comparable pressure drop and continued structural integrity of the nozzle. The overall design of the DFBN is similar to the . current design except it is shorter and has a thinner top plate to allow for fuel rod growth. Axial Blankets Although noted-as a new mechanical feature of the VANTAGE 5 design and
. licensed in Reference 2, axial blankets have been and are currently operating in-Westinghouse plants to reduce neutron leakage and improve fuel utilization. . A description and design application of this feature are contained in Reference 2, Section 3.0. The axial blankets utilize a chamfered pellet physically different from the enriched pellet in the fuel stack to help prevent accidental mixing with the enriched pellet during manufacturing.
I l 4494F:6-900926 13
/"N Mechanical- Comoatibility of Fuel Assemblies Based on the evaluation of the VANTAGE 5/LOPAR design differences and hydraulic test results, References 1 and 2, it is concluded that the two designs are mechanically compatible with each other. The VANTAGE 5 fuel rod mechanical design BASES remain unchanged from that used for the LOPAR fuel assemblies. Rod Bow It is predicted that the 17x17 VANTAGE 5 rod bow magnitudes, like those of the
- Westinghouse OFA fuel, will be within the bounds of existing 17x17 LOPAR assembly rod bow data. The current NRC approved methodology for comparing rod bow for two different fuel assembly designs is given in Reference 7.
Rod bow in fuel rods containing IFBAs is not expected to differ in magnitude or frequency from that currently observed in Westinghouse LOPAR fuel rods h unde,r similar operating conditions. No indications of abnormal rod bow have been observed on visual or dimensional inspections performed on the test IFBA rods. Rod growth measurements were also within ' predicted bounds. Fuel Rod Wear Fuel rod wear is dependent on soth the support conditions and the flow environment to which the fuel rod is subjected. Due to the LOPAR and VANTAGE 5 fuel assembly designs employing different grids, there is an unequal axial pressure distribution between the assemblies.- Crossflow resulting from this unequal pressure distribution was evaluated to determine the induced rod vibration and subsequent wear. Hydraulic tests (Reference 2, Appendix A.I.4) were-performed to -verify that the crossflows were negligible and also to check
. hydraulic cosipatibility of the LOPAR and VANTAGE 5 designs. The VANTAGE 5 fuel assembly'was flow tested adjacent to a 17x17 0FA, since vibration test results indicated that the crossflow effects produced by this fuel assembly combination would have the most detrimental effect on fuel rod wear.
O 4494F:6 900926 14
Results of the wear inspection and analysis discussed in Reference 2, Appendix A.I.4, revealed that the VANTAGE 5 fuel assembly wear characteristic was similar to.that of the 17x17 0FA when both sets of data were normalized to the test duration time. It was concluded that the VANTAGE 5 fuel rod wear would be less than the maximum wear depth established, Reference 8, for the 17x17 0FA at EOL. Seismic /LOCA Imoact on Fuel Assemblies An evaluation of the VANTAGE 5 fuel assembly structural integrity considering the lateral effects of LOCA and seismic loadings has been performed using time-history numerical techniques based on the Vogtle plant specific Safe Shutdown Earthquake (SSE). The VANTAGE 5 fuel assembly is structurally equivalent to the LOPAR fuel design. The main differences between the two designs are six zircaloy grids replacing inconel mid-grids in the LOPAR design, three additional intermediate flow mixers, and optimized fuel rods. The load bearing capability f,or the t (u )l zircaloy grids and flow mixers under the faulted condition loadings has been analyzed. The results indicated that 17x17 VANTAGE 5 grid loads are below the allowable grid strengths. Based on the grid load results, the 17x17 VANTAGE 5 zircaloy grid is capable of maintaining the core coolable geometry under the SSE and asymmetric pipe rupture transients in either homogeneous or transition core operation. The 17x17 VANTAGE 5 fuel assembly is structurally acceptable for both VEGP Units 1 and 2. This is also true for a transition core composed of both VANTAGE 5 and LOPAR fuel assembly core configurations. The grids will not buckle due to the combined imo nt lueos of a seismic and LOCA event. The coolable geometry requirement is met. The stresses in the fuel assembly components resulting from seismic and LOCA induced deflections are within acceptable limits. [ \ V I l 4494F:6-900926 15
(D \'"J Core Components The core components are oe:igr.Gd to be compatible with VANTAGE 5 and LOPAR fuel assemblies. The reduced diameter VANTAGE 5 thimble tube provides sufficient clearance for insertion of control rods, WABA rods, source rods or dually compatible thimble plugs to assure the proper operation of these core components. O-O 4494F:6-900926 16
( 3.0 NUCLEAR EVALUATION The evaluation of the transition and equilibrium cycle VANTAGE 5 cores presented in Reference 2, as well as the VEGP specific transition core evaluation, demonstrate that the impact of implementing VANTAGE 5 does not cause a significant change to the physics characteristics of the VEGP cores beyond the normal range of variations seen from cycle to cycle. The methods and core models used in the VEGP reload transition core evaluations are described in Reference 2, 4, 9, and 10. These licensed methods and models have been used for Vogtle and other previous Westinghouse reload designs using the 0FA and VANTAGE 5 fuel. No changes to the nuclear design philosophy, methods, or models are necessary because of the transition to VANTAGE 5 fuel. For the nuclear design area, the following VEGP Units 1 and 2 Technical Specifications changes are proposed: n Q., 1) Increased FAH limits. These higher limits serve to increase nuclear design flexibility and allow loading patterns with reduced leakage which in turn will allow longer cycles.
2) Increased gF limit. The increased F g limit will provide greater flexibility with regard to accommodating the axially heterogeneous cores (axial blankets and part length burnable absorbers).

3) Relaxed Axial Offset Control (RA0C) band implementation, and revised surveillance requirements on the heat flux hot channel factor, Fg (z).
The methods used in the core analysis incorporating these Technical Specifications changes are licensed and described in Reference 9. Implementation of a RA0C band will increase plant operating flexibility, and performing surveillance directly on F9 (z), rather than Fxy(z) the radial component of the total peaking factor, will more directly monitor the parameter of interest. The steady state Fg(z) is measured and increased by applicable uncertainties. This quantity is further 4494F:6 900926 17
s. increased by an analytical factor called W(z) which accourts for possible increases in'the steady-state Fn (z) resulting from operation within the allowed axial flux difference limits. The resulting F n(z) is compared to the Fg(z) limit to demonstrate operation below the heat flux hot channel factor limit. The Nuclear Design analysis was performed for a RAOC AFD band of +10/-20% AFD at 100% RTP and +30/-35% AFD at 50% RTP; the actual RAOC operating bands to be used for future specific reloads will be the same or narrower and will be specified in a Core Operating Limits Report.
4) Core Operating Limits Report. The AFD operating bands for RAOC operation will be specified in the Core Operating Limits Report, instead of in the Technical Specifications. This eliminates the necessity of Technical Specification amendments for future reload cycles, while providing assurance that the correct operating limits will be followed.
1 Power distributions and peaking factors show slight changes as a result of the incorporation of reduced length burnable absorbers, axial blankets, and increased peaking factors limits, in addition to the normal variations experienced with different loading patterns. The usual methods of enrichment variation and burnable absorber usage can be employed in the transition and full VANTAGE 5 cores to ensure compliance with the peaking factor Technical Specifications. The key safety parameters evaluated for the Vogtle reactor as it transitions
= to an all VANTAGE 5 core show little change relative to the range of parameters experienced for.tne all LOPAR fuel core. The changes in values of the key safety parameters are typical of the normal cycle-to-cycle variations experienced as loading patterns change. As is current practice, each reload core design will be evaluated to assure that design and safety limits are satisfied according to the reload methodology. The design and safety limits will be documented in each cycle specific Reload Safety Evaluation report which serves as a basis for any significant changes whicn may require a future NRC review.
N x) 4494F:6-900926 18
(n) ' ' 4.0 THERMAL AND HYORAULIC CVALUATION This section describes the calculational methods used for the thermal hydraulic analysis, the DNB performance, and the hydraulic compatibility during the transition from mixed-fuel cores to an all VANT4GE 5 core. ThL Westinghouse transition core DNB methodology is given in References 1 and 11 and has been approved by the NRC via Reference 12. Using this methodology, transition cores are anlayzed as if the entire core consisted of one assembly type (full LOPAR or full VANTAGE 5), and the resultant ONBR values are reduced by the appropriate transition core penalty. The LOPAR and the VANTAGE 5 fuel assemblies were shown to be hydraulically compatible in Reference 2. The ONBR analyses for VEGP Units 1 end 2 were based on parameters which conservatively bound the licensing values. Table 4 1 summarizes the pertinent thermal and hydraulic design parameters used in the analyses as well as the ( ) licensing values. The improved THINC-IV PWR design modeling method, Reference 13, was used for the DNBR analyses of the VANTAGE 5 and LOPAR fuel. For high core power density applications, the improved model yields more conservative values of minimum DNBR than the present model, Reference 14. No changes to the basic THINC !V models and correlations were made for the improved core modeling scheme. The DNBR analyses of the VANTAGE 5 fuel and LOPAR fuel are based on the Revised Thermal Design Procedure (RTOP), Reference 15. The primary ONB correlation used for the LOPAR fuel is the WRB 1 ONB correlation which is described in Reference 16. The VANTAGE 5 fuel DNBR ar,alyses use the WRB 2 DNB correlation which is described in Reference 2. The WRB 2 DNB correlation takes credit for the reduced grid to grid spacing of the VANTAGE 5 fuel assembly mixing vane grids resulting from the use of the Intermediate Flow Mixer (IFM) grids. Both the WRB 1 and WRB 2 DNB correlations have a g3 correlation limit of 1.17. Q 1 4494r:6 9009ts 19
\") The W 3 DNB correlation, References 17 and 18, is used for both fuel types where the primary DNB correlations are not applicable. The WRB 1 and WRB 2 DNB correlations were developed based on mixing vane data and, therefore, are only applicable in the heated rod spans above the first mixing vane grid. The i DNB correlation, which does not take credit for mixing vane grids, is used to ulate DNBR values in the heated region below the first mixing vane grid, in adoition, the W+3 DNB correlation is applied in the analysis of accident conditions where the system pressure is below the range of the primary correlations. For system pressures in the range of 500 psia to 1000 psia, the W ; ONB correlation limit is 1.45, Reference 19. For system pressures greater than 1000 psia, the W 3 DNB correlation limit is 1.30. A cold wall f actcr, Reference 20, is applied to the W 3 DNB r yrelation to account for the presence of the unheated thimble surfaces. Also, a 0.88 multiplier is applied to the W 3 DNB correlation to account for the 17x17 fuel rod diameter effect, Reference 21. With the RTOP methodology, uncertainties in plant operating 5.arameters, nuclear and thermal parameters, fuel fabrication parameters, compute. codes and DNB correlation predictions are considered statistically to, obtMr. ONB uncertainty
] factors. Based on the DNB uncertainty factors, RTOP design i n.it DNBR values are determined such that there is at least a 95 percent probability at a 95 percent confidence level that DNB will not occur on the most limiting fuel rod during normal operation and operational transients and during transient conditions arising from faults of moderate frequency (Condition I and 11 events as defined in ANSI N18.2).
Uncertainties in the plant operating parameters (pressurizer pressure, primary coolant temperature, reactor power, and reactor coolant system flow) have been evaluated for the VEGP Units 1 and 2 with Resistance Temperature Detector (RTO) bypass loops, Reference 22, and for the RTD bypass loops eliminated, Reference
23. In the DNBR analyses with RTOP, a set of plant operating parameter uncertainties were used which are bounding for operation with RTO bypass loops or for RTO bypass loops eliminated. Only the random portion of the plant operating parameter uncertainties is included in the statistical combination.
Instrumentation bias is treated as a direct DNBR penalty. Since tne parameter o uncertainties are considered in determining the RTOP design limit DNBR values, the h plant safety tr.Llyses are performed using input parameters at their nominal values. 4494F:6 900926 20
- . _ . . . _ - _ . . - - _ . - . = . .__ . - . . .. ,.m 'I h v
TABLE 4 1 THERMAL And HYDRAULIC DESIGN PARAMETERS FOR YEGP UNITS 1 AND 2 The**al and Hveraul'e bet tea e seemettet Acatytte eseeeeteen pee *e ue Parameteen (Usin9 R10P) Reactor Core Heat Output. *vt 3566 3411 Reactor Core Heat Output. 106 BTU /hr !!!$4 11639 Heat Generated in fuel. 1 97.4 97.4 fressurtrer Pressure, hostmal, pata !!!0 !!$0 F 3g. Nuclear Enthalpy Rise Hot Channel Factor (LDPAR) 1.67(1+.3(1 P)) 1.67(1+.3(1 P)) (V 6) 1.66(1+.3(1 P)) 1.65[1. 3(1 P)] Minimum CNBR at hominal Condittoes typical Flow Channel (LDPAR) 2.32 st.32 p (V 5) !.41 *t.41 g Thtable (Cold Wall) Flow Channel (LOPAR) 2.!! et.!!
\
(v.5) 2.26 $2.28 Dest 9n Limit DhbR Typical Flow Channel (LDPAR) 1.23 1.23 (V 5) 1.24 1.!4 Thimble (Cold Wall) Flow Chai. 41 (LDPAR) 1.!! 1.!! (V 5) 1.23 1.23 Ch8 Correlation (LOPAR) VRB 1 WRB 1 (V 5) WRB ! WR6 2 a 4494F:6 900926 21
m I
V) TABLE 4a1 (continued) THERMAL ?ND HYDRAULIC DESIGN PARAMETERS FOR VEGP UNITS 1 AND 2 Hre heetaa 1 teelsat teadit ican saatvate encamete t' Liteastas ea *nmeteet" Vessel Minimum Measured Flow Rate (thcluding Bypass). 106 lbm/hr 142.3(a) 145.!(b) spm 382.400 391.!!$ vessel thermal Design Flow Rate (including lypass). 106 lbm/hr 139.4 142.1 gpm 374.400 382.600 Core Flow Rate (eacluding Bypass, based on thermal Design Flow) 106 lbm/hr 127.7 130.2 gpm 342.950 350.545 Fuel Assembly Flow Area O for heat Transfer. ft! ' (LOPAh) $1.08 51.0B V (V 5). 54.13 64.!? Core inlet Mass Velacity. 106 lbm/hr+ftt (Based on 70F)+ (LOPAR) 2.60 2.55 (V 6) 2,36 f.41
+ Assumes all LOPAN or VANTAG( $ Core
Analysts flow. rates are based on 10% steam generator tube plugging
" Licensing flow rates are based on 01 steam generator tube plugging (a) Inlet temperature * $57.4'F (b) Inlet temperature e 559.3'F O
V 4494F:6 900926 22
1 l' t TABLE 4 1 (continued) THERMAL AND HYDRAULIC DESIGN PARAMETERS FOR VEGP UNITS 1 AND 2 the*aal sad Hyd*sul6e Bestea e senmeteen Acalysis e s eemeters Liceastae es*emeters (lated en 10F) nomtral Vessel / Core Inlet Temperature. 'F $16.8 558.7 Vessel Average temperature. *F 588.4 588,4 Core Average temperature. 'F $93.0 597.6 Vessel Outlet Temperature. 'F $20.0 618.1 Averste Temperature Rise im Vessel. 'F 63.2 59.4 Average temperature Rise in Core. 'F 68.2 64.! we st Treatte* Active neat fransfer Surface Area. ftI * (L0 EAR) 59.742 59.742 (v.5) 57.505 57.505 Average Meat Flus. BTU /hr ftI * (LOPAR) 198.370 189.800 (V 5) 206,085 197,t00 Average Lieser Power, km/ft** 5.69 5,45 Peak Lieear Powet for Normal Operation, km/ft*** 14.2 13.6
-r ^ g Temperature Limit for h Prevent ton of Center 1tne welt. 'F 4.700 4.700 + Assumes all LOPAR or VANTAGE 5 core ++ Based on densified active fuel length +++ Based on 2,50 Fo peaking factor O
4494F:6 900926 23
4 l The RTDP design limit DNBR values are 1.24 and 1.23 for the typical and thimble cells respectively for VANTAGE 5 fuel, and 1.23 and 1.22 for the
j. typical and thimble cells respectively for LOPAR fuel.
l t in addition to the above considerations, plant specific DNBR margin was ; maintained by performing the safety analyses to DNBR limits higher than the ; design limit DNBR values. A fraction of the available DNBR margin is utilized to accommodate tne transition core penalty. For VANTAGE 5 fuel, this transition core penalty is a function of the number of VANTAGE $ fuel assemblies in the core as given in 4 Reference 24. There is no transition core penalty for the LOPAR fuel. Additional margin is used to offset the_ rod bow DNBR penalty. Based on Reference 7, the fuel rod bow DNBR penalty is less than 1.5% for both LOPAR , and VANTAGE $ fuel in the 20 inch grid spans. No rod bow penalty is required in the 10 inch grid spans of the VANTAGE 5 fuel. The remaining DNBR margin, after consideration of these' penalties, is available for operating and design i flexibility. The option of ~ thimble plug removal has been included in all of the DNBR analyses performed for the VANTAGE 5 and LOPAR fuel. The primary effect of 1 thimble plug removal is an increase in the core bypass flow. This increased core bypass flow is reflected in the core flow rates and the DNBR values presented in Table 4-1. l Operation with thimble plugs in place reduces the core bypass flow through the fuel assembly thimble tubes. The reduction in core bypass flow for operation with the thimble plugs in place-is a DNBR benefit. The increased DNBR margin associated with the use of a full compliment of thimble plugs can be used to offset DNBR penalties.
The Standard Thermal Design Procedure (STDP) is used for those analyses where the RTDP methodology is not appitcalle. In the STDP method, the parameters used in DNBR analyses are treated in a conservative way to give the lowest minimum DNBR. _ Sufficient DNBR margin to cover appropriate DNBR penalties is preserved whenever the STDP is used.
4494F:6 900926 24 1 i
, _ _ . , _ . .. - _ _ _ -- _ . - , _ _ . . . _ . _ - . . . _ _ , _ . . . . _ . _ - _ _ _ , _ , _ _ - , _ = , _ . , _ , . . - .
The fuel temperatures used in safety analysis calculations for the VANTAGE 5 and LOPAR fuel were calculated with the improved fuel performance code, Reference 6. This code was used to perform both design and safety calculations. These fuel temperatures were used as initial conditions for LOCA and non-LOCA transients. I l O i 4494F:6 900926 25
,*aww--,.m,_-, - _
M 5.0 ACCIDENT EVALUATION-5.1 Non-LOCA Accidents This section addresses the effects of the VANTAGE 5 design features and safety , 4 analysis assumptions for the VEGP Units 1 and 2 non LOCA accident analyses, i 5.1.1 YANTAGE 5 Design Features The following VANTAGE 5 fuel assembly design features were considered in the non LOCA analyses and evaluations: o Fuel Rod Dimensions o- IntermediateFlowMixer(IFM) Grids , o Axial Blankets o Integral Fuel Burnable Absorbers (IFBAs) o Reconstitutable Top Nozzle (RTN) o Debris Filter Bottom Nozzle (DFBN)
o. Extended Burnup t/ o Zircaloy Grids A brief description of each of these and its consideration in the non LOCA safety analyses follows.
Fuel Rod Dimensions The VANTAGE 5 fuel rod dimensions which determine the safety analysis temperature versus linear. power density relationship include rod diameter, ' pellet' diameter, initial pellet to clad gap size, and stack height. The non-LOCA safety-analysis fuel temperature and rod geometry-assumptions t consider this geometry change and bound both the LOPAR and VANTAGE 5 fuel, IFM Grids-The IFM grid feature of the ' VANTAGE 5 fuel design increases Departure from Nucleate Boiling (DNB) margin. The fuel safety analysis limit DNB values contain significant DNB margin (see Section 4.0). 4494F:6 901001 26
. . _ . _ . . _ _ _ _ _ _ _ - , _ . . _ - . . _ _ _ . _ _ _ - _ ~ . . _ . _ . . .
1 e 1 The IFH g id feature of the VANTAGE 5 'uel design increases the core pressure drop. This results in an increased control rod scram time to the dashpot from 2.2 to 2.7 seconds. This increased drop time primarily affects the fast reactivity transients which were analyzed for this report. The revised control rod drop time was incorporated in all the analyzed events requiring this parameter and the remaining transients have been evaluated. , 1 Axial Blankets and IFBAs Axial blankets reduce power at the top and bottom of the rod which increases axial power peaking at the center of the rod. This effect is offset by the presence of part length IFBAs which flatten the power distribution. The net effect on the axial shape is a function of the number and configuration of IFBAs in the core and , the time in core' life. The effects of axial blankets and IFBAs on the reload safety analysis parameters are taken into account in the reload design process. The axial power distribution assumption in the safety analyses kinetics calculations have been determined to be sufficiently conservative to accommodate the introduction of axial blankets and/or IFBA in the VEGp Units 1 and 2. Reconstitutable Too Nozzle (RTN) and Debris Filter Bottom Nozzle (DFBN)- RTNs and DFGNs have been used extensively in Westinghouse designs' . Analyses / tests i~ were performed to confirm the hydraulic. compatibility of the Westinghouse nozzle
-designs-to the existing designt therefore, these components will not affect any parameters important to-the non LOCA safety analyses, t Extended Burnuo L
The VANTAGE 5 fuel assemblies were modified for extended burnups by reducing the thickness of both the top and bottom nozzle end plates, decreasing the height of the bottom nozzle and increasing the length of the fuel rod. The effects of these extended burnup features _have been accounted for in the non LOCA analyses, u' Zircalov Grids Zircaloy grids have replaced inconel grids in the VANTAGE 5 fuel assembly with the exception of the top and bottom grids which remain inconel. The effects of Zircaloy grids have been incorporated in the safety analysis, j 4494F:6 901001 27
.,.:. . . ~ . . _ ._-_...,_.__,,._a..__..._.__..._ _ -
______._._ ,_____ _ _ ~
5.1.2' Safety Analysis Assumptions O Listed below are the safety analysis assumptions which represent a departure i from those currently used for VEGP Units I and 2. j p 1 o !ncreased Core Thermal Power o Reduced Thermal Design Flow o Revised Thermal Design Procedure (RTDP) for Appropriate DNB Events o Increr. sed Uncertainties for Reactor Coolant System (RCS) Temperature , and Pressure 2 o 10% Steam Generator Tube Plugging o Relaxed Axial Offset Control (RAOC) o !ncreased End of Life Moderator Density Coefficient o Removal of Thimble Plugs o increased Design Enthalpy Rise Hot Channel Factors (FAH and Fg) o Steam Generator tap relocation A brief descriptisen of each of these assumptions follows. O- Increased Core Thennal Power An increase in the nominal core thermal power from 3411 MWt to 3565 MWt was considered in the non LOCA safety analyses for the potential rerating of "EGP , Units 1 and 2. :The non LOCA safety analyses performed at 3565 MWt will 1 conservatively bound the current rated core thermal power level of 3411 MWt. Reduced Thermal Desian Flow A decrease in the RCS thermal design flow from 382,800 gpm to 374,400 gpm was considered in the non LOCA safety analyses for the potential rerating of VEGP
' Units _1 and 2. The non LOCA safety analyses performed at 374,400 gpm will conservatively bound'the current thermal design flow of 382,800 gpm.
l O 4494F26 901001 28
Revised Thermal Desion Precedure (RTDP)for Acerooriate DNB Events The calculational method utilized to meet the DNB design basis is the RTDP, which is described in Reference 15. Uncertainties in the plant operating parameters are statistically incorporated in the design limit DNBR value as discussed in Section 4.0. Since the parameter uncertainties are considered in determining the design DNBR value, the associated plant safety analyses are
.r -
performed using nominal initial conditions. Increased Uncertainties for RCS Temeerature and Pressure The RCS temperature uncertainity has been increased from 1 4*F to i 6*F, j and the RCS pressure uncertainty has been increased from 30 psi to r 50 psi. 10% Steam Generator Tube Pluccino The. nominal primary and secondary side conditions have been established which
include up to 10% steam generator tube plugging (or the hydraulic equivalent of plugs and sleeves) in each steam generator. It is assumed that no one steam O generator exceeds 10% tube plugging. These conditions were assumed for all of the analyses performed. l Relaxed Axial Offset Con'rol (RAOC)
RAOC operation with a +10/.20% Axial Flux Difference (AFD) band at 100% Rated Thermal Power (RTP) operation formed the-basis for the-safety evaluations. increased End of Life Moderator Density Coefficient In order to accommodate longer fuel cycles and extended fuel burnup, a . moderator density coefficient of 0.50 Ap/gm/cc corresponding to end of , cycle, full. power conditions was conservatively incorporated into the safety analyses performed for this report, f I O 4494F:6 901001 29 1
Thimble Plua Remqyal Thimble plug removal affects core pressure drops and increases core bypass flow. These effects have been conservatively incorporated into the non LOCA analyses performed for this report. The analyses are applicable for either the thimble plugs in place or removed. Increased F 6H dQ The design FAH for the LOPAR and VANTAGE 5 fuel is 1.57 and 1.65, respectively. The non LOCA calculations applicable for the VANTAGE 5 core have assumed a full power Fgg of 1.70. This is a conservative safety analysis assumption for this report. ihe increase in the Technical Specification maximum LOCA Fg from 2.3 to 2.5 is bounded in the non-LOCA transients. A maximum Fg of 2.55 was conservatively assumed in the non LOCA safety analyses.
;O Steam Generator Tao Relocation v .
The non LOCA analyses account for the relocation of the steam generator level tap. The tap was lowered from 438 inches to 333 inches from the top of the tubesheet. O(~~ 449u:s sotoot 30
l I l l 5.1.3 Hon LOCA Safety Evaluation Methodology The non LOCA reload safety evaluation methodology is described in Reference 4. The methodology confirms that, if a core configuration is bounded by existing safety analyses, the applicable safety criteria are satisfied. The methodology systematically identifies both parameter changes on a cycle by cycle basis which may exceed existing safety analysis assumptions and the transients which require evaluation. This methodology is applicable to the evaluation of VANTAGE 5 transition and full cores. Any required evaluation identified by the reload methodology is one of two types. 'If the identified parameter is only slightly out of bounds, or if the transient is relatively insensitive to that parameter, a simple evaluation may be made which conservatively evaluates the magnitude of the effect and explains why the actual analysis of the event does not have to be repeated. Alternatively, should the deviation be large and/or expected to have a significant or not easily cuantifiable effect on the transients, analyses are required. The analysis app.oach will utilize Westinghouse codes and methods which have been accepted ty the NRC and have been used in previous submittals to the NRC. These methods are those which have been presented to the NRC for a specific plant, reference SARs or reports for NRC approval. The analysis methods and codes are described in Appendix A. The key safety parameters are documented in Reference 4. Values of these safety parameters which bound both fuel types (LOPAR and VANTAGE 5) were assumed in the non LOCA safety analyses. For subsequent fuel reloads, the key safety parameters will be evaluated to determine if violations of these l bounding values exist. An evaluation of the affected accidents will take place as described in Reference 4. l 5.1.4 Conclusions l Descriptions of the non-LOCA accidents analyzed for this report, method of analysis, results, and conclusions are contained in Appendix A. Appendix A l conforms to the format of the VEGP Units 1 and 2 FSAR. It was found that the 4494F:6 900926 31
i appropriate safety criteria were met for each of the transients analyzed. In k addition, an evaluation was performed regarding the effect of VANTAGE 5 fuel l on the steamline break mass and energy release analyses, both inside and outside j containment. The results of this evaluation verify that the mass and energy releases previously calculated are not adversely affected by the transition to VANTAGE 5 fuel.
- Based on the plant operating limitations given in the Technical Specifications
and the proposed Technical Specifications changes given in Enclosure 3 of this l
subnittal, the results show that the transition from LOPAR to VANTAGE 5 fuel can be accommodated with margin to the applicable FSAR safety limits. 4. 5.2 LOCA Accidents This section addresses the effects of the VANTAGE 5 design features and modified safety analysis assumptions for the VEGP Units 1 and 2 LOCA analyses. 5.2.1, Large Break LOCA O 5.2.1.1 Description of Analysis / Assumptions for 17x17 VANTAGE 5 Fuel , The large break Loss Of Coolant Accident (LOCA) analysis for VEGP Units 1 and 2, applicable to a full core of VANTAGE 5 fuel assemblies, was performed to develop VEGP Units 1.and-2 specific peaking factor limits. This 16-consistent with the methodology employed in the Reference Core Report for 17x17 VANTAGE 5, Reference 2. The Westinghouse 1981 Evaluation Model with BASH, References 25 and 26, was utilized and a spectrum of cold leg breaks was analyzed for VEGP
- Units 1.and 2 that bound nomical opera \ing conditions.
Other pertinent large break LOCA analysts assumptions include: o An uprated core thermal power of-3365 MWt, o 10% steam generator tubes plugged for each of the four steam generators, c An FAH of 1.65 for VANTAGE 5 and 1,57 for LOPAR, and a total peaking factor, Fg of 2.5, for VANTAGE 5 and LOPAR fuel in the
-transition core o A two.line segment K(z) curve with a value for K(z) = 1.0 up to 6 ft.
and a linear ramp up to a value of 0.92 at 12 ft, 4494F:6 900926 32 i s~., , ~ + - -,
- -. .- . . . _ - . - - . - = . - . . . - . - - . -
l 1 a o A reduced thermal design flow of 93.600 gpm per loop, o A 10% reduction of ECCS safety injection flow, i o A delay time increase to 40 seconds for the ECCS safety injection, o A RWST minimum water temperature of 40'F, o A RCS temperature operating band of 1 6'F, o A RCS pressure uncertainty of t 50 psia, o A widened accumulator water level range of 860 cubic feet to 940 cubic feet.with a nominal water level of 900 cubic feet, o The fuel temperature and rod internal pressure data based on the improved fuel thermal model, Reference 6, o Containment Mini Purge Isolation, o Thimble Plug Removal, o RAOC operation with a +10/ 20% AFD band at 100% RTP, o Steam Generator tap relocation. The VANTAGE 5 fuel features, as applied at the VEGP Units 1 and 2, result in a fuel assembly that is more limiting than the LOPAR fuel currently impler.ented p at VEGP Units 1 and 2 with respect to large break LOCA ECCS performanca,
't/ Reference 2. As such, VANTAGE 5 fuel has been analyzed herein.
5.2.1.2 Method of Analysis The method used in analyzing the VEGP Units 1 and 2 for VANTAGE 5 fuel, including the computer codes used and the assumptions are described in detail > in Appendix B, Section 15.6.5.. 5.2.1.3 Results The results of the large break LOCA analysis for VEGP Units 1 and 2, it cluding tabular and plotted results of the break spectrum analyzed are provided in
' Appendix B.-Section 15.6.5.
I-l
-4494F:6-90092s 33 ~ _ - - - . _ .
! Reference 25 stated three restrictions related to the use of the 1981 J
Evaluation Model (EM) + BASH calculational model. The application of these
~
restrictions to the plant specific large break LOCA analysis was addressed with the following conclusions: ) i VEGP Units 1 and 2 are neither an Upper Head Injection (VHI) or Upper Plenum Injection (UPI) plant; so restriction 1 does not apply, The VEGP Units 1 and 2 plant specific large break LOCA analysis considers both minimum and maximum ECCS safeguards to address restriction 2. The Cd =0.6 Double Ended Cold Leg Guillotine (DECLG) with minimum ECCS flows was found to result in the most limiting consequences. Concerning restriction 3, a chopped cosine power shape was used in the large break LOCA analysis for the VEGP Units 1 and 2. l 5.2.1.4 Conclusions i The large break LOCA' analysis performed for:the VEGP Units 1 and 2 has demonstrated that for breaks up to a double ended severance of the reactor - coolant piping, the Emergency Core Cooling System (ECCS) will meet the acceptance criteria of Title 10 CFR Part 50 Section 46. That is: 4
1. The calculated peak cladding temperature will remain below the required 2200*F.
l
2. The amount of fuel cladding that reacts chemically with the water or steam does not exceed one percent of the total amount of fuel rod zircaloy cladding in the reactor. 4

3. . The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching, i
L 4494F:6 900926 34 l
4. The core remains amenable to cooling during and after the LOCA. )
i 1
5. The core temperature is rerJuced and decay heat is removed for an extended period of time. This is required to remove the heat goduced by the long lived radioactivity remaining in the core.
The large break LOCA results for all breaks analyzed is shown in Table l 15.6.5 2 of Appendix B, Section 15.6.5. The large break LOCA analysis for the VEGP Units 1 and 2 assuming a full core of VANTAGE 5 fuel, utilizing the 1981 Evaluation Model (EM) with the BASH calculational model, resulted in a peak cladding temperature of 2058'F for the limiting DECLG break at a total peaking factor of 2.50. The maximum local metal water reaction was 5.62%, and the total core wide metal water reaction was less than 0.3% for all cases analyzed. The clad temperature transients turn around at a time when the core geometry was still amenable to cooling. r Also the effec,t of the transition core' cycles is conservatively evaluated to be at most 50'F higher for the calculated peak cladding temperature which would yield a transition core PCT of 2108'F, The transition core penalty can be accommodated by the margin to the 10CFR50.46, 2200'F limit. It can be determined from the results contained in Appendix B, Section 15.6.5 that the ECCS analysis for the VEGP Units 1 and 2 remain in compliance with the requirement of 10CFR50.46 including consideration for transition core configurations. l . 4494F:6 9ecers 35
q V 5.2.2 Small Break LOCA 5.2.2.1 Description of Analysis and Assumptions for 17x17 VANTAGE 5 Fuel The small break LOCA was analyzed assuming a full core of VANTAGE 5 fuel to determine the peak cladding temperature. This is consistent with the methodology employed in WCAP 10444 P A, Reference 2, for the 17x17 VANTAGE 5 transition. The currently approved NOTRUMP small break ECCS evaluation model, References 27 and 28, was utilized for a spectrum of cold leg breaks. Appendix B, Section 15.6.5, includes a full description of the analysis and assumptions utilized for the Westinghouse VANTAGE 5 ECCS LOCA analysis. Pertinent assumptions for the VEGP Units 1 and 2 small break LOCA analysis include: o An uprated core thermal power of 3565 MWt, o 10% steam generator tubes plugged for each of the four steam generators, p o An F39 of 1.70 for VANTAGE 5 and LOPAR fuel, and a total peaking f actor for Fg of 2.58, o A two line segment K(z) curve with a value for K(z).= 1.0 up to 6 ft. and a linear ramp to a value of 0.92 at 12 ft, o A reduced thermal design flow of 93,600 gpm per loop, o A 10% reduction of ECCS safety injection flow, o A delay time increase to 40 seconds for the ECCS safety injection, o A RWST minimum water temperature of 40*F, o A RCS temperature operating band of i 6'F, o A RCS pressure uncertainty of i 50 psia, o A widened accumulator water level range of 860 cubic feet to 940 cubic feet with a nominal water level of 900 cubic feet, o The fuel temperature and rod internal pressure data based on the improved fuel thermal model Reference 6. o Thimble plug removal, o RAOC operation with a +10/ 20". AFD band at 100% RTP, o Steam Generator tap relocation. f3 O l 4494 :6 900926 36
(oy Sensitivity studies performed using the NOTRUMP small break evaluation model have demonstrated that VANTAGE 5 fuel is more limiting than the OFA fuel in the calculated ECCS performance. Similar studies using the WFLASH evaluation model have previously shown the OFA fuel is more limiting than the LOPAR fuel. For the small break LOCA, the effect of the fuel difference is most pronounced during core uncovery periods and, therefore, shows up predominantly in the LOCTA IV calculation in the evaluation model analysis. Consequently, the previous conclusion drawn from the WFLASH studies, regarding the fuel difference, may be extended to the NOTRUMP evaluation model analysis. On this basis, only VANTAGE 5 fuel was analyzed, since it is the most limiting of the two types of fuel (LOPAR and VANTAGE 5) that would reside in the VEGP Units 1 and 2 cores. 5.2.2.2 Method of Analysis The methods of analysis, including codes used and assumptions, are de scribed in detail in Appendix B. Section 15.6.5. f Q) 5.2.2.3 Results The results of the small break LOCA analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Appendix 8, Section 15.6.5. 5.2.2.4 Conclusions The small break VANTAGE 5 LOCA analysis for the VEGP Units 1 and 2, utilizing the currently approved NOTRUMP Evaluation Model resulted in a calculated Peak Cladding Temperature (PCT) of 2056'F for the 3 inch diameter cold leg break. The analysis assumed a limiting small break power shape consistent with a LOCA Fg(z) envelope of 2.58 at the core midplane elevation and 2.368 at the top of the core. The maximum local-water reaction is 7.747., and the total core metal-water reaction is less than 0.3Y. for all cases analyzed, The clad temperature transients turn around at a time when the core geometry is still amendable to cooling. f3 4494F:6 900926 37 l
1. Analyses presented in Appendix B, Section 15.6.5 show that one centrifugal pump and one high head pump, together with the accumulators, provide sufficient core flooding to keep the calculated peak clad temperature well below the required limits of 10CFR50.46. It can also be seen that the ECCS analysis remains in compliance with all other requirements of 10CFR50.46 and the peak cladding temperature results are well below the peak cladding , temperatures calculated for the large break LOCA. Adequate protection is therefore afforded by the ECCS in the event of a small break LOCA. 5.2.3 Transition Core Effects on LOCA s W6en assessing the effect of transition cores on the large break LOCA analysis, it must be determined whether the transition core can have a greater calculated PCT than either a complete core of the LOPAR fuel assembly design or a complete core of the VANTAGE 5 design. For a given peaking factor, the only mechanism available to cause a transition core to have a greater
~
calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch. Hydraulic resistance mismatch will exist only for a transition core and is the only unique difference between a cory lete core of either fuel type and the transition core. 5.2.3.1 Large Break LOCA The large break LOCA analysis was performed with a full core of VANTAGE 5 and-conservatively applies the blowdown results to transition cores. The VANTAGE 5 differs hydraulically from the LOPAR fuel assembly design it replaces. The differences in the total assembly hydraulic resistance between the two designs is approximately 10% higher for VANTAGE 5. O 4494F:6 9009rs 38
An evaluation of hydraulic mismatch of approximately 10% showed an , insignificant effect on blowdown cooling during a LOCA. The SATAN.VI computer code models the crossflows between the average core flow channel (N 1 fuel assemblies) and the hot assembly flow channel (one flow assembly) during a blowdown. To better understand the transition core large break LOCA blowdown transient phenomena, conservative blowdown fuel clad heatup calculations have been performed to determine the clad temperature effect on the new fuel design for mixed core configurations. The effect was determined by reducing the axial flow in the hot assembly at the appropriate elevations to simulate the effects of the transition core hydraulic resistance mismatch. In addition, the Westinghouse blowdown evaluation model was modified to account for grid heat transfer enhancement during blowdown for this evaluation. The results of this evaluation have shown that no peak cladding temperature penalty is observed during blowdown for the mixed core. Therefore, it is not necessary to perform a blowdown calculation for the VANTAGE 5 transition core configuration because the evaluation model blowdown calculation performeo for the full VANTAGE 5 core is conservative and bounding. Since-the overall resistance of the two types. of fuel is essentially the same during blowdown, only the crossflows during core reflood due to.the IFM grid need to be evaluated. The LOCA analysis uses the BASH computer code to calculate the reflood transient, Reference 25, which utilizes the BART code, Reference 29. A detailed _ description of the BASH code is given in Appendix B. Fuel assembly design specific analyses have been performed with a version of the BART computer code, which accurately models mixed core i configurations during reflood. Westinghouse transition core designs, including specific 17x17 LOPAR to VANTAGE 5 transition core cases,-were
-analyzed. For this case, BART modeled both fuel assembly types and predicted-the reduction in axial flow at the appropriate elevations. As expected, the increase in hydraulic resistance for the VANTAGE 5 assembly was shown to produce a reduction in reflood steam flow rate for the VANTAGE 5 fuel at mixing vane grid elesations for transition core configurations, This reduction in steam flow rate is partially offset by the fuel grid heat transfer enhancement predicted by the BART code.durine reflood. The various
, fuel assembly specific transition core analyses perfcemed resulted in peak cladding temperature increases of up to 50*F for core axial elevations 4494h 6 900926 39
i-1 that bound the location of the PCT. Therefore, the maximum PCT penalty possible for VANTAGE 5 fuel residing in a transition core is 50'F, Reference 2. Once a full core of VANTAGE 5 fuel is achieved, the large break LOCA analysis will apply without the transition core penalty. 5.2.3.2 Small Break LOCA j The NOTRUMP computer code, Reference 27, is used to model the core hydraulics during a small break event. Only one core flow channel is modeled in the NOTRUMP code, since the core flow during a small break is relatively slow, providing enough time to maintain flow equilibrium between fuel assemblies j (i.e., no crossflow). Therefore, hydraulic resistance mismatch is not a
" factor for small break. Thus, it is not necessary to perform a small break evaluation for transition cores, and it is sufficient to reference the small break LOCA for the complete core of the VANTAGE 5 fuel design, as bounding for all transition cycles.
5.2.4 Blowdown Reactor Vessel and Loop Forces f O The forces created by a hypothesized break in the RCS piping are principally caused by the motion of the decompression wave through the RCS. The strength of the decompression wave is primarily a result of the assumed break opening time, break area and RCS operating conditions of power, temperature and pressure. These parameters will not be significantly affected by a change in fuel at the VEGP Units 1 and 2 from 17x17 LOPAR to VANTAGE 5 fuel. The forces in the vicinity of the core are affected by the core flow area / volume. Since
- there will be no significant change in the core flow area / volume for VANTAGE 5 fuel, there will not be an adverse change in the forces calculated for a hypothesized LOCA. Forces acting in the RCS loop piping as a result of a hypothesized LOCA are not influenced by changes in fuel assembly design.
Reviewing LOCA hydraulic forcing functions used in the plant design for the VEGP Units, it has been determined that LOCA forcing functions for limited displacement inlet and vitlet nozzle breaks are acceptable for the evaluation of VANTAGE 5 fuel. Taking ciadit for leak Before Break (LBB) technology, it has been determined thit any differences due to the implementation of VANTAGE
' 5 can be offset by the margin that exists between limited displacement v'essel 4494F;6 900926 40 - m e er ,,-w,, ,,,n.,
1 nozzle breaks and RCS branch line breaks. Thus the implementation of VANTAGE 5 fuel at VEGP Units 1 and 2 will not result in an increase of the calculated , consequences of a hypothesized LOCA on the reactor vessel internals or RCS loop piping. 5.2.4.1 Evaluation of RCCA Insertion in order to assess the feasibility of taking credit for the RCCA insertion during a postulated rupture of one square foot cf the Reactor Pressure Vessel (RPV) inlet nozzle break and one square foot of tae RPV outlet nozzle break, the dynamic loads that were calculated for these tuo breaks were combined with the cross flow. loads resulting from the RPV it.".et and outlet nozzle breaks on the VEGP Units 1 and 2. The evaluation showed that the maximum guide tube deflection for the RPV inlet and outlet nozzle breaks was within the allowable deflection limits which were established during the scram deflection testing for the 17x17 guide tube.
q. Consequently, the control r'ods in the VEGP Units 1 and 2 can be inserted C/ following the postulated LOCA breaks described above.

5.2.5 Post LOCA Long-Term Core Cooling - ECCS Flows, core suberiticality.
and switchover of the ECCS to hot leg recirculation The implementation of VANTAGE 5 fuel at the VEGP Units 1 and 2 does not affect the assumptions for decay heat, core reactivity or boron concentration for sources of water residing in the containment sump Post LOCA. Thus, the post LOCA long term core cooling ECCS flows, and switchover of the ECCS to hot leg recirculation are not significantly affected by the implementation of , VANTAGE 5 fuel. Additionally Westinghouse during a specific reload design, performs an independent check on core suberiticality for each cycle operated at VEGP Units 1 and 2.- , O 4494h6 900926 41
._ . . ~ _ _ . _ _ __ ,. _ _ _ _ _ . _ . _ ._ _._- __ _- _._- __ - _ - . _ . . . - . -
l l c i Currently, to show that the core temperature is reduced and decay heat is ' removed for an exter.ded period of time, it has been determined that the core - must remain suberitical. (Keff <l.0) after a postulated LOCA. As a result of the effect of LOCA forces on the control rods (RCCAs) for breaks greater than ! 1.0 ft2 in the Reactor Coolant System (RCS), an additional requirement has been needed to show that subcriticality can be maintained while taking no credit for the control rods. This set of circumstances led to a requirement i that-the borated ECCS water provides sufficient negative reactivity to keep the' reactor' core suberitical (Keff <1.0), All Rods Out (ARO), No Xenon, at'the . I most reactive time in life for reactor coolant system temperatures s 212'F. This requirement has been easily satisfied for years by the Refueling Water
. Storage Tank (RWST) minimum Technical Specification requirement of 2000 ppm for the boron concentration. Consequently, much of the plant equipment design and~ qualification. assumed an RWST boron concentration in the range of 2000 to 2600 ppm. 'The current trend to a longer core cycle life and a Positive Moderator Temperature Coefficient (PMTC) has required increased RWST boron concentrations to meet the_ post LOCA long term suberiticality requirement,
[ ' because the adepd reactivity from the RCCAs is.not-considered. Increasing the
- RWST boron contentration, in: turn, increases the demands on related LOCA ,
requirements and equipment. Assuming that_ the control rods do not enter the core dtring the LOCA event is one area of conservatism identified in the curr c.+ post LOCA long term core cooling methodology. This is based on the assumption that the LOCA is , initiated instantaneously and may cause enough damage to the upper internals that the control rods cannot enter the core. -Currently, this assumption is made for' reactor coolant-system breaks greater than 1.0 ft2 when-performing the ECCS LOCA thermal analyses, For the postulated LOCA with
~
breaks in the reactor coolant system smaller than 1.0 2ft , credit for control rods is used in.the ECCS thermal analysis of the small break LOCA. However, taking credit for control rods during the long term core cooling ? phase of.the I.0CA is a departure from the licensing commitment stated in Section 4.5 of WCAP 8339, Westinghouse ECCS Evaluation Model - Summary." Border 1on,-F. M., et al., June 1974. This commitment states that the core will- be maintained in a shutdown condition with borated water, with no mention
- 4494F:6 900926 42 1 - e . _ . . .w. +.-..w.3,.._4 --6.., %,,. ,,mr.,m.,y.n,. . _ , .-w..n.m.,,,,, ,.,._,_p. ,_,y,-._ryy.m..y.,,,,g.,.. . , . _ _ . , .[ , y y,% 7
1 l l l O. of taking credit for control rods. The following information will provide a l discussion on the use of LBB technology to satisfy the long term core cooling : portion of the postulated LOCA and replace the licensing approach in WCAP 8339. General Discussion j Using the LBB technoiogy has demonstrated that safe shutdown of the reactor can be accomplished based on detection of a leak before a significant rupture in the RCS piping can occur. LBB technology has ; eliminated analyses of main RCS piping breaks in the evaluation of the mechanical and structural integrity of the reactor coolant system for a postulated LOCA and this tas led to a reduction in break sizes considered in analysss to determine the LOCA hydraulic forcing functions on reactor vessel internals and reactor coolant system loop supports. Generally, the analysis to generate the LOCA hydraulic forcing functions is analyzed or evaluated for reactor coolant system branch line breaks which are less than 1.0 ft2 in break area. Since these breaks have been considered , acceptable for use when LBB technology is applied to demonstrate the structural and mechanical integrity of the reactor coolant system, taking credit for a mechanical feature such as control rods (RCCAs) when determining the post-LOCA boron requirements should also be. acceptable as long as control rod insertion can be demonstrated for branch line breaks. Using LBB technology to satisfy 10CFR50.46 Acceptance Criterion 5 for long term cooling would be an extension of the same approach used for the calculation of grid loads from a' postulated LOCA to demonstrate a coolable geometry. For post LOCA long term core cooling, the mechanical-integrity evaluation based on branch line break LOCA hydraulic forcing functions for the reactor vessel components, and specifically insertion of the control rods (RCCAs), has been evaluated in Section 5.2.4.1 to establish that credit for control rods can be taken for the post accident long term core cooling aspect of the postulated LOCA. However, for the large break ECCS LOCA thermal analyses, no credit for control rods will contirue to be assumed to demonstrate compliance with the first three acceptance criteria of 10CFR50.46. l 4494F:6 900926 43
i ) c conclusions
Large break LOCAs are considered hypothetical. events and are analyzed to ]
demonstrate the effectiveness of the ECCS and limit the core peaking factors. p i Analysis of _the large break LOCA for VEGP Units 1 and 2 currently limits the core peaking factors to assure that the calculated PCT is less than ~ _2200'F,:a specific requirement of 10CFR50.46. Limiting the core peaking ) factors by the Large Break LOCA ECCS thermal analysis has not created ; conditions undesirable to safety. The large break LOCA ECCS analysis would still be performed without taking credit for RCCA insertion to determine acceptable thermal limits. To meet the post LCCA long term core cooling
-requirement with'RCCA insertion .the borated ECCS water should provide
, sufficient negative reactivity to keep the reactor core suberitical (Keff ! <l.0, A11. Rods In minus 2 (ARI-2), No Xenon, most reactive time in life, at ) r 'RCS temperatures s 212'F). The ARI 2 requirement-is to address the loss I -of_one rod due to a rod ejection event which would create a LOCA and the failure of one rod to insert. 9
~
5.2.6 Steam Generator Tube Rupture 5.2.6.1 Introduction - Design ~ basis analyses of.a Steam Generator Tube Rupture (SGTR) event at the -l VEGP Units 1 and 2 have been performed to assess the effect of the transition j to a core with VANTAGE 5 fual assemblies. The analyses performed include a o ' demonstration of margin to steam generator overfill in the event of a tube' ; rupture and an analysis'which demonstrates that the calculated offsite radiation doses are within the limits set forth in 10CFR100. The. analyses performed bound operation of VEGP Units 1 and 2 at an uprated
NSSS-power of 3579 MWt with a LOPAR/ VANTAGE 5 fuel transition core, LOPAR_ fuel core or VANTAGE 5 fuel core with up to 10% uniform steam generator tube i
plugging. The analyses also considered a 90 second delay time for auxiliary
.feedwater flow delivery and;the steam generator lower narrow range level tap -relocation. Since the assumption that the initial primary coolant activity at the Standard Technical Specifications limit will not change for VEGP Units 1 O
and 2 due to the proposed change in fuel, the parameters which effect the i 4494h6 900926 44
(D offsite radiation doses calculated for the FSAR SGTR analysis are primary to C/ secondary break flow and the steam released from the ruptured steam generator ; to the atmospheie. Therefore, the analyses to support the transition to 1 VANTAGE 5 fuel assess the effect of the fuel change on primary to secondary i break flow and steam released via the ruptured steam generator. ! 5.2.6.2 Methodology The steam generator tube rupture analyses were performed for VEGP Units 1 and 2 using the methodology and the assumptions described in WCAP ll731 Reference 30. Plant response to the SGTR event was modeled using the LOFTTR2 computcr code with conservative assumptions of break size and location, condenser availability and initial secondary water mass in the ruptured steam generator. The analysis methodology includes the simulation of the operator actions for recovery from a steam generator tube rupture event based on VEGP Units 1 and 2 Emergency Operating Procedures, which were developed from the Westinghouse Owners Group Emergency Response Guidelines. Since the limiting single failure is different for the overfill analysis and the offsite radiation dose analysis, the two analyses were performed using different single failure assumptions. For the margin to overfill analysis, the single failure was assumed to be that the auxiliary feedwater flow control valve located in the flow path from the turbine driven auxiliary feedwater pump to the ruptured steam generator fails to close when the isolation of the ruptured steam generator is being performed. In the offsite radiation dose analysis, the ruptured steam generator PORV was assumed to fail open when the isolation of the ruptured steam generator was performed. 5.2.6.3 Margin to Overfill Analysis Results The LOFTTR2 analysis to determine the margin to overfill was performed for the time period from the steam generator tube rupture until the primary and secondary pressures are equalized and break flow is terminated. The water volume in the secondary side of the ruptured steam generator was calculated as
; a function of time to demonstrate that overfill does not occur. The results of 4494F:6 900976 45
1 the ar.alysis demonstrate that the transition to VANTAGE 5 fuel does not change the r.oncluston that there is margin to overfill calculated for VEGP Units 1 and 2 in the event of a tube rupture. 5.2.6.4 offsite Radiation Dose Analysis Results f for the offsite radiation dose analysis, the primary to secondary break flow and the steam ielease to the atmosphere from both the ruptured and intact generators were ulculated for use in determining the activity released to the atmosphere. The c ss releases were calculated with the LOFTTR2 program ~ from the initiation of the event until termination of the break flow. For the time period following break flow termination, steam releases from and feedwater flows to the ruptured and intact steam generators were determined from a mass and energy balance using the calculated RCS and steam generator conditions when primary to secondary tube leakage was terminated. The mass release information was used to calculate the radiation doses at the exclusion area boundary and at the low population zone assuming that the primary coolant O -activity is at the Standard Technical Specifications limit prior to the accident. The results of the analysis to support the transition to VANTAGE 5 fuel'show that the offsite doses for VEGP Units 1 and 2 are well within the allowable guidelines specified in the Standard Review Plan, NUREG 0800, FSAR Chapter 15.6.3, and 10CFR100. Refer to Appendix C' for the Offsite radiation dose analysis results. A more' complete description of the SGTR radiological. assessment and results is provided in Appendix B. 5.2.7' Containment Integrity Mass and Energy Release for the short term mass and energy release analysis presented in Section 6.2.1.4.3 of the .Vogtle FSAR, there is no impact on this analysis, because changes in fuel type have a very small effect on the transient. For subcompartment analyses, approximately only the first 3 seconds of the blowdown are considered. Therefore, the current shor. term mass and energy release analysis remains valid and applicable. 4 4494F:6*800926 46
i For the long term mass and energy analysis presented in the Vogtle FSAR, there Q(3 is no adverse effect on the analysis, because the plant T avg remains essentially the same for the change in fuel. Additionally the VANTAGE 5 fuel closely resembles the 17x17 LOPAR fuel. Since the fuel type dimensionally rmains basically the same, there is no difference in initial core stored energy, and hence no additional energy would be available for release to containment. Thus, the current analysis remains valid and applicable. 5.3 Radiological Evaluation The effect of the transition to VANTAGE 5 fuel on the radiological source terms and, subsequently, on the releases, both normal and accidental, is primarily due to the extension of maximum fuel burnup levels for both the VANTAGE 5 and LOPAR fuel types. This safety evaluation and Appendix C address the fuel burnup to g maximum level of 60,000 MWD /MTV for the lead rod average. This safety evaluation and Appendix C are based on the current NRC position regarding extended fuel burnups as set forth in the Federal Register
m. (Reference 31; and NUREG/CR-5009 (Reference ?2). These reviewt considered U average burnups of up to 60,000 MWD /MTV for the' lead rod average which bounds both the VANTAGE 5 and LOPAR fuel designs. The radiological consequences in the VEGP FSAR remain valid for both LOPAR and VANTAGE 5 fuel assembly designs to 60,000 MWD /MTV lead rod average burnups. The only exceptions are the revised results far the Fuel Handling Accident and SGTR Event which are presented in Appendix C. A reduction in the conservatism in the assumptions of Reg. Guide 1.25 was utilized for the Fuel Handling Accidents, however, the calculations of doses remain conservative. The results in Appendix C for tnese two accident scenarios are within the guideline values, as referenced in the NRC Standard Review Plan.
u 4494F 6 900926 47
/
,.m %)- 6.0
SUMMARY
OF TECHNICAL SPECIFICATIONS CHANGES Table 6-1 presents a list of the Technical Specifications changes and the justification for the changes The changes noted in Table 61 are given in the proposed Technical Specifications page mark ups (see Enclosure 3 of this submittal). 28 4 l l 7s
\v 4494F:6 900926 4g
TABLE 6-1 l,-~T
SUMMARY
AND JUSTIFICATION FOR THE V0GTLE UNITS 1 AND 2 V TECHNICAL SPECIFICATIONS CHANGES FOR VANTAGE 5 FUEL bag Section Descrietton Justiftention 22 2.1 Change to core limite, This change is a result of changes 2 a,2 8 2.2 to the OTDT and OPDT associated with the VANTAGE 5 fuet-2 9,2a10 Setpoints. and to provide operational 2 11 flexibility. B 2*1 2.1.1 Basis Addition of the Witt ) This change reflects the DNS B22 and WRs.2 correlation correlation used in analyses. 8 3/4 2+1 3/4.2 Seels 3 3/4 2 2 8 3/4 2 4 8 3/4 4 1 3/4.4 seals 3/4 1 11 3.1.2.5 twr. minisun solution this change is to allow 3/4 1*12 3.1.2.6 tenversture operational flexibility. 3/4 1 13 3/4 5 10 3.5.4 8 3/4 5 2 3/4.5.4 sesis avst Bases this change is to allow operational flexibility. 3/4 1 19 3.1.3.4 Revised rod drop time This change is a result of changes to less than or og,aal to associated with the VANTAGE 5 fuel. 2.7 seconde The effect of this increase on the tafety erlyste has toen considered, 3/4 2 1, 3/4,2.1 4xist Flux Difference This change is made to give the [m 3/4 2 2 and Fo;Z) changes plant operatina flexiblLity and V 3/4 2 4 3/4 2 6 3/4.2.2 also se a result of changes associated with the VANTAGE 5 fuel. 3/4 2 7 8 3/4 2 1 8 3/4.2 Basis B 3/4 2 2 8 3/4 2 4 , 3/4 2 13 3/4.2.5 DN8 parameter changes this change is made to give the B 3/4 2 5 3/4.2.5 Basis plant operating flexibility and *
. elso se a result of changes essociated with the VANTAGE 5 fuel.
3/4 3 33 3/4.3.2 -Pressurlier pressure This change le to provide trip setpoint operational flexibility. 3/4 5+1 3/4.5.1 Acctmutator Water Level This change is to provide aange operational flexibility. 6 21 6.8.1.6 - Core operating Limits This change le to provide j 6 21e operational flexibility. 1 l l l i t: _ D y . 49 I
i.
7.0 REFERENCES
l. .Davidson, S. L. and lorii, J. A., " Reference Core Report - 17x17 Optimized. Fuel Assembly," WCAP-9500 A, May 1982.

2. Davidson, S. L. and Kramer, W. R. (Ed.) " Reference Core Report VANTAGE 5 Fuel Assembly," WCAP 10444-P A, September 1985.
13 . Davidson, S. L. (Ed.) et al., " Extended Burnup Evaluation of Westinghouse Fuel," WCAP 10125 P A, December 1985.
4. Davidson, S. L. (Ed.), et al., " Westinghouse Reload Safety Evaluation
. Methodology," WCAP 9272-P-A, July 1985.
5. ; Miller, Ji V., " Improved Analytical Models Used in Westinghouse Fuel Rod
' Design Computations," WCAP-8720, October 1976.
6. Weiner, R. A., et al., " Improved Fuel Performa'nce Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.

7. Skaritka, J.,. (Ed.), " Fuel Rod Bow Evaluation," WCAP-8691, Revision 1, July 1979.
8.- Davidson, S. L., and lorii,.J. A. (Eds.), " Verification Testing a'nd Analyses of the 17x17 Optimized Fuel Assembly," WCAP 9401-P-A, August 1981.
9. Miller, R. W., et al., " Relaxation of Constant Axial Offset Control-FqSurveillance Technical Specification," WCAP-10216-P-A, June 1983.
-10. Davidson, S. L. (Ed.), et al., "ANC: Westinghouse Advanced Nodal Computer Code," WCAP-10965-P-A, September 1986.
11. Letter from E. P. Rahe (W) to Miller (NRC) dated March 19, 1982,

\ _-
NS EPR 2573, WCAP 9500 and WCAPS-9401/9402 NRC SER Mixed Core Compatibility Items. 4494F:6 900926 50
Y r " Supplemental- Acceptance No. 2
12. Letter from C. O. Thomas (NRC) to Rahe (W) 1] . for. Referencing Topical Report WCAP 9500," January 1983.

13. Friedland, A.J., and Ray, S., " improved THINC IV Modeling for PWR Core Design," WCAP-12330 P, August 1989.

14. Hochreiter, L. E., and Chelemer, H., " Application cf the THINC-IV Program to PWR Design," WCAP 8054 (Proprietary) and WCAP-8195 (Non proprietary),
September 1973.-
15. Friedland,-A. J. and Ray, S., " Revised Thermal Design Procedure,"
. WCAP-ll397-P-A, April 1989.
16. Motley, F. - E., et al.,'"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP 8762 P, July 1984.
f-s 17. Tong, L. S., " Critical Heat Fluxes in Rod Bundles, Two Phase Flow and Heat 'b Transfer in Red Bundles," Annual Winter Meeting ASME, November 1968, p.:3146.
18. Tong, L.' S., " Boiling Crisis. and Critical Heat Flux...," AEC Office of Information Services, TID-25887, 1972.

19. letter from A. C.'Thadani (NRC) to W. J. Johnson (Westinghouse), Jan. 31,
-1989,
Subject:
Acceptance for Referencing of Licen' sing Topical Report, WCAP-9226-P/WCAP 9227-NP, " Reactor Core Response to Excessive Secondary Steam Releases."
20. Motley, F. E., Cadek, F. F., "DNB Test Results for. R Grid Thimble Cold Wall Cells," WCAP-7695-L Addendum 1, October 1972.

21. Hill, K. W., Motley, F. E. , Cadek, F. F. , and Wenzel, A. H. , "Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8296, March 1974.

22. Moomau, W. H., " Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Georgia Power Vogtle 1 & 2 Nuclear Power Stations O (For RTD Bypass Loop)," WCAP-12460 (Proprietary), December 1989.
4494F:6 900926 51
23. . Moomau, W. H., " Westinghouse Revised Thermal Design Procedure Instrument;
[
~ Uncertainty Meth'odology for Georgia Power Vogtle 1 & 2 Nuclear Power !
Stations (For RTD Bypass Loop Elimination)," WCAP 12462 -(Proprietary),. December 1989.
- 24. >Schueren,1 P. and McAtee, K. R., " Extension of Methodology for Calculating Transition Core DNBR Penalties," WCAP ll837 P-A, January 1990. -1 - 25. Kabadi, J. N.,' et al., "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the-BASH Code,-" WCAP 10266-P A (Proprietary),
March 1987. 1
26. Eicheldinger, C., " Westinghouse ECCS Evaluation Model,1981 Version,"
WCAP 9200-P A, 1981, Revision.l.
27. J Lee, N.n Rupprecht, S! D., Schwarz, W. R. and Tauche, W. D.,
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," ,
WCAP-10054-P A ' August 1985.
28. -Meyer,_ P. lE., "NOTRUMP - A . Nodal Transient Small Break and General Network Code," WCAP-10079-P A, August 1985.
29., Young,LM.:et-a1.,'"BART--A1: A Computer Code for the Best. Estimate Analysis of Refloo'd Transients," WCAP 9561-P A,; with Addenda 2, 1 March ~1984. 30.' - Lewis,- R. N. , MEndler, 0. J. ,-Miller, T. A. and Rubin, "LOFTTR2 Analysis - for A Steam Generator Tube Rupture Event for_ the Vogtle Electric Generating Plant Units 1 and 2," WCAP 11731,; January 1988.. I
, 31. Federalf Register /Vol. 53. No. 39/ Monday, February 29, 1988/pages- ., _6040-6043.
y I i .
32. --Baker, D.-A., Bailey, W. J Beyer, C. E. , Bold F. C. and Tawil, J. J.,
" Assessment of the'Use of atended:Burnup Fuel in Light Water Power Reactors," NUREG/CR-5009, February 1988.
4494F:6-900926 52
_ - - ._. . . ~ _ . - . . . . . . _ _ . . . l Q,,G ' l Accendix A Hon-LOCA Accident Analyses for the Vogtle Electric Generating Plant Units 1 and 2 Transition to Westinghouse 17x17 VANTAGE-5 Fuel Assemblies lO I O :
Table of Contents A)
, Section Descriotion Pace List of Tables A-4 List of Figures A-5 15.0 Introduction A-15.0-1 15.0.1 Classification of Plant Conditions A-15.0-2 15.0.2 Optimization of Control Systems A-15.0-3 15.0.3 Plant Characteristics and Initial Conditions Assumed in the Accident An: lyses A-15.0-3 15.0.4 Reactivity Coefficients Assumed in the Accident Analyses A-15.0-3 15.0.5 Rod Cluster Control Assembly Insertion Characteristics A-15.0-4 15.0.6 Trip Points and Time Delays to Trip Assumed in Accident Analyses A-15.0-4 15.0.7 Instrumentation Drift and Calorimetric Errors, Power Range Neutron Flux ,A-15.0-5 /m (j '
15.0.8 Plant Systems and Compo' n ents Available for Mitigation of Accident Effects A-15.0-5 15.0.9 Fission Product Inventories A-15.0-5 15.0.10 Residual Decay Heat A-15.0-5 15.0.11 Computer Codes Utilized A-15.0-5 15.0.12 Limiting Single Failures A-15.0-7 15.0.13 Operator Actions A-15.0-7 15.1 Increase in Heat Removal by the Secondary System A-15.1-1 15.1.1 Feedwater System Malfunctions that Result in a l Decrease in Feedwater Temperature A-15.1-1 15.1.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow A-15.1-2 15.1.3 Excessive Increase in Secondary Steam Flow A- 1., .1 15.1.4 Inadvertent Opening of a Steam Ger.erator Rallef or Safety Valve A-15.1-8 O A-1 l
t i-Table of Contents-(continued) Section Descriotion Paae 15.1.5: . Steam. System Piping Failure A-15.1-8
15.25 Decrease in Heat Removal by the Ocondary System A-15;2-1 4
'15. 2.1-' Steam Pressure Regulator Malfunction or' Failure 'that Results in-Decreasing-Steam Flow A-15.2-1 .15.2.2 Loss of Electrical Load A-15.2-1 15.2.3 Turbine Trip A-15.2-1 -15.2.4' Inadvertent Closure of Main Steam Isolation Valves A-15.2 15.2.5 Loss of Condenser Vacuum and Other Events Resulting in a Turbine Trip A-15.2-5 15.2.6. Loss of Non-Emergency AC Power to the Plant . Auxiliaries A-15.2-5 l 1
15.2.7i Loss.of Normal Feedwater Flow - A-15.2-5 15.2.8 _Feedwater' System. Pipe Break A-15.2-5 15.3 10ecrease in Reactor Coolant : System Flowrate A-15.3-1
' f, ;15.3.'l- Partial Loss of' Forced- Reactor Coolant Flow ' A-15.3-1
15. 3 . 2. . Complete . Loss of Forced Reactor Coolant Flow A .15.3-3 15.3;3 Reactor. Coolant Pump Shaft Seizure (Locked. Rotor) -A-15.3-4 l i
L15;3.44 Reactor Coolant Pump' Shaft Break A-15;3-7
15. 4 f- LReactivity and Power Distribution l Anomalies A-15~.4-1 Uncontrolled: Rod Cluster Control ( Assembly Bank--
15.4.1
. Withdrawal from_ a Suberitical or Low . Power Startup Condition' A-15.4-1 115.4.2 : Uncontrolled Rod Cluster Control Assembly Bank JWithdrawal At Power A-'15.4-4 -- 15. 4. 3 - -Rod Cluster Control Assembly Misalignment -(System: Malfunction or Operator Error) A-15.4-7 '15.4.4: Startup-of an inactive Reactor- Coolant-Pump at an L
incorrect: Temperature A-15.4-ll 15.4.5 A Malfunction >or Failure of .the Flow Controller in a Boiling Water. Reactor Loop that Results in an Increased Reactor Coolant Flowrats 'A-15 4-13 p G A-2
- w we 5 -t , s
Table of Contents (continued) [\ \-J Section Descriotion Pace 15.4.6 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant A-15.4 13 15.4.7 -Inadvertent loading and Operation of a Fuel Assembly in an Improper Position A-15.4-16 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents A-15.4-16 15.4.9 Steamline Break With Coincidental Rod Withdrawal at Power A-15.4-22 15.5 Increase in Reactor Coolant Inventory A-15.5-1 15.5.1 Inadvertent Operation of the Emergency Core Cooling System Ouring Power Operation A-15.5-1 15.5.2 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory A-15.5-3 15,5.3 A Number of Boiling Water Reactor Transients A-15.5-4 15.6 Decrease in Reactor Coolant Inventory A-15.6-1 .I G 15.6.1 Inadvertent Opening of a Pressurizer Safety or . '-') Relief Valve A-15.6-1 15.R References A-15 R-1 m i
>I A-3
. - - ~ ~ . . ~ . . . - . . - . . - . . . - . . - - . . ~ . . - _ . . _ - . - . - . - - . _ . - . . .
1
. List of Tables O Tab 1'e Title Paae> I 15.0.3 . Nuclear Steam' Supply System Power Ratings A-15.0-8 i
15.0.3-2: Summary of' Initial Conditions and Computer Codes Used- A-15.0-9 15.0.3-3 - Nominal Values of Pertinent- P1 ant Parameters EUsed-in'the Accident Analyses A-15.0-12 - 15.0.6 Trip Points and Time Delays to Trip Assumed in ! Accident Analyses A-15.0-13 ; 15.1.2-1 Time Sequence of Events for Incidents That Result , in.an increase in Heat Removal by'the Secondary ~ System = A-15.1-9 115.2.3-1 Time Sequence of Events- for Incidents Which Result in a Decrease in Heat Removal- by the Secondary System _ A-15.2-6 15.3.1 Time Sequence of Cvents for Incidents Which Result in a Decrease in Reactor Coolant System Flowrate ~ A-15.3-8 15.3.3-1 Summary of Results fer the Locked Rotor Transient - g (Four 1. oops Operating Initially) A-15.3-9 15:4.1-1 Time Sequence.of. Events for Incidents Which Result- ' in Reactivity and . Power Distribution Anomalies 'A-15.4-26 15.4.8-1 Parameters Used in the Analysis'of the Rod. Cluster'
-Control Assembly Ejection. Accident A-15.4-29 '15.5?l-1 Time Sequence of Events for incidents -Which Result' in-an. Increase in Reactor Coolant Inventory A-15.5-5 15.6.1-1 Time Sequence of Events for Incidents Which Cause a -
Decrease in Reactor Coolant Inventory. A-15.6 , I O A-4
= = . - _ . -
. , , _ , . _. -- _ -.. - , __. .---.__m.._._.__.-__.- _. . . . - - . . . _
l List of Ficures i 1 D) (,, Fiaure Title ___ Pace.___ 15.0.4 Doppler Power. Coefficient Used in Accident
. Analyses A- 15. 0 1.4 --
15.0.4-2 Moderator Density Coefficient Used in Accident Analyses A-15.0-15
-15.0.5-1 Rod Position vs. _ Time on Reactor Trip A-15.0-16 15.0.5-2 ' Normalized RCCA Reactivity _ Worth vs. Fraction Full Insertion A-15.0-17 i 15.0.5-3 No malized RCCA Bank Worth vs. Time After Trip A-15.0-18 15.0.6-1 Osertemperature and Overpower AT Protection A-15.0-19 15.0.6-2 f(AI). Penalty A-15.0-20 15.1.2 Transients ' for Feedwater Control Valve' Milfunction A-15.1-10 15.1.2-2' Transients.for Feedwater Control-Valve Malfunction A-15.1-11 15.1.3-1 10% Step Load Increase, Min. Moderator Feedback,-
Manual Rod-Control A-15.1-12 i i O' '15.1.3-2 10% Step. Load Increase, Min Moderator Feedback.. Manual Rod Control A-15.1-13 15.1.3-3 10% Step Load Increase, Max Moderator Feedback, . .
- Manual Rod Control A-15.1-14 15.1.3-4 10% Step Load. Increase, Max. Moderator Feedback, ' ' Manual Rod Control A-15.1-15 15.1.3-5 10% Step Load Increase, Min. Moderator Feedback, -
- - Automatic Rod Control A-15.1-16 , 15.1.3-6 10% Step Load Increase, Min. Moderator Feedback,
- Automatic Rod Control A-15.1-17 15.1.3-7 - 10% Step Load Increase, Max. Moderator Feedback, Automatic Rod Control A-15.1-18' 15.1.3-8 10% Step load . Increase, Max. Moderator Feedbac(,
Automatic Rod Control A-15.1-19 15.2.3-1 Turbine Trip Accident With Pressurizer Spray and Power-Operated . Relief Valves, Min. Moderator Feedback A-15.2-8 L O A-5
, _ . _ _ _ . . _ . . . .. . _m . . _ _ m .__ _.-
List of Fiaures (continued): Fioure- Title Paae 15.2.3-2 . Turbine Trip. Accident With Pressurizer Spray and Power-0perated Relief Valves, Min. . Moderator Feedback A-15.2-9 15.2.3-3 Turbine Trip Accident With Pressurizer Spray and Power-Operated Relief Valves, Max. Moderator Feedback A-15.2-10 15.2.3-4 Turbine Trip Accident With Pressurizer Spray and Power-0perated Relief Valves, Max. Moderatu Feedback- A-15.2-Il
~15.2.3-5 Turbine Trip Accident Without Pressurizer Spray and Power-0perated Relief Valves, Min. Moderator Feedback A-15.2-12 15.2.3-6 -Turbine Trip Accident Without Pressurizer Spray and Power-0perated Relief Valves, Min. Moderator feedback A-15.2-13 15;2.3-7f Turbine Trip Accident Without Pressurizer Spray and Power-0perated Relief Valves, Max. Moderator -Feedback A-15.2-14 15.2.3-8' Turbine-Trip Accident Without Pressurizer Spray and > .- Power-Operated Relief Valves, Max. Moderator . Feedback A-15.'2-15 '15. 3.1-l = Flow Transients.for~4 Loops in Oferation,.
2 Pumps Coasting Down. A-15.3-10 15.3.1-2 Nuclear Power and Pressurizer Pressure Transients for.4 Loops in Operr. tion, . 2 Pumps Coasting Down A-15.3-ll
^'
15.3.1-3 Average.and Hot Channel Heat ' lux Transients. *- t for 4 Loops in Operation, 2 Pumps Coasting Down A-15.3-12 15.3.1 DNBR Versus Time for.4 Loops in Operation, 2 Pumps Coasting Down A-15.3-13
-15. 3. 2 Flow Transients for 4 Lotps in Operation. ~
4 Pumps-Coasting Down A-15.3-'14 15.3.2-2 Nuclear-Power and Pressurizer, Pressure Transients for 4 Loops 'n Operation, . 4 Pumns Coasting Down A-15.3-15
'15 3.2-3' Averfge and Hot Channel Heat Flux Transients for 4 Loops in Operation, 4 Pumps Coasting Down A-15.'3-16
( A-6
-, . _ = _ _ _ _
-. . . - - - . . . - - . - - . - . . ~ _ . .-
l l
' List'of Fioures ~
(continued) l Fioure Title Pace-15.3.2 DNBR-Versus Time for 4' Loops in Operation, 4 Pumps Coasting Down A-15.3-17 , 15.3.3-la Flow Transients.for.4 Loops in Operation, 1 Locked' Rotor With Offsite Power Available A-15.3-18
- 15.3.3-lb - Flow Transients for. 4 Loops in Operation, 1 Locked Rotor Without Offsite Power Availabie A-15.3-19 15.3.3-2a Peak Reactor Coolant Pressure _for 4 Loops in '
Op'eration, 1 Locked Rotor With Offsite Power Available A-15.3-20 15.3.3-2b Peak Reactor Coolant Pressure'for 4-Loops in Operation, 1. Locked Rotor Without Offsite Power Available .A-15.3-21
- 15.3.3-3a- Average and Hot Channel Heat Flux Transients for 4 Loops in Operation,:1 Locked Rotor With Offsite . 7 Power Available A-15.3-22 q
15.3.3-3b- Average and Hot Channel Heat Flux Transients .for
- 4 Loops'in Operation, 1 Locked Rotor Without Offsite Power Available A-15.3-23 . 15.3.3-4a - Nuclear Power and Maximum Clad Temperature.at Hot . Spot Transients for 4 Loops in Operation, 1 Locked Rotor With Offsite Power Available A-15.3 15.3.3-4b Nuclear Power and Maximum Clad--Temperature at Hot Spot Transients:for 4 Loops in Operation, 1 Locked Rotor Without Offsite Power Available A-15.3-25 , ~ - 15.4.1-1 Nuclear Flux Transient for Uncontrolled Rod Withdrawal From a Subcritical Condition A-15.4-30 15.4.1-2 - . Thermal Flux Transient .for Uncontrolled Rod Withdrawal From a Subcritical Condition A-15.4-31 15l. 4 .' 1 Fuel and Clad Temperature for Uncontrolled eled Withdrawal From'a Subcritical Condition A-15.4 W 15.4.2-1 Uncontrolled RCCA Bank Withdrwal From Full Poder -With. Minimum Reactivity. Feedback (80 pcm/s - L Withdrawal Rate) .A-15.4-33 - 15.4.2-2 Uncontrolled RCCA Bank Withdrawal From Full Power
. 'With Minium Reactivity feedback (80 pcm/s Withdrawal Rate) A '.".4-34 l W V' A-7 o
...,,,L,
~. - . . . . - - - . . . - . - --.-. - . - . . - . . . - - - - -. ' List of Fioures '
(continued) Fioure Title ' pace 15.4;2-3. Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity feedback (80 pcm/s
.Withdraul Rate) A-15.4-35 15.4.2-4 Uncontrollei RCCA Bank Withdrawal From Full Power With Minimum 9eactivity Feedback (3 pcm/s Withdrawal Rate) -A-15.4-36 15.4.2 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (3 pcm/s-Withdrawal Rate) A-15.4-37 15.4.2-6 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (3 pcm/s Withdrawal Rate)- A-15.4-38 '15.4.2-7 Minimum DNBR vs. Reactivity Insertion Rate for Rod Withdrawal from 100% Power .A-15.4-39 15.~4 . 2-8 Minimum DNBR vs. Reactivity' Insertion. Rate for Rod '
Withdrawal from 60% Power A-15.4 15.4.2-9 Minimum DNBR vs. Reactivity Insertion Rate for Rod Withdrawal-from 10% Power A-15.4-41
,.s. 15.4.3-1 ~
Nuclear Power Transient and Core Heat Flux Transient for Oropped RCCA A-15.4-42
'15.4.4-1 Improper Startup of an inactive Reactor. Coolant Pump A-15,4 15.4.4-2 ' Improper Startup of an Inactive Reactor Coolant' Pump A-15.4-44 '15.4.4-3 Improper Startup of an Inactive Reactor. Coolant Pump A-15.4-45
15. 4 -. 4-4 . Improper Startup of an Inactive Reactor Coolant Pump A-15.4-46 15.4.4-5 Improper-Startup~of an Inactive ~ Reactor Coolant Pump A-15.4-47 15.4.8-1 Nuclear Power Transient BOL Full-Power A-15.4-48 15.4.8-2 Hot Spot Fuel & Clad Temperature-vs. Time BOL Full Power A-15.4-49 15.4.8-3 Nuclear-Power Transient EOL Zero Power A-15.4 l A-8
~__ -
List of Fiaures (continued) Fiaure Title Paae 15.4.8-4 Hot Spot Fuel & Clad Temp, vs. Time E0L Zero Power A-15.4-51 15.4.9-1 Steamline Break Coincident With Control Rod Withdrawal: Core Heat Flux, Core Average Temperature, and Steam Flow vs. Time A-15.4-52 15.4.9-2 Steamline Break Coincident With Control Rod Withdrawal: RCS Pressure, Reactivity, and Steam Pressure vs. Time A-15.4-53 15.4.9-3 Steamline Break Coincident With Ccntrol Rod Withdrawal: DNBR vs. Time A-15.4-54 15.5.1-1 Inadvertent Operation of ECCS During Power Operation A-15.5-6 15.5.1-2 Inadvertent Operation of ECCS During Power Operation A-15.5-7 15.5.1-3 Inadvertent Operation of ECCS During Power Operation A-15.5-8 15.6.1-1 Nuclear Power and DNBR Transients for inadvertent Opening of a Pressurizer Safety Valve A-15.6-4 15.6.1-2 h essurizer Pressure Transients and Core Avg. Temp. Tra.'sient for Inadvertent Opening of a Pressurizer A Safety Valve A-15.6-5 () rs O A-9 l
J Appendix A Non-LOCA Ac'ident Analysis O 15.0- Introduction i This 'section' addresses the impact of the complete transition of the Vogtle-Electric Generating Plant (VEGP) Units l. and 2.from Westinghouse 17x17
-LOPAR fuel to Westinghouse 17x17 VANTAGE =5 fuel on the FSAR Chapter 15 - non-LOCA. accident analyses. Section'15.0.11~of this report discusses the methods used for accident evaluation.
VEGP's licensing basis includes the analyses or evaluations of the non-LOCA accidents as listed below. q .,' The evaluations of all -non-LOCA accidents performed to- determine the E impact of the VANTAGE 5 fuel transition are documented in this report.
"Ihe specific design-features associated with the VANTAGE 5 fuel and the modified' safety analysis assumptions considerad in the non-LOCA safety analyses are descri. bed elsewhere in this report.
Accidents Analyzed The transients affected by the VANTAGE 5 fuel design features or modified
- safety.' analysis. assumptions as discussed-elsewhere in this report-that where analyzed =are the following: -1. Feedwater System Malfunctions That Result'in a Decrease in Feedwater Temperature (See Section 15.1.1) '\ '2. Feedwater System Malfunctions .that Result in an increase in - Feedwater Flow.(See Section 15.1.2)
3. - Excessive-Increase-in Secondary Steam Flow (See Section 15.1.3)

4. Turbine Trip-(See Section 15.2.3)

5. Partial Loss of' Forced Reactor Coolant Flow (See Section 15.3.1)

6. Complete Loss of Forced Reactor Coolant Flow (See Section 15.3.2)-
7.- Reactor. Coolant Pump Shaf t Seizure (Locked Rotor) (See Section 15.3.3)
8. Reactor Coolant' Pump Shaft Break (See Section 15.3.4) ,

9. - Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low-Power Startup Condition (See'Section'15.'4.1) 1

10. Uncontrolled Rod' Cluster ~ Control Assembly Bank Withdrawal at Power (See~Section 15-4.2) .

11. Rod Cluster Control Assembly Misalignment (System Malfunction or Operator-Error) (See Section 15.4.3)-
O
-A-15.0-1 h
a yg + q w -m ,,yw --mv , +- s w - >- *, - , - f-- r*f>
r
- -12; -Startup ~ofian:Inac tveiReactor Coolant Pump at an Incorrect '
Q Temperature (See fection-15.4.4) l 13 '. Chemical and Volsme Control System' Malfunction That Results in a 7
Decrease in:the Boron Concentration in the Reactor ' Coolant (See
-Section 15.4 A)-
14. ' Spectrum of Rod Cluster Control Assembly Ejection Accidents (See
-SectionL15,4.8)'
15. 'Steamline Break With Coincidental Rod Withdrawal at Power (See
-Section 15.4.9)
16. ' Inadvertent Operation of the Emergency Core Cooling System During Power Operation (See Section 15.5.1)

17. : Inadvertent. Opening of a Pressurizer Safety-or Relief Valve (See Section_15.6.1)
Accidents Evaluated sThe remaining non-LOCA accidents are as follows:
'l . IM dvertent Opening of a Steam Generator Relief or Safety Valve -(See Section;15.1.4);
c 2 '. : Steam System Piping' Failure (See Section 15.1.5)-
3. Steam P'ressure Regulator Malfunction or Failure That Results in.
, , -Decreasing-Steam flow.(See Section 15.2.1) :
O 4. Loss -of' External Electrical Load (See Section l'5.2.2) >
'5. 'Inadsertent Closure of Main Steam Isolation Valves-(See Section 15 2.4)~ , 6.< LossLof Condenser Vacuum and Other Events Resulting in a' Turbine Trip,:(See Section 15.2 5) . . Loss!of- Nonemergency AC Power to the Plant Auxiliaries -(See 7 '. .Section 15.2.6)
(8. Loss of Normal Feedwater Flow (See Section:i5.2.'7) . 19.1 .Feedwater System Pipe ~ Break (See Section~15.2.8)- m.' 710.c' Inadvertent. Loading-and Operation of a Fuel. Assembly-in an LImproper. Position (See.Section 15.'4~.7) c m 11. JChemical and: Volume Control System Malfunction That Increases i 1 Reactor ~ Coolant ' inventory (See .Section 15.5.2) Referetorthe: appropriate' sections for a discussion of the analysis or evaluation for each off the. accidents listed -previously. 15'0.1
; Classification of Plant Conditions-None of the1 VANTAGE 5 fuel. design features or modified safety analysis assumptions as discussed elsewhere in this report affect this section.
O
- A-15.0-2 ,a'. .s .e -. . . . , _.-..i - -,. r 1,.,r,...-e-
, . .. -- . - - . . ~ . ~. - . - .-- -. , ~ ..-
15.0.21 .Ootimization of-Control Scstems None of.the VANTAGE 5 fuelL' design features or' modified safety analysis
% -assumptions as discussed elsewhere in this report. affect this section, ~ ~
l'5.0.3 Plant Characteristics and Initial Conditions Assumed-in the Accident Analyses Table 15.0.3-1 lists the power rating values assumed in analyses performed
.for this report; The power rating values listed in Table 15.0.3-1 are-based'on the= design nuclear steam supply system.(NSSS) thermal-power output which includes the thermal power generated by the reactor coolant pumps (RCPs).
The initial conditions employed in the' accident analyses are conservative
' to bound' aL futuro plant'uprating. .The analyses of departure from nucleate '
m boiling-(DNB) accidents'use the Revised Thermal' Design Procedure '(RTOP) to define the initial conditions. Initial conditions for other accidents are obtained by applying the maximum steady-state errors to rated values (this ,
.. procedure is commonly known as Standard Thermal' Oesign Procedure or STOP). The following-steady-state errors are considered in the analyses:
o- Core power 2 percent allowance for calorimetric error (note'that this error is conservatively. applied in the positive direction in non-LOCA accident analyses) of Average, reactor. 16'F allowance for deviation'from O cool, ant system temperature programmed:Tavg (includes measurement error) 10 Pressurizer pressure- 50 psi allowance for steady-state fluctuations and measurement errors
Accidents employing = RTOP assume minimum measured flow (MMF); accidents employing STOP assume thermal design flow (T0F). - Table 15.0.3-2

summarizes the initial-condition.s:and computer codes used in the. accident Lanalyses. 1
,;The values of.other pertinent plant parameters used in the accident analyses are given in Table 15.0.3-3. '15.0.4 Reactivity Coefficients Assumed in the Accident Analyses . ~ The transient response 'of the-reactor coolant system (RCS) depends on -reactivity. feedback effects,- particularly .the moderator temperature 1 coefficient and the Doppler power coefficient.
In Lthe analysis.of all. events, conservative values of beginning of life Land endiof Llife reactivity coefficients are used. Figure 15.0.4-1 shows
y. the Doppler power coefficients, as-a-function of power, used in the
~ transient analyses. Figure 15.0.4-2 shows the moderator' density coefficient, as a function of-temperature, useo in the transient analyses. The values used in each event are given in Table 15.0.3-2.
O A-15.0-3
- -- - - - - - .- -- - . ~ -
I 15.0.5" Rod Clust er Control Assembiv Insertion Characteristics i The negative reactivity insertion following a reactor trip is a function of the. acceleration of the rod cluster control assemblies (RCCAs)-and the variation in rod worth as a function of rod position. With respect to accident analyses, the RCCA insertion time from the start of insertion up to the dashpot entry, approximately 85 percent of the rod-cluster travel, is assumed to be 2.7 seconds. The RCCA position versus time is shown on Figure 15.0.5-1. Figure 15.0.5-2 shows the fraction of total negative reactivity insertion-versus normalized rod insertion. There is inherent conservatism in the use of this curve in that its basis is a bottom-skewed axial power distribution. For cases other than those associated with axial xenon. oscillations, the more favorable axial power distribution existing before trip results in significant negative reactivity to be inserted. The normalized RCCA-negative reactivity insertion versus time used in the safety analysis is shown on Figure 15.0.5-3. The curve shown on this figure results from the combination of Figure 15.0.5-1 and-Figure 15.0.5-2. The transient analyses, except where specifically noted otherwise, assume a total negative reactivity. insertion following a trip , of 4.0% Ak/k. This assumption is verified to be conservative with respect to the core design. For analyses requiring the use of a multi-dimensional spatial neutron kinetics code-(TWINKLE, Reference 1), the code directly calculates the negative react.ivity insertion resulting from reactor trip which is not separable from other reactivity feedback effects. In this case, the code
-( models the RCCA position versus time of Figure 15.0.5-1.
15.0.6 Trio Points and Time Delays to Trio Assumed in Accident Analyses A reactor trip signal opens two trip breakers connected in series, which feeds power to the: control rod drive mechanisms. The loss of power to the mechanism coils causes the mechanisms to release the RCCAs which then fall-lby gravity into the core. There are various instrumentation delays associated with each trip function, including delays -in signal actuation, in opening the trip breakers, and in the release of the rods by the mechanisms. The time delay from the time that the reactor reaches trip setpoint conditions to the time the rods lare free to fall defines- the total'. delay to trip. Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in Table 15.0.6-1.
. Table 15.0.6-1 refers- to the overtemperature AT (OTAT) and the overpower AT (0 PAT) reactor trip setpoints shown on Figure 15.0.6-1. The-e trip setpoints bound the transition cores and a full core of VANTAGE 5 fuel . The associated OTAT f(AI) penalty is shown on Figure 15.0.6-2.
For all'the reactor trips, the difference between the trip set' points assumed in the analysis and the nominal trip setpoints account for Instrumentation channel error and'setpoint error. The plant Technical = g Specifications specify the nominal trip setpoints. The calibration of protection system channels and the periodic determination of instrument response times- are in accordance with the plant Technical Specifications. A-15.0-4 I I. -
, .~ . . - - . . . _ _ , .-- , - . - . . - - . _ . - - . . ~ . ~ - . . -~ . ~ . . . . . - -. ]
n
' Instrumentation Orift and Calorime+ric Errors. Power Ranoe ~
15.0.7
.Ugutron Flux The ' VANTAGE 5' fuel design features and the modified safety analysis assumptions _as discussed elsewhere in this report with respect to.these ' changes are covered'in WCAP-12460 and_WCAP-12462.
15.0.8' Plant ~ Systems and Comoonents Available for Mitiaation of Accident Effects None of the VANTAGE _5 fuel design features or modified safety analysis
- assumptions as discussed elsewhere in -this report affect plant systems and components available. for mitigation _ of accident effects.
15.'0. 9 ' -Fission Product-Inventories The VANTAGE 5 fuel design-features and the modified safety-analysis 1
' assumptions-as discussed elsewhere in this report affect the fission product inventories and are.addresseo in Appendix C of this report.
15.0.10 Residual Decay Heat None_ of, the VANTAGE.5 fuel . design features or modified safety analysis-Lassumptions as discussed'elsewhere in this report affect
- residual; decay-_ heat. .'15'.0.11_ lComouter Codes Utilized Summary descriptions of the' principal computer codes used-in.the transient '
analyses are given below. : Table =15.0.3-2 lists the codes'used in the
. analysis of each transient.
E 15 ~. 0.114.' 1 FACTRAN Comouter Code-
'FACTRAN calculates the-transient temperature distribution-in a. - . _ _
cross-section of a Tmetal' clad U02 fuel . rod-and the transient heat flux-
at.the.>surfaceLof the clad,-using as input the nuclear power and the Stime-dependent coolant parameters. (pressure, flow, temperature, density).
The code.uses a fuel;model;which simultaneously contains the.following
. features: .o A sufficiently large1 number of radial spar.e: increments to handle fast- ' transients such as a rod ejection acciderts J o' Material _propertie's which are: functions or' temperature and a l sophisticated fuel-to-clad gap heat tr.nsfer calculation L oi - The7necessary calculations to' handle post-DNB transients: film , - boiling-heat itransfer correlations; zircaloy-water reaction;-and partial- melting of the fuel FACTRAN is further discussed in Reference 2. ! A-15.0-5 + -v vw- w em-- r -- # . . . _ __ - _. .__ =_. _ _ . _ _ . . -_m_____ - . ___a +___
- - , . . . . - . . .- ..- - - _ - . - . - - . - . - - ~ _ _ . - . - - . . .
t 15.0.11.2; LOFTRAN ComDuter Code-Transient response studies of_a pressurized' water reactor (PWR) system to
-* specified perturbations in process parameters use the LOFTRAN program. , The LOFTRAN program models.all four reactor coolant loops. __This code < simulates a' multi-loop system by a _model containing the reactor vessel, hot and cold leg piping, steam generators-(tube and shell sides), and the pressurizer. .The pressurizer heaters, spray, relief valves, and safety = valves are also considered in the program, LOFTRAN also includes a point neutron kinetics model and reactivity effects of the moderator, fuel, boron, and rods. The secondary side of the steam generator uses a homogeneous, saturated mixture for the thermal transients and a water level correlation for. indication:and control. The code. simulates the reactor protection system which includes reactor trips on high neutron flux, OTAT, OPAT, high and. low pressurizer pressure, low flow, and - high pressurizer level. Control systems are also simulated including rod - control, steam dump, feedwater control, and pressurizer _ pressure control.
The ECCS, including;the accumulators, is also modeled. LOFTRAN also can calculate the transient value of DNBR based on the input fromLthe core thermal safety limits. LOFTRAN is further discussed in Reference 3.
'15.0;11.3 ANC Comouter Code .i - ANC.is an~ advanced nodal. code capable of two-dimensional.and three-dimensional neutronics calculations. ANC is the reference model for 4 -O all safety analysis calculations,-power distributions, peaking factors, V critical boron. concentrations, control rod worths, reactivity-coefficients, etc., in addition, three-dimensional ANC validates
_ one-dimensional and two-dimensional .results and provides information about
' radial (x-y) peaking factors __as a function -of__ axial position. It can calculate discrete pin powers from-nodal -information as well. - ANC is further discussed in Reference 10, o -15.0.11.4 TWINKLE'Comouter Code-g The TWINKLE program is a: multi-dimensional. spatial . neutron kinetics code. 'The code uses an implicit finite-difference method to . solve the two-group transient neutron diffusion equations -in one, two, and three dimensions.
4 The code'uses six delayed neutron _ groups and contains a: detailed multi-region fuel-clad-coolant heat transfer model for calculating pointwise Doppler'and moderr. tor feedback effects. The code handles.up to
'2000 spatial points and. performs its own steady-state initialization.
Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as-input basic driving functions such as inlet temperature, L pressure, flow, boron concentration, control rod motion, and others. The-l- , : code provides various output edits, e.g., channelwise power, axial offset, H enthalpy, volumetric surge, pointwise power and fuel temperatures. The-TWINKLE code predicts the kinetic behavior of a reactor for transients
' which cause a major perturbation in the spatial neutron flux distribution. . TWINKLE is further described in Reference 1.
l l A-15.0-6
15.0.11.5 THINC Computer Code The THINC computer program performs thermal-hydraulic calculations. This !
' f-)
(
'v code calculates coolant density, mass velocity, enthalpy, void fractions, static-pressure, and departure from nucleate boiling ratio (DNBR) distributions along flow channels within a reactor core. Safety Evaluation Section 4.0 describes the THINC code. ;
15.0.12 Limitina Sinole Failures None of the VANTAGE 5 fuel design features or modified safety analysis assumptions as discussed elsewhere in this report affect the limiting single failures currently assumed in VEGP's safety analyses. 15.0.13 Ooerator Actions None of the VANTAGE 5 fuel design features or modified safety analysis assumptions as discussed elsewhere in this report affect the operator actions.
. i G" .
as A-15.0-7
l Table 15.0.3-1
;O Nuclear Steam Sucolv System Power Ratinos Parameter Valug Design NSSS thermal power output (MWt) 3579 Minimum thermal power generated by the RCPs (MWt) 14 Maximum NSSS thermal power output (MWt) 3585 Maximum thermal power generated by the RCPs (MWt) 20 Design core thermal power (MWt) 3565 O.
L-] { (~h 1 L) A-15.0-8
g () C/ - O Table 15.0.3-2 (Sheet I of 3) Summary of Initial Conditions and Computer Codes Used Reactivity Coefficients Assumed Moderator initial NSSS Computer. Density Thermal Power Output Section Faults Codes Utilized (Ak/am/cm3 ) Dopoler Assumed (MWt) 15.1 Increase in heat removal by the secondary system Feedwater system malfunctions LOFIRAN 0.50 tower (see Figure 0anja)(f) that result in an increase in (also refer to 15.0.4-1) 3579 feedwater flow Section 15.4.1) Excessive increase in secondary LOFIRAN 0.0 and 0.50 Lower and upper 3579(a)(f) steam flow (see figure 3 15.0.4-1) 4
Decrease in heat removal by the 15.2
?
e secondary-system Loss of external electrical LOFTRAN Figure 15.0.4-2 Lower and upper 3579(a)(f) load and/or turbine trip and 0.50 (see Figure 15.0.4-1)
15.3 Decrease in reactor coolant l system flowrate Partial and complete loss of LOFIRAN, FACTRAN, Figure 15.0.4-2 Upper (see 3585(CI(f) forced reactor coolant flow. TillNC Figure 15.0.4-1)
Reactor coolant pump shaft LOFTRAN, FACIP.AN Figure 15.0.4-2 Upper (see 3657(c)(d)(e) seizure (locked rotor) Figure 15.0.4-1) I
O O O I' Table'15.0.3-2 (Sheet 2 of 3) Summary of Initial Conditions and Computer Codes Used l Reactivity Coefficients
' Assumed Moderator Initial NSSS Computer Density Thermal Power. Output-Sectia Faults Codes Utilized (Ak/om/cm3 ) Doppler Assumed (NWt) t 15.4- Reactivity and power distribution anomalies Uncontrolled RCCA bank with- TWINKLE, FACTRAN, Refer to Coefficient'is 0 - drawal from a subcritical or low ' IHINC Section consistent with power startup condition 15.4.1.2 Doppler Defect of , -0.94% AK Uncontrolled RCCA i 1k with- LOFTRAN Figure 15.0.4-2 tower and upper 358.5, 2151 drawal at power and 0.50 (see' Figure and 3585(CIII) p 15.0.4-1)
RCCA misalignment THINC, ANC N/A N/A 3565(b) b Startup of an inactive reactor LOFIRAN, FACTRAN, 0.50 Lower (see Figure 2577(a)(f) coolant pump at an incorrect THINC 15.0.4-1) -!
' temperature Chemical and volume control N/A N/A N/A 0 and 3565(b) ;
system malfunction that'_results in a decrease in the boron concentration in the reactor Spectrum of RCCA ejection -lWINKLE, FACIRAN, Refer to Coefficient is 0 and 3636(d) accidents THINC Section consistent with 15.4.8.2 Doppler Defect of ,
-0.94% AK 1
mi
. _ -- _ _ _m.- i-%
L l O ~1; v . l , Table 15.0'.3-2 (Sheet 3 of 3) Summary of initial Condit ions and Computer Codes Used i
. Reactivity Coefficients Assumed Moderator Initial NSSS-Computer _ Density Thermal Power Output Section Faults - Codes Utilized lak/qukm 3) Doppler Assumed (MWt)-
Steamline' break with coincidental L0f1RAN- 0.50 f.ower (see 3585(C)(II rod withdrawal at power' Figure 15.0.4-1)-
..I 15.5 Increase in reactor coolant '
inventory Inadvertent operation of the' .LOTTRAN 0.0 Lower (see _3565(b)(f} emergency core cooling system Figure 15.0.4-1) during power operation
$ 15.6 Decrease in reactor coolant <n inventory o :
I h ' Inadvertent opening of a LOFTRAN Figure 15.0.4-2 Upper _(see 3585(C)(f) pressurizer safety or relief- Figure 15.0.4-1) valve-(a) iiinimum pump heat of 14 MWt assumed (b) No' pump heat (core thermal power) assumed (c) Maximum pump heat of 20 MWt assumed - (d) 2% margin applied . (e) - Standard Thermal Design Procedure (SIDP) with a thermal design flow (TDF) of 93600 gps / loop assumed ! (f) Revised Thermal Design Procedure (RlDP) with a minimum measured flow (N9F) of 95600 3pe/ loop assumed - t t 4 k v s "
' ~ '
Table 15.0.3-3 f% Nominal Values of Pert'nent Plant Parameters Used in the Accident Analyses () STDP RTisP Parameter Values V31ni < Thermal output of NSSS (including minimum thermal power generated by the RCPs, MWt) 3579 3579 Core inlet temperature ('F) 556.8 557,4 Vessel average temperature (*F) 588.4 588.4 RCS pressure (psia) 2250 2250 Reactor coolant flow per loop (gpm) 93600(a) 95600(b) Steani flow from NSSS, total (lb/hr) 15,920,000 15,920,000
-Steam pressure at steam generator outlet (psia) 950 950 Maximum steam moisture content (%) ,
0.25 0.25 Assumed feedwater temperature at steam generator inlet (*F) 446 446 Average core heat flux (Btu /hr-f t2) 206,085 206,085 x l-(a) Thermal design flow (TOF) (b) Minimum measured flow (MMF) A-15.0-12 l
i Table 15.0.6-1 g Trio Points and Time Delavs 'o Trio Assumed in Accident Ana'vses i Limiting Trip Time Delays Trio Function Point Assumed in Analyses (s) Power range high neutron flux, high setting 118% 0.5 Power range high neutron flux,
low setting 35% 0.5 Hign neutron flux, P-8 84% 0.5 Source range neutron flux N/A 0.5 OTAT Variable (See Figure 15.0,6-1) 6.0(a)
OPAT Variable (See Figure 15.0.6-1) 6.0(a) High pressurizer pressure 2425 psig 2.0 4 Low pressurizer pressure 1920 psig 2.0 Low reactor coolant flow (from loop flow detectors) 87% loop flow 1.0 RCP undervoltage trip 68% nominal 1.5 O C Turbine trip N/A 2.0 Low-low steam generator level 16.0% of narrow range level span (D) 30.0% of narrow range level span 2.0 (c)2.0 High steam generator level trip of the feedwater pumps and closure of feedwater system 100% of narrow range level span 2.5(d) valves and turbine trip 7.0(')
a. Total time delay (including resistance tweerature detector (RTD) bypass loop fluid transport delay effect. bypass loop piping thermal capacity, RTD time response, and trip circuit channel electronics delay) from the time the temperature difference in the coolant loop exceeds the trip setpoint until-the rods are free to fall,

b. Low-low level setpoint assumed for "Feedwater System Pipe Break" (15.2.8);
the setpoint includes environmental errors,
c. Low-low level setpoint assumed for " Loss of Non-Emergency AC Power to the Plant Auxiliaries ' (15.2.6) and " Loss of Normal Feedwater Flow" (15.2.7);
the setpoint does not include Any environmental errors,
d. From time setpoint is reached to turbine trip.

e. From time setpoint is reached to feedwater isolation.
A-15.0-13
-- - . .. .- ..-.. ~ - .. .. - . - . - . .. . . . -.-. - - .-..- -- . - - -.
. t I , a : !O , t i i a
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i N0tt 1 p-
-14 =
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>=
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a.
4 NOTE 1: " UPPER CURVE" MOST NEGATIVE DOPPLER i
ONLY POWER DEFECT = -1.6% ak , i
+2 - l NOTE 2: ' LOWER CURVE" LEAST NEGATIVE DOPPLER +
ONLY POWER OEFECT = -0.78% aK 0 , , , 0 20 40 60 80 100 PERCENT POWER W voorts DOPPLER POWER COEFFICIENT USED IN O CeOrgia Power s Lscraic on=enari=o PLANr u=ir i *= u=ir : ACCIDENT ANALYSES FIGURE 15.0.4-1 n. A-15,0-14
l i
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$ 0.5 NOTE 1 E
d 0.4 - ' S NOTE 1: " Upper curve" most r...o ve moderator density b 0.J = Coefficient for end of Cycle life 8ssumptions. 5
,, , . NOTE 2: " Lower curve" most negative moderator density O coefficient for beginning of cycle life ' - assumptions. !
y o.i - ; w i I
'o-3 _ NOTE 2 ~ } -o.1 - l $ -o.: ,
350 550 630
-TEMPERATURE (*F)
(Pressure is at 2250 PSIA) 4 i voorts MODERATOR DENSITY COEFFICIENT USED A stacTaic oswa mATiwa rt4=7 IN ACCIDENT ANALYSES Georgia Power ma v=ir i awa vwir : FIGURE 15.0.4-2 A-15.0-15
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0 i i i i 1 , , , , , , 3 0.0 0.2 0.4 0,6 0.8 1.0 1.2 DROP TIME / DROP TIME-to DASHPOT ENTRY t v ROD POSITION VS TIME ON REACTOR e Lacraic on us n Art.o A NT Georgia Power a$ m u=oorts-r i 4= v ir : TRIP FIGURE 15.0.5-1 ' A-15.0-16 1
l O 4.0 3.0. O d 2.0.. O O ;~ U g 1.0 . l i ; 2 1 2 ! 0.0 . . . 0 .20 .40 .60 .80 1.0 ROD POSITION (FRACTION OF FULL INSE8'.T10t!) 4 voans RCCA REACTIVITY WORTH VS. FRACTION Georgia Power A E~IrTlio'"'r"*""T u FULL TNSERTION FIGURE 15.0.5-2 A-15.0-17
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f. FIGURE 15.0.5-3 A-15.0-18 I~ - - - - . . - - - _ . . . - _ _ _ _ _ _ _ _ _ _ _
O 75.< g , t \ LOCUS OF
\ \ CON 0lil0NS 72.5-3 WHERE DNBR = LIMITING yAtyg \7--..
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45.- \ \ GENERATOR SAFETY 4 VAlvt$ OPEN g
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42.5 . . ' 598 595 598 595 688 685 6i8 IS 628 625 658 655 T avg ('I) voorts !( l l_ Geog {g pO%'er tL'"T"'""'"'""I'* P' ANT UNIT 1 AND UNIT : OVERTEMPERATURE AND OVERPOWER AT PROTECT 10N FIGURE 15.0.C-1 A-15.0-19 __. _ _ -.-.._ ~. - . . .__ . __ _ _
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Georgia Power d Edif,7l'f,'l^"'LANT FIGURE 15.0.6 -2 i l A-15.0-20
-15.1 . Increase in Heat Removal by the Secondary System Many events can result in an increase in heat removal from the RCS by the V secondary system. This section presents several limiting cases of such events.
15.1.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temoerature 15.1.1.1 Introduction Reductions ~in feedwater temperature will result in an increase in core power by initially decreasing reactor coolant temperature. The thermal capacity of the secondary plant and of the RCS attenuates such transients. The high neutron flux trip, OTAT trip, and OPAT trip - - _ prevent any power. increase which could lead to a ONBR less than the limit . ; value.
-A low-pressure feedwater train or a high-pressure heater out of service may cause a reduction in feedwater temperature. If a spurious heater drain pump trips, there is a sudden reduction in feedwater inlet-temperature to the steam generators. At power, this increased subcooling will create a greater load demand on the RCS.
With- the plant at no-load conditions, the addition of cold feedwater will I causr, a decrease in RCS temperature and a reactivity insertion due to the eff6 cts of the negative moderator temperature coefficient of reactivity; Lhowever, the rate of energy change is reduced as load and feedwater flow O- decrease, so the transient is less severe than the full power case. 4 The net effect on the RCS due to a reduction in feedwater temperature is similar to the effect of increasing secondary steam flow; i.e., the - reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator AT.
-This is an American Nuclear Society '(ANS). Condition !! incident.
15.1.1.2 Method of Analysis This transient is analyzed by computing conditions at the feedwater pump.
. inlet following the removal of a low-pressure feedwater train or a high-pressure heater'from service. These feedwater conditions are then -used to recalculate a heatt halance for that RCS loo) containing a reduced . number of feedwater heaters in service. This heat salance gives-the new .feedwater conditions at the steam generator inlet.
The following assumptions are made: f
'A. Plant initial . power level corresponds to design NSSS thermal output j at the uprated condition B. One string of feedwater heaters is isolated This accident analysis employs the RTDP with the initial conditions shown- ~
in Tables 15.0.3-2 and 15.0.3-3. E A-15.1-1
~.-_ ;. __ _ ._._,- _ .._..__,.
1 No single active f ailure in any plant systems or equi? ment will adversely affect the consequences of the accident. g 15.1.1.3 Results Isolation of a string of low-pressure feedwater heaters causes a reduction in feedwater temperature, which increases the thermal load on the primary system. The calculated reduction in feedwater temperature is lets than 30'F. This reduction in feedwater temperature results in an inerease in heat load on the primary system of less than 10 porcent of 'ull power. Thus, increased thermal load due to a spurious heater W ain ramp trip would result in a transient vry siwilar (but of a reduceo magnitude) to that presented in Section 15.1.3 for an excessive increase in secondary steam flow transient. The consequences of a 10 percent step load increase are evaluated in Section 15.1.3; therefore, there is no presentation of the results of this analysis. 15.1.1.4 Conclusions The decrease in feedwater temperature transient is less severe than the increase in secondary steam flow event (Section 15.1.3). Based on results presented in Section 15.1.2 and Section 15.1.3, the applicable acceptance criteria for the decrease in feedwater temperature event have been met. The conclusions presented in the FSAR remain valid. 15.1.2 Feedwater system Malfunctions that Desult in an increase CN in Feedwater Flow w) 15.1.2.1 Introduction Addition of excessive feedwater will cause an increase in core power by decreasing reactor coolant temperature. The thermal capacity of the secondary plant and of the RCS attenuates such transients. The high neutron flux trip, OPAT trip and OTAT trip prevent any power increase which could lead to DNBR less than the minimum allowable value in the event that the steam generator high-high water level protection does not actuate. The full opening of a feedwater control valve due to a feedwate.a control system malfunction or an operator error may cause excessive feedwater flow. At power conditions, this excess flow causes a greater load demand on the RCS due to increased subcooling in the steam generator. With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator temperature coefficient of reactivity. The steam generator high-high water level trip signal does the following: o Closes the feedwater valves o Closes the feedwater pump discharge valves o Trips the turbine A-15.1-2 1 .
e Trips the main feedwater pumps g o Prevents continuous addition of excessive feedwater V The turbine trip signal initiates a reactor trip. This is an ANS Condition !! incident. 15.1.2.2 Method of Analysis The analysis of the excessive heat removal due to a feedwater system malfunction transient uses the detailed digital computer code 1.0FTRAN (Reference 3). This code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, steam generator safety valves, and feedwater control system. The code computes pertinent plant variables including temperatures, pressures, and power level. The purpose of the system analysis is to demonstrate acceptable plant behavior in the event of an excessive feedwater addition due to a control system malfunction or operator error which allows a feedwater control valve to open fully. Cases analyzed assuming a conservatively large negative moderator temperature coefficient follow: o Accidsntal full opening of a feedwater control valve with the reactor just critical at zero power conditions (Mode 3) with the reactor in manual rod control o Accidental full opening of a feedwater control valve with the reactor O just critical at zero power conditions (Mode 2) with the reactor in both automatic and manual rod control o Accidental full opening of one feedwater control valve with the reactor at full power assur$ng automatic and manual rod control The calculation of the reactiv'..y insertion rate following a feedwater system malfunction uses the following assumptions: A. For the feedwater control valve accident at full power, one feedwater control valve malfunctions, which results in a step increase to 157 percer,t of nominal feedwater flow to one steam generator. B. For the feedwater control valve-accident at no-load conditions (Mode 2), a feedwater control valve malfunction occurs which results in a step increase in flow to one steam generator from zero to 225 percent of the nominal full load value for one steam generator. C. For the feedwater control valve accident at no-load conditions (Mode 3), a feedwater control valve malfunction occurs which results in a step increase in flow to one steam generator from zero to 50 percent of the nominal full load value for one steam generator. The flow is limited by a 6 inch pipe in the line from the condensate pump to the steam generators through which all the feedwater to the steam generators must pass. A_ 0. For the no-load condition, feedwater temperature is at a () conservatively low value of 32*F. L l A-15.1-3
p - E. No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown. ; F. The feedwater flow resulting from a fully open cetrol valve terminates by a steam generator high-high level trip signal which closes all feedwater control and isolation valves, trips the main - feedwater pumps, and trips the turbine. The turbine trip signal initiates a reactor trip. G. The analysis assumes the RCS flow equivalent to the operation of four RCPs. The analysis uses RTOP methodology in the determination of initial reactor power, pressure, and RCS temperature (see Tables 15.0.3-2 and 15.0.3-3) . The analysis does not require normal reactor control system and engineered safety systems to function.. The reactor protection system function at full power trips the reactor due to overpower or turbine trip on high-high steam generator water level conditions.
'No single active failure in any plant systems or equipment will adversely affect the consequences of the accident.
15.1.2.3 Results in the case of an accidental full opening of one feedwater control valve with the reactor at zero power (Mode 3) and the above mentioned assumptions,'the maximum reactivity insertion rate is less than'the maximum reactivity insertion rate analyzed in Section 15.4.1 for uncontrolled rod cluster < control assembly bank withdrawal from a
.O suberitical-or. low-power'startup condition; therefore, this section does not present the results of the analysis. Note that the analysis in Section 15.4.l' conservatively assumes only two RCPs in operation for the DNBR calculation. This assumption bounds the feedwater malfunction event with respect to the DNBR.
The analysis of the Mode 2 case.uses the TWINKLE,' FACTRAN, and THINC codes to evaluate the resulting reactivity insertion rato as described in Section 15.4.1. The Mode 2 analysis assumes all- four RCPs to be in operation.- The resulting minimum calculated DNBR is above the limit value and is less limiting than the analysis presented in Section 15.4.1; therefore, this section does not present the results of the analysis. Note that-if -the incident occurs with the unit just critical at no-load, the reactor may trip by the power range high neutron flux trip (low setting)- set at approximately 25 percent nominal full power. The full power case-(maximum reactivity feedback coefficients, manual rod ; control) results in the greatest power increase. Assuming automatic rod-control results in a less severe transient.
'When the steam generator water level in the faulted loop reaches the high-high level setpoint, the feedwater control valves and feedwater pump discharge valves are automatically closed and the main feedwater pumps trip. This prevents continuous addition of feedwater. A turbine trip and a resulting reactor trip are also initiated.
A-15.1-4
Transient results show the increase in nuclear power associated with the increased thermal load on the reactor (See figure 15.1.2-1 and Figure 15.1.2-2). The DNBR does not drop below the limit value. : Of Following the reactor trip, the plant approaches a stabilized condition; standard plant shutdown procedures then apply to further cool down the plant. Since the power level rises during the excessive feedwater flow incident, ' the fuel temperatures will also rise until after reactor trip occurs. The core heat- flux lags behind the neutron flux response due to the fuel rod thermal time constant; hence, the peak heat flux does not exceed 118 percent of its nomini value (i e., the assumed high neutron flux trip setpoint). Thus, the peak fuel temperature will remain well below the fuel melting temperature.
.The transient results show that DNB does not occur at any time during the excessive feedwater flow incident; thus, there is no reduction in the ability of the primary coolant to remove heat from the fuel rods. The fuel cladding temperature, therefore, does not rise significantly above its initial value during the transient. .
. The calculated sequence of' events.for the increase in feedwater flow for - the full power case is shown in Table 15.1.2-1. 15.1.2.4 Conclusions The results of the analysis shcw that the DNBRs encountered for an excessive feedwater addition at power are above the limit-value. Additionally, for an excessive feedwater addition in Mode 3, the resulting O- reactivity insertion rate is less than the value used in the rod withdrawal from subcritical analysis (Section 15.4.1). For an excessive feedwater addition in Mode 2, the resulting reactivity insertion rate was analyzed and the minimum 0NBR is above the limit value-and is less limiting than the analysis _ presented in Section 15.4.1. The conclusions
. presented in_the FSAR remain valid.
15.l.3 ' - Excessive Increase in Secondary Steam Flow L15.1.3.1 Introduction A rapid' increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand defines an h excessive load increase incident. The reactor control system desigt accommodates a 10 percent step load increase and a 5 percent per minute ramp: load. increase in thc range of 15 to 100 percent of full power. Any
-loading rate more than these values may cause a reactor' trip actuated by the reactor protection system. Sections 15.1.4 and 15.1.5 discuss steam flow increases greater than 10 percent.
This accident could result from either an administrative violation such as t- excessive loading Lby the operator or an equipment malfunction in the steam L dump control or turbine speed control. Reactor coolant condition signals control turbine bypass to the condenser during power operation, i.e., high reactor coolant temperature indicates a A-15.1-5
l need .for turbine bypass by ut ng steam dumps. A single controller malfunction does not cause tt bine bypass; an interlock blocks the opening of:the steam dump valves unit s a large turbine load decrease or turbine O trip occurs. The following reactor _ protect on system signals protect against an ' excessive load increase accic nt: o- OPAT o OTAT 3 o Power range high neutror flux o Low pressurizer pressure This is an ANS Condition 11 i cident. 15.1.3.2 Method of Analysis The analysis of this accideht uses the LOFTRAN code (Reference 3). This code simulates the neutron k' etics,- RCS, pressurizer, pressurizer relief and safety valves, pressurizt spray, steam generator, steam generator safety valves, and_ feedwater ystem. The code computes pertinent plant variables including temperatt es, pressures, and power level. The analysis includes four et es that demonstrate the plant behavior following a'10 percent step 1 ad increase from rated load. These cases are as follows: A'. Manual-rod control with -inimum reactivity feedback
'B. Manual _ rod control with aximum reactivity feedback C. Automatic rod control w' h minimum reactivity feedback i
D .; Automatic rod control w h maximum reactivity feedback l for the minimum moderator fee back cases, the analysis assumes that the core has a zero moderator ten erature coefficient.of reactivity and the r least negative Doppler-only r wer coefficient curve. This results in the least11nherer.t transient rest nse capability. The zero moderator temperatureLeoefficient of rt etivity bounds a positive moderator temperature coefficient for t is cooldown event. For the maximum moderator feedback cases, the moderator temperature coefficient of reactivity has its ' highest at olute value and the most negative
' Doppler-only power coefficier curve. This results in the largest amount of reactivity-feedback due to changes in coolant temperature. For the cases--with automatic rod con' 01, no credit was taken for AT trips on- 1 overtemperature-or_ overpower n order to demonstrate the inherent !
transient capability of the l ant. Under actual operating conditions, such a trip may occur after t ich the plant would quickly stabilize. " h The analysis assumes a 10 pe ent step increase in steam demand, and the E analysis of all the cases do- not take credit for pressurizer heaters. The analysis of th accident uses RTDP as described in
~ Reference 7. The analysis a: umes nominal values for the initial reactor l- A-15.1-6 .
L---- - - -
l power, pressure, and RCS temW rature. The limit DNBR includes u1 certainties in initial conditions. Tables 15.0.3-2 and 15.0,3-3 show the plant characteristics and initial conditions. The analysis does not require normal reactor control systems and engineered safety systems to function. The analysis assumes the reactor protection system to be operable; however, due to the error allowances assumed in the setpoints, a reactor trip does not occur. No single active failure will prevent the reactor protection system from performing its intended function. No single active failure in any pant systems or equipment will adversely affect the cons 2quences of the acc. dent. 15.1.3.3 Res ul,tt Figures 15.1.3-1 through 15.1.3-4 ill s trate the transient with the reactor in the manual rod control modt. As expected, for the minimum moderator feedback case there is a slight power increase, and the average core temperature shows a large decrease. This results in a DNBR which increases above its initial value. For the maximum moderator feedback manual-rod controlled case, there is a large increase in rear. tor power due to the moderator feedback. A reduction in DNBR occurs, but DNBR remains above the limit value. Figures 15.i.3-5 through 15.1.3-8 illustrate the transient assuming the reactor is in ihe automatic rod control mode. Both the minimum and maximum moderatoi' feedback casts show that core power increases, thereby C i reducing the rate if decrease in coolant average temperature and pressurizer pressuri. For both of these cases, the minimum DNBR remains above the limit value. For all cases, the plant rapidly reaches a stabilized condition at the higher power level where a reduction in power can occur by following normal-plant operating procedures. Note that due to the measurement errors assumed in the setpoints, it is possible that reactor trip could occur for the automatic rod control cases. The plant would then reach a l stabilized condition following the event. The excessive load increase incident is an overpower transient for which the fuel temperatures will rise. Reactor trip does not occur for the cases analyzed, and the plant reaches a new equilibrium condition at a higher power level corresponding to the increase in steam flow. Since DNB does not occur at iny time during the excessive load increase transients, there is no reduction in the ability of the primary coolant tr remove heat from the fuel rod. Thus, the fuel cladding temperature does - not rise significantly above its initial value during the transient. The calculated sequence of events for the excessive load increase incident is shown in Table 15.1.2-1. 15.1.3.4 Conclusions The analysis presented shows that for a 10 percent step load increase, the DNBR remains above the limit value. The plant reaches a stabilized condition following the load increase. The conclusions presented in the FSAR remain valid, A-15.1-7
15.1.4 Inadvertent Openino of a steam Generator Relief or Safetv Valve / Section 15.1.4 of the FSAR describes the inadvertent opening of a steam (]/ generator relief or safety valve accident. Performance of the analysis is at hot Zero power conditions with the control rods fully inserted in the core and the most reactive rod stuck out. Since the analysis of this accident does not use the RTOP, the only significant impact of the design changesassociatedwightheVANTAGE5fuelreloadtransitionisthatof the increase in the F gg peaking factor. The analysis perfermed for steam system giping failure (Section 15.1.5) which evaluates the increase in the F w peaking factor, bounds this transient. The safety analysis DNBR limit is met and the conclusions of the FSAR remain valid. 15.1.5 Steam System Pioina Failure Section 15.1.5 of the FSAR describes the steamline break accident. The analysis is performed at hot zero power conditions with the control rods fully inserted in the core and the most reactive rod stuck out. Since the analysis of this accident does not use the RTOP, the only significant impact of the design changes associated wit transitionisthatoftheincreaseintheFgtheVANTAGE5fuelreload gy peaking factor. The evaluation of the increase in the F w peaking factor concludes that the safety analysis DNBR TTmit is met; therefore, the conclusions of the FSAR remain valid. O A-15.1-8
Table 15.1.2 1
.\ Time Secuence of Events for Incidents ibat Ruult in an increase in Heat Removal by the Setondaty System Time Delays Accident Event (s)
Feedwater system malfunctions One main feedwater control valve that result in an increase in fails fully open 0.0 feedwater flow (full power) Minimum DNBR occurs 83.0 High-high s+aam oenerator wter level signal generated 86.2 Turbine trip occurs due to high-high steam generator level 88.7 Reactor trip occurs 90.7 feedwater isolation valves close automatically 93.2 Excessive increase in secondary steam flow
/
1. Manual rod control 10 percent step load increase 0.0 (minimum moderator feedback)
Equilibrium conditions reached (approximate time only) 150
2. Manual rod control 10 percent step load increase 0.0 (maximum moderator feedback)
Equilibrium conditions reached (approu mate time only) 50
3. Automatic rod control 10 percent step load increase 0.0 (minimum moderator feedback)
Equilibrium conditions reached (approximate time only) 125
4. Automatic rod control 10 percent step load increase 0.0 (maximum moderator feedback)
Equilibrium conditions reached (approximate time only) 50 O A-15.1-9
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15.2 D, crease in Heat Removal by the Secondary System ( Several postulated transients and accidents which result in a reduction of the capacity of the secondary system to remove heat generated in the RCS are discussed in this section. 15.2.1 Steam Pressure Reoulator Malfunction or Failure that Results in_ Decreasina Steam Flow As stated in the FSAR, there are no steam pressure regulators in either VEGP units whose failure or malfunction could cause a steam flow : transient. I 15.2.2 Loss of Electrical load As discussed in FSAR Section 15.2.2, the analysis for the turbine trip event bounds this transient. Section 15.2.3 describes the turbine trip event., 15.2.3 Turbine Trio 15.2.3.1 Introduction For a turbine trip event, the reactor trips riirectly (unless below approximately 50 percent power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop valves. The turbine stop valves close rapidly (typically in 0.1 second) on loss of trip fluid oressure actuated by one of several possible turbine trip signals. Turbine trip initiation signals include: o Generator trip o Low condenser vacuum o Loss of lubricating oil o Turbine thrust bearing failure o Turbine overspeed o Manual trip Upon initiation of stop valve closure, steam flow to the turbine stops abruptly. Sensors on the stop valve detect the turbine trip and initiate turbine bypass through steam dump valves and, if above 50 percent power, a reactor trip. The loss of steam flow results in an almost immediate rise in secondary systam temperature.and pressure with a resultant increase in l primary system temperature and pressure. A slightly more severe transient ! than the loss of electrical load event occurs for the turbine trip event I due to a more rapid loss of steam flow caused by the more rapid valve O l A-15.2-1 l
closure. FSAR Sections 15.2.2 and 15.2.3 contain more information on the loss.of_ load and turbine trip events, O ;A turbine trip accident is more limiting.than.the loss of external electrical / load-(Section 15.2.2), . inadvertent closure of main steam
-isolation valves (Section 15.2.4), and loss'of condenser vacuum and other Svents resulting in a turbine trip (Section 15.2.5); therefore, this event .ha., L:an analyzed-in detall.
This is.an ANS Condition II incident. 15.2.3.2 Method'of Analysis In'this analysis, evaluation of the behavior of the units is for a complete. loss of steam. load from ominal full -power, with a turbine trip
-not causing a direct reactor trip. This demonstrates the adequacy of the pressure-relieving devices and the core protection margins . This assumption delays reactor trip until conditions in the RCS result in a trip due to other signals. Thus, the analysis models a worst-case transient. In addition, no credit is taken for the turbine bypass system. Main feedwater flow terminates at the time of turbine trip with no credit taken for auxiliary feedwater (except for long-term recovery) to mitigate the' consequences of the transient.
The P alysis of the turbine trip transients employs the detailed digital computer program LOFTRAN (Reference 3). The program simulates the neutron kinetics, RCS, pressurizer,-pressurizer relief and safety valves, pressurizer.' spray,. steam generator, and steam generator safety valves. O
' LOFTRAN' computes pertinent plant variables including temperatures, pressures', and power level .
The acalysis of this accident uses RTOP. methodology. Plant characteristics-and initial conditions are shown.in Tables 15.0.3-2 and 15.0.3-3 . The-following' summarizes'the major assumptions used in the analysis: 4 A. Initial Operating' Conditions
.The analysis assumesinominal values of core power, reactor coolant average temperature, and nominal reactor. coolant average pressure.
The limit DNBR includes uncertainties-in initial conditions as described in the" Safety Evaluation (Section 4). Previous. studies
-have .shown that the peak pressurizer pressure reached- for the turbine trip eventLis insensitive to the initial conditions of temperature Land pressure, and the peak pressurizer pressure is only slightly sensitive to the. initial power condition. Therefore, the use of these initial conditions is appropriate for this event. 'B. Moderator and'00ppler Coefficients of Reactivity
, The analysis of the turbine trip is with both maximum and minimum
reactivity feedback. With maximum feedback, the analysis assumes a large negative moderator temperature coefficient and the most-7 negative Doppler-only power coefficient. With minimum feedback, the analysis assumes the most positive moderator temperature coefficient
. and the least-negative Doppler-only power coefficient. A-15.2-2
.,, - _ _ _ . _ __ ~ _ _ _ -- . . _ . _ . . . - _ _ _ _ . _ _ ~. .
l C. Rod Control
'O lt is cunservative to assume that the reactor is in manual rod control with respect to the maximum pressures attained. .If the reactor were in automatic rod control, the-control rod banks would move before the. trip and reduce the severity of the transient.
D. Steam Release , No credit is taken for the operation of_ the steam dump system or steam generator power-operated relief valves. The steam generator
. pressure rises to the safety valve setpoint where steam release through the safety valves limits secondary steam pressure at the setpoint value.
E. Pressurizer . Spray and Power-Operated Relief Valves The following analyses are the two cases for both the minimum and maximum reactivity feedback cases examined:
1. Full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are also available.

2. No credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are operable.
O F. Feedwater Flow d The analysis assumes main feedwater' flow to the steam generators to be lost at the time of turbine trip. No credit is taken for auxiliary: feedwater flow since che plant will reach a stabilized condition before auxiliary feedwater initiation is normally assumed to occur; however, the auxiliary feedwater pumps will start on a trip of the main feedwater pumps. The auxiliary feedwater flow would remove' core decay heat following plant stabilization.-
- G. _ Reactor .Tr_ip Reactor trip. actuates by the first reactor protection system trip a setpoint- reached, with no credit taken for direct reactor trip on turbine trip. Trip _ signals are expected due to high pressurizer pressure, 0 TAT, high pressurizer water level', and low-low steam -generator water level.
No single active failure in any plant systems or equipment will adversely affect the consequences of the accident. 15.2.3.3 Results. The. transient responses for a turbine trip from nominal full power operation are shown for the following four cases: two cases with minimum
- - ~ reactivity feedback and two cases with maximum reactivity feedback s (Figures 15.2.3-1 through 15.2.3-8).
A-15.2-3 i
i
. Figures 15.2.3-1 and 15.2i3-2 show the transient responses for the turbine trip event with minimum reactivityL feedback assuming full credit for the
[O pressurizer spray and pressurizer power-operated relief valves. No credit
.is taken for the turbine bypass through the steam dumps. The reactor trips on the high pressurizer pressure trip channel. The minimum DNBR remains well above the limit value. The pressurizer safety valves and steam generator safety valves prevent overpressurization in the primary and secondary systems, respactively.
Figures 15.2.3-3 and 15.2.3-4 show the responses for the turbine trip event with maximum' reactivity feedback. All other niant parameters are the same as the above. The DNBR increases throughoist the transient and never drops below its initial value. The pressurizer power-operated relief valves-and steam generator safety valves prevent overpressurization in the primary and secondary systems, respectively. The reactor trips on the high pressurizer pressure trip channel. The pressurizer safety valves do not actuate for this case. The turbine trip-accident was also studied assuming the plant to De initially operating at nominal full power with no credit taken for the pressurizer spray,. pressurizer power-operated. relief valves, or turbine bypass-system. The reactor trips on the high pressurizer _ pressure-signal.- Figures 15.2.3-5 and 15.2.3-6 show the transient respontes with minimum reactivity ftedback. 'The neutron flux remains constant at nominal full power until the reactor trips.- The DNBR never goes below its initial value throughout the transient._ In this case the pressurizer safety valves'and steam generator safety valves actuate to maintain the RCS and main' steam system pressure below-110 percent of their respective design values. ., Figures 15.2.3-7 an.d 15.2.3-8 show the' transient. responses with maximum reactivity feedback with the other assumptions being the same as_ in the preceding case. The-reactor trips on-the high' pressurizer pressure signal , and the DNBR increases throughout the-transient, The pressurizer safety: valves and steam generator safety valves : actuate to limit primary and secondary _ system pressures,~respectively, _; N e calculated sequence of events for the turbine trip event is shown in iable 15.2.3-1. 15.2.3.4 Conclusions
- Results of the analyses show that- the plant design is such that a turbine trip without a direct reactor trip does not present any hazard to the.
integrity of the RCS or the main-steam system. Pressure-relieving devices incorporated in the two systems are adequate to limit the maximum i pressures to within the design 1imits. -The integrity of the. core .is also
. maintained since the DNBR remains'above the-limit value. Thus, the -conclusions presented in-.the FSAR remain valid for the turbine trip event.
15.2.4 Inadvertent Closure of Main Steam isolation Valves Inaovertent closure of the main steam isolation valves would result in a turbine trip'with no credit taken for the turbine bypass system. Section
-O- -15 2.3 discusses turbine trips. The analysis performed for the turbine ' trip event applies to this category. of events. . A-15.2-4
.. ..m - . __ _ _ . . ~ , _ . . . - _ - _ _ . . _ . _ _ _ _ ~ _ _._ - ._--._. _ _ _
3
' 15. 2.- 5 -Loss of Condensei' Vacuum and Other Events Resultina
in a Turbite Trio Loss _of condenser _ vacuum is one of the. events that can cause a turbine trip. 'Section 15.2.3 describes turbine trip initiating events. FSAR ,
Section 15.2.5 provides additional information for-this category. The analysis performed for the turbine trip event applies to this category of events.- ! 15.2.6' -Loss of Non-Emeroency AC Power to the plant Auxiliaries Section 15.2.6 of the'FSAR describes the loss of non-emergency AC power to the plant auxiliaries accident. This is a long-term heat removal event
- analyzed to determine-if the auxiliary feedwater (AFW)-heat removal-capacity is sufficient to ensure that the peak RCS. pressure does not exceedallowablelimits-andthenaturalcirculationis'gufficientto remove residual heat'from the core. Th6 ;ncreases in-F gg and Fn do not affect. system transients, and thus have no impact on these e7ents. The remaining effects'. of the VANTAGE 5 fuel transition will also have no discernible impact on these transients; therefore, the results and T conclusions presented in the FSAR' remain valid for this event.
15.2.7J Loss of Normal Feedwater Flow Section'15.2.7' of the FSAR describes the loss of normal feedwater flow accident. This is a long-term heat removal event. analyzed to determine if q the AFW heat removal capacity.is sufficient in removing long term decay
~\ heat-andpreventingexcessiveheatupoftheRCSwithpossiblereguitant RCS overpressurization or loss of RCS water. The increases in F AH and F do not affect system transients, and thus have no impact on these events.g The. remaining effects'of the' VANTAGE'S fuel transition will also have no discernible impact.on these transients; therefore,'the results and-conclusions presented'in the FSAR remain valid for this event. ~15.2.8 Feedwater System Pioe Break. .Section15.2.8offtheFSARdescribesthefeedwater'systempipebreak accident;- This is a long-term heat removal: event' analyzed to. determine if the AFW heat removal? capacity is sufficient to ensure that the' peak RCS' pressurefdoesinot-exceedallowablelimits,andghecoreremainscovered ;.and Lin, a coolable n geometry. The increases inLF g and F 'do not' affectosystem transients, and thus have no impac on the e events. The
. remaining effects -of- the VANTAGE 5 fuel transition will also have no
-discernible impact on-these transients;-therefore,_the results and . conclusions' presented in the FSAR remain valid for this event.
4 3 A-15.2-5
Table 15.2.3-1 (Sheet 1 of 2) Time Seouence of Events for Incidents Which Result in a Decrease in Heat Removal by the Secondary System Time Delays Accident Event (s) Turbine Trip
1. With pressurizer control Turbine trip; loss of main feed-(minimum reactivity water flow 0.0 feedback)
High pressurizer pressure reactor trip point reached 7.1 Initiation of steam release from steam generator safety valves 7.5 Rods begin to drop 9.1 Peak pressurizer pressure occurs 10.5 Minimum DNBR occurs 11.0
2. With pressurizer control Turbine trip; loss of main feed-(maximum reactivity water flow 0.0 O feedback)
Initiation of steam release from steam generator safety valves 7.5 High pressurizer pressure reactor trip setpoint reached 8.4 Peak pressuri2.er pressure occurs 9.0 Rods begin to drop 10.4 Minimum DNBR occurs (a)
3. Without pressurizer control Turbine trip; loss of main feed-(minimum reactivity water flow 0.0 feedback)
High pressurizer pressure reactor trip point reached 4.6 Rods begin to drop 6.6 Initiation of steam release from steam generator safety valves 7.5 Peak pressurizer pressure occurs 8.0 Minimum DNBR occurs (a) A-15.2-6
- . Table 15.2.3-1 (Sheet 2 of 2)
Cl Time Seauence of Events for incidents Which Result in a Decrease in Heat Removal by the Secondary SVstem Time Delays Accident Event (s) 4.'Without pressurizer control Turbine trip; loss of main (maximum reactivity feedwater flow 0.0 feedback) High pressurizer pressure reactor trip point reached 4.6 Rods begin to drop 6.6 Peak pressurizer pressure occurs 7.5 Initiation of steam release from steam generator safety valves 7.5 Minimum DNBR occurs (a) n v
a. DNBh does not decrease below its initial value.
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i 15.3 Decrease'in Reactor Coolant System Flowrate Several~ faults can result in_ a decrease in the RCS.flowrate. This section Ddiscusses these events.
'15.3.1 Partial loss of Forced Reactor Coolant Flow 15.3.1.1 Introduction.
A partial loss-of-forced-reactor-coolant flow accident can result from a mechanical or electrical' failure in an RCP or from a fault in the power supply to the pump or pumps supplied by an RCP bus. If the reactor is at power at the time of the accident, the'immediate effect of the. loss-of-forced-reactor coolant flow is .a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor does not trip promptly. Two buses connected to the generators supply power to the pumps. When a generator trip occurs,.the buses are automatically transferred to a ; transformer supplied from external power lines, and the pumps continue to operate. .Following any turbine trip where there- are no electrical faults which require tripping the generator from the r.etwork, the generator remains connected to the network for approximately 30 seconds. The RCPs
' remain connected to the generator, thus ensuring full flow for * .approximately 30 seconds after the reactor trip before any transfer is-made.
The low primary. coolant flow reactor trip signal, which actuates in any reactor coolant loop by.two out of three-low-flow signals, provides the necessary protection' against this_ event.L Above permissive P-8, low flow
-in any. loop will actuate a reactor-trip. Between approximately .10' percent power (permissive P-7)- and the power level corresponding to permissive P-8=, low flow'in .any -two loops will actuate a reactor trip. Above . permissive P-7, two or more RCP circuit breakers from:the same bus will open'whichi will actuate the corresponding undervoltage relays. This results in a reactor- trip which serves as backup to the flow trip.
Thi_s is a ANS Condition 11 incident, m 15.3.1.2 ' Method of Analysis o p This analysis examines partial loss-of-forced-reactor-coolant flow }, involving loss of two ' pumps with four loops in operation. , Thisianalysis_uses three. digital computer codes. First the LOFTRAN code (Reference-3)' calculates the loop and core-flow during the transient, the
~
y ' time of reactor' trip based on the calculated flows, the_ nuclear power
, .-transient,.andJthe primary system pressure and temperature' transients.
The FACTRAN' code (Reference 2) then calculates the heat' flux transient based on the nuclear power and flow from LOFTRAN. Finally, the THINC code
, calculates the DNBR during the transient based on the heat flux from -. 'FACTRAN and flow. from LOFTRAN. The DNBR transients presented represent i []- ' the minimum of the typical. or thimble fuel assembly cell .
L l' L A-15.3-1
;;; = . .
This analysis employs RTDP methodology; therefore, the initial conditions A assume nominal values of power, reactor coolant average temperature, and (") RCS average pressure (see Tables 15.0.3-2 and 15.0.3-3). The limit DNBR includes uncertainties in the initial conditions. The analysis assumes a conservatively large absolute value of the Doppler-only power coefficient (see Figure 15.0.4-1). This is equivalent to a total integrated Doppler reactivity from 0 to 100 percent power of 0.016 Ak. The analysis assumes the most positive moderator temperature coefficient (minimum moderator density coefficient) since this results in the maximum core power during the initial part of the transient when the transient ' reacnes minimum DNBR (see Figure 15.0.4-2).
-These analyses use the curve of trip reactivity inserti<,n versus time (Figure 15.0.5-3).
The basis for the flow coastdown analysis is a momercum balance around each reactor coolant loop and across the reactor care. This momentum balance is combined with the continuity equation, a pump momentum balance, and the pump characteristics and is based on high estimates of system pressure losses. No single active failure in any plant systems or equipment will adversely affect the consequences of the accident. 15.3.1.3 Resulti (G ~) Figures 15,3.1-1 through 15.3.1-4 show the transient response for the loss of power to two RCPs with four loops in operation. The reactor trips on the low-flow signal . Figure 15.3.1-4 shows the DNBR to be always greater than the safety analysis limit value for the most limiting fuel assembly cell. Since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not significantly reduced. Thus, the average fuel and clad temperature do not increase significantly above their respective initial values. The time sequence of events is shown in Table 15.3.1-1 for the partial loss of flow event. 15.3.1.4 Conclusions The analysis shows that the minimum DNBR always remains above the limit value during the transient. Thus, all applicable acceptance criteria are met. The conclusions presented in the FSAR remain valid. f% A-15.3-2
i 15.3.2 Comolete loss of Forced Reactor Coolant Flow (h
\ ') 15.3.2.1 Introduction A loss-of-forced-reactor-coolant flow may result from a simultaneous loss of electrical power to all RCPs. If the reactor is at power at the time of the accident, the immediate effect of a loss-of-forced-coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent adverse effects to the fuel if the reactor does not trip promptly. The reactor trip together with flow sustained by the inertia of the pump impeller will be sufficient to prevent RCS cverpreerization and the DNBR from exceeding the limit values. The trip systems available to mitigate the consequences of this accident are the following:
o RCP power supply bus undervcitage or underfrequency o Low reactor coolant loop flow These trip functions are fully described in the FSAR. This is an ANS Condition 111 incident. 15.3.2.2 Method of Analysis The method of analysis and the assurrptions made regarding initial A. operating conditions and reactivity coefficients are identical to those V discussed in Section 15.3.1, except that following the loss of power supply to all pumps at power, a reactor trip actuates by either RCP power supply undervoltage or underfrequency. 15.3.2.3 Results Figures 15.3.2-1 through 15.3.2-4 show the transient response for the loss of power to all RCPs with four loops in operation. The reactor trips on the undervoltage signal. Figure 15.3.2-4 shows the DNBR to be always greater than the safety analysis limit value for the most limiting fuel assembly cell. Since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not significantly reduced. Thus, the average fuel and clad temperature do not increase significantly above their respective initial values. Besides the complete loss-of-forced reactor-coolant flow (loss of power to four pumps), an underfrequency event with a frequency decay rate of 5 Hz/sec was also analyzed. For this event, the reactor trip occurs on an underfrequency signal. The DNBR analysis of the underfrequency event verified that the ONBR remains above the safety analysis limit value. If the maximum grid frequency decay rate is less than approximately 2.5 Hz/sec, the low flow signal will actuate a reactor trip which will protect the core from underfrequency events. The time sequence of events is shown in Table 15.31-1 for the complete loss of flow event. A-15.3-3 1-___________________----- --- - _ _ . _ _ . - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - - - - - _ _ --------------_--__________ _ .
a l 15.3.2.4. Conclusions ,
~
i' The analysis shows that the minimum DNBR always remains above the limit value during the transient. Thus, the analysis does not predict any adverse fuel. effects or clad rupture and all applicable acceptance i criteria are met. The conclusions presented in the FSAR remain valid. , 15.3.3 Reactor Coolant Pumo Shaft Seizure flocked Rotor) .j 15.3.3.1 Introduction For the instantaneous seizure of an RCP rotor, flow through the affected reactor coolant loop is rapidly reduced, leading to a reactor trip on a i low flow signal . Following the trip, heat stored in the fuel rods continues to be transferred into the core coolant, causing the coolant to expand. At the same time, heat transiar to the shell side of the steam generator reduces, first because the reoc ed flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cools down while the shell side temperature k,rreases (turbine steam
' flow reduces to.zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with the reduced heat transfer in the steam generator, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer causes a pressure increase, which .in turn actuates the automatic spray system, opens the power-operated relief valves, and opens the pressurizer safety valves .in that sequence. The power-operated relief valves are safety grade and would be expected to-function properly during an accident; however, for
(' conservatism, the, analysis does not use the pressure-reducing effect of the-power-operated' relief. valves and the pressure-reducing effect of the spray. The analysis of the locked rotor event demonstrates that overpressurization~of.the'RCS does'not occur and that the core remains in a coolable geometry, included in the analysis are the design changes
-associatad:with the transition to VANTAGE 5 fuel and other modified safety analysis assumptions.
This is an ANS~ Condition IV-incident. 15.3.3.2 Method of Analysis The analysis of this transient uses two digital computer codes. The
,t.0FTRAN code (Reference 3) calculates: 1) the resulting loop and core flow-transients- following the pump seizure; 2) the time of reactor trip based on the loop- flow transients; 3) the nuclear power following reactor-trip; and 4) the. peak RCS pressure. The thermal behavior of the fuel located'at~the: core hot spot is investigated using the FACTRAN code -(Reference 2) based on the core flow and the nuclear power calculated -by LOFTRAN. The FACTRAN-code uses a film boiling heat transfer coefficient.
c The plant characteristics and the initial conditions are shown in
' - Table.-15.0.3-2 and Table 15.0.3-3. The analysis evaluates the transient with and'without offsite power available.
No single active-failure in any slant systems or equipment will adversely affect the consequences of the accident. A-15.3-4
- _ - - - - .- . . ..- - - - -- - -. - ~.- .. - - - .- 15 .' 3 ~. 3 '. 2 . 2 - - Evaluation of the -pressure Transient t - After pump seizure, the neutron flux is rapidly reduced by control rod insertion due to reactor trip on low coolant flow in the affected loop. Rod motion begins 1 second after the flow in the affected loop reaches 87= percent' of nominal flow. No credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow af ter plant trip. Although these components will operate and will. result in a lower peak RCS pressure, ignoring their effect provides an additional degree of conservatism. The analysis conservatively bounds the pressurizer safety valves opening at 2500 psia and achieving rated flow at 2575 psia.
.15.3.3.2.3 Evaluation of DNB in the Core Ourino the Accident Because DNB occurs in the core for this accident, there is an evaluation of the consequences with respect to fuel rod thermal transients. Results obtained from analyses of this. " hot spot" condition represent the upper 1imit with respect to clad temperature and zirconium-water reaction.
In the evaluation, the rod power at the hot spot is conservatively assumed to be 2.55. times the average rod power (i.e., Fq = 2.55) at the initial
. core power. level .
15.3.3.2.4 Film'Boilina Coefficient To'model the effect of ONB occurring, the FACTRAh code calculates the film boiling coefficient using the Bishop-Sardberg-Tong film boiling-correlatien. . Fluid properties are evaluated at film temperatures -(average between wall and bulk temperatures). The program calculates the film coefficient at every time step. based upon the actual heat transfer conditionsJat the time. The neutron flux, system pressure, bulk density, and mass flowrate as a function of time are program inputs. This analysis uses the initial values of the pressure and .the bulk density throughout the transient since they are the most-conservative with respect to clad temperature response. For conservatism, the anaijsis assumes ONB
.to: start at the beginning of the accident to maximize the fuel rod thermal transient.
15.3.3.2.5 Fuel Clad Gao Coefficient The magnitude and time dependence of the heat transfer coefficient-between fuel and clad (gap coefficient) have a. pronounced influence on the thermal results. The larger the value of the gap coefficient, the more heat transferred between pellet and clad. Based on investigations of the effect;of the gap coefficient upon the maximum clad temperature during the transient,' the. analysis assumes the gap coefficient to increase from a steady-stgte .value ' consistent with initial fuel temperature to 10,000 Btu /hr-ft *F at the-initiation of the transient. Thus', the large amount of ene.'gy stored in the fuel because of the small initial value releases to the clad at' the initiation of the transient.
.A-15.3-5
15.3.3.2.6 Zirconium-Steam Reaction The zirconium-steam reaction can become significant a5ove 1800*F (clad temperature), in order to take this phenomenon into account, the models (Reference 4) introduced the following correlation which defines the rate of the zirconium-steam reaction, 2 d(w ) = 33.3 x 106 x , - [(45000.)/(1.986 T)) dt where: w = amount reacted, mg/cm2 t - time, sec T = temperature, 'F The reaction heat is 1510 cal /g 15.3.3.3 Results The transient results for the locked rotor accident are shown in Figures 15.3.3-1 through 15.3.3-4. Table 15.3.3-1 also summarizes the results of the locked rotor calculations. The peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted. condition stress limits of the American Society of Mechanical Engineers Code, Section Ill. Also, the peak clad temperature is considerably less than 2700*F. Note that the clad temperature was conservatively calculated assuming DNB occurs at the initiation of the 'v transient. These results represent the most limiting conditions of the locked rotor event or RCP shaft break. As a result of this accident, a fraction of the fuel rods will undergo DNB and will release gap inventory to the reactor coolant. Fewer than 5 percent of the fuel rods in the core will have clad damage, Appendix C discusses the evaluation of the radiological consequences of a postulated locked rotor accident. The calculated sequence of events for the locked rotor event is shown in Table 15.3.1-1. 15.3.3.4 Conclusions o The maintenance of the integrity of the primary coolant system occurs because the peak RCS pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted conditions stress limits, o Since the peak clad surface temoerature calculated for the hot spot during the worst transient remains considerably less than 2700*F (the temperature at which clad embrittlement may occur), the core will remain in a coolable geometry. The conclusions presented in the FSAR remain valid. A-15.3-6
15.3.4- Reactor Coolant Pumo Shaft Brqah r' l The analysis for the locked rotor event described in Section 15.3.3 N- represents the most limiting event with respect to a locked rotor or pump shaft break. With a failed shaft, the impeller could conceivably be free to spin in the reverse direction as opposed to the assumption of being in a-fi ed position as for a locked rotor. The effect of such reverse spinning is a slight decrease in the end point (steady-state) core flow when compared to the locked rotor. The analysis described in Section 15.3.3 modeled the most limiting conditions for the locked rotor and shaft break; therefore, this section does not present a separate analysis.
.O V
I l l' l l-O V A-15.3-7 l
l 1 Table 15.3.1-1 i
\ Time Seouence of Events for incidents Which Result in a Decrease in Reactor Crolant System Flowrate Time Delays Accident Event (s) __
Partial loss of forced reactor coolant flow Loss of two pumps with four Coastdown begins 0.0 loops in operation Low-flow reactor trip 1.4 Rods begin to drop 2.4 Minimum DNBR occurs 3.6 Complete loss of forced reactor coolant flow Loss of four pumps with All operating pumps lose power with four loops in operation and begin coasting down 0.0 RCP undervoltage trip point A reached 0.0 t,j Rods begin to drop 1.5 Minimum DNBR occurs 3.2 RCP shaft seizure (locked rotor) One locked rotor with four loops Rotor on one pump locks 0.0 in operation with offsite power available Low-flow trip point reached 0.0 Rods begin to drop 1.0 Maximum RCS pressure occurs 3.3 Maximum clad average temperature occurs 3.5 One locked rotor with four loops Rotor on one pump locks 0.0 in operation without offsite power available low-flow trip point reached 0.0 Rods begin to drop. 1.0 Maximum RCS pressure occurs 3.4 O Maximum clad average temperature kJ occurs 3.6 A-15.3-8
1 Table 15.3 3-1 Summary of Results for the Locked Rotor Transient (Four tooos Operatina Initially) Value for Value for With Offsite Without Offsite Parameter Power Available Power Available Maximum RCS pressure (psia) 2669,0 2669.0 Maximum clad average temperature, core hot spot ('F) 2048.0 2054.0 Zr-H O reaction, core hot spot 2 (percent by weight) 0.6 0.7 O O A-15.3-9
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15.4 Reactivity and power Distribution Anomalies t
" Several postulated faults can result in reactivity and power distribution anomalies. Control rod motion, control rod ejection, boron concentration changes, or addition of cold water to the RCS results in reactivity changes. Control rod motion, control rod misalignment, control rod ejection, or fuel assembly mislocation results in power distribution changes. This section discusses these events.
15.4.1 Uncontrolled Rod Cluster Control Assemb1v Bank Withdrawal from a Suberitical or Low-Power Startuo Condition , 15.4.1.1 Introduction An RCCA withdrawal incident is an uncontrolled addition of reactivity to the reactor core caused by withdrawal of RCCA banks resulting in a power excursion. While the occurrence of a transient of this type is highly unlikely, a malfunction of the control rod drive system can cause such a transient. This could occur with t* reactor either suberitical, low power startup, or at power. Sectiu '.5.4.2 discusses the "at power" case. RCCA bank withdrawal adds reactivity at a prescribed and controlled rate to bring the reactor from a suberitical condition to a low power level during-startup. Although the initial startup procedure uses the method of boron dilution, the normal startup is with RCCA bank withdrawal. RCCA bank motion can cause much faster changes in reactivity than can be made *
'. by changing bo_on r concentration. .
p . The control rod drive mechanisms wire into preselected banks which remain
, unchanged during the core life. The circuit design is such that RCCAs can .* not be withdrawn in other than their procer withdrawal sequence. Control of the power supplied to the rod banks is such that no more than two banks can be withdrawn at any time. The RCCA drive mechanism is the magnetic ,
latch type, and the coi', actuation sequencing provides variable speed
~ travel. -The analysis :f the maximum reactivity insertion rate includes the assumption of the simcitaneous withdrawal of the two sequential banks having the maximum: combines worth at maximum speed. ' Should a continuous control rod assembly withdrawal initiate, the-transient will terminate by the following reactor trip functions:
7 1 o Source range high neutron flux reacor trip is actuated when either of two independent source range channe's indicates a flux level above a preselected, manually adjustable setpo:nt. This trip function may be manually bypassed when the intermediate range flux channel indicates a flux level above the source range cutoff level. It is automatically reinstated when both intermediate range channels indicate a flux level below a specified setpoint. o Intermediate range high neutron reactor flux trip is actuated-when either of two independent intermediate range channels indicates a flux level above a preselected, manually adjustable setpoint. This trip function may be manually bypassed when two of the four power i l A-15.4-1 1 ~ . _ . . . - - . . - . - _ -. -. - - - , . . . . ~, - . . . , . - , _ . , - - _ - . _ . . - _ - . -
l range channels are reading above aoproximately 10 percent of full power / flux level and is automatically reinstated when three of the Q four power range channels indicate a power / flux level below this s
/ setpoint.
o Pvwer range high neutron flux reactor trip (low setting) is actuated when two out of the four power range channels indicate a flux level above approximately 25 percent of full power / flux level. This trip function may be manually bypassed when two of the four power range channels indicate a flux level above approximately 10 percent of full power / flux level and is automatically reinstated when three of the four channels indicate a power / flux level below this setpoint, o Power range high neutron flux reactor trip (high setting) is actuated when two out of the four power range channels indicate a flux level above a preset setpoint. This trip function is always active. o High nuclear flux rate reactor trip is actuated when the positive rate of change of neutron flux on two out of four nuclear power range channels indicates a rate above the preset setpoint. It is always active. In addition, control rod stops on high intermediate range flux (one out of two) and high power range flux (one out of four) serve to cease rod withdrawal and prevent the need to actuate the intermediate range flux trip and the power range flux trip, respectively. This is an ANS Condition 11 incident. p , V 15.4.1.2 Method of Analysis , The following three stages comprise the analysis of the uncontrolled RCCA bank withdrawal from subtritical accident: first, an average core nuclear power transient calculation; then, an average core heat transfer calculation; and finally, the DNBR calculation. The spatial neutron kinetics computer code TWINKLE (Reference 1) performs the average core calculation to determine the average power generation with time including the various total core feedback effects, i.e. Doppler reactivity and moderator reactivity. FACTRAN (Reference 2) performs a fuel rod transient heat transfer calculation to determine the average heat flux and temperature transients. The average heat flux is next used in THINC for transient DNBR calculations, in order to give conservative results for a startup incident, the following additional assumptions are made con;erning the initial reactor conditions: A. Since the magnitude of the neutron flux peak reached during the initial part of -the transient for any given rate of reactivity insertion is strongly dependent on the Doppler power reactivity coefficient, the analysis employs a conservatively low value for Doppler power defect (-940 pcm). B. The contribution of the moderator reactivity coefficient is g negligible during the initial part of the transient because the heat transfer time constant between the fuel and the moderator is much (v) longer than the neutron flux response time constant; however, after l A-15.4-2
I the initial neutron flux peak,- the moderator temper 4ture retctivity coefficien' affects the succeeding rate of power increase. The analysis assumes a moderator temperature coefficient which is O +7 pcm/'F at the zero power nominal temperature. 1 C. . The analysis-assumes the reactor to be at hot zero power (557'F). This assumption is more conservative than that of a , lower initial system temperature. The higher initial system temperature yields a larger fuel-to-water heat transfer coefficient, a larger fuel-specific heat, and a less-negative (smaller absolute magnitude) Deppler coefficient; these reduce the Doppler feedback effect, thereby increasing the neutron flux peak. The high neutron flux peak combined with a high fuel specific heat and larger heat transfer coefficient yields a larger peak heat flux. The analysif assumes the initial effective multiplication factor (k f) to be I 1.0 since this results in the maximum neutron flux pea D. - The most adverse combination of instrumentation error, setpoint error, delay for trip signal actuation, and delay for control rod assembly release.are taken into acqount. The analysis assumes a 10-percent increase in the power range flux trip setpoint, raising it from the nominal value of 25 percent to a value of 35 percent and not taking any credit for the source and intermediate range protection. , Figure-15.4.1-1 shows that the rise in nuclear flux is so rapid that ' the effect.of error in the trip setpoint on the. actual time at which I the rods release is negligible. Besides the above, the assumption ) that the highest worth control rod assembly is stuck in'its fully withdrawn position is the basis of the rate of negative reactivity g -insertion corresponding to the reactor trip. U E. The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of=the two. sequential control banks having the greatest combined worth at maximum speed
. (45 in./ min).
F. The'DNB' analysis assumes the most limiting axial and radial power l
- shapes associated with having the two highest combined worth banks in their high-worth position.
G. The analysis assumes the initial power level .tg be below the-power level expected for any shutdown condition (10~ fraction of nominal power). The combination of highest reactivity insertion rate and low initial power produces the highest peak heat' flux. H. The analysis assumes two RCPs to be in operation (Mode 3 Technical
- Specification allowed operation). This is conservative with respect to the DNB transient. - The accident analysis employs the STDP with the initial conditions shown in Tables 15.0.3-2 and 15.0.3-3.
No single active failure in any plant systems or equipment will adversely affect the consequences of the accident. O A-15.4-3 k
-- - ,,--,~,-.,._,.~n, -.-,.n _ - . . . , - . , - _ , , _ _ _ , , , _ _ , , , , __
15.4.1.3 Results The nuclear power, heat flux, fuel average temperature, and clad O' temperature versus time transient results are shown in Figures 15.4.1-1 d' through 15.4.1-3. In addition, Table 15.4.1-1 presents the time sequence of events. For all regions of the core, the DNB design basis is met. 15.4.1,4 Conclusions The minimum DNBR remains above the safety analysis limit value; therefore, the conclusions presented in the FSAR remain valid. 15.4.2 Uncontrolled Rod Cluster Control Assembiv Bank Withdrawal At Power 15.4.2.1 Introduction An uncontrolled RCCA withdrawal at power results in an increase in core heat flux. Since the heat extraction from the steam generator lags behind the core power generation until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually
result it. DNB; therefore, to avert damage to the fuel clad the reactor protection system is designed to terminate any such transient before the ONBR falls below the limit value.
(v] . The automatic features of the reactor protection system which prevent core damage in an RCCA bank withdrawal incident at power include the following: o Power range neutron flux instrumentation actuates a reactor trip on high neutron flux if two out of four channels exceed an overpower setpoint. o Reactor trip actuates if any two out of four AT channels exceed an OTAT setpoint. This setpoint is automatically varied with axial power distribution, coolant average temperature, and coolant average pressure to protect against DNB. o Reactor trip actuates if any two out of four AT channels exceed an OPAT setpoint. This setpoint is automatically varied with coolant average temperature so that the allowable heat generation rate (kW/ft) is not exceeded, o A high pressurizer pressure reactor trip, actuated from any two out of four pressure channels, is set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves, o Any two out of three level channels when the reactor power is above l approximately 10 percent (permissive P-7) uill actuate a high l pressurizer water level reactor trip. V A-15.4-4 i l L. _. __ _ _ _ _ _ _ -. __. - --
Bssides the above listed ranctor trips, there are the following RCCA withdrawal blocks: 0 High neutron flux (one out of four) I o OPAT (two out of four) i
o. OTAT (two out of four)
FSAR Chapter 7 describes _the manner in which the combination of OPAT and OTAT trips provide protection over the full range of RCS conditions. Figure 15.0.6-1 presents allowable reactor coolant loop average temperature and AT for the design power distribution and flow : as a function of primary coolant pressure. J The purpose of this analysis is to demonstrate the manner in which the above protective systems function for various reactivity insertion rates 1 from different initial conditions to prevent fuel damage. Reactivity ! insertion rates and initial conditions influence which protection function 1 actuates first. This is an ANS Condition 11 incident. 15.4.2.2 Method of Analysis
.The analysis of this transient employs the LOFTRAN code (Reference 3).
LOFTRAN uses the core limits.as illustrated in Figure
15.0.6-1 as input to ,
determine the minimum DNBR during the transient. l The analysis of this accident _ uses the RTOP described in Reference _7. Os Section 15.0.3 discusses.the plant characteristics and initial conditions.- For an uncontrolled RCCA bank withdrawal at power accident, , the analysis assumes the following conservative assumptions: A' . - Nominal values form the basis of the initial reactor power, pressure, l and RCS temperature assumption.. The limit DNBR includes uncertainties _in initial _ conditions as described in Reference 7. B. Reactivity coefficients -- two_ cases analyzed: J
1. A +7 pcm/'F moderator temperature coefficient of reactivity and a least-negative Doppler-only power coefficient form the-basis of_ the beginning of life minimum reactivity fetdback assumption.

2. A conservatively large positive moderator density coeificient (corresponding to a_large negative moderator temperatore-coefficient) and a most-negative Doppler-only power er, efficient form the basis of the end of life maximum reactivity feedback i assumption.
C.- A conservative value'of -118 percent of nominal full core power actuates the reactor trip on high neutron. flux. The AT-trips include all adverse: instrumentation and'setpoint errors while maximum values form the basis of the delays for the trip signal actuation assumption. O A-15.4-5 l l
) ; i t D.- The assumption that the highest worth assembly is stuck in its fully 'j withdrawn position forms the basis of the RCCA trip insertion
. Characteristic. i E. The analysis examines a range of reactivity insertion rates. The maximum positive reactivity insertion rate is greater than that for '
the simultaneous withdrawal of the two control banks having the maximum combined worth at maximum speed assuming normal overlap. 1 No single active failure in any plant systems or equipment will adversely affect the consequences of the accident. i
. 15.4.2.3 Results a,
Figures 15.4.2-1-through 15.4.2-3 show the transient response for a rapid RCCA bank withdrawal incident starting from full power with minimum feedback. Reactor trip on high neutron flux occurs shortly after the start of the accident. Because of the rapid reactor trip with respect to the thermal time constants of the plant, small changes in T ayg and pressure result, and the margin to DNB is maintained. , The transient response for a slow RCA bank withdrawal from full power
-with minimum feedback is shown in Figeres 15.4.2-4 through 15.4.2-6.
Reactor trip on OTAT occurs after a lone r period and the rise in temperature and pressure is consequently targer than for rapid RCCA bank withdrawal'. Again, the minimum DNBR is greater than the limit value. Figure 15.4.2-7 shows the. minimum 0NBR as a function of reactivity , insertion rate from initial full power operation for both' minimum and u O maxim a reactivity feedback, it can be'seen that the two reactor trip
- functions (high neutron flux and OTAT functions) provide DNB protection over the whole range of reactivity insertion rates. The minimum DNBR ts always greater than the limit value.
i; Figures 15.4.2-8 and 15.4.2-9 show the minimum DNBR as a function of reactivity insertion-rate-for RCCA-bank withdrawal incidents starting at ~
60 and-10 percent power, respn tively. The results are similar to the 100 percent power case; however, as the initial power decreases, the' range
- over which the OTAT trip'is effective is increased. In neither case does the DNBR fall below the limit value, The~ shape of the curves of minimum DNBR versus reactivity insertion rate in'the referenced figures is G e both to reactor core and coolant system ,
transient response and to protection system action in initiating a reactor t ri p .'- 4 The reactor trips sufficiently fast during the RCCA bank withdrawal at
oower_ transient to ensurt. that there is not a reduction in the ability of the primary coolant to remove haat from the fuel rods. Thus, the fuel cladding temperature does not rist significantly above its initial value
- during_the transient.
Table 15.4.1-1 shows the calculated sequence of events for the I uncontrolled RCCA bank withdrawal at power incident. O A-15.4-6 i w,. % h -c y g , . ~ ew---, ,~,,y,. -e,ywge@.--, ,-r.. .,.,-,.,,y ,,- ,,o.,--w.gwo,wr%,me.,,,,--,-w,.-w -, , , , - . - - , , . , , . w---w--omav,,,*-r,mie-.~+-3.-,,-.r,wy ep 4D
'15.4.2.4 Conclusions , The high neutron- flux- and OTAT trip functions provide adequate protection over the entire range of possible reactivity insertion rates O (i.e., the minimum value of DNBR is always larger than the limit value for all fuel types); therefore the conclusions presented in the FSAR remain -valid.
15.4.3 Rod Cluster Control Assembly Misalionment (System Malfunction or Qgerator fy_tpd 15.4.3.1 Introduction RCCA misoperation accidents include the followingt o One or more dropped RCCAs within the same group o A dropped RCCA bank o Statically misaligned RCCA o Withdrawal of a. single RCCA i Each RCCA has a position indicator channel which displays the position of the assembly in a display grouping that is convenient to the operator. Fully inserted assemblies are also indicated by a rod at bottom signal i which actuates a local alarm and a control rot)m annunciator. Group demand position is also, indicated. RCCAs move in preselected banks, and the banks move in the same preselected sequence. Each bank of RCCAs consists of two groups. The i rods comprising a group operate in parallel through multiplexing
-thyristors. -The two groups in a bank move sequentially such that the ;
first group is always within one step of the second group in the bank. A 1 definite schedule of actuation (or deactuation of the stationary gripper, i movable gripper,.and lift coils:of a mechanism) withdraws the RCCA l attached to the. mechanism. Mechanical failures are in the direction of l insertion or immobility.:
.No single electrical or mechanical failure in the rod control system could r cause the accidental withdrawal of a single RCCA from the inserted bank at
"> full power operation. The event analyzed must result from multiple wiring failures, multiple-significant operator errors, or subsequent and repeated : L operator disregard-of event indication. The-probability of such a L combination of conditions is low such that the limiting consequences may include slight fuel damage. The following indicators detect one or more dropped RCCAs, RCCA group, or 'i RCCA bank: i o Sudden drop in the core power level as seen by the nuclear instrumentation system o Asymmetric power distribution as seen on out-of-core neutron detectors or coro exit thermocouples O. -o Rod at bottom signal l A-15.4-7 -l p
o Rod deviation alarm o Rod position indication The following indicators detect misaligned RCCAs: o Asymmetric power distribution as seen on out-of-core neutron j detectors or core exit thermocouples o Rod deviation alarm
.o Rod position indicators The resolution of the rod position indicator channel is 25 percent of span (17.Cin.). Deviation of any RCCA from its group by twice this distance (10 percent of span or 15 in.) will not cause power distributions worse ] i than the design limits. The. deviation alarm alerts the operator to rod :
deviation with respect to the group position in excess of 5 percent of span.- If the rod deviation alarm is not operable, the Technical Specifications require the., operator.to take action. If' one or more rod position. indicator channels is out of service, the operator must fellow detailed operating instructions to ensure the , alignment of the nur.-indicated RCCAs. The operator is also required to l take action, as required by the Technical Specifications. In the unlikely event of simultaneous electrical failures which could
result in single RCCA withdrawal, the plant annunciator will display both '
the rod deviation and rod control urgent failure; and the rod position ! indicators will indicate the. relative pnsitions of the RCCAs in the bank. The _ urgent failure alarm also inhibits automatic rod motion in'the group oO in which it occurs. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, would result in
)
activation of the same' alarm and the same visual indication. The OTAT j reactor trip provides automatic protection for this event, although due to ! the increase in local power density, it is not possible to always provide i assurance that the core safety limits will not.be exceeded. l i 15.4.3.2 = Method of Analysis A. One or More Oropped RCCAs from'the-Same Group l The LOFTRAN computer code (Reference 3) calculates the transient ! system response for_ the evaluation of the dropped RCCA event. The o code simulates the neutron kinetics, RCS, pressurizer, pressurizer , H relief and safety valves, pressurizer spray, steam generator, and j steam generator safety valves _. The code computes pertinent plant i
, variables including temperatures, pressures, and power level.
Calcul'ated statepoints and nuclear models form the basis used to I obtain a hot channel factor consistent with the primary system i 4 conditions and reactor power. _By incorporating the primary ; L..# conditions from the transient and the' hot channel factor from the
nuclear analysis, the DNB design basis'is shown to be met using the .'
" THINC code. The transient response analysis, nuclear peaking factor analysis, and performance of the DNB design basis confirmation are in O A-15.4-8 r b
l l accordance with the methodology described in Reference 5. Note that the analysis does not takt credit for the negative flux rate reactor trip.
) B. Dropped RCCA Bank A dropped RCCA bank results in a symmetric power change in the core.
As discussed in Reference 5, assumptions made for the dropped RCCA(s) analysis provide a bounding analysis for the dropped RCCA bank. C. Statically Misaligned RCCA FSAR Table 4.1-2 describes the computer codes used in the analysis of steady-state power distributions. The peaking factors are then used as input to the THINC code to calculate the DNBR. The analysis examines the case of the worst rod withdrawn from bank D inserted at the insertion limit with the reactor initially at full power. The analysis assumes this incident to occur at beginning of life since this results in the minimum value of moderator temperature coefficient. This assumption maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution. D. Single RCCA Withdrawal at Full Power FSAR Table 4.1-2 describes the computer codes used in the calculation of power distributions within the core. The peaking factors are then used by THINC to calculate the minimum DNBR for the event. The plant's analysis is for the case of the worst withdrawn rod from D bank inserted at the insertion limit, with the reactor initially at full power. The analysis assumes the transient to occur at beginning (V ) of life since this results in the minimum value of moderator temperature coefficient. This assumption maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution. 15.4.3.3 Results A. One or More Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion. The core is not adversely affected during this period, since power is decreasing rapidly. Either reactivity feedback or control bank withdrawal will reestablish power. Following a dropped rod event in manual rod control, the plant will establish a new equilibrik ondition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a c9ncern and establishing the automatic rod control mode of operatioi as the limiting case. For a dropped RCCA event 11 the automatic rod control mode, the rod control system detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by the p automatic rod controllee after which the control system will insert (j the control bank to rester" nominal power. Figure 15.4.3-1 shows a A-15.4-9
i l l E typical transient response to a dropped RCCA (or RCCAs) in automatic i rod control mode. In all cases, the minimum DNBR remains above the ; limit value. l plant stabilization, the operator may manually retrieve the Followinfollowingapprovedoperatingprocedures. RCCA by B .- Dropped RCCA Bank A dropped RCCA bank results in a negative reactivity insertion i L greater than 500 pcm. The core is not adversely affected during the . insertion period. since power is decreasing rapidly. The transient will proceed as described in Part A; however, the return to power will be less due to the greater worth of the entire bank. The power transient for a dropped RCCA bank is symmetric. Following plant stabilization, normal procedures are followed. 4 C. Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at significant power levels arise from cases in which one RCCA is fully inserted or where bank D is fully inserted with one RCCA fully [
' withdrawn. Multiple independent alarms, including a bank insertion limit alarm, alert the operator well-before the transient approaches the postulated conditions. The bank can be inserted to its insertion limit with any-one assembly fully withdrawn' without the DNBR falling below the limit value.
,' The insertion limits in the Technical Specifications may vary from
-time to time. depending on several limiting criteria. The full-power insertion limits on control bank D must be chosen to be above that O- position which meets the minimum DNBR and peaking factors. The full power insertion limit is usually dictated by other criteria.
1 Detailed results will vary from cycle to cycle depending on fuel arrangements.- For this RCCA misalignment, with bank D inserted to~its-full-power insertion limit and one RCCA fully withdrawn, DNBR does not fall below the limit value. The analysis of this case assumes that the initial reactor power, pressure, and RCS temperature are at their nominal values with the increased radial peaking factor associated with the misaligned RCCA. c For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall.below the limit value. The analysis of this case assumes that initial reactor power, pressure, and RCS temperatures are at their nominal values, with the increased radial peaking factor associated with the misaligned RCCA. DNB does not occur for the RCCA misalignment incident, thus there is. no reduction in the ability _of the primary coolant to remove heat
*from the fuel rod. The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution. The resulting linear heat generation rate is well below that which would cause fuel melting.
O A-15.4-10
1. _ , _ . .
After identifying an RCCA group misalignment condition, the operatar must take action as required by the plant Technical Specifications and operating instructions. D. The analysis of the single rod withdrawal event considers the following two events:
1. If the reactor is in the manual rod control mode, continuous withdrawal of a single RCCA results in both an increase in core power and coolant temperature and an increase in the local hot channel factor in the area of the withdrawing RCCA. Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip may not occur quickly enough to prevent the minimum DNBR from falling below the limit value. Evaluation of this case at the power and coolant conditions at which the OTAT trip would trip the plant shows that an upper limit for the number of rods with a DNBR less than the limit value is 5 '
percent.. 1
2. If the reactor is in the automatic rod control mode, the multiple failures that result in the withdrawal of a single RCCA cause immobility of the other RCCAs in the controlling bank. '
The transient will then proceed in the same manner as case 1 described above. For such cases as above, a reactor trip will ultimately ensue, although not quickly enough in all cases to prevent a minimum DNBR in the core of less than the limit value. Following reactor trip, normal shutdown procedures are followed'. O . 15.4.3.4 Conclusions For cases of dropped RCCAs or dropped banks, the DNBR remains greater than t the limit value; therefore, the DNB-design criterion is met. For all_ cases of any RCCA fully inserted, or bank D inserted to its rod insertion limits with any single RCCA in that bank fully withdrawn (static misalignment). the DNBR remains greater than the limit value. For the case of the accidental withdrawal of a single RCCA, with the
, reactor in'the automatic or manual control mode and initially operating at U full power with 0 bank at the insertion limit, an upper bound of the number of fuel rods experiencing DNBR is 5 percent of the total fuel rods in the core.
L o 15,4.4 Startuo of an Inactive Reactor Coolant Pumo at an incorrect i' Temeerature l
'15.4.4,1 Introduction Technical Specification 3/4.4.1 does not permit VEGP Units 1 and 2 operation in Modes 1 and 2 with less than four loops operating; however, this analysis assumes approximately 75 percent power in Mode 1 in order to bound Mode 3 operation where the Technical Specifications permit operation with less than ~ *O four loops.
A-15.4-ll i
- - - .-l,._..-..,-.. , , _ _ . . _ , _ , , , - _ _ . . _ _ . _ - - . - _ _ _ _ _ _ _ . - . _ . _ . - _ . . - . - - - . - _ _ . - -
If the plant operates with one RCP out of service, there is reverse flow through the inactive loop due to the pressure difference across the reactor vessel. The cold leg temperature of the inactive loop is identical to the cold CT leg temperature of the active loops. If the reactor is operated at power and assuming there is no isolation of the secondary side of the steam generator in (/ the inactive loop, there is a temperature drop across the steam generator in the inactive loop and, with the reverse flow, the hot leg temperature of the inactive loop is lower than the reactor core inlet tempers.ture. If the startup of an inactive RCP accident occurs, the transient terminates automatically by a reactor trip on low coolant loop flow when the power range neutron flux (two out of four channels) exceeds the P-8 setpoint which has been previously reset for three-loop operation. This is an ANS Condition 11 incident. 15.4.4.2 Method of Analysis The analysis of this transient uses three digital computer codes. The LOFTRAN computer code (Reference 3) calculates the loop and core flow, nuclear power, and core pressure and temperature transients following the startup of an idle pump. FACTRAN (Reference 2) calculates the core heat flux transient based on core flow and nuclear power from LOFTRAN. The THINC code is then used to calculate the DNBR during the transient based on system conditions calculated by LOFTRAN and heat fluxes calculated by FACTRAN. Section 15.0.3 discusses plant characteristics and initial conditions, in order to obtain conservative results for the startup of an inactive pump accident, the following assumptions are made (this analysis employed STDP): ? ) v A. Initial conditions of maximum core power and reactor coolant average temperatures and minimum reactor coolant pressure resulting in minimum initial margin to DNB. These values are consistent with the maximum steady-state power level that would be permitted with three loops in operation. The high initial power gives the greatest temperature difference between the core inlet temperature and the inactive loop hot leg temperature. B. Following initiation of startup of the idle pump, the inactive loop flow reverses and accelerates to its nominal full-flow value in approximately 9 seconds. C. The analysis assumes a conservatively large negative moderator temperature coefficient. D. The analysis assumes a least-negative Doppler-only power coefficient. E. The initial reactor coolant loop flows are at the appropriate values for one pump out of service. F. The reactor trip occurs on low coolant flow when the power range neutron flux exceeds the P-8 setpoint. The P-8 setpoint is conservatively assumed to be 84 percent of rated power, which corresponds to the nominal setpoint plus 9 percent for nuclear instrumentation errors. No single active failure in any plant systems or equipment will adversely p) affect the consequences of the accident. A-15.4-12
l 15.4.4.3 Results The results following the startup of an idle pump with the above listed assumptions are shown in Figures 15.4.4-1 through 15.4.4-5. These curves show l that during the first part of the transient, the increase in core flow with ! cooler water results in an increase in nuclear power and a decrease in the core l l- average temperature. The minimum DNBR during the transient is greater than the ' safety analysis limit values. Reactivity addition for the inactive loop startup accident is due to the i decrease in core water temperature. During the transient, this decrease is due < both to the increase in reactor. coolant flow and, as the inactive loop flow . reverses, to the colder water entering the core.from the hot leg side of the ; steam generator in the inactive loop. Thus, the reactivity insertion rate for i this transient changes with time. The resultant core nuclear power transient, computed with consideration of both moderator and Doppler reactivity feedback 3 effects, is shown in Figure 15.4.4-1. > The calculated sequence of events for this accident is shown in Table 15.4.1-1. The transient results illustrated in Figures 15.4.4 1 through 15.4.4-5_ indicate that a stabilized plant condition, with the reactor tripped, - is rapidly approached. By following normal shutdown procedures, the plant can subsequently achieve cooldown. - 15.4.4.4 Conclusions The transient resultt show that the core is not adversely affected. There is considerable margin 'o the ". Sty analysis limit DNBRs; thus, the DNB design basis ts met and the c e Nsion? presented in the FSAR remain valid.
15.4.5 A Malfunction or Failure t f the Flow Controller in a Boilina Water '
Reactor Loop that Result L n an Increased Reactor Coolant Flowrate Thi_s section is not applicable to VtGP Units 1 and 2. 15.4.6 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15,4.6.1 Introduction
-Feeding primary grade water into the RCS via the reactor makeup portion of the chemical and volume control system (CVCS) adds reactivity to the core. Baron dilution is a manual operation under strict administrative controls with r . procedures calling for a limit on the rate and duration of dilution. A boric acid blend system permits the operator to match the boron concentration of reactor coolant. makeup water during normal charging to that in the RCS. Even under various postulated failure modes, the design of the CVCS limits the potential rate of dilution to a value which gives the operator sufficient time to correct the situation in a safe and orderly manner.
The opening of.the primary water makeup control valve supplies water to the RCS which can dilute the reactor coolant. Inadvertent dilution can be readily terminated by closing this valve. In order to add makeup water to the RCS at A-15.4-13
pressure, at least one charging pump in addition to the primary makeup water pumps must be running. Normally, only one primary water supply pump is o operating while the other is on standby. The boric acid from the boric acid tank blends with primary grade water at the mixing tee, and the preset flowrates of boric acid and primary grade water on the control board determine the composition. Information on the status of reactor coolant makeup is continuously available to the operator. Lights on the control board indicate the operating condition of pumps in the CVCS. Alarms actuate to warn the operator if boric acid or demineralized water flowrates deviate from preset values as a result of system malfunction. This is an ANS Condition !! incident. 15.4.6.2 Method of Analysis To cover all phases of the plant operation, this analysis considers boron dilution during refueling, cold shutdown, hot shutdown, hot standby, startup, and power operation. The analysis assumes conservative values for the critical parameters, i.e., high RCS critical boron concentrations, high boron worths, minimum shutdown margins, and small RCS volumes. These result in conservative calculations of the time available for the operator to determine the cause of the addition and take corrective action before shutdown margin is lost. A. Dilution During Refueling 7m This analysis evaluates boron dilution events during refueling (Mode 6). During refueling, a very small amount of unborated chemical solution is-
!] '
allowed to enter the RCS for water chemistry quality control. The opening, of CVCS valves 176 and 177 provides the dilution flow path. The maximum flowrate possible through this flow path is less than 3.5 gpm. Any other chemical makeup solution required during refueling will be borated water supplied from the refueling water storage tank by the low head safety injection pumps. Valves 175 and 183 in the CVCS will be locked closed during refueling operations. These valves will block additional flow paths which could allow more than 3.5 gpm of unborated chemical makeup water to reach the RCS, B. Dilution During Cold Shutdown, Hot Shutdown, and Hot Standby This analysis evaluates boron dilution events during cold shutdown (Mode 5), hot shutdown (Mode 4), and hot standby (Mode 3). The analysis uses failure modes and effects analysis, human error, and event tree analysis to identify credible boron dilution initiators and to evaluate the plant response to these events. For the initiators identified, the calculation of the time intervals from alarm to loss of shutdown margin helped determine the length of time available for operator response. These calculations depended on dilution flowrates, boron concentrations, and RCS volumes specific to the event and mode of operation. The technique modeled realistic plant conditions and responses, including both mechanical f ailure and human errors, m A-15.4-14
The analysis identified four events which are the most likely initiators. FSAR Section 15.4.6.2.1.2 provides details on the events. (O ( The examination of Mode 5b (mid-loop operation) for the addition of small amounts of unborated chemical solution into the RCS for water quality chemistry control was included. The maximum flowrate possible for this flow path is approximately 3 percent of that associated with the limiting flaw path for Modes 3, 4, and Sa (RCS loops filled). C. Dilution During Full Power Operation, Including Startup For the dilution during startup (Mode 2), the analysis assumes an initial maximum critical boron concentration of 1800 ppm based on the rods being inserted to the insertion limits. The analysis assumes the minimum change in the boron concentration from this initial condition to a hot zero power critical condition to be 300 ppm. The analysis also asc"m full rod insertion to occur due to reactor trip, minus the most reactive stuck rod. The analysis assumes the dilution flow to be the combined capacity ofthetwoprimarywatermakeuppumgs(asproximately242gpm)anda T11s volume corresponds to the minimum RCS water volume of 9583 ft . active volume of the RCS minus the pressurizer and accounts for 10 percent steam generator tube plugging. During power operation (Mode 1), the plant operates under either manual or automatic rod control. While the plant is in manual control, the analysis assumes the dilution flow to be a maximum of 242 gpm, which is the combined capacity of the two primary water makeup pumps. While in automatic control, the maximum letdown flow (approximately 125 gpm) limits the dilution flow. The analysis assumes an initial maximum critical boron j concentration, corresponding to the rods inserted to the insertion limits ! ( at hot full power, of 1800 ppm. The analysis also assumes the minimum change in the boron concentration from this initial condition to a hot zero power critical condition to be 300 ppm. The analysis assumes full t l rodinsertiontooccurduetoreactortrip,minusthemosjreactivestuck rod. The analysis uses a minimum water volume of 9583 ft in the RCS, corresponding to the active volume of the RCS minus the pressurizer volume and accounts for 10 percent tube plugging. No single active failure in any plant systems or equipment will adversely affect the consequences of the accident, j l 15.4.6.3 Results L During refueling, the maximum flowrate associated with the available dilution flow path is very small. The total time from the initiation of the event to the eventual complete loss of shutdown margin is significantly large compared to the minimum required operator action time; therefore, the operator has sufficient time to terminate the RCS water chemistry adjustments before the loss of shutdown margin. Additionally, assuming the availability of the high l flux at shutdown alarm (set at 2.3 times background), it is shown that the l Technical Specification shutdown margin requirement for Mode 6 is sufficient to ensure that the operator has 30 minutes from the time of alarm to terminate the dilution before shutdown margin is lost. i For dilution during cold shutdown (RCS loops filled), the Technical Specifications provide the required shutdown margin as a function of RCS boron
'f concentration. The specified shutdown margin ensures that the operator has 15 A-15.4-15
1 minutes from the .ime of the high flux at shutdown alarm to the total loss of shutdown margin. For mid-loop operation during cold shatdown, the results are ( similar to those discussed above for refueling. For dilution during hot shutdown and hot standby, the Technical Specifications provide the required shutdswn margin as a function of RCS boron concentration. The specified shutdown ma/ gin ensures that the operator has 15 minutes from the time of the high flux at shutdown alarm to the total loss of shutdown margin. In the event of an unp'<anned approach to criticality or dilution during power escalation while in tte startup mode, the operator is alerted to the event by a source range reactor tt % After the tsactor trip, there are more than 15 minutes for operator action prior to loss of shutdown margin.
During full power operation with the rear. tor in manual control, an OTAT reactor trip alerts the operator to an uncontrolled dilution. At least 15 minutes are available after the reactor trip for operator action before the loss of shutdown margin.
During full power operation with the reactor in automatic control, the rod insertion limit alarms alert the operator to an uncontrolled dilution. At least 15 minutes are available after the low-low rod insertion limit alarm for operator action before the loss of shutdown margin, 15.4.6.4 Conclusions The results presented above demonstrate that adequate time is available for the operator to manually terminate.the source of dilution flow. Following O V. termination of the dilution flow, the operator can initiate boration to establish adequate shutdown margin, i 15.4.7 Inadvertent loadina and Ooeration of a Fuel Assembiv in an Imorocer Position Fuel type does not affect the ability of the in-core instrumentation to detect the inadvertent loading and operation of a fuel assembly in an improper position; therefore, the conclusions of the FSAR remain valid, 1 15.4.8 Soectrum of Rod Cluster Control Assembiv E.iection Accidents l 15.4.8.1 Introductiqn This accident is a mechanical failure of a control rod mechanism pressure housing resulting in the ejection of an RCCA and drive shaft. The consequence L of this mechanical failure is a rapid positive reactivity insertion together l with an adverse core power distribution, possibly leading to localized fuel rod damage. FSAR Section 15.4.8 further discusses this accident. The limiting criteria are as follows: o Average fuel pellet enthalpy at hot spot below 200 cal /g for unirradiated and irradiated fuel l \ Peak reactor coolant pressure less than that which could cause stresses to 4
\ o exceed the faulted condition stress limits A-15.4-16 I
o fuel melting less than 10 percent of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits of the first (~T criterion listed above Thi is an ANS Condition IV incident. 15.4.8.2 Method of Analysis The performance of the calculation of the RCCA ejection transient is in two stages: first an average core channel calculation and then a hot region calculation. The average core calculation uses spatial neutron-kinetics methods to determine the average power generation as a function of time including the various total core feedback effects, i.e., Ocppler reactivity and moderator reactivity. Enthalpy and temperature transients at the hot spot are then determined by multiplying the average core energy generation by the hot channel factor and performing a fuel rod transient heat transfer calculation. The power distribution calculated without feedback is conservatively assumed to persist throughout the transient. Reference 6 provides a detailed discussion of the method of analysis. Averaae Con Analysis The average core transient analysis uses the spatial kinetics computer code, TWINKLE (Reference 1). This code solves the two group neutron diffusion theory kinetic equation in one, two or three spatial dimensions (rectangular coordinates) for six delayed neutron groups and up to 2000 spatial points. The computer code includes a detailed multi-region, transient fuel-clad-coolant heat transfer model for calculation of pointwise Doppler and moderator feeoback (7) effects. This analysis uses the code as'a one-dimensional axial kinetics code
\ since it allows a more realistic representation of the spatial effects of axial moderator feedback and RCCA movement. However, since the radial dimension is missing, it is still necessary to emoloy very conservative methods (described below) of calculating the ejected rod worth and hot channel factor. Further description of TWINKLE appears in Sedion 15.0.11.4.
Hot Soot Analysis in the hot spot analysis, the initial heat flux is equal to the nominal value times the design hot channel f actor. During the transient, the heat flux hot channel f actor is linearly increased to the transient value in 0.1 second, the time for full ejection of the rod. Therefore, the assumption is made that the hot spots before and after ejection are coincident. This is conservative since the peak after ejection will occur in or adjacent to the assembly with the ejected rod, and before ejection the power in this region will be depressed due to the inserted rod. The hot spot analysis uses the detailed fuel and cla; instent heat transfer computer code, FACTRAN (Reference 2). This co...suter code calculates the transient temperature distribution in a cross section of a metal clad UO2 fuel rod, and the heat flux at the surf ace uf the rod, using as input the nuclear power versus time and local coolant conditions. The zirconium-water reaction is explicitly represented, and all material properties are represented as a function of temperatures. The code uses a conservative radial power distribution within the fuel rod. U FACTRAN uses the Dittus-Boelter or Jte-Lottes correlation to determine the film heat transfer before DNB, end the Bishop-Sandberg-Tong i correlation (Reference 8) to dE! amine the film boiling coef ficient af ter A-15.4-17
l ONB. The Bishop-Sandberg iong correlation is conservatively used assuming-zero bulk fluid quality. The code does not calculate the DNBR, instead specifying a conservative DNS heat flux which forces the code into DN8.
- (("m The code can calculate the gap heat transfer coefficient; however, it is *~
ad, lusted to force-the full power, steady-state temperature distribution to agree with fuel heat transfer design codes. Further description-of FACTRAN appears in-Section 15.0.11.1. System Overoressure Analysis. Because the transient does not exceed the safety limits for fuel damage specified earlier, there is little : likelihood of fuel dispersal into the coolant; therefore, the basis of the pressure surge calculation may be conventional heat transfer from the fuel and prompt heat generation in the coolant. The pressure surge .4 calculated by first performing the' fuel heat transfer calculation to determine the average and hot spot heat flux versus time. A THINC calculation uses this heat fkx data to determine
'the volume surge. Finally, the LOFTlAN computer code (Reference 3) simulates the volume ~ surge. This code calculates the pressure transient taking into account fluid transport in the RCS and heat transfer to the steam generators. . No credit is taken for the possible pressure reduction caused by the assumed failure of the control rod pressure housing.
15.4.8.3 Inout Parameters . Input parameters for the analysis are conservatively selected on the basis l - (' of values calculate'd for this type of core. The discussion of the-more important parameters is' presented below. Table 15.4.8-1 presents the I
\
parameters used in this analysis,
1. Ejected P.ed Worthf and Hot Channel Factors The values for ejected rod worths and hot channel- factors are calculated using either three-dimensional static methods- or a synthemis method employing one-dimensional and two-dimensional calculations. The analysis uses m ndard nuclear design codes. No credit is taken'for the flux flatten 499 effects oi reactivity l ' feedback. The calculation is performed for the maximum allowed bank insertion at a given power level, .as determined by the rod insertion limits. The analysis assumes adverse xenon distributions to provide worst case results.
The ejected rod worth and hot channel factors include appropriate margins-to account for any calculational uncertainties, including an allowance for nuclear power peaking due to densification. Ouring initial plant startup physics testing, the measurement of ejected rod worths and p>wer distributions is performed in the zero
-and part power. configuration and compared to the values used in the analysis. Experience has shown that the ejected rod worth and power peaking _ factors are consistently overpredicted in the analysis.
O . A-15,4-18
~_ .- - -- . - . - . - - . - ~ .- - - -- - -- -.
2.. Reaciivity' Fee'dback Weinhting Factors The largest. temperature increases, and hence, the largest reactivity
.( feedbacks,' occur in channels where the power.is higher than average.
Since the weight of a region.is dependent on flux, these regions have high weights. This means tt 2t the reactivity feedback is larger than-tnat indicated by a simple cnannel. analysis. Physics calculations have been performed for temperature changes with a flat temperature distribution and with a large number of axial and radial temperature distributions. The analysis compares reactivity changes and determines effective mighting factors. These wei hting factors take the form of multipliers which, when applied to sin le channel feedbacks, correct them to effective whole-core feedbacks for the appropriate flux shape. Axial weighting is not necessary because this analysis employs a .one-dimensional (axial) spatial kinetics method (i.e., the initial condition is made to match the ejected rod configuration). In addition, this analysis does not apply any weighting to the moderator feedback. The analysis applies a conservative radial weighting factor to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension. These weighting factors have also been shown to be conservative compared to three-dimensional analysis (Reference 6).
3. Moderator and Doppler Coefficient The nuclear code adjusts the critical boron concentrations at the beginning of life and end of life in order to obtain moderator density coefficient curves which ara ' conservative when compared to p
the actual design conditions %r the plant.- As discussed above, Q these results do not have any weighting factor applied to them. The resulting moderator temperature coefficient is at 1 east +7 pcm/*F at the appropriate zero or full power nom:aal average temperature for the beginning of life cases. The calculation determines the Doppler reactivity defect as a function of power level using a one-dimensional steady-state computer code;with a Doppler weighting factor of 1.0. The Doppler weighting factor will increase uncar accident conditions, as discussed above. L L 4. Delayed Neutron Fraction, Seff To allow for future cycles, the analysis used-conservative Seff estimates of 0.54 percent at beginning of life hot zero power; 0.57 , percent.at beginning of life hot full power; and 0.46 percent for both end of life cases. 1 l -5. Tr.ip. Reactivity Insertion The trip reactivity insertion assumed is given in Table 15.4.8-1 and includes the effect of one stuck RCCA adjacent to the ejected rod. L The reactivity of the ejected rod reduces these values. Dropping a l' rod of the required worth into the core simulates the shutdown reactivity. The start of rod motion occurred 0.5 seconds after reaching the high neutron flux trip point. It is assumed that Msertion to dashpot does not occur until 3.3 seconds after the rods G beg h to f all . The choice of such a conservative insertion rate (G A-15.4-19
1 J$ means that there is over 1 second after reaching the trip point . before significant shutdown reactivity is inserted into the core. l This is a particularly -important conservatism for hot full power l
' [, . accidents. ~ .The minimum design shutdown margin available for this plant at hot zero power (HZP) may'only occur at end of life in.the equilibrium ;
cycle. This value includes an allowance for the worst stuck rod, an-adverse xenon distribution, conservative Doppler and moderator ; defects,- and an allowance for calculational- uncertainties. Physics a calculations'have shown that two stuck RCCAs (one of which is the worst ejected rod) reduce the :'hutdown by= about an additional l
'1% Ak/k. Therefore, followir.g a reactor trip resulting from an i RCCA ejection accident, the reactor will be subcritical when the core returns to HZP.
6. Reactor Protection High neutron flux trip (high and low setting) and high positive rate of neutron flux increase trip, although the analysis only modeled the high neutron flax trip (high and low setting), provide reactor.
protection for a rod ejection. These protection functions are part of the reactor-trip-system. No single failure of the reactor trip system will negate the protection functions required 'or the rod ejection accident, or adversely affect the consequences of the i accident. , No single active ' failure in- any plant systems or equipment will adversely affect the-consequeni:es of the accident. t t L\ 15.4.8.4 Results f Table 15.4.8-lJsummarizes the results. The results of the analysis
'present cases for.both beginning and end of life at zero and full power.
A. Beginning of- Life, Full Power This case assumed Control Bank D at its insertion limit. The worst-
. ejected rod worth and hot channel factor were conservatively calculated to be 0.24% Ak/k and 5.5, respectively. The peak hot spot fuel centerline temperature reached melting, conservatively
. ' assumed at 4900'F; however, fuel melting was well below the-limiting criteriori of 10 percent of the pellet volume at the hot spot. B. Beginning of Life, Zero Power
-For this condition, the analysis assumed Control Bank 0 to be fully ' inserted with Banks B and C at their insertion limits. The worst ejected rod:is in Control Bank 0 and has a worth of 0.75% Ak/k and a hot channel factor of 11.0. The fuel centerline temperature was 3985'F.
C. End of Life, Full Power The analysis assumed Control Bcnk 0 at its insertion limit. The
'O ejected rod worth and hot char.nel factors were conservatively A-15.4-20
calculated to be 0.25% Ak/k and 6.0 respectively. The peak hot spot fuel centerline temperature reached rnelting, conservatively assumed at 4800*F;'however, fuel meltingi a well below the limiting criterion of 10 percent of the pellet volume at the hot 4 A. spot. D .- End of Life, .Zero Power , The analysis of this case determines the ejected rod worth and hot channel factor by' assuming Control Bank D to be fully inserted with Banks B and C at.their insertion limits. The results were 0.84% Ak/k and 26.0, respectively. The fuel centerline temperature was 3891*F. The Doppler weighting factor for this case is significantly higher than for the other cases due to the very large transient hot channel factor. For all the cases analyzed, average fuel pellet enthalpy at the hot spot ' remained belu 200 cal /g. Table 15.4.1-1 presents the calculated sequence of events for the worst , case rod ejection. accidents. Figures 15.4.8-1 through 15.4.8-4 show the i worst case rod ejection accident results. For all cases, rod insertion occurs after the nuclear power excursion is terminated by Doppler feedback. As discussed previously, the reactor remains subcriticci following reactor trip. Fission Product Release The analysis conservatively assumes that the gaps of all rocs entering DNB ' release fission products. In all cases considered., less than 10 percent ( of the rods entered ONB based on a detailed three-dimensional THINC analysis (Reference 6). Although the analysis predicts-limited fuel melting _ at the hot spot for the full-power cases, in practice melting is not likely since the analysis conservatively assumed that the hot spots before and after ejection were coincident. Pressure Surae A detailed calculation of the pressure surge for an ejection rod worth of
-1 dollar at .beginning of life, hot full power,' indicates that the peak pressure does not exceed that which would cause reactor pressure vessel stress to exceed the faulted condition stress limits (Reference 11).
Since the severity of the present analysis does not exceed the worst case analysis, the accident for this plant will not result in an excessive pressure-rise or further adverse effects to the RCS. Lattice Deformations A large temperature gradient exists in -the region of the hot spot. Since the fuel rods are free to move-in the vertical direction, differential
- expansion between separate rods cannot produce distortion. However, the temperature = gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rod toward the hotter side of the rod.. Calculations indicate that this bowing results in a negative ~
reactivity effect at the hot spot since Westinghouse cores are undermoderated, and bowing tends to increase the undermoderation at the
' hot spot, in practice, significant bowing is not expected since the O' structural rigidity of the core is more than sufficient to withstand the A-15.4-21 L _ - _ . .
c. . _ . _ . . - __ _ .. .-
forces' produced. Boiling in the hot . spot region produces a net flow away from.that region; however, the fuel releases heat to the water slowly, and
-it :is considered inconceivable that cross flow is sufficient to produce significant lattice forces. Even if massive and rapid boiling sufficient
to distort the lattice is hypothetically postulated, the large void
- fraction in the hot spot region produces a reduction in the total core moderator-to fuel ratio and a large reduction in this ratio at the hot spot. Therefore, the net effect is negative. feedback which leads to the - conclusion that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation. In fact, a small negative . feedback may result. The effect is conservatively ignored in the analysis.
Radiolooical Consecuences Appendix C contains the evaluation of the radiological consequences of a m postulated control rod ejec' ion accident. 15.4.8.5 Conclusions The results of the analyses do not exceed the described fuel limits; therefore, there is no likelihood of sudden fuel dispersal into the coolant. Since the' peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, there is no likelihood of further consequence to the RCS. The analyses have demonstrated that the fuc1 rods entering ONB are less than 30 percent of the fuel rods in the core; d.:9 fore, the Msumption gf 10 percent of the fuel rods in the. core entering DNB for the fission product release .
'W. calculation is conservative. The conclusions presented in the FSAR remain Q. valid.
15.4.9 Steamline Break With Coincidental Rod Withdrawal at Power 15.4.9.1 Introduction The coincidental and consequential occurrence of an uncontrolled RCCA bank withdrawal at power following a steamline break event is one of four potential interaction scenarios resulting from adverse environmental conditions (either inside or outside of containment) following a high energy line break; these scenarios are identified.in "!E Information Notice.79-22."' The premise of this concern is.that~during a high energy line break (such as steamline rupture), certain sensors used in the control systems could be exposed to an adverse environment. If the equipment is not qualified for the adverse environment, a control system
. malfunction might occur.
The automatic rod-control system is one of the control systems that could malfunction. The rod control system relies on the measurement of T a nuclear . power, and turbine impulse pressure to determine if control Nd' motion is required. A small steamline rupture may occur outside containment near the turbine impulse pressure transmitters or inside containrant in the vicinity of the excore detectors, thus exposing equipme..t used in the rod control system to an adverse environment. If p this equipment is not properly qualified for -these conditions, a d consequential RCCA withdrawal following a stt:amline rupture may occur. A-15.4-22 1
- - , ~ . .- - - , . . - . - - -. . - . ,
The steamline' break affects the rod control system (via either an inside containment break near the excore detectors or an outside containment break near the-turbine impulse transmitters) and causes the control rods to withdraw following the initiation of the transient. This causes an
.[V) increase in reactor power and core heat flux to the point at which an OPAT:or OTAT trip setpoint is reached. This trip terminates the most adverse part of the transient. .The steamline break causes-increased heat' removal and subsequent decrease in primary pressure simultaneous with 1 the increase in reactor power. Secondary pressure also decreases until the low steamline pressure setpoint is reached, initiating steamline and feedwater isolation.-
Because of the lower RCS pressure coincident with the' increase in reactor power, the consequences at the point of peak heat flux may be more adverse than the RCCA bank withdrawal at power transient analyzed in the FSAR. The most' limiting part of this transient-pertinent to this study is. immediately before reactor trip (i.e., rod motion). The most limiting case is that for the largest steamline break that trips on OPAT prior i
to reaching a reactor trip on a safety injection signal (e.g., low steamline pressure). 'Therefore, the analysis assumes the largest steamline break size for which a low steamline pressure signal will not occur prior to the OPAT reactor trip, and the analysis terminates 5 seconds after reactor trip. " Steam System Piping Failure" presented in FSAR Section 15.1.5 bounds the return to power following reactor trip and steamline isolation. If the low steamline pressure setpoint is reached, a reactor trip on safety injection actuation would result and terminate the event. Therefore, like the analysis performed for " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power" (Section 15.4.2), to m demonstrate protection by the AT trips, only- the applicable range of these trips needs to be considered. Also note that no credit is taken in I:- [U) --
L the steamline break with coincident rod withdrawal at power analysis for a reactor trip via the high neutron flux overpower protection signal, since this trip function may be inoperable due to adverse environmental conditions associated with'a steamline break inside containment. The performance of the analysis for a steamline break with coincident withdrawal of the control rods due to an adverse environment demonstrates that the corresponding minimum DNBR does not decrease below the appropriate safety analysis limit ONBR value and no fuel or clad damage occurs. Additionally, no system _overpressurization is expected since the steamline break results in a RCS depressurization as described above. This is an ANS Condition Ill/IV incident. 15.4.9.2 Method of Analysis l' The analysis of this transient uses the LOFTRAN corguter code l
(Reference 3). The following assumptions were made for this transient

A. ,The' analysis employs RTDP methodology in dete mining initial conditions of maximum core power, reactor cooiant average temperature, and minimum reactor coolant pressure. 1 L B. End of life shutdown margin and equilibrium xenon conditions. The ) (qj analysis assumes the most reactive RCCA stuck in its fully withdrawn position for conditions following reactor trip. A-15.4-23
C. The analysis uses a negative moderator coefficient corresponding to the end of life unrodded core. This maximizes the reactivity insertion caused by the cooldown during the steamline break. D. The analysis assumes the reactor trip setpoints on OPAT and i OTAT at a conservative value. The AT trips include all adverse instrumentation and setpoint errors; the delays for trip actuation are et the maximum values. E. The analysis bases the RCCA trip insertion characteristic on the assumption that the highest worth assembly is stuck in its fully withdrawn position. F. The break size assumed for this transient is 2.24 ft2 (0.560 ft2 per sttam generator). This is the largest break size for which a low steamline pressure signal will not occur before the reactor trip on OPAT. Before the eventual steamline isolation on low steamline pressure, all four steam generators feed this break. Following steamline isolation, one steam generator feeds the break causing an asymmetric transient. G. The calculation of the steam flow during a steamline break uses the Moody Curve for fL/0 0. H. The reactivity insertion rate is 900 pcm/ min. No single active failure in any plant systems or equipment will adversely affect the.censequences of the accident. 15.4.9.3 Results The calculated sequence of events for this transient is shown in Table 15.4.1-1. Figures 15.4.9-1 and 15.4.9-2 show RCS transient and core heat flux following the steamline rupture with coincident RCCA bank withdrawal. The steamline break affects the turbir.e impulse transmitters and causes the control rods to withdraw at the init1& tion of the transient. This causes an increase in reactor Nwer and core hiist flux to the point at which the OPAT trip setpoint is reached. The reactor trip terminates the most adverse part of the transient. The steamine break causes increased heat removal and subsequent decrease in prieary pressure simultaneous with the increase in reactor power. If the transient extends beyond post-reactor trip, secondary pressure will decrease until the low steamline pressure setpoint is reached, initiating stemline and feedwater isolation. The analysis of the steamline break with coincident RCCA bank withd awal demonstrates that the DNBR limit is met. The most limiting part of this transient pertinent to this study was immediately before reactor trip (i.e., rod motion). The transient for the steamline break presented in FSAR Section 15.1.5 bounds the return to power following reactor trip and steamline isolation. The other FSAR steamline break analysis assumed a larger break size and initial conditions corresponding to no-load temperatures (i.e., less stored energy in the RCS and reactor fuel). l A-15.4-24
The DNBR is always greater than the limit value. Figure 15.4.9-3 shows the'0NBR as a function of time for this transient. ( 15,4.9.4 Conclusions The analysis demonstrates that the DNBR does not decrease below the limit value and no fuel or clad damage occurs. Additionally, no system overpressurization will occur, thus all applicable safety criteria are met. As stated in the results, the large steamline break analysis presented in FSAR Section 15.1.5 bounds the return to power following a reactor trip and steamline isolation; therefore, there is adequate protection to ensure plant safety for this transient.
$J O
v A-15.4-25
' Table 15.4.1-1 (Sheet 1 of 31-Time'Seauence of Events for incidents =Which Result in 'O Reactivity ;nd Pow,er Distribution Anomalies-Time Delays (s)
Accident Event Uncontroll'ed RCCA bank withdrawal-from a subcritical or low-power startup condition initiation of uncon withdrawal from 10 grolled rod-fraction of nominal power 0.0 Power range high neutron flux low setpoint reached 12.5 Peak nuclear power occurs 12.7 Rods begin to fall into-core 13.0 Minimum DNBR occurs- 14.5 Peak heat flux occurs '14.9' 7 Peak average clad temperature occurs 15.1 Peak. average fuel temperature O occurs 15.4
. Uncontrolled RCCA bank with-drawal at power (full power ~with minimum feedback)
l. Case A Initiation of uncontrolled RCCA withdrawal .at a high-reactivity
-insertion rate (80 pcm/sec) O Power range high neutron flux-high-setpoint reached 1.4 Rods begin to fall into core 1.9 Minimum DNBR occurs 2.9
2. Case'B Initiation of uncontrolled RCCA withdrawal at a small reactivity -
insertion' rate (3 pcm/sec) 0 OTAT setpoint reached 34.0 Rods begin to fall into core 36.0 Minimum DNBR o(. curs 37.1 A-15.4-26
i m Table 15.4.1-1 (Sheet 2 of 3)
.. Time Secuence of Events for Incidents Which Result in l . Reactivity and Power Distribution Anomalies >
Time Delays Accident Event (si-Startup of an inactive reactor coolant' pump at an incorrect
-temperature Initiation of pump startup 0.0 Power reaches P-8 trip setpoint 3.5 i Rods begin to drop 4.0 Minimum DNBR occurs 5.0 CVC'S malfunct'.on that results '-
in a'decreasre in the boron
'concentratirn'in the_ reactor coolant-
1. Diluttor during startup' Power range - low setpoint-reactor trip due to dilution 0
Shutdown margin' lost-(if dilution- ; continues'after trip) . 2010
2. Dilution:during full-power operation a.-Automatic reactor control Operation receives low-low rod insertion limit alarm due to dilution 0 1
Shutdown margin lost 3700 b Manual. reactor' . Reactor trip on OTAT due to dilution 0 Shutdown margin .is' lost (if dilution continues.after trip)' 1860 l l < l l O A-15.4-27
Table 15.4.1-1 (Sheet 3 of 3) Time Secuence of Events for Incidents Which Result in Reactivity and Power Distribution Anomalies Time Delays Accident _ Event (s) RCCA ejection accident
1. End of life, zero power Initiation of rod ejection 0.0 Power range high neutron flux low setpoint reached 0.22 Peak nuclear power occurs 0.24 Rods begin to fall into core 0.72 Peak clad average temperature occurs 1.79 Peak heat flux occurs 1,79 Peak fuel average temperature occurs 1.98

2. Beginning of life, full power Initiation of rod ejection 0.0 Power range high neutron flux O high setpoint reached Peak nuclear power occurs 0.05 0.13 Rods begin to f all into core 0.55 Peak fuel average temperature occurs 2.44 Peak clad average temperature occurs 2.52 Peak heat flux occurs 2.53 Steamline break with coincidental rod withdrawal at power Steamline ruptures, RCCA bank begins to withdraw 0.0 OPAT reactor trip setpoint reached 8.8 Rods begin to fall 10.8 Minimun. DNBR occurs 11.6 0 .
A-15.4-28
Table 15.4.8-1
~
Parameters Used in the Analysis of the Ro Cluster 3 Control Assembly E.iection Accident LJ HZP HFP HZP HFP Time in life Beainnina Beainnina M M Power level (%) 0 102 0 102 Ejected rod worth (% AK) 0.75 0.24 0.84 0.25 Delayed neutron fractio,i (%) 0.54 0.57 0.46 0.46 Doppler feedback reactivity weighting 1.744 1.30 3.55 1.30 Trip reactivity (% AK) 2.0 4.0 2.0 4.0 Fg before rod ejection 2.55 2.55 Fg after rc/ ejection 11.0 5.5 26.0 6.0 Number of operational pumps 2 4 2 4 Maximum' fuel pellet average temperature at the hot spot (*F) 3425 4091 3412 3901 Maximum fuel center temperature at the hot spot (*F) 3985 >4900 3891 >4800 Maximum fuel stored energy at the hot spot (cal /g) 144.9 179.2 144.2 169.2 Percent of fuel mel; at the hot spot 0 <10 0 <10 l A-15.4-29
O : I 101, n 5 100 5g a E E t-- O~ sC
=g 3o-1. .
b 10-2 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 TIME (s) vooTLE NEUTRON FLUX TRAN51ENT FOR UNCONTROLLED R00 i ELECTRIC QEN$ RATING PLANT WITHDRAWAL FROM A SUBCRITICAL CON 0! TION J GeorgiaPower uairi a=o uair FIGURE 15.4.1-1 A-15.4-30
,r~ .
i.
.9 .8 -
2 g .7 - 5 'O dwC
- . .5 ~
c%)
== 2 %s N cm 6
_ .3 -
.2 - .1 0.
0 2 4 6 8 10 -12 14- 16 18 20 22 24 26 28 30 TIME (s) V007Ls THERMAL FLUX TRANSIENT FOR UNCONTROLLED R00 m GeorgiaPowerkh I l ELECTn6C GENE RAT 1=o PLA.1T WITHDRAWAL FROM A SUSCRITICAL CON 0! TION [v) unit i 4=o unir FIGURE 15.4.1-2 A-15.4-31
1h.
% 800 C
750 w
$ 700 3
w I w 650
$ 600 2 ^ J $ $$0 $00 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30
} p 2000. l 1800. C.
~
1600. E E 1400. 5 w I '1200. w - d g 1000. W g 800. l 2 l 600. J 400. 2, 6. 8. 10. 12, 14, 16, 18. 20. 22. 24 26. 28, 30. G. 4 TIME (s)
t. V087'8 FUEL AND CLAD TEMPERATURE FOR UNCONTR01 LED R00 I s f' ELac,mic caNERA11NG Pt. ANT WITHDRAWAL FROM A $UbCRITICAL CON 0! TION COgla Wer uN,1, ANo oN.1, FIGURE 15.4.1-3 l A-15.4-32
g4,y b -s %_- A- - , , , - ,_A--,en---e a kut- --.4% n as h11 &MR e.'4 H-b ai a a 4 m- a A i- ---e k , a s iM4A&&e k- $
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$U .6 "$.~
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g , f E 3 4 5 6 7 l TIME (s) : O' E!$$l!}$7!yEEy"NEj,"lALFROMry((peygg K (80 PCM/s)
***'"'"*7'"o esaur WITH0hl MTE)
Ccorgia power UNIT 1 ANG UM87 2 b FIGURE 15.4.2.} A-15.4-33
.- . . . . . . . -. ~ ..... . . - .-._.. - - -. .- - - -. . - . , . - _ . - . - - - _ ..
4 t 2600. . \,_/
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5
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6. 7. 8. 9. 10.

0. 1. 2. 3. 4. 5. .
TIME (s) V0GTLg UNCONTROLLED RCCA BANK WITHORAWAL FRCH FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (80 PCN/$ b N. .Geo3la Power E LECTRN: GENERATINO PLANT unir i aNo unit : wrmavat wo FIGURE 15.4.2-2 A-15.4-34
620 610 C
, 600<
gg 590 IE
$4 ww 580 g 570 >
l 560 , 550 - - f k 4 v , 3.5 , 3. 2.5 i
2. ,
I 1.5 l l 1. 0 1 2 3 4 5 6- 7 8 9 to l TIMC (s) i voGTLE UNCONTROLLED RCCA BANK WITHORAW#.L FRCH FULL POWER WITH MINIMUM REACTIVITY FEEDBACK ELECTalC OENE R ATINC PLANT .
\ (so PCx/s WitwoaAWAL m E) l- CCOrgiaPower u.if i u. T .
FIGURE 15.4.2 3 A-15.4-35
h i . ,D i i ' \v)_ ' 1.4 1.2 A
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.2 0. '0' 5 10 15 20 25 30 35 40 TIME (s)
UNCONTROLLED RCCA BANK WITHORAWAL FROM FULL, POWER
/* voo7Ls ELEC1MH:0ES4 R Af1NG PLANT WITH MINIMUM REACTIVITY FEED 8ACK (3 PCM/S WITH0RAWAL RATE)
GeorgiaPower UMY 1 AN0 umT 2 FIGURE 15.4.2 4 A-15.4-36
~~~ _
O 2600, i i 2500. g 2400. k ' g 2300. O 2200.
=.
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5e y 2000, a. 1900. 1800. O 1400. ig 1300. 1 b 1200. g 1100, M y 1000.
I :
5 900*
t 5
G 300. u a. 700. 600. _ _ _
' . 10. 15. 20, 25. 30. 3$' 'O' TIME (s) unconinottto acta aANK wtTuonAwAt racM rutt pewtR vocTsa " l "PC O (CC@l3 kWCf ELECTRIC CENE MAftNG PLANT UNt? 1 AND UNtf 2 FIGURE 15.4.2 5 A-15.4 37
. .- + >-+f - J.W.,a 4 - 4A -.WJ.u.=4.. . .++ 4-, .ACh 4 5 4 ..J.-ir1w-e '
f -
's '620 610 C
., 600 w
B 5g 590 SE y# uw 580 b
-g 570 l 560 550 .
5 I 4 V 3.5
3. t l
l-2.5 2, 1.5 1. 10 15 20 25 30 35 40 0 5 TIME (s) vocTLE UNCONTROLLED RCCA BANK WITHORAWAL FROM FULL POWER O CeOgiaPower ELsCTRIC GENERATING PLANT unit i A=o u=iT : WITH MINIMUM REACTIVITY FEEDBACK (3 PCM/S vimonAWAt RATE: i FIGURE 15.4.2-6 A-15.4-38
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N - - . 18NO WOWINIW VOGTLg I ' MINIMUM ONBR VS. REACTIVITY INSERTION RATE FOR R00 ELSCTRIC QENERAftNG PLANT WITH0RAWAL FROM 100% POWER GCOgia ber UNIT 1 ANO UNIT 2 FIGURE 15.4.2-7 A-15.4-39
p g n d V U O SG SAFETY VALVE OPENS 8 PRIOR TO TRIP 6a 24 iir r' l il '. i: i zij .
* ~ ~ ir: .;I:l.!l..n p a,. '.
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til Flux TRIP
$ $ i ,! !!! lill l:
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- - MINIMUM FEEDBACK - -
l i. ; ip'. IL
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t O O- O d O SG SAFEIY VALVE OPENS I i b .. PRIOR TO 1 RIP Si .! '- "i { f !! ' ll l . j !
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l. '
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OTAT TRIP l 3 g i ~5 5 % I.;},]
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._g i j gj MINIMUM FEED 8ACK .p i i y l . MAXIMIM IEEDBACK l l ; : !
A 96 2.0' i 'Although the 56 valves are open ! ',l *]l l' !! 'l' l
!!' l' llllIf"l ,
i
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0 I 2 10'I 10 10 10 g REACTIVITY INSERIl0N RATE (pcm/sec) 5
- - - ----- -- ---- J
O 1.2 g g j g 600 g j j w _ 1.1 s 5
- 580 V- $2 1.0 0.9 is -
s W@0.8
$ 540 g
8*N0.7 2 - y [ $ 520 0.6 - y w 0.5
$ 500 1.2 2400 l
g 11 - - p2300 t O. $ x2 1 ,0 - y -
~- ~
3g a 2200
*2 0 .9 8 w $$ $2100 -
[ $ 0.8 5 m tf 2000 - O N 0.7 - 5
= a ~
0.6 - $ a. l l l l I l l 0.5 1800 0 50 100 150 200 0 50 100 150 200 TIM E (SECONOS' TIME (SECONDS) NUC! EAR POWER TRANSIENT vocTLE AND CORE HEAT FLUX TRANSIENT
'O Georgia Power E',1" ?fJ 'u"'v"* " "' FOR DROPPED RCCA FIGURE 15.4.3-1 -~
433.. A-15.4-42
. .. _. -.. . _ _ , - . - . . . _ . . . - . .. - . = . . = . _ . = . _ . . . _ . - . - -
I O 1.2 1.0 < 8 y .8 . 5
0. 6 < -
W f 5 0. 4 - y l.
$ 0 .2 -
0 2 4 6 4 10 12 14 16 18 20 22 24 26 28 30 TIME (SEC) L L hnic osNenArmo etANT!MPROPER STARTUP OF AN INAC'!VE REACTOR COOLANT PUMP I (> GeorgiaPower d u mri asumt:
- FIGURE 15.4.4-1 l
A-15.4-43 l'
1
^
l l b
'w/
1.2 , g 1.0- - I Q.8 . b l o.6 ! y 0.2 - 0 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 TIME (SEC)
' IMPROPER STARTUP OF AN INACTIVE REACTOR COOLANT PUMP A a$cinic cassRATING Pt. ANT O GeorgiaPower ma unir i *=o uair FIGURE +5.4.4-2 A-15.4-44
7%
\.
2.0 C.
< 1. 8. '!
m 1.6
$ 1.4 w
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\
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0. 8 l C 0.6
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8
. 4 l
u C. 2 i L o 0 2 4 6 8 10 12 14 16. 18 20 22' 24 26- 28 30 i TIME -(SEC) i 1 VOCTLE A TlVE REACTOR CC0LANT pyMp _i T Ano T 0tOQl8 VOWCf FIGURE 15.4.4 3 A-15.4-45
O 2600 2500 k 2400 W 300 E2200. / s U 2100 E G u 20c0 1900 1800 620 O -
,, 6 0 0 w
g 590 5 g 540 5
570 2 560 W
", 550 8
v 540 8 g 530 < C
s s20 4 6 8 to 12 14 16 16 20 22 24 26 7 30 0 2 TIME (s)
O RIC o tNE R ATING PLANT IMPROP(k sTARTVP OF AN lhACTIV[ R(ACTOR C00 TANT PUMP CCOgIC" rower unit i Aho vwit FIGURE 15.4.4 4 A-15.4-46
i O \ f l' 3.2
3. l 2.8 2.6 ! !
i 2.4 1 t l' ! 2.2 2. 1.8 1.6 ! 1.4 j 1.2 0 2 4 6 8 10 12 14 16 18 20 22 24 26 TIME (SED i I l l l.
' INPROPER $fARTUP OF AN thACTIVt RIACTOR COOLANT PWMP!
stactns onwanAtimo
TANT i Georgia Power u= 7 i *=o u=it :
FIGURE 15.4.4 5 A-15.4-47 , i
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15.$ Increase in Reactor Coolant inventorv [\ ( This section discusses the events which could result in an increase in the RCS inventory. 15.5.1 Inadvertent Ooeration of the Emeroency Core Coolino System Durina Power Operation 15.5.1.1 Introduction Operator error or a false electrical actuation signal can cause spurious ECCS operation at power. A spurious signal may originate from any of the safety injection (SI) actuation channels as described in FSAR Section 7.3. The suction of the coolant charging pumps diverts from the volume control tank to the refueling water storage tank following the actu& tion signal. The valves isolating the charging pumps from the injection header then automatically open. The charging pumps force the borated water from the RWST through the header, the injection line, and into the cold leg of each loop. The 51 pumps also start automatically, but they do not provide any flow when the RCS is at normal pressure. The passive accumulator tank safety injection system (SIS) and the low head system also do not provide any flow at normal RCS pressure. An $1 signal normally results in a reactor trip followed by a turbine trip; however, any single fault that actuates the SIS will not necessarily f- produce a reactor trip. If an $1 signal generates a reactor trip, the operator should determine if the signal is spurious, if the SI signal is ' (V) determined to be spurious, the operator should terminate $1 and maintain the plant in the hot standby condition as determined by appropriate recovery procedures, if repair of the ECCS actuation instrumentation is necessary, future plant operation will be in accordance with the Technical Specifications, if the reactor protection system does not produce an immediate trip as a result of the spurious SIS, the reactor experiences a negative reactivity excursion due to the injected boron, which causes a decrease in reactor power. The power mismatch causes a drop in T av and consequent coolant shrinktge. Thepressurizerpressureandwaterieveldecrease. Load decreases due to *,he effect of reduced steam pressure on load after the turbine throttle valve is fully open. These effects will lessen until the rods have moveo out of the core by using automatic rod control. The transient is eventually terminated by the reactor protection system low pressurizer pressure trip or by manual trip. Initial operating conditions affect the time to trip. The initial conditions include the core burnup history which affects initial boren concentration, rate of change of boron concentration, and Doppler and moderator coefficients. O j A-15.5-1
. _ . _ . ._ _ . . . _ , . _ _ ~ . _ . . _ . _ _ _ _ --
t
> r Recovery is made in the same tranner as described for the case where the $1 :
signal results directly in a reactor trip. The only difference is the !
' lower T and pressure associated with the power mismatch during the i
transieNg The time at which reactor trip occurs does not affect this transient. At lower loads, coolant contraction will be slower which will , result in a longer time to trip. ]
j. This is an ANS Condition It incident.
15.5.1.2 Method of Analysis , ' The analysis of the spurious operation of the ECCS employs the detailed digital computer program LOFTRAN (Reference 3). The code simulates the L neutron-kinetics, RCS, pressurizer, pressurizer relief and safety valves, > pressurizer spray, steam generator, steam generator safety valves, Md effect of the $15. The program computes pertinent plant variables including temperatures, pressures, and power level. Because of the power and temperature reduction-during the transient, . operating conditions do not approach the core limits. Previous analysis of several cases has shown that the results are independent of time to .r c
! trip, i
l The analysis presents a typical transient representing minimum reactivity r ' feedback. Results with maximum reactivity feedback indicate no. significant transient. The analysis assumes zero injection line purge " Volume for calculational simplicity; thus, the toration transient begin' immedi_ately when the appropriate valves open.
.The' assumptions are as follows: >
j. A. Initial Operating Conditions The analysis.of this transient employs RTOP methods. The analysis
- assumes nominal values consistent with' the steady-state, full power operation for the initial reactor power and RCS temperature. This results in the maximum power difference for the load loss. The analysis assumes-a nominal value consistent with steady-state, full power operation for initial RCS pressure, Tables 15.0.3-2 and 15.0.3-3 show the initial conditions assumed in the analysis. ,
B. Moderator and Doppler Coefficients of Reactivity . The analysis assumes a zero moderator temperature coefficient. The cooldown of the RCS would cause a faster decrease in reactivity and the reactor.would trip sooner with the use of a positive moderator temperature coefficient. The analysis assumes a low (absolute value) Doppler power coefficient. C. Reactor Control The reactor is-in manual rod control. D. Pressurizer Heaters Pressurizer heaters are inoperable. This assumption yields a higher rate of pressure decrease. A-15.5-2
l J E. Boron injection At time zero, two charging pumps inject 2600 ppm borated water intc the cold leg of each loop. t F. Turbine Load Turbine load is constant until the governor drives the throttle valve wide open, and then drops as steam pressure drops. G. Reactor Trip Lew pressurizer pressure initiates reactor trip.- No single active failure in any plant systems or equipment will adversely affect the consequences of the accident. 15.5.1.3 Results . Figures 15.5.1-1 through 15.5.1-3 show the transient response to inadvertent operations of the ECCS during power oparation. Neutron flux starts decreasing immediately due to boron injection, but steam flow does :
- not decrease untti later in the transient when the turbine throttle valve goes wide open. The mismatch between load and nuclear power causes i ,- pressurizer water level, and pressurizer pressure to drop. The T,y0tortripsandcontrolrodsstartmovingintothecorewhenthe rea pressurizer pressure reaches the pressurizer low pressure trip setpoint.
The DNBR increases throughout the transient. Table 15.5.1-1 shows the calculated sequence of events. Pressure and temperature slowly rise after the reactor trip, because the. turbine also trips and the reactor produces some power due to delayed neutron fissions and decay heat. Paragraph 15.5.1.1 discussed the recovery from this accident. 15.5.1.4 Conclusions . Results of.the analysis show that spurious ECCS operation without immediate reactor trip does not present any hazard to the integrity of the RCS. If_the reactor does not trip immediately, the low pressurizer pressure reactor trip actuates. This trips the turbine and prevents excess cooldown, which expedites recovery from the incident. The conclusions presented in the FSAR remain valid. 15.5.2 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory Section 15.4.6, " Chemical and Volume Control System Malfunction that Results in a Decrease'in Boron Concentration in the Reactor Coolant," analyzes an increase in reactor coolant inventory which results from the A-15.5-3 L - - . - - - _ _ - _ - . _ . - - - _ _ - - - . - - ..
I addition of cold, unborated water to the RCS. The preceding section, 1 15.5.1, analyzes an increase in reactor ctolant inventory which results
I from the injection of highly borated water into the RCS.
15.5.3 A Number of Boilino Water Reactor Transients This section is not applicable to VEGP Units 1 and 2. i e O A-'15.5-4
Table 15.5.1-1 O Time Secuence of Events for Incidents Which Result in an increase in Reactor Coolant Inventory Time Delays Accident Event (s) Inadvertent actuation of the ECCS during power operation Spurious SI signal generated; two centrifugal charging pumps begin injecting borated water 0.0 i Low pressurizer pressure reactor trip setpoint reached 57.1 Control rod motion begins 59.1 Turbine throttle valve wide open; load begins to drop with steam pressure 60.0 O O A-15.5-5
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15.6 Decrease in Reactor toolant Inventory 1
~This section discusses the event which could result in a decrease in the RCS inventory.
I 15.6.1 Inadvertent Ooenino of a Pressurizer Safety or Relief Valve 15.t.1.1 intr 3 duction An accidental depressurization of the RCS could occur as a result of an ' inadvertent opentng of a pressurizer relief or safety valve. Since a pressurizer safety valve is sized to relieve approximately twice the steam flowrate of a relief valve and will allow a much more rapid depressurization upon opening, the most severe core _ conditions resulting
from an accidental depressurization of the RCS are associated with an inadvertent opening of a pressurizer safety valve. Initially the event results in a rapidly decreasing RCS pressure. The effect of the pressure decrease is to increase power via the moderator density feedback. The average coolant _ temperature remains approximately the same, but the pressurizer level increases unt.11 reactor trip.
The reactor may be tripped by the following reactor protection system signals: o OTAT o Pressurizer low pressure , This is an ANS Condition 11 incident, , 15,6.1.2 . Method of Analysis The accidental depressurization transient is analyzed by using the detailed digital computer code LOFTRAN (Reference 3). This code simulates the neutron-kinetics, RCS, pressu.rizer, pressurizer relief and safety , valves, pressurizer spray, steam generator, and steam generator safety > valves. -The code computes pertinent plant variables including temperatures, pressures, and power level. In order to produce conservative results in calculating the DNBR during the transient, the following assumptions are made: , A. Nominal values are assumed for the initial reactor power, pressure, and RCS temperatures.- Uncertainties'in initial conditions are included in the limit DNBR as described in Reference 7 (see Tables 15.0.3-2 and 15.0.3-3). B. A most' positive moderator temperature coefficient of reactivity is assumed. The spatial _ effect of voids resulting from local or c subcooled boiling is not considered in the analysis with respect to j reactivity feedback or cora power shape. O C. A small (absolute value) Ooppler coefficient of reactivity, such that the resultant amount of negative feedback is conservatively low, maximizes any-power increase due to moderator feedback. A-15.6-1
Normal reactor control systems are not required to function. The reactor r protection system functions to trip the r3 actor on the appropriate ( signal. No single active failure will prevent the reactor protection system from functioning properly.. 15.6.1.3 Results The system responses to an inadvertent opening of a pressurizer safety valve are shown in Figures 15.6.1-1 and 15.6.1-2. Figure 15.6.1-1 illustrates the nuclear power transient following the depressurization. Nuclear power increases slowly until reactor trip occurs on OTAT. The pressure decay transient and average temperature transient following are given in Figure 15.6.1-2. Pressure drops more rapidly while core heat generation is reduced via the trip and then slows once saturation temperature is reached in the hot leg. The DNBR decreases initially, but increases rapidly following the trip as shown in Figure 15.6.1-1. The DNBR remains above the limit value throughout the transient. The calculated sequence of events is shown in Table 15.6.1-1. 15.6.1.4 Conclusions The results of the analysis show that the pressurizer low pressure and OTAT reactor protection system signals provide adequate protection against the RCS depressurization event. The conclusions presented in the FSAR remain valid. l O A-15.6-2
Table 15,6,1-1 Time Secuence of Events for Incidents Which Cause a d Decrease in Reactor Coolant .:nventory Time Delays Accident Event .. (s) Inadvertent opening of a pressurizer safety valve Pressurizer safety valve opens fully 0.0 OTAT reactor trip setpoint reached 24.5 Rods begin to drop 26,5 Minimum DNBR occurs 27,0 O O A-15.6-3 i,, ,, , . . - . . , , - . . - - . . , - . , . , . , , - . . . , , , , , ,_.
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gs 15.R References
1. D. H. Risher, Jr. and R. F. Barry, TWINKLE -- A Multi-Dimensional Neutron Kinetics Computer Code, WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Nonproprietary), January 1975.

2. H. G. Hargrove, FACTRAN -- A FORTRAN-IV Code for Thermal Transients in a 002 Fuel Rod, WCAP-7908-A, December 1989.

3. T. W. T. Burnett, et.al., LOFTRAN Code Descriotion, WCAP-7907-A, April 1984.

4. L. Baker and L. Just, Studies of Metal Water Reactions of Hiah Temoeratures. til Excerimental and Theoretical Studies of the Zirconium-Water Reaction, AHL-6548, Argonne National Laboratory, May 1962.

5. R. L. Hoessler, et.al., Methodoloav for the Analysis of the Drocoeo Rod Event, WCAP-ll394-P-A (Proprietary) and WCAP-ll395-A (Nonproprietary), January 1990.
' p 6. D. H. Risher, Jr., an Evaluation of the Rod E.iection Accident of Westinohouse Pressurized Water Reactors Usina Soatial Kinetics Methods, WCAP-7588, Revision I-A, January 1975.
7. A. J. Friedland and S. Ray, Revised Thermal Desian Procedure, WCAP-ll397-P-A, April 1989.

8. A. A. Bishop, R. O. Sandbarg, and L. S. Tong, Forced Convection Heat Transfer at Hiah Pressure After the Critical Heat Flux, ASME 65-HT-31, August 1965.

9. S. D. Hollingsworth and D. C. Wood, Reactor Core Resoonse to Excessive igeondary Steam Releases, WCAP "26-RI, January 1978.

10. Y. 5. Liu, A. Meliksetian, J. A. Rathkopf, D. C. Little, F. Nakano, and M. J. Poploski, ANC-A Westinabouse Advanced Nodal Comouter Code, WCAP-10965-P-A (Proprietary) and WCAP-10966-A (Nonproprietary),
December 1985.
11. T. G. Taxelius, ed. , Annual Reoort-Soert Pro 'act. October 1968.
Sectember 1969, Idaho Nuclear Corporation, IN-Id o. June 1970. A-15.R-1
O Accendix B LOCA and SGTR Accident Analyses for the Vogtle Electric Generating Plant Units 1 and 2 Transition to Westinghouse 17x17 VANTAGE-5 Fuel Assemblies O O
VEGP-FSAR TABLE OF CONTENTS PAGE 6.2.1.! Minimum Containment Pressure Analysis for Performance B-7 Capability Studies on Emergency Core Cooling System i 15.6.3 Steam Generator Tube Rupture (SGTR) B-21 15.6.5 Loss of Coolant Accidents B-71 O
\
l O B-1
VEGP-FSAR LIST OF TABLES 6.2.1-69 Reflood Mass and Energy Releases DECLG (Pages B B17) (C0 0.6) 6.2.1-70 Broken Loop Injection Spill During Blowdown DECLG (CD-0.6) 6.2.1-71 Active Heat Sinks for Minimum Containment Pressure Analysis 6.2.1-72 Passive Heat Sinks 6.2.2-2 Containment Fan Cooling Hest Removal Capacity (Post-LOCA Modes) 15.6.3-1 Operator Action Times For Design Basis (Pages B B-55) SGTR Analysis , 15.6.3-2 Sequence Of Events 15.6.3-3 Mass Releakes Results 15.6.3-4 Parameters Used in Evaluating Radiological Consequences 15.6.3-5 lodine Specific Activities in the Primary and Secondary Coolant 15.6.3-6 lodine Spike Appearance Rates 15.6.3-7 Noble Gas Specific Activities in the Reactor Coolant Based on W Fuel Defects 15.6.3-8 Atmospheric Dispersion Factors and Breathing Rates 15.6.3-9 Thyroid Dose Conversion Factors 15.6.3-10 Average Gamma and Beta Energy for Noble Gases 15.6.3-11 Offsite Radiation Doses 15.6.5-1 Input Parameters Used in the ECCS Analysis (Pages B B-91) 15.6.5-2 Large Sreak Results 15.6.5-3 Small Break Results o B-2
~
VEGP-75AR O l LIST OF FIGURES l 6.2.1-35 Containment Pressure DECLG (CD=0.6) (Pages B B-20) 6.2.1-36 Containment Temperature DECLG (CD=0.6) 6.2.1-35 Containment Wall Condensing Heat Transfer DECLG (CD=0.6) I 15;6.3-1 Steam Generator Tube Rupture Pressurizer (Pages B B-70) Level 15.6.3-2 Steam Generator Tube Rupture RC5 /ressure i 15.6.3 Steam Generator Tube Rupture Secondary Pressuie 15.6.3-4 Steam Generator Tube Rupture Intact Loop Thot and Tcnid l ! 15.6.3-5 Steam Generator Tube Rupture Ruptured Loop Thot and Tcold ! 15.6.3-6 Steam Generator Tube Rupture Differential Pressure q 15.6.3-7 Steam Generator Tube Rupture Primary To Secondary Break' Flow : 15.t 1-8 Steam Generator Tube Rupture Ruptured SG Water Volume 15.6.3-9 Steam' Generator Tube Rupture Ruptured SG Water Mass l 15.6.3-10 Steam Generator Tube Rupt'ure Ruptured SG Atmospheric Mass Release 15.6.3-11 Steam Generator Tube Rupture Intact SGs Atmospheric Mass Release 15.6.3-12 Steam Generator Tube Rupture lodine Transport Model , 15.6.3-13 Steam Generator Tube Rupture Break Flow Flashing Fraction 15.6.3-14 Steam Generator Tube Rupture'SG Water Level Above Top Of Tubes 15.6.3-15' Steam Generator Tube. Rupture lodine Scrubbing Efficiency l 15.6.5-1 Sequence of Events for large Break (Pages B B-163) L LOCA Analysis 15.6.5-2A _Small Break Safety Injection Flowrate 15.6.5-2B ECCS Pumped Safety Injection During 3-Inch Small Break 15.6.5-3 Smail Break Hot Rod Power Shape 15.6.5-4 Code Interface Description for Large Break Model 15.6.5-5 Code Interface Description for Small Break Model 15.6.5-6' Peak Clad Temperature DECLG (CD 0.4)- 15.6.5 -Peak Clad: Temperature DECLG (CD 0.6) 15.6.5-8 Peak Clad Temperature DECLG (CD 0.8) l lO B-3
. _- . . _ _ _ . - _ - . . _ . - - . _ . - ._ ~
VEGP-FSAR LIST OF FIGURES (continued) 15.6.5-9 Peak Clad Temperature DECLG (CD=0.6 Maximum Safety Injection) 15.6.5-10 Core Pressure DECLG (CD=0.4) 15.6.5-11 Core Wenure DECLG (C0 0.6) 15.6.5-12 Core Pressure DECLG (CD=0.8) 15.6.5-13 Reflood transient - Core and Downcomer Water Levels DECLG (CD=0.4) 15.6.5-14 Reflood Transient - Core and Downcomer Water Levels DECLG (CD=0.6) 15.6.5-15 Reflood Transient - Core and Downcomer Water Levels DECLG (CD=0.8) 15.6.5-16 Reflood Transient - Core and Downcomer Water Levels DECLG (C0=0.6 Maximum Safety injection) 15.6.5-17 Reflood Transient Core Inlet Velocity DECLG (CD 0.4) 15.6.5-18 Reflood Transient Core inlet Velocity OECLG (CD=0.6) 15.6.5-19 Reflood Transient Core inlet Velocity DECLG (CD=0.8) 15.6.5-20 .Reflood Transient Core Inlet Velocity DECLG (CD=0.6 Maximum Safety injection) 15.6.5-21 Core Power Transient DECLG (C0=0.4) 15.6.5-22 Core Power Transient DECLG (C0=0.6) 15.6.5-23 Core Power Transient DECLG (C0 0.8) 15.6.5 24 Containment Pressure Transient DECLG (C0 0.4) 15.6.5-25 Containment Pressure Transient DECLG (C0-0.6) 15.6.5-26 Containment Pressure Transient DECLG (C0 0.8) 15.6.5-27 Containment Pressure Transient DECLG (CD=0.6 Maximum Safety Injection) l 15.6.5-28A Core flow (Top and Bottom) DECLG (CD = 0.4) 15.6.5-288 Core Flow (Top and Bottom) DECLG (CD = 0.6) 15.6.5-28C Core Flow (Top and Bottom) DECLG (CD = 0.8) l llitll^":::;;:"ll::l:lllll:::llllllll:::ll O O 1 B-4
VEGP-FSAR O LIST OF FlCURES (continued) 15.6.5-29C Meat Transfer Coefficient DECLG (CD - 0.8) 15.6.5-290 Heat Transfer Coefficient DECLG (CD = 0.6 Maximum Safety Injection) 15.6.5-30A Fluid Temperature DECLG (CD = 0.4) 15.6.5-30B Fluid Temperature DECLG (CD 0.6) 9.6.5-30C Fluid Temperature DECLG (CD = 0.8) 15.6.5-300 Fluid Temperature DECLG (CD - 0.6 Maximum Safety injection) 15.6.5-31A Break Flowrate DECLG (CD - 0.4) 15.6.5-318 Break Flowrate DECLG (CD 0.6) 15.6.5-31C Break Flowrate DECLG (CD 0.8) 15.6.5-32A Break Energy Released to Containment DECLG (CD 0.4) 15.6 5-32B Break Energy Released to Containment DECLG (CD 0... 15.6.5-32C Break Energy Released to Containment DECLG (CD 0.8) 15.6.5-33A Fluid Quality DECLG (CD 0.4) 15.6.5-33B Fluid Quality DECLG (CD 0.6) ', 15.6.5-33C Fluid Quality DECLG (CD - 0.d) 15.6.5-33D Fluid Quality .DECLG (CD = 0.6 Maximum Safety injection) 15.6.5-34A Mass Velocity DECLG (CD = 0.4) 15.6.5-34B Mass Velocity DECLG (CD 0.6) 15.6.5-34C Mass Velocity DECLG (CD 0.8) 15.6.5-340 Mass Velocity DECLG (CD 0 6 Maximum Safety injection) 15.6.5-35A Accumulator Flow (Blowdown) DECLG (CD 0.4) 15,6.5-358 Accumulator Flow (Blowdown) DECLG (CD - 0.6) 15.6.5-35' Accumulator Flow (Blowdown) DECLG (CD - 0.8) 15.6.5-364 Minimum Safety injection (Pressure Versus Flow) 15.6.5-36B Maximum Safety injection (Pressure Versus Flow) 15.6.5-36C Combined Accumulator and ECCS Pumped Safety injection (C?-0.6, Minimum Safeguaros) 15.6.5-360 Combined Accumulator and ECCS Pumped Safety injection (CD=0.6, Maximu.n Safegt.ards) B-5
. ._-.m. .mm - . . . .._. _ ~ . . _ . _ _ _ . . . - . . . - - . _ _ _ _ _ _ _ _ . . . , _ - _ _ _ . . _ . _ _ _ . _ . . . . . . _ - . . . _
r VEGP-FSAR O LIST OF FIGURES (continued) 15.6.5 RCS Depressurization Transient (3-inch Break) 15.6.5-38 RCS Depressurization Transient (4-inch Break)
-15.6.5-39 RCS Depressurization Transient (2-inch. Break) 15.6.5-40 Core Mixture Level (3-inch Break) 15.6.5-41 Core Mixture Level (4-inch Break) .
15.6.5 Core Mixture Level (2-inch Break)
- 15.6.5-43 Clad Average-Temperatur. - Hot Rod (3-inch Break) 15.6.5-44 Clad Average Temperature - Hot Rod (4-inch Break) 15.6.5 l. Clad Average Temperature - Hot Rod (2-inch Break) 15.6.5-46~ Core Exit Steam Flow-(3-inch Break) 15.6.5 Fluid Temperature - Hot Spot-(3-inch Break) -
15.6.5-48 Heat Transfer Coefficent - Hot Assy (3-inch Break) , (. I-1D L . l l l t. O B-6
a
- VEGP.FSAR 6 p ~6.2.1.5 -Minimum containment Pressure Analysis for Performance Canability Studies-on Emeraency Core Coolino Svstem The. containment b'acxpressure and temperature and the containment wall condensing-heat transfer' coefficient, used for the limiting case,CD = 0.6 double-ended cola leg guillotine' break for the ECCS analysis found-in-Subsection 15.6.5, are presented in Figures 6.2.1-35, 36.-and 37. The i containment buckpressure'is calculated using the methods and assumptions described _in Westinghouse' Emergency Core Cooling-System Evaluation Model.--
Summary,'WCAP-8339.-Appendix A. Jnput parameters includin? the containment ' initial conditions,_ net free containment volume, passive heat sink materials,
' thicknesses, surface areas, starting. time, and number of containment cooling.
systems used in the analysis are described below. The inputs used remain the same as that presented in. Chapter 6 of the'FSAR except for the RWST water
-temperature which was changed from.50 to 40 degrees F. . j 6.2.1.5.1- Mass-and Energy. Release Data
[ L The mass ~and energy releases to the containment during the reflood portions of the limiting break transient are presented in Table 6.2.1-69. No credit
'is taken for blowdown. The broken loop mass and energy releases to the containment'for_the limiting break are given in Table 6.2.1-70. ;
The mathematical models which calculate the_ mass and energy releases to the- " containment are described in Subsection 15.6.5. Since the requirements of Appendix K ofl10'CFR 50 are very specific in regard to the modeling of the RCS during blowdown and since the models used are in conformance with Appendix K,~no alterations to those models have been made in regard to the nass and energy releases. A break spectrum analysis is performed. (See the double-ended cold leg guil10 tines which affect the mass and energy released tothecontainment.) This etfeet is considered for each case analyzed. During refill, the mass and energy released to the containment is assumed to be zero, which minimizes the. containment pressure. During reflood, the. effect of steam-water mixing between the safety injection water and the steam flowing through the RCS intact loops reduces the available energy released to p o0326 1:10 092490 B-7
1 i VEGP.FSAR 6 (m) v 'the containment vapor space and, therefore, tends to minimize containment pressure. 6.2.1.5.2 Initial Containment Internal Conditions The following initial values were used in the analysis: Containment pressure (psia) 14.7 Containment temperature (*F) 90 Pefueling water storage 40 tanktemperature(*F) NSCWtemperature('F) 40 Outside temperature ('r) 17 The containment initial conditions of 90'F and 14.7 psia are representative 1y low values anticipated during normal full-power operation. The initial relative humidity was conservatively assumed to be 99 percent. i 6.2.1.5.3 Containment Volume
'The volume used in the analysis is 2.95 x 106 ft3 6.2.1.5.4- Active Heat Sinks The containment spray system and the containment fan coolers operate to remove-heat from the containment. Pertinent data for these systems which were used in the analysis are presented in Table 6.2.1-71. The heat removal capability of each fan cooler is presented in Table 6.2.2-2.
Because the fan coolers use nuclear service cooling water (NSCW), the lowest normal NSCW temperature (40*F) was used in the analysis. The containment. sump temperature was not used in the analysis because the maximum peak cladding temperature occurs prior to initiation of the recirculation mode for the containment spray system. In addition, heat 00326 1,10 492490 B-8
VEGP FSAR 6 O transfer between the sump water and the containment vapor space was not considered in the analysis. 6.2.1.5.5 Steam-Water Mixing Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment (computer program COCO) code calculational model. 6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis, with their thermophysical properties, are given in Table 6.2.1-72. The passive heat sinks and the thermophysical properties were dividec. compliance with Branch Technical Position CSB6-1, Minimum Containment Presu . . , , _ for Pressurized-Water Reactor (PWR) ECCS Performance Evaluat1on. 6.2.1.5.7 Heat Transfer to Passive Heat Sinks The condensing heat transfer coefficients used for heat transfer to the steel contr.inment structures are given in Figure 6.2.1-37 for the limiting break. The containment temperature transient for the limiting break is shown in Figure 6.2.1-36, 6.2.1.5.8 Other Parameters The effect of having containment purge in operation at the onset of the double-ended cold leg guillotine break was evaluated. With a 5.0-s valve closure time and a 1.5-s signal delay time, containment pressure would be expected to drop less than 0.19 psi. In terms of the LOCA results presented in Chapter 15, this would indicate no penalty in FQ and only a small increase (less than 10'F) in peak clad temperature. This is well within the margin for conformance to Appendix K requirements. No other parameters have a substantial effect on the minimum containment pressure analysis. O 00326.t.10 092490 B-9
A VEGP FSAR 6 - a 6.2.1.5.9 ' Standard Revirw Plan Evaluation The VEGP does-not employ the heat transfer coefficients; supplied in the Standard Review Plan. The heat transfer coefficients were calculated in conformance with WCAP-8339, Appendix A, which has received Nuclear Regulatory Commission approval. This reference is for the COC0 code and predates CSB6-1. Results using the heat transfer coefficients have oeen found acceptable in other plant applications. G i m O 00326 I.10 492490 B-10
VEGP.FSAR 6 TABLE 6.2.1-69 REFLOOD MASS AND ENERGY RELEASES (DECLG BREAK /CD=0.6) Time Mass Energy
,,1,11 (1bm/s) (Btu /s) 43.89 0.00 0.00 44.79 4.93 6404.72 1}}

ML20065P204

Text

SUMMARY

Background

Attachment:

Attachment:

3.5.1 EadhReactorCoolantSystem(RCS)accumulatorshallbeOPERABliwith

1.0 INTRODUCTION

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

7.0 REFERENCES

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