ML20246K446

Proposed Tech Specs,Incorporating Increased Rod Drop Time & Increased Peaking Factors & Changing DNB Design Basis to WRB-1 Correlation

ML20246K446
Person / Time
Site:	Beaver Valley
Issue date:	05/09/1989
From:	DUQUESNE LIGHT CO.
To:
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ML20246K429	List: ML20246K425 ML20246K436 ML20246K446 ML20246K454
References
NUDOCS 8905180059
Download: ML20246K446 (21)

Text

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+u-ATTACHMENT A g Revise the Technical Specifications as follows: j 1

I Remove Pace Insert Pace B 2-1 B 2-1 B 2-2 B 2-2

'B 2-4 B'2-4 B 2-6 B 2-6 3/4 1-22 3/4 1-22 .!

3/4 2-5 3/4 2-5 ,

3/4'2-7 3/4 2-7 '

3/4 2-8 3/4 2-8 l' B 3/4 2-1 B 3/4 2-1 B 3/4 2-4 B'3/4 2-4  :

B 3/4 2-5 B 3/4 2-5 B'3/4 2-6 B 3/4 2-6 B 3/4-4-1 B 3/4 4-1 B 3/4 9-4 B 3/4 9-4 q l

l I" l l

l l

l L

L 1

.I l

8905180059 890509

PDR ADOCK 05000334 P PDC

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2.1 SAFETY LIMITS

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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.

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 WRB-1 correlation. The WRB-1 DNB correlation has been l developed to predict the DNB flux and the location o'f DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting fuel rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNDR is at the DNBR limit (1.17 for the WRB-1 correlation).

In meeting this design basis, uncertainties in nuclear and thermal parameters, and fuel fabrication parameters were combined statistically with the DNB correlation uncertainties to determine the plant DNBR uncertainty and establish the design DNBR limit such that there is at least a 95% probability with 95% confidence level that the minimum DNBR for the limiting fuel rod is greater than or equal to the DNBR limit. For this application, the design DNBR limit is 1.21. This DNBR value must be met in plant safety analyscs using nominal values of the input parameters that were included in the DNBR uncertainty evaluation. In addition, margin has been maintained in the design by meeting a safety analysis DNBR limit of 1.33 in performing safety analyses.

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

BEAVER VALLEY - UNIT 1 B 2-1 PROPOSED

_ _ _ _ _ _ _______ ______ _ _ _ _ _ . - l

r , - - . _ - - - _ _ _ . . _ _ - _ _ _ _ _ _ _ _ - _ _ _ ___ _____ - - _ _ _ _ _ -

I SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, FN H' Of 1.62 and a reference cosine with a peak of 1.55 for axial power l shape. An allowance is included for an ' increase in' F H at A

reduced power based on the expression:

F[H $ 1.62 [1 + 0.3 (1-P) ] l where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated' for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(aI) function of the overtemperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature AT trip will reduce the I setpoint to provide protection consistent with core safety limits. i 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The- reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping and fittings are designed to ANSI O 31.1 and the valves are designed to ASA 16.5 which permit a maximum transient pressure of 120% (2985) psig of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.

1 BEAVER VALLEY - UNIT 1 B 2-2 PROPOSED

)

g LIMITING SAFETY SYSTEM SETTINGS c

BASES The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR limit for l control rod drop accidents. At high power a single or multiple rod drop accident could cause flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. For those transients on which reactor trip on power range negative rate trip is not postulated, it is shown that the minimum DNBR is greater than the design DNBR limit.

Intermediate and Source Ranae. Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor start-up. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels. Th Source Range Channels will initiate a reactor trip at about 10+g counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional i capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature AT The overtemperature a T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes ,

corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figures 2.1-1, 2.1-2 and 2.1-3. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

f BEAVER VALLEY - UUIT 1 B 2-4 PROPOSED

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' LIMITING SAFETY SYSTEM SETTINGS l r

BASES through the pressurizer safety valves. No credit was taken for operation of .this trip in the accident analyses; functional capability at the however, its this specification to enhance the overall reliability of the specified trip setting is required by 1 Protection System.

Reactor j Loss of Flow 1

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

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip 1

will occur if the flow i full loop flow.

in any two loops drop below 90% of nominal Above 31% (P-8) of RATED THERMAL POWER, automatic {

reactor trip will occur if the flow in any single loop drops below l 90% of nominal full loop flow. l minimum value of the DNBR from This going latter trip will prevent the below the. design DNBR limit J

during normal operational transients and anticipated transients when l l 2 loops are in operation and the overtemperaturea T trip setpoint is adjusted to the value specified for all loops-in operation.

overtemperature a T trip setpoint adjusted to the value specified With the 2 loop operation, for stop valves open and at 71% the P-8 trip at 66% RATED THERMAL POWER with loop valve closed will prevent RATED THERMAL POWER with a loop stop below the design DNBR limit during normal the minimum value of the DNBR from going; operational transients and l anticipated transients with 2 loops in operation.

Steam Generator Water Level The- Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum specified setpoint provides volume required for adequate heat removal capacity. The water inventory in the steamallowance that there will be sufficient generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not accident analyses but is included in Table used in the transient and 2.2-1 to ensure the functional capability of the specified trip enhance the overall settings and thereby BEAVER VALLEY - UNIT 1 B 2-6 PROPOSED

1

^

j REACTIVITY CONTROL SYSTEMS o

ROD DROP TIME LIMITING CONDITION FOR OPERATION q l

3.1.3.4- The individual full length (shutdown and control)l rod drop q

time from the fully withdrawn position shall be 5 2.7 seconds from } a beginning of decay of stationary gripper coil voltage to dashpot 0 entry with- -

t

a. T avg 2 541*F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3.

' ACTION:

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

b. With the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to:

1. 5 61% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, or

2. 5 66% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed.

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

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

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

BEAVER VALLEY - UNIT 1 3/4 1-22 PROPOSED

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-Fg (Z)

LIMITING CONDITIONS FOR OPERATION 3.2.2 Fg(Z) shall be limited by the following relationships:

Fg(Z) 1 2.40 [K(Z)] for P > 0.5 l P

Fg(Z) $ [4.80] [K(Z)] for 5 0.5 l where P = THERMAL POWER RATED THERMAL POWER and X(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: Mode 1 ACTION: i With Fg(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% F exceeds the limit within 15minutesandsimilarlyreduOe(Z)the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower A T Trip Setpoints have been reduced at least 1%

for each 1% FO(Z) exceeds the limit. The Overpower A T Trip Setpoint feduction shall be performed with the reactor subcritical.

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided FO(Z) is demonstrated through incore mapping to be within its limit.

i BEAVER VALLEY - UNIT 1 3/4 2-5 PROPOSED j

l.2 m

g 1.0 (0.0.I.0) (s.o.l.o)-

o (10.e,0.94) 1 O O.8 w

N m

d 0.6 (12.o.o.e4) - 1 I

CC o

Z O.4 )

'] y-O.2 j

l O.0 O 2 4 6 8 10 12 14 CORE HEIGHT (PT) i Figure 3.2-2. K(2)- Normalized Fo(2) as a Function of Core Height BEAVER VALLEY - UNIT 1 3/4 2-7 PROPOSED

'y 4 -POWER' DISTRIBUTION LIMITS HUCLEARENTHALPYHOT-CHANNELFACTOR-Fh "

LIMITING CONDITION FOR OPERATION p

3.2.3 (Hshallbelimitedbythefollowingrelationship

l F[H $L1.62 (1 + 0.3 (1-P)]

where P = THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With F3N exceeding its-limit:

a. Reduce THERMAL POWER to less than 50% of RA'I'ED THERMAL POWER witnin. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High-Trip Setpoints to $ 55% of RATED' THERMAL POWER within.the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,

b. Demonstrate thru in-core mapping that F[H is within its limit 'within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

c. . Identify and correct the cause of the out of limit condition prior to increasing THERMAL' POWER, subsequent POWEF.

OPERATION may proceed provided that Fa H is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to. exceeding this THERMAL-power, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

1 BEAVER VALLEY - UNIT 1 3/4 2-8 PROPOSED l

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.- 3/4.2 POWER DISTRIBUTION 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: (a) maintaining the minimum DNBR in the core 2 the design DNBR limit during normal operation and in short l term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS I acceptance criteria limit of 2200*F is not exceeded.

The definitions of hot channel 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.

FfH 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 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the Fg(Z) upper bound envelope of 2.40 times the normalized axial peaking factor is l not exceeded during either normal operation or in the event of xenon redistribution following power changes.

l Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are BEAVER VALLEY - UNIT 1 B 3/4 2-1 PROPOSED

WER DISTRIBUTION LIMITS BASES l 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL ,

FACTORS - F;(Z) andPyll l The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) . the design limits on peak local power density'and minimum DNBR are not exceeded 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200*F.

Each of these hot channel factors are 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 insure that the hot channel factor limits are maintained provided:

a. Control rod in a single group move together with no individual rod insertion differing by more than i 12 steps from the group demand position,

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5'.

c. The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.

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

The relaxation in F$H as a function of THERMAL POWER allows changes in tge radial power shape for all permissible rod insertion limits. F AH will be maintained within its limits provided conditions a thru d above, are maintained.

When an F measurement is taken, both experimental error and  !

manufacturing tolerance must be allowed for. 5% is the appropriate experimental error allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance'for manufacturing tolerance.

The specified limit of FN contains an 8% allowance for uncertainties which meags thak normal, full power, three loop operation will result in F g aH $ 1.62/1.08.

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BEAVER VALLEY - UNIT 1 B 3/4 2-4 PROPOSED {'

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,' ' POWER DISTRIBUTION LIMITS BASES Fuel rod bowing reduces the value of DNB ratio. Margin has been maintained between the DNBR value used in the safety analyses (1.33) and the design limit (1.21) to offset the rod bow penalty and other penalties which may apply.  !

The radial peaking factor F (Z) is measured periodically to 1 provide assurance that the xhot channel factor, Fo(Z),

within its limits. The F remaggg as limit for Rated Thermal Power F provided in the RadYl Peaking Factor Limit Report per I specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.  !

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during start-up testing and periodically during power operation.

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

The two-hour time allowance for operation with greater a tilt condition than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0 l

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

EpWER DISTRIBUTION LIMITS BASES 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 maintain a minimum DNBR 1 the design DNBR limit throughout each analyzed transient. l The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

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BEAVER VALLEY - UNIT 1 B 3/4 2-6 ,

PROPOSED  !

3/404 REACTOR COOLANT SYSTEM BASES l

3/4.4.1 REACTOR COOLANT IDOPS l The plant is decigned to operate with all reactor coolant loops in operation and maint.ain DNBR above the design DNBR limit during all l normal operations and anticipated transients. In Modes 1 and 2, with one reactor coolant loop not in operation, THERMAL POWER is restricted to $ 31 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset. Either action ensures that the DNBR will be maintained above the design DNBR limit. A loss of flow  !

in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER).

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical condition, two operating coolant loops are required to meet the DNB design basis for this condition II event.

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

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 275'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part

50. The RCS will be protected against overpressure transients and will not exceed the limits.

BEAVER VALLEY - UNIT 1 B 3/4 4-1 PROPOSED

REFUELING OPERATIONS BASES The results of the spent fuel pool criticality analysis (August 1986) for Westinghouse STD/ Vantage 5H and OFA/ Vantage 5 fuel in three of l four storage locations show that there is more than 0.3% margin to the k egg limit of 0.95 with all uncertainties included. Based on ,

the sensitivity study completed with this analysis, an increase in '

the maximum allowed enrichment for fuel stored in the spent fuel storage racks from 4.00 to 4.05 w/o will increase the maximum rack ke rf by less than 0.002.

STD/ Vantage SH and OFA/ Vantage 5 fuel enriched at 4.05 w/o stored in Therefore, with Westinghouse 17 x 17 l

the spent fuel racks in three of four storage locations and with all )

of the assumptions and conservatism presented in the criticality analysis, the maximum rack keff will be less than 0.95.

3/4.9.15 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS The OPERABILITY of the control room emergency habitability system ensures that the control room will remain habitable for operations personnel during and following all credible accident conditions. The ambient air temperature is controlled to prevent exceeding the allowable equipment qualification temperature for the equipment and instrumentation in the control room. The OPERABILITY of thin system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control roam to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

+&

BEAVER VALLEY - UNIT 1 B 3/4 9-4 PROPOSED

ATTACHMENT B Safety Analysis Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Chance No. 162 Description of Amendment Request: The Cycle 8 reload fuel will incorporate upgraded Westinghouse fuel design features and operate with increased peaking factors. The attached report

" Plant Safety Evaluation for Beaver Valley Power Station Unit 1 Fuel Upgrade and Increased Peaking Factors" summarizes the safety I evaluations that were performed to confirm the acceptable use of these options for the safe operation of the plant. This Plant Safety Evaluation (PSE) will serve as a reference safety evaluation / analysis report for the transition from the present core to a core containing the upgraded fuel features and will be used as a basic reference document to support future reload safety evaluations using upgraded fuel designs. The proposed technical specification amendment supports the PSE by incorporating an increased rod drop time, increased peaking factors, and replacing reference to the DNB limit and W-3 R-Grid correlation noted in the Bases with the safety analysis DNBR limit and WRB-1 correlation, respectively. The PSE does not address N-1 loop operation, therefore, the proposed technical specification changes will be implemented for N loop operation only.

The upgraded fuel design features include the VANTAGE SH design features, reconstitutable top nozzles, debris filter bottom nozzles, snag resistent grids and standardized fuel pellets. The upgraded fuel assemblies will continue to use the Integral Fuel Burnable Absorber (IFBA) and axial blanket features currently used in the Cycle 7 reload fuel.

The VANTAGE SH fuel assembly design incorporates the use of Zircaloy grids with standard diameter Westinghouse fuel rods. The Zircaloy grid metal is thicker than the current Inconel grid design due to the difference in material strength properties. The use of thimble tubes with a reduced diameter identical to those used in the 17 X 17 Optimized Fuel Assembly (OFA) and VANTAGE 5 fuel assembly designs is required with the Zircaloy grids due to the increased grid metal thickness. The reduced thimble diameter will increase the design rod drop time from the current maximum of 2.2 seconds to 2.7 seconds. Therefore, Specification 3.1.3.4 has been revised to incorporate the 2.7 second rod drop time. This slower rod drop time will affect the results of the limiting FSAR transients affected by rod drop time such as Loss of Forced Reactor Coolant Flow, RCCA Bank Withdrawal from Subcritical and Rod Ejection. These and the other applicable FSAR accidents have been re-evaluated using the slower rod drop time. Demonstration fuel assemblies with Zircaloy grids have been used in Westinghouse cores, including Beaver Valley Unit 1, and Zircaloy grids have been used in many Westinghouse reload cores since the early 1980's. Therefore, the use of Zircaloy grids has been proven based on a successful and sa , operating history.

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4

, Attachment B Safety Analysis Proposed Tech. Spec. Change No. 162 Page 2 The Cycle 8 reload and future reload cores will contain the upgraded fuel features as well as the Vantage SH design features described above. These design changes are currently part of the licensing basis in other plants and meet all fuel assembly and fuel rod design criteria.

The Debris Filter Bottom Nozzle (DFBN) is designed to inhibit debris from entering the active fuel region of the core and thereby improves fuel performance by minimizing debris related fuel failures. This is a low profile bottom nozzle design made of stainless steel with a reduced end plate thickness and leg height.

This low profile design is structurally and hydraulically equivalent to the existing bottom nozzle design.

The Reconstitutable Top Nozzle (RTN) differs from the current design in that (a) a groove is provided in each thimble thru-hole in the nozzle plate to facilitate attachment and removal, and (b) the nozzle plate thickness is reduced to provide additional space for fuel rod growth. Along with the RTN, a long tapered fuel rod bottom end plug is used to facilitate removal and reinsertion of the fuel rods.

The standardized fuel pellets are a refinement to the current pellet design with the objective of improving manufacturability while maintaining or improving performance. This design incorporates a reduced pellet length, modification to the previous dish size and the addition of a chamfer. i The snag-resistant grids contain outer grid straps which are modified to help prevent assembly hangup from grid strap interference during fuel assembly removal. This was accomplished by changing the grid strap corner geometry and adding guide tabs on the outer grid strap.

The change from the current standard fuel core to an upgraded fuel core vill not change the FSAR nuclear design bases. However, the design bases will be modified due to the increased peaking factor limits. The peaking factors are primarily loading pattern dependent and the variations in the loading pattern dependent safety parameters are expected to be typical of the normal cycle to cycle variations for the standard fuel reloads.

Specification 3.2.2 has been revised by increasing the current Fg(Z) limit to:

F (Z) $ 2.4/P [K(Z)] for P > 0.5 and F (Z) $ 4.8 [K(Z)] for P $ 0.5

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l

, Attachment B l Safety Analysis )

Proposed Tech. Spec. Change No. 162 )

Page 3 l

l Figure 3.2-2 has also been revised to provide the K(Z) function )

appropriate for the increased Fq(Z) limit. Specification 3.2.3 has been revised by increasing the current FN delta H limit to:

FN delta H = 1. 62 (1 + 0. 3 (1-P) . j 1

l The increased peaking factors will allow additional flexibility in i fuel management and determining core loading patterns. The reload f cores will employ the usual methods of enrichment variation and j' burnable absorber usage to ensure compliance with the new peaking factor limits. The cycle specific reload core analysis will be performed in accordance with the current methodology.

No changes to the nuclear design philosophy or methods are required because of the upgraded fuel product or the use of increased peaking factors. The reload design philosophy includes the evaluation o' the reload core key safety parameters which comprise the nuclear design dependent input to the FSAR safety evaluation for each reload cycle. The key safety parameters are evaluated for each reload cycle and if one or more of the parameters fall outside the bounds assumed in the safety analysis, the affected transients will be re-evaluated and the results documented in the cycle specific reload safety evaluation.

The DNB design basis has been modified to address both the 17 X 17 standard and VANTAGE SH fuel assemblies using the WRB-1 DNB correlation, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids" WCAP-8762-P-A, and MINI-RTDP, " MINI Revised Thermal Design Procedure (MINI RTDP)" WCAP-12178-P. The WRB-1 DNB correlation is based entirely on rod bundle data and takes credit for a significant improvement in the accuracy of the critical heat flux predictions over previous DNB correlations. With the MINI-RTDP methodology, peaking factor uncertainties are combined statistically with the DNB correlation uncertainties to obtain the overall DNBR uncertainty factor which is used to define the design limit DNBR that satisfies the DNB design criterion. This criterion states that the probability that DNB will not occur on the most limiting fuel rod is at least 95% (at 95% confidence level) for any Condition I or II event. The 95/95 limit DNBR using the WRB-1 DNB correlation for the 17 X 17 standard and VANTAGE SH fuel assemblies is 1.17. The design limit DNBR for typical and thimble cells is 1.21. For use in the DNB safety analyses, the limit DNBR is conservatively increased to provide DNB margin to offset the effects of rod bow and any other DNB penalties that may occur and provides flexibility in the design and operation of the plant. The safety analysis limit DNBR with

, Attachment B

- Safety Analysis Proposed Tech. Spec. Change No. 162 Page 4 9% margin is equal to (Design limit DNBR)/(1.0 .09) = 1.33. The current maximum rod bow penalty is 1.3% DNBR, this penalty is also applicable to VANTAGE SH fuel assemblies based on the similarities between the 17 X 17 standard and VANTAGE SH fuel assemblies including fuel rod diameter, fuel rod pitch and grid spacing.

Therefore, adequate margin in the safety limit DNBR is available to cover any rod bow penalties and, in addition, all current thermal hydraulic design criteria are satisfied.

The spent fuel pool criticality analysis remains conservative since the fuel assemblies were modeled without taking credit for flux reduction due to neutron absorption in the grids. Technical Specification Bases Section 3/4.9.14 Fuel Storage - Spent Fuel Storage Pool has been revised to document the suitability of the criticality analysis to applicable optimized and VANTAGE SH fuel.

The transient analyses in Section 14 of the FSAR have been re-analyzed where required or re-evaluated to include the increased peaking factors, increased rod drop time and revised fuel assembly design parameters to ensure the safety analysis limits are satisfied. UFSAR changes to reflect the revised '

Section 14 accident analyses are attached to the Plant Safety i Evaluation. In addition, UFSAR Section 3 has been revised to I include a description of the fuel assembly design changes and Section 14D has been revised to update the description of the transient analyses methodology and computer codes used.

The VANTAGE SH fuel assembly design has been approved by the NRC. The other upgraded fuel features described in the Plant Safety Evaluation have been implemented in other Westinghouse reload cores in accordance with 10 CFR 50.59 and do not require prior NRC approval, however, the changes in RCCA drop time and increased peaking factors require changes in the technical specifications.

The proposed technical specification changes and supporting l Plant Safety Evaluation are provided for NRC review and approval .

and to document the safety analysis and evaluations performed to ensure the proposed changes are consistent with accepted methodology and required safety limits. The attached UFSAR i changes are provided for background information and will be '

incorporated into the UFSAR in a future update following approval of the proposed technical specification changes. The proposed technical specification changes have been evaluated in accordance with approved methodology and have been shown to satisfy the applicable acceptance criteria. Therefore, based on the above, the proposed changes will not reduce the safety of the plant.

. ATTACHMENT C No Significant Hazard Evaluation Beaver Valley Power Station, Unit no. 1 Proposed Technical Specification Chance No. 162 Basis for Proposed No Significant Hazards Consideration Determination: The Commission has provided standards for determining whether a significant hazards consideration exists in I accordance with 10 CFR 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve  !

a significant reduction in a margin of safety.

The proposed changes do not involve a significant hazard consideration because:

1. The Cycle 8 reload fuel will incorporate upgraded Westinghouse fuel design features and operate with increased peaking factors. The upgraded fuel design features include the VANTAGE SH design features, reconstitutable top nozzles, l debris filter bottom nozzles, snag resistent grids and I

standardized fuel pellets. Along with the upgraded fuel design features, Integral Fuel Burnable Absorbers and axial blankets will continue to be used as in the Cycle 7 reload fuel. >

The VANTAGE SH and the Standard (STD) 17 X 17 fuel assembly l are hydraulically equivalent. Implementation of the VANTAGE SH fuel design will not significantly change the core physics characteristics. The proposed changes have been assessed from a core design and safety analysis standpoint. No increase in l

the probability of occurrence of any accident was identified.

Extensive reanalyses were undertaken to demonstrate compliance with the revised technical specifications. The methods used to perform the analyses have been previously approved by the NRC. The results, which include transition core effects, show changes in the consequences of accidents previously evaluated. However, the results are all clearly within NRC acceptance criteria and demonstrate the capability to operate the plant safely. The major components that determine the structural integrity of the fuel assembly are the grids.

Mechanical testing and analysis of the VANTAGE SH Zircaloy grid and fuel assembly have demonstrated that the VANTAGE SH

, structural integrity under seismic /LOCA loads will provide l margins comparable to the STD 17 X 17 fuel assembly design and will meet all design bases. Therefore, the proposed amendment does not result in an increase in the probabilities or consequences of a previously evaluated accident.

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, Attachm:nt C Safety Analysis Proposed Tech. Spec. Change No. 162 Page 2

2. The proposed changes do not significantly affect the overall method and manner of plant operation and can be accommodated without compromising the performance or qualification of safety related equipment.

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

3. The evaluations and analyses to support the proposed technical specification changes and operation of the plant with VANTAGE SH fuel show some changes in the consequences of previously analyzed accidents. In some cases, an increase in event consequences occurs. However, in all cases the results of the changes are clearly within all plant design and NRC safety acceptance criteria.

Therefore, the proposed amendment does not significantly reduce the margin of safety.

Based on the above, the proposed changes have been determined to be safe because (1) the core reload will use VANTAGE 5H fuel with upgraded design features which vre not significantly different from previous reload cores. (2) the technical specification changes result from the core reload and not from any significant change to the acceptance criteria for technical specifications, and (3) the analytical methods used in the required reload analysis have been previously found acceptable by the NRC. Therefore, it is proposed that this amendment application does not involve a significant hazard consideration.

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