ML20235T981

Application for Amend to License NPF-42,revising Tech Specs to Reflect Revised RCS Total Flow Rate.Fee Paid

ML20235T981
Person / Time
Site:	Wolf Creek
Issue date:	10/02/1987
From:	Withers B WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20235T984	List: ML20235T981 ML20235T995
References
WM-87-0254, WM-87-254, NUDOCS 8710130311
Download: ML20235T981 (23)

Text

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. 1 W6) LF NUCLEAR CREEKOPERATING l U M n "o Chief ExecutNo Officer October 2,1987 i

WM 87-0254 q U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 /

Subjecu de...;t No. 50-482: Revision to Technical Specification 3.2.3 - Reduced Thermal Design Flow Gentlemen:

The purpose of this letter is to transmit an application for amendment to-l Facility Operating License No. NPF-42 for Wolf Creek Generating Station; Unit No. 1. This license amendment request proposes revising Technical Specification 3.2.3, Figure 2.1-1, Figure 3.2-3, and Table 2.2-1. This application for amendment modifies these TechnicalSpecificatjonstoreflect a revised Reactor Coolant System total flow -rate and F AH part power l multiplier. i A complete Safety Evaluation and Significant Hazards Consideration are 3 l

provided as Attachments I and II respectively. The proposed changes to the 1 Technical Specifications are provided in Attachment III. _l l

In accordance with 10 CFR 50.91, a copy of this application with attachments )

is being provided to the designated Kansas state official. Enclosed is a ~

check (No.1019)forthe$150.00applicationfeerequiredby10CFR170.21.

The proposed revision to the Wolf Creek Generating Station Technical Specifications will be fully implemented within 30 days of- formal Nuclear Regulatory Commission approval.

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8710130311 -" h" h~B2

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P.O. Box 411 / Burhngton, KS 66839 / Phone: (316) 364-8831 l An Equal Opportunity Employer M0HGvET i

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WM 87-0254 Page 2 of 2 October 2 ,1987 If you have any questions concerning this matter, please contact me or Mr.

O. L. Maynard of my staff. ,

Very truly yours,

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Bart D. Withers President and Chief Executive Officer Enclosure Attachments: I - Safety Evaluation II - Significant Hazards Consideration l III - Proposed Technical Specification Changes cc: P. O'Connor (2)

R. Martin J. Cummins G. Allen i

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STATE OF KANSAS )

) SS j COUNIT OF COETEY ) .

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Bart D. Withers, of lawful age, being'first duly sworn upon oath says that-he is President and Chief Executive Officer of Wolf Creek Nuclear Operating i Corporation; that he has read the foregolng document and knows' the content {

thereof; that he : has . : executed that -same for and-.on . behalf.'of~ said j Corporation with full power and authority to do so; 'and that-the facts therein ' stated are true and correct to the best of. his knowledge, information and belief. -

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BY Bart D. Withers- i President and Chief Executive Officer l -

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SUBSCRIBED and sworn to before me this .2 day.of [C M , 1987. l

.$;/ %:l5.< o'e 911ala W Notary Public l I 2l4%%$dA l

,I Expiration Date M N,/998 Rk' ' S..hk!

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ATTACllENT I SAFETY EVALUATION

Attachment I to WM 87-0254 page 1 of 17 October 2, 1987 i DESCRIPTION OF PROPOSED AMENDMENT This proposed Technical Specification amendment reduces theRgactorCoolant System (RCS) total flow rate requirement and revises the F part power multiplier. As a result of these two changes, there are adoinstrument '

allowance changes in Table 2.2-1 for Overtemperature Delta-T (0 TDT), Overpower  !

Delta-T (0PDT)and Reactor Coolant Flow-Low. Also the K constant in the Overtemperature Delta-T equation is being revised. Theaksociatedtechnical specification BASES are also being revised to reflect these changes.

The reduction in RCS total flow rate is being made to increase the margia between the technical specification flow rate limit and the calculated RCS total flow rate.

The proposed revision to the F N part power multiplier is being submitted for j reactor core design considerNionsforsubsequent cycles. The part power l multiplier change will help enhance the efficiency of future cycles at Wolf Creek Generating Station (WCGS). I PART POWER MULTIPLIER Increasing the allowable F,k with decreasing )ower has been permitted for all previously spprovedWestingNouse designs. T1e increase is permitted by the DNB protection set points and allows for radial power distribution changes with rod insertion to the insertion limit. The current Wolf Creek Unit 1 TechnfcalS)ecifications(Section3/42.3)recuirea0.2partpowermultiplier on F T1e NRC has in recent years ap3rovec a 0.3 part power multiplier for i anum $r. of operating plants such as Tur(ey Point, R.E. Ginna, Trojan, D.C.

Cook Unit 1, Zion, Indian Point Unit 3, Point Beach and Surry. Described in this submittal is a proposedchagge to the WCGS Technical Specifications l reflecting a design change in the FAH part power multiplier.

The results of the WCGS F N Technical Specification limit analysis indicate that thelimitmaybemodibedbychangingthelimitslopefrom0.2to0.3at reduced power, resulting in the following relationships:

N FAHs1.55[1.0+0.3(1-P)]

where: P = fraction of rated thermal power The only change from the current F N Technical Specification is theN multiplier onthequantity(1-P)from0.2togH3. No change was made in the F 3g limit at full power.

This change is proposed for WCGS to allowN optimization of the core loading pattern by minimizing restrictions on the F o at low power. This change will-also minimizetheprobabilityofmakingrod$nsertionlimitchangesinfuture reload cycles to satisfy peaking factor criteria at lower power.

This evaluation concludes that the modification of the F N part power multiplier from 0.2 to value of 0.3 does not resultinanunreDiewedsafety question, nor does it involve a significant hazards consideration.

Attachment I to WM 87-0254 page 2 of 17 October 2,1987 PART POWER MULTIPLIER - THERMAL-HYDRAULIC DESIGN EVALUATION The proposed WCGS N Technical Specification change which impacts DNBR evaluation is the value of F AH determined from the following equation:

N F H < 1.55 [1 + K(1-P)]

where: K (part power multiplier) has increased from 0.2 to 0.3 1 N

F g = measured radial peaking factor with appropriate uncertainties P = fractional core power level at less than 100% Rated Power or, P = 1.0 at greater than or equal to 100% Rated Power The increase in the part power multiplier (K) from 0.2 to 0.3 has a direct impact on DNBR calculations. The core limits for WCGS (reflected in Technical Specification Figure 2.1.1) represent restrictions of average enthalpy at the vessel exit and minimum DNBR.

The average enthalpy at the vessel exit must be less than the enthalpy of ,

saturated liquid to assure the proportionality between vessel temperature and core power. The exit enthalpy restriction is more limiting than DNBR at low l power, and vessel exit limit lines are not impacted by the radicai peaking factor as shown in the following relation:

hout = hin + Q/G < hsat where.

l hout = average coolant enthalpy at vessel exit (BTU /lbm) h in = vessel inlet coolant enthalpy (BTU /lbm)

Q = total core power (BTV/hr)

G = total core coolant flow (1bm/hr)

Therefore, a change in radial peaking factor will not impact core limits at i

power levels restricted by vessel exit boiling limits.

At power levels greater or Ne qual to 100% rated power F N is not impacted by K(i.e.,1-P=0) and the peak F us levels is unchanged. Therekdre,edtogeneratethecoreNmitsatthesepowe the core limit restricted by DNBR at these power levels will not change. */he core limits at power levels less than 100%

rated power which are not restricted by vessel exit boiling limits will be impacted by the change in the part power multiplier.

PART POWER MULTIPLIER - NUCLEAR DESIGN EVALUATION l

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Attachment 1 t'o WM 87-0254 page 3 of 17 October 2,1987 The proposed technical specification basis changes do not impact the other nuclear design bases used to evaluate the reload cores. The standard calculation methods described in WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," are still valid. Currently, each reload core design is evaluated to assure that design and safety limits are satisfied according to this reload methodology. The 0.3 multiplier in conjunction with the technical specification control rod insertion limits insures that peaking factor limits are not exceeded during anticipated power control maneuvers.

PART POWER MULTIPLIER - ACCIDENT EVALUATION Tge Reactor Core Safety Limits were reevaluated due to the proposed increased F limit to ensure adequate core protection. It has been concluded from saletyevaluationsthatthereactorcoresafetylimitsarenotchangedandtge USAR non-LOCA safety analyses remain a)plicable for the increased F multiplier. A change to the OTDT tecinical specification reflects 3N a increased allowance for instrumentergor. Since non-LOCA accident analysis l are not impacted by the proposed F multiplier change no reanalysis is required for non-LOCA accident events. AH l

The currgntlargeand small break LOCA analysis of record remain applicable increase at partial powers as long as the product of power times fgrtheFisle$sthan the full power value. The LOCA analyses are performed at Fm%

102 of rated power and use the technical specification up)er limit for F3N .

TgustheLOCAanalysisboundsallpartialpowerconditionsw1entheproductOf F times power decreases from the full power value with decreasing power.

Sicethetechnicalspecificatiogchangeforthepartialpowermultiplierdogs not increase the full power F and the product of the partial power F 3 tgmes power is less than the AH full power limit; then thepartialpoweY F q multiplier increase has no impact on LOCA analyses.

THERMAL DESIGN FLOW The original loop Thermal Design Flow (TDF) used in analyzing the Wolf Creek Generating Station was 95700 gpm. This evaluation supports a reduction in loop TDF to 93200 gpm. This evaluation considers the impact of the reduced i

TDF and associated changes to the RCS design parameters on the Updated Safety l Analysis Report (USAR) safety analysis results and conclusions. Full power, 1 l steady state reactor coolant pressure and vessel average temperature are '

maintained for the reduced TDF. An increase in vessel and core outlet I

tem)eratures within 1 degree F, a decrease in vessel / core inlet temperature l

wit 1in 1 degree F and a 4 psi reduction in steam pressure result from the evaluation.

THERMAL DESIGN FLOW - LOSS OF COOLANT ACCIDENT (LOCA) EVALUATION CONTAINMENT LONG TERM MASS AND ENERGY RELEASE AND CONTAINMENT SUBCOMPARTMENT ANALYSES - USAR SECTION 6.2 The containment analyses are described in the 9SAR sections 6.2.1.2 and ,

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I Attachment I to WM 87-0254 page 4 of 17 Octooer 2, 1987 i

6.2.1.3. These sections consider the containment subcompartment release and )

long term mass and energy release analyses for postulated loss-of-coolant accidents. A review was performed of the background information utilized for the current USAR Section 6.2 containment analyses. The results of the review identify that the proposed reduction in TDF will have a negligible impact on the RCS temperature assumptions modeled, therefore the current Section 6.2 containment mass and energy releases continue to be valid.  !

STEAM GENERATOR TUBE RUPTURE - USAR SECTION 15.6.3 The USAR Steam Generator Tube Rupture (SGTR) analysis was performed using the LOFTRAN (Reference 9) program. The primary to secondary break flow was assumed terminated at 30 minutes after initiation of the SGTR event. The major factors that affect the radiological doses of an SGTR event are the l amount of fuel failure; the amount of primary coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube after reactor trip; and the steam released from the ruptured steam generator to the atmosphere. An evaluation was completed to determine the impact on the USAR SGTR analysis of a reduced thermal design flow (TDF).

The evaluation completed to determine the impact of reduced TDF also included the necessary safety analysis 0 TDT reactor trip setpoint change required to support the part power multiplier technical specification change. The reduced TDF and 0 TDT reactor trip setpoint change were conservatively estimated to yield a bounding ev61uation. The results of the evaluation indicated that during the time between reactor trip and ruptured SteamGenerator(SG) isolation both the primary to secondary break flow and the steam released via the ruptured SG will increase. However, based on conservatism included in the USAR SGTR analysis, the consequences associated with the increase in mass releases for this evaluation will be bounded by the results reported for the USAR SGTR analysis. Therefore, the results of this evaluation are bounded by the results of the USAR SGTR analysis. The DNBR for the SGTR analysis with a reduced TDF has also been evaluated. It has been determined that the DNBR for the SGTR analysis is bounded by the " Inadvertent Opening of a Pressurizer Safety or Relief Valve" analysis.

The results of an evaluation completed by Westinghouse to determine the impact of a reduced TDF on the " Inadvertent Opening of a Pressurizer Safety or Relief Valve" analysis indicate that the DNB limits will not be violated. Therefore, it can be concluded that the core DNB limits for the SGTR analysis will not be violated due to a TDF reduction.

Based on the above evaluation, it is concluded that a reduction of TDF will not change the conclusions reported in the USAR. Therefore, no reanalysis or USAR changes are required.

i POST-LOCA LONGTERM CORE COOLING - USAR SECTION 15.6.5 l The reduction in TDF has no impact on long term core cooling following a large break LOCA since this is controlled by the safety systems; the reactivity of the core; and the total mass of primary coolant and soluble boron that are collected in the containment sump. Since the temperatures of the primary loop

O t Attachment I to WM 87-0254 page 5 of 17 October 2,1987 have been changed by less than one de ree due to the reduction in TDF, the change in primary coolant mass is negli ible.

ROD EJECTION MASS AND ENERGY RELEASE FOR DOSE CALCULATIONS - USAR SECTION 15.4.8 A small break LOCA is .modeled for this analysis, with a break in the upper head the size of a control rod drive shaft in order to determine the primary.

coolant mass' released to the containment through the break and the steam released from the steam generator safety valves. This information is then used to compute the radiological' consequences of a rod ejection accident. -Minor changes could be expected in rod ejection mass and energy . releases with a reduction in TDF, primarily as a result. of a slight decrease (less than one i

degree) in vessel inlet temperature. -The current values listed in USAR Table l 15.4-4 and Figure 15.4-27 are:

Primary Coolant Mass, 1bm .

530,000 Steam Released Through Safety Valves, 1bm 48,600 Based on previous sensitivities, the computed impact of a one degree F decrease in vessel inlet temperature would change the values to:

Primary Coolant Mass, 1bm 531,060 Steam Released Through Safety Valves, lbm 47,579 These results show only a very small effect on the long term mass releases due to the decrease in TDF.

R00 EJECTION RADIOLOGICAL EVALUATION - USAR SECTION 15.4.8 The radiological concerns for the rod ejection accident include releases from the containment and from the secondary side via the steam generator safety valves. Because coolant activity is directly proportional to coolant mass, the 0.2% increase in coolant mass . associated with the reduced thermal design flow corresponds to a 0.2% increase in coolant activity. Per analysis )

assumption, the coolant activity released to the containment is assumed to be mixed instantaneously. Therefore, releases are estimated to increase by 0.2%.

The additional activity released however, is well within the limits of 10 CFR 100. Whereas the coolant mass increases as a result of the reduced thermal design flow, the steam discharged via the safety valves decreases thus reducing offsite release. Therefore, the conclusions presently in the USAR remain valid.

HOT LEG SWITCHOVER TO PREVENT POTENTIAL BORON PRECIPITATION - USAR SECTIONS 6.3.2.5 AND 15.6.5 Post-LOCA hot leg recirculation switchover time is determined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel 1 following boiling in.the core. This time is dependent on power level, and the RCS,. Refueling Water Storage Tank (RWST) and accumulator water volumes and boron concentrations. Since the reduction of thermal design flow does not affect the power level or the maximum boron concentrations or volumes assumed j

I Attachment I to WM 87-0254 page 6 of 17 October 2 ,1987 for the RCS, RWST and accumulators, there is no impact on the post-LOCA hot leg switchover time.

SMALL BREAK LOCA - USAR SECTION 15.6.5 The small break LOCA analysis consists of a thermal hydraulic RCS and a hot rod analysis. These analyses were performed with the WFLASH and LOCTA computer codes respectively. The impact of the reduction in TDF on the transient characteristics is discussed below. j For a small break LOCA analysis, a reactor trip occurs when the low pressurizer pressure reactor trip setpoint is reached. As primary coolant i inventory spills out the cold leg, the RCS depressurizes linearly until the reactor trip setpoint is reached. For equal size breaks, break flow is dependent on the delta-pressure across the break and on the density of the water in the cold leg. TDF reduction will not affect primary pressure thus the ,

delta-pressure across the break will remain the same for this evaluation. 1 However the TDF reduction will affect the temperature in the cold leg and thus j the density of the water in the cold leg. J For the reduction in loop TDF from 95700 to 93200 gpm the temperature in the cold leg was calculated to be less than one degree F lower than the case analyzed at the original TDF. Thus the density of the cold leg with the reduced TDF would be slightly greater than the case analyzed. Because the differences in densities of water in the cold leg between the two cases is very small (less than 0.2 %), there would be no significant difference in RCS depressurization rate or reactor trip time.

After reactor trip there would be no significant difference in the thermal hydraulic response between the original analysis and the reduced TDF case.

Additionally, for a small break LOCA, a minor perturbation in initial operating conditions should not have an impact on calculated peak cladding temperature (PCT) because of the event sequence. The peak cladding temperature  !

of a small break LOCA occurs after loop seal clearing. The coolant inventory '

and core mixture level after loop seal clearing are strongly dependent on steam generator operation conditions and loop seal clearing oscillations during the transient. These factors are not affected by the initial loop flow rate.

The peak hot rod cladding and fuel temperatures are dependent upon the initial fuel rod temperatures that are reinitialized at core uncovery. From the discussion above it has been concluded that, the thermal hydraulic conditions at core uncovery will not be significantly different for the reduction in TDF o case. The only parameter in LOCTA that is not reinitialized at core uncovery l is the radial ga) between the fuel rod and the fuel. However calculations have 4 been performed t1at show that the radial gap between the fuel and the fuel rod is not sensitive to changes in TDF. Consequently, no difference in the thermal hydraulic conditions at core uncovery means there will be no difference in 1 peak cladding temperature. !i Based on the above discussions, the reduction in TDF from 95700 gpm to 93200 l

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Attachment I to WM 87-0254 page 7 of 17 October 2 ,1987 g )m will have no significant impact on the small break LOCA analysis results.

T1us, the reduction in TDF will have no significant impact on the WCGS small break margin to the PCT limit of 2200 degrees F. The reduced TDF would also have no impact on the conclusions of WCAP-11145-P-A, which demonstrated that the previous NRC approved WFLASH small break-LOCA evaluation model results were conservative when compared with the new NRC approved NOTRUMP small break

'LOCA evaluation model.

LARGE BREAK LOCA USAR SECTION 15.6.5 The analysis of record for the .WCGS was performed with the revised BART evaluation model. The pertinent assumptions.for this analysis. included.: core licensed -power of 3411 MWt; 17 x 17-STD fuel; 10% steam generator tube plugging level; total core peaking factor of 2.42; hot assembly average rod  ;

radial peaking factor of 1.55; accumulator water volume of 850 cubic feet per j accumulator; and a loop TDF rate of 93200 gpm.- l Therefore, since the analysis incorporates the reduced TDF value there is no impact on the large break results.

THERMAL DESIGN FLOW - NON-LOCA EVALUATION The purpose of this section is to provide technical justification from a i non-LOCA safety analysis standpoint, for a reduction.in the WCGS TDF from 382800 gpm to 372800 gpm.

All of the WCGS non-LOCA transients have been reviewed to determine the impact i of this reduction in TOF. For evaluation purposes, the non-LOCA transient j analyses have been reviewed on the basis of both DNB and non-DNB related j acceptance criteria. The following assumptions have been made for the non-LOCA transient evaluation: 4

1. Current analyses of -record are consistent with Wolf Creek USAR Chapter 15, Rev. O and USAR Section 6.2.4.1, Rev. O. I

2. Rated Core Thermal Power = 3411 MWt.  ;

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3. Nominal full power vessel average temperature = 588.5 F.

4. 1.65 degrees F steam generator tube fouling margin.

5. Zero percent steam generator tube plugging margin. (372800gpmis the current Large Break LOCA TDF which is assumed to correspond to 10% steam generator tube plugging. Consideration of tube plugging which involves adjustments to pressure drops, flow areas and heat transfer is not- considered in this evaluation. This non-LOCA evaluation will remain consistent with. the _ current non-LOCA assumption of zero percent tube plugging as reflected in the USAR.)

6. LOCA qF = 2.32 Fh=1.55andF$g part power multiplier = 0.3 i

Attachment I to WM 87-0254 page 8 of 17 October 2,1987

7. Unless otherwise noted, safety analysis assumptions regarding protection system setpoints are consistent with those used in the current analyses of record.

Ti!ERMAL DESIGN FLOW - DNB CONSIDERATIONS The proposed 2.7% reduction in TDF was accommodated by a reduction in available DNBR generic margin applied to all DNBR limiting conditions. A total generic margin allocation of 3.4% was used to keep the DNBR limiting portions of the core limit curves unchanged and to compensate for the reduced flow in DNBR limiting transients (i.e., Loss of Flow, RCCA Malfunctions, l etc.). The vessel exit boiling and quality limit lines of the core limits were re-evaluated and adjusted appropriately for the reduced vessel flow.

EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTION -

USAR SECTION 15.1.1. AND 15.1.2 I Feedwater system malfunctions that result in a decrease in feedwater temperature are bounded by those which result in increased feedwater flow.

The increase in feedwater flow event is considered for both the no-load and full power conditions. The no-load case is bounded by the Uncontrolled RCCA Bank Withdrawal from Subcritical and the full power casc is explicit!y analyzed and presented in the USAR. ,

In the USAR full )ower event, automatic rod control has been assumed ano a reactor trip on tie high-high steam generator water level signal is at 36.2 seconds. The reduced TDF will not change the time at which this signal is  !

initiated. Minimum DNBR occurs immediately after reactor trip. Up until the i point of reactor trip, core power matches the increased load demand. Since the T-avg and system 3ressure to which the plant would be controlled have not l changed, there would )e no significant change to these parameters at the time of minimum DNBR. The heat flux at the )oint of minimum DNBR would be the same as well. Thus, the only significant clange in the RCS conditions at the time of minimum DNBR would be flow and this is accounted for by the allocated generic DNBR margin. Therefore, the USAR conclusion that the DNBR acceptance criterion is met remains valid.

EXCESSIVE LOAD INCREASE - USAR SECTION 15.1.3 This transient is characterized by a rapid increase in steam flow that causes a power mismatch between reactor core power and steam generator load demand.

Four cases are analyzed and presented in the USAR. These are manual and automatic rod control for minimum and maximum reactivity feedback. In all-cases, equilibrium is reached corresponding to the increased load demand.

Only the manual control maximum feedback case, results in a reduction in DNBR from that associated with steady state o)erating conditions. The calculated minimum DNBR is significantly greater taan the safety analysis limit. All cases are conservatively covered by the allocation of generic DNBR margin.

Therefore, the USAR conclusion that the DNBR acceptance criterion is met remains valid.

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Attachment I to WM 87-0254 page 9 of 17 October 2,1987 STEAMLINE DEPRESSURIZATION - USAR SECTION 15.1.4 The steamline depressurization transient is analyzed for the following cases:

1. Inadvertent opening of a steam generator relief or safety valve,

2. Major rupture of a main steam line with and without offsite power.

Steamline depressurization results in a primary side cooldown which in the presence of a negative moderator temperature coefficient causes a reactivity insertion.

A reduction in TDF reduces primary to secondary heat transfer and the reactivity insertion due to the negative moderator temperature coefficient.

In the main steamline rupture event, the transient is mitigated by borated safety injection (SI) flow initiated by a low steamline pressure signal.

Initiation of safety injection is therefore unaffected by the reduction in TDF. This coupled with the reduction in reactivity insertion results in a DNBR benefit. A similar argument is applied for the inadvertent opening of a steam generator relief or safety valve for which SI delivery is initiated by a low pressurizer pressure signal. Any effect on the minimum DNBR statepoints for either case is accounted for by the allocation of generic DNBR margin. On this basis it is concluded that the DNB acceptance criterion continues to be met for the current USAR steamline depressurization transients.

I LOSS OF LOAD / TURBINE TRIP - USAR SECTION 15.2.2 AND 15.2.3

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Four cases are presented in the USAR. They are minimum and maximum reactivity l feedback with and without pressurizer sprays and PORVs. Three of the four ,

cases have increasing DNBRs. The case with the decreasing DNBR is the minimum i reactivity feedback case with pressurizer control. The calculated minimum ]

DNBR which occurs shortly after reactor trip on OTDT is well above the i accident analysis design limit. The reduction in TDF may have a slight impact on the RCS conditions at the time of minimum DNBR. The allocation of generic DNBR margin, however, conservatively compensates for any penalty resulting from the reduction in TDF. Therefore, the DNB acceptance criterion continues '

to be met.

PARTIAL AND COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW -

USAR SECTION 15.3.1 AND 15.3.2 The complete loss of forced reactor coolant flow is the limiting of the two loss of flow DNBR transients. Reactor trip on undervoltage for this event is unaffected by the reduced TDF. Any effect on the RCS conditions which occur at time of minimum DNBR is compensated for by the allocation of generic DNBR margin. A similar argument applies to the less limiting Partial Loss of Reactor Coolant Flow transient. Therefore, the DNB acceptance criterion continues to be met.

REACTOR COOLANT PUMP SHAFT SEIZURE - USAR SECTION 15.3.3 AND 15.3.4 The locked rotor transient is classified as a Condition IV event. As such,

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Attachment I to WM 87-0254 page 10 of 17 October 2,1987 i

i compliance with the guidelines of 10 CFR 100 regarding the release of radioactive materials must be demonstrated, although DNB is not precluded.

For conservative clad temperature calculations, DNB is assumed to occur at the beginning of the transient. A conservative calculation is performed to determine the percentage of fuel rods in the core that experience DNB. The calculated number of rods-in-DNB is a function of the transient flow and heat  :

flux. The allocation of generic DNBR margin will offset any penalty resulting l form the reduced initial condition TDF such that the previously calculated I percentage of rods-in-DNB may continue to be reported as one percent. ),

UNCONTROLLED RCCA BANK WITHDRAWAL FROM A SUBCRITICAL CONDITION -

USAR SECTION 15.4.1 The point of minimum DNBR for the USAR rod withdrawal from sd critical event occurs at 12.5 seconds which is 1.7 seconds after reactor trip on high neutron flux - low setpoint. Since loop transport time is approximately 12 seconds, the reduced TDF would have little impact on the calculated temperature at the time of minimum DNBR. Additionally, the reduced TDF would result in reduced fuel-to-coolant heat transfer; the associated heat flux reduction would be a DNB benefit.

In summary, the transient RCS conditions would not be significantly affected by the TDF reduction and any resulting penalty is conservatively offset by the  ;

allocation of generic DNBR margin. Therefore, the DNB acceptance criterion ,

continues to be met. '

UNCONTROLLED RCCA BANK WITHDRAWAL AT POWER - USAR SECTION 15.4.2 The USAR rod withdrawal at power transient analysis is performed at three power levels for a spectrum of reactivity insertion rates. The more rapid reactivity insertion cases have reactor trip on high neutron flux at 118%

power regardless of the flow. Any effects on RCS conditions at the time of minimum DNBR will be offset by the allocated generic DNBR margin.

The slower reactivity insertion cases trip on OTDT. Lower flow may cause earlier reactor trip on OTDT unless conditions are such that the steam generator safety valve setpoint is reached. In this case, OTDT trip may be delayed and transient results may worsen. The limiting rod withdrawal at power cases which trip on 0 TDT were examined. For these cases, there is sufficient margin to the steam generator safety valve setpoint such that the reduced flow would not cause delay in reactor trip on OTDT. Additionally, the current safety analysis setpoint value is conservative with respect to the newly calculated setpoint required to protect the core limits reflected in the

)roposed technical specification Figure 2.1-1. Any DNB penalty will be offset

]y the allocated generic DNBR margin. Therefore, the DNB acceptance criterion continues to be met.

RCCA MISOPERATION - USAR SECTION 15.4.3 The limiting Condition II RCCA misoperation transient is the dropped rod event

Attachment I to WM 87-0254 page 11 of 17 October 2,1987 assuming automatic rod control for which a power overshoot after the rod is dropped may occur. The T-avg and pressure for this transient are insensitive to small variations in reactor coolant flow. Heat flux, primarily a function of power, dropped rod worth, control bank ~ worth and moderator feedback, will not be significantly affected by the reduction in TDF. On this basis, flow will be the only significant ' change to 'the transient RCS conditions.

Allocation of generic DNBR margin will com)ensate for this DNB penalty. The same argument applies for the balance of tie Condition II RCCA misalignment accidents; therefore, the DNB acceptance criterion continues to be met.

Similarly, any DNBR penalty to the Condition III Statically Misaligned aCCA transient is offset by.the generic DNBR' margin allocation. _The calculated percentage of fuel failure will not change.

STARTUP 0F AN INACTIVE REACTOR COOLANT LOOP - USAR SECTION 15.4.4 Up until the point of minimum DNBR, the heat flux increases from 72% to'84.6%

of nominal. Concurrent with the increasing heat flux is an increase in core flow to 119.4% of~the initial transient value, an increase in RCS pressure and a relatively constant core average temperature. The net effect- is a small decrease in DNBR from the initial value. The minimum DNBR is significantly above the safety analysis design limit. Therefore, any DNB penalty due to the reduction in TDF is conservatively accommodated by the allocation of generic DNBR margin and the DNB acceptance criterion continues to be met.

BORON DILUTION EVENT - USAR SECTION 15.4.6 The boron dilution event in the USAR is analyzed for the six technical specification modes (i.e., dilution during REFUELING, COLD SHUTDOWN, HOT SHUTDOWN,HOTSTANDBY, START-UP,andPOWEROPERATION). A boron dilution event during refueling is precluded by administrative controls. The time to. lose shutdown margin, which is a measure of the boron dilution event's severity, is a function of dilution flowrate, RCS volume, boron concentration, boron worth, and shutdown margin. Therefore, the boron dilution event is insensitive to variations in thermal design flow. The conclusions currently in the USAR remain valid.

ACCIDENTAL DEPRESSURIZATION OF THE REACTOR COOLANT SYSTEM -

USAR SECTION 15.6.1 The assumption of automatic rod control maintains T-avg .at a constant value

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throughout this transient. The pressure transient is primarily a function of the safety valve relief rate and therefore is insensitive'to small variations in TDF. Heat flux remains _ constant throughout the transient. Therefore, there will be no significant changes to the transient RCS conditions except for flow. The current analysis minimum DNBR is well above the safety analysis design limit. Therefore, any DNB penalty due to the reduction in TDF is conservatively accommodated by the allocation of generic DNBR margin and the DNB acceptance criterion continues to be met.

THERMAL DESIGN FLOW - NON-DNB CONSIDERATIONS

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Attachment I to WM 87-0254 page 12 of 17 October 2, 1987 The following evaluations are presented for those incidents which are not DNB related or for which DNBR is not the only relevant safety criterion to be met: ,

i LOSS OF LOAD / TURBINE TRIP - USAR SECTION 15.2.2 AND'15.2.3 )

In addition to the Condition-II DNBR requirement., the USAR analysis for this-event must demonstrate that peak RCS pressure and peak ' steam pressure stay within 110% of design, 2750 psia and 1320 psia, respectively. Whether from loss of external electrical load or turbine trip, this- transient is characterized by an increase in core power .which exceeds the secondary side power extraction. The result is a primary side heat up and RCS pressure increase.

Of the four cases presented in the USAR, the maximusa RCS pressure is attained assuming minimum reactivity feedback and no availability of pressurizer spray or PORVs. . Reanalysis of this case using the LOFTl!AN (Reference 9) computer code with assumption changes reflectirg the reduced TDF indicate that the peak RCS and steam pressures continue to be well' within the . design limits.

Therefore, the current USAR conclusion that all appropriate safety analysis acceptance criteria are met, remains valid.

LOSS OF NON-EMERGENCY AC POWER.TO THE STATION AUXILIARIES -

USAR SECTION 15.2.6

. The purpose of this USAR transient is to demonstrate the adequacy of core I decay heat removal via natural circulation aided by auxiliary feedwater.

i Additional acceptance criteria are peak RCS pre'ssure and peak pressurizer water level.

l The USAR transient is initiated by the termination of feedwater flow. After 51.5 seconds, is reached the lo-lo steam generator level trip setpoint followed shortlybyreactorcoolantpump(RCP) coastdown. At 112.5 seconds, auxiliary feedwater is delivered. By 1850 seconds, the auxiliary feedwater heat removal capability exceeds core decay heat generation. The maximum pressurizer water level occurs at 2112 seconds.

The non-DNB acceptance criteria would not be significantly affected by the reduction in TDF. Peak pressurizer water vollume and the point in the transient where heat removal via auxiliary feedwater exceeds core decay heat both occur well after RCP coastdown. Peak pressurizer pressure is limited by pressurizer PORVs and spray which are assumed to be available in order to-maximize pressurizer water volume. Therefore,. it is concluded that the existing USAR results remain valid for the reduction in TDF.

LOSS OF NORMAL FEEDWATER - USAR SECTION 15.2.7 The ' current USAR analysis results demonstrate that peak RCS pressure remains within 110% of design (2750 psia), the pressurizer does not become water solid and the auxiliary feedwater is sufficient to remove core decay heat. The transient is initiated with the loss of normal feedwater followed by reactor trip on 10-10 Steam generator level and delivery of auxiliary feedwater.

Reactor coolant flow remains at the nominal value throughout the transient. A

i Attachment I to WM 87-0254 page 13 of 17- '

October 2,1987 i

reduction a coolant density decrease I in TDF would increase RCS heatup and thus a pressurizer volume increase.- The Wo catnin$f Creek USAR transient was f reanalyzed using the LOFTRAN code with the reduced TDF as well as conservative (

core residual heat generation based on the 1979 version of ANS-5.1 (Reference {

10). The resulting reduction in decay heat offsets any transient penalties to j the non-DNB acceptance criteria. The current USAR results and conclusions  !

remain valid for the reduction in TDF. l FEEDWATER SYSTEM PIPE BREAK - USAR SECTION 15.2.8 j The current USAR analysis results demonstrate that the peak RCS pressure is within 110% of design, the auxiliary feedwater.is sufficient to remove decay heat and the activity release requirements of 10 CFR 100 are met. Of the two _,

cases presented in the USAR, the case with offsite power available is H limiting. This case was reanalyzed using the LOFTRAN code with the reduced TDF and conservative core residual heat generation based on the 1979 version  ;

of ANS-5.1. The resulting reduction in decay heat offsets any transient penalties to the non-DN8 acceptance criteria. The current USAR results and conclusions remain valid for the reduction in TDF.

The USAR case without offsite power available has reactor trip on lo-lo steam generator level followed shortly by. reactor coolant pump coastdown at approximately 70 seconds. The non-DNB acceptance criteria critical points occur, in this case, after RCP coastdown. A check was made to ensure that -

sizeable margin exists, during the time in the transient before RCP coastdown, I between the analysis pressurizer relief and the maximum relief capacity. On these bases it is concluded that the reduction in TDF has no significant impact on this case and the USAR conclusions remain valid.

REACTOR COOLANT PUMP SHAFT SEIZURE - USAR SECTION 15.3.3 i This transient is characterized by an instantaneous seizure of a reactor 1 coolant pump rotor at 100% power. The flow reduction in the affected loop is i so rapid that the time of reactor trip on low flow does not change due to the )

reduction in TDF. However, the TDF reduction may result in increases in  ;

system pressure and fuel cladding temperature. The current USAR analysis demonstrates that peak RCS pressure remains within 110% of design and peak  ;

cladding temperature is less than 2700 degrees F. Reanalysis of this  ;

transient for the purpose of peak pressure and cladding temperature 4 calculations was made using the LOFTRAN and FACTRAN (Reference 11) com) uter codes. Results of the study showed an insignificant impact on peat RCS pressure and an increase in peak cladding temperature which remains well within the bounds of the safety analysis acceptance criteria limit. It is therefore determined that the conclusions in the USAR remain valid.

UNCONTROLLED RCCA BANK WITHDRAWAL FROM SUBCRITICAL - USAR SECTION 15.4.1 principal acceptance criterion for The this transient is DNB. Other calculated transient parameters include peak fuel cladding average temperature and peak fuel average temperature. A reduction in TOF would degrade heat transfer from the fuel and thus increase peak fuel and cladding temperatures.

Current USAR analysis values for these parameters are significantly below i

l Attachment I to WM 87-0254 page 14 of 17 October 2 , 1987 l

acceptance criteria limits. The peak clad average temperature is 711 degrees F and the peak average fuel temperature is 1935 degrees F. The reductions in TDF will not change the conclusions presented in the USAR.

SPECTRUM 0F RCCA EJECTION ACCIDENTS - USAR SECTION 15.4.8 The current USAR analysis' results demonstrate that the following acceptance criteria are met for the rod ejection transient:

Peak Cladding Average Temperature <2700 degrees F Fuel Melt <10% at hot spot Peak Fuel Pellet Enthalpy <200 cal /gm, irradiated

<225 cal /gm, fresh Four cases are presented in the USAR. They are beginning and end of liie hot full power and hot zero power. In the limiting hot full power case beginning j of life conditions are assumed and in the limiting hot zero power case end of

' life conditions are assumed. These two cases were reanalyzed with the reduced TDF using the TWINKLE (Reference 12) and FACTRAN computer codes. The results of the beginning of life - hot full power and end of life - hot zero power cases indicate that the current USAR conclusions remain valid for the reduced  !

TDF.

INADVERTENT ECCS AND CVCS MALFUNCTION - USAR SECTION 15.5.1 AND 15.5.2 Inadvertent operation of the Emergency Core. Cooling System during power ,

operation is analyzed to demonstrate that spurious ECCS operation without immediate reactor trip presents no hazard to the integrity of the RCS. The transient is characterized by decreasing nuclear power, decreasing core water temperature, decreasing pressurizer pressure and water level, and increasing DNBR. Small changes to steady state RCS flow would have no significant impact on the transient behavior. Therefore, the USAR conclusions remain valid for the reduction in Thermal Design Flow.

Malfunctions in the Chemical and Volume Control System which increase reactor coolant inventory are analyzed in the USAR to demonstrate that the operator has sufficient time to mitigate pressurizer filling. This transient is i

characterized by increasing pressurizer level and pressure and constant boron concentration. All the USAR cases exhibit relatively constant core power and  ;

RCS average temperature and increasing RCS pressure. Reduction of the TDF will not impact the transient behavior or change the conclusions presented in the USAR.

MASS AND ENERGY RELEASE ANALYSIS FOR POSTULATED SECONDARY PIPE RUPTURES INSIDE AND OUTSIDE CONTAINENT - USAR SECTION 6.2.1.4 AND WCAP-10961 The objective of the USAR analysis is to maximize the release of high energy fluid to the containment environment during a steamline ru)ture. A reduction in TDF reduces 3rimary to secondary heat transfer and t1erefore the energy release. As suc1, the energy releases presented in the USAR bound analysis results for a reduced TDF. The same conclusion applies for the steamline breakenergyreleasesoutsidecontainment(Reference 14).

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Attachment I to WM 87-0254 page 15 of 17 October 2, 1987 THERMAL DESIGN FLOW - PROTECTION SYSTEM SETPOINT CONSIDERATIONS Calculations have been performed to determine if the current safety analysis Overtemperature Delta-T (0 TDT), Overpower Delta-T (0PDT) and steam generator safety valve setpoints prevent violation of the revised core limits.

Adjustment to the safety analysis value of the OTDT K value is required to demonstrate protection of the revised core limits.y The balance of the setpoint assumptions made in the USAR safety analyses have not changed.

CONCLUSION The WCGS USAR safety analyses have been evaluated for the reduction in Thermal Design Flow from 382800 gpm to 372800 gpm and the revision of the part power multiplier from 0.2 to 0.3. Based on the above evaluation it has been determined that:

1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report has not increased. The proosed changes affect reactor core design parameters associated wit 1 accident mitigation or operational transients. These parameters are determined by design to protect the reactor core from exceeding safety limits. These changos do not affect initiators of an event that would change the probability of occurrence. The consequences have not been increased. The impact of these changes on the safety analyses has been evaluated. The evaluation results indicate that the safety analyses continue to meet required safety limits. .

2) The possibility for an accident or malfunction of a different type than any evaluated 3reviously in the safety analyses report has not been created. T1e reduced thermal design flow and the revised part power multiplier have been evaluated and found acceptable.

The changes affect parameters used in the reactor core design to protect the reactor core from exceeding safety limits assuming an accident or malfunction.

3) The margin of safety as defined in the basis for any technical specification has not been reduced. The changes proposed impact the analysis assumptions and inputs used in the safety analyses and the core power limits used in the reactor design. The affected safety analyses and core power limits have been evaluated and/or reanalyzed as necessary and it has been determined that all applicable safety criteria are met. Therefore, the margin of safety as defined in the bases has not been reduced. Affected Technical Specifications have been revised to reflect the reduced thermal design flow and the revised part power multiplier changes.

Based on the above discussions and the considerations presented in Attachment II, the proposed revisions to the WCGS Technical Specifications do not l

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-Att'achment I to WM 87-0254 page 16 of 17 October - 2,1987 ,

increase the probability of occurrence or the consequences of an accident or q malfunction of equipment important to safety previously evaluated in the' safety analysis report; or. create a possibility for an accident or malfunction of a. different type than any previously evaluated in the safety analysis

. report; or - reduce the margin of safety as defined in' the basis for.any technical specification. Therefore,. the proposed revisions do not' adversely affect or endanger the health or safety of the general public or involve a

=significant safety hazard.

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I Attachment I to WM 87-0254 page 17 of 17 October 2 ,1987 REFERENCES

1. Wolf Creek Updated Safety Analysis Report, Revision 0, March 11, 1987.

2. WCAP-9220-P-A (Proprietary), WCAP-9221 (Non-Proprietary), Eicheldinger, C., " Westinghouse ECCS Evaluation' Model - 1981 Version", Revision 1, 1981.

3. WCAP-8339 (Non-Proprietary),Bordelon, F.M., et al., " Westinghouse ECCS Evaluation Model - Summary", June 1974.

4. WCAP-9561-P-A (Proprietary) and WCAP-9695-A (Non-Proprietary), Young, M.Y. et al., "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," January 1980.

5. WCAP-9561-P-A Addendum 2, McIntyre, B.A., " Addendum To BART-A1: A Computer Code For The Best Estimate Analysis Of REFLOOD Transients (SpecialReport: NS-NRC-85-3025-A), July 1985. .

6. WCAP-9561-P-A Addendum 3, Revision 1, Young, M.Y., " Addendum To BART-A1:

A Computer Code For The Best Estimate Analysis Of REFLOOD Transients i (Special Report: Thimble Modeling In The Westinghouse ECCS Evaluation Model), July 1986.

7. WCAP-8471, (Proprietary) and WCAP-8472 (Non-Proprietary), Bordelon, F.M.,

et al., " Westinghouse ECCS Evaluation Model-Supplementary Information,"

1975.

8. WCAP-89'/0 (Proprietary) and WCAP-8971 (Non-Proprietary), " Westinghouse Emergency Core Cooling System Small Break October 1975 Model", April 1977.

9. Burnett, T.W.T, et al., "LOFTRAN Code Description", WCAP-7907-P-A, April l 1984. j 1

10. ANSI /ANS-5.1-1979, "American National Standard for Decay Heat Power in l Light Water Reactors", August 1979.

11. Hargrove, H.G., "FACTRAN - A FORTRAN-IV Code for Thermal Transients in a U02 Fuel Rod", WCAP-7908, June 1972.

12. Richer, D.H.,Jr., and Barry, R.F., " TWINKLE - A Multi-Dimensional Neutron 4 Kinetics Computer Code", WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary), January 1975.

13. WCAP-11145-P-A (Proprietary), Ruppercht, S.D. et al., " Westinghouse Small l' Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code", ,

October 1986, i

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ATTACIMENT II I i

SIGNIFICANT llAZARDS CONSIDERATION l l

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Attachment 11 to WM 87-0254 page 1 of 1 October 2,1987 j 1

SIGNIFICANT HAZARDS CONSIDERATION This amendment request revises Wolf Creek Generating Station, Unit No. 1, Technical Specifications to reduce the ReactorCoolagtSystemThermalDesign )

Flow from 382800 gpm to 372800 gpm and to revise the F part power multiplier j from 0.2 to 0.3. The proposed technicalspecificadon changes have been 1 evaluated and it has been concluded that:

1) The proposed changes do not involve a significant increase in the I probability or consequences of an accident previously evaluated.  !

The proposed changes affect reactor core design parameters associated with accident mitigation or operational transients.

These parameters are determined by design to protect the reactor l core from exceeding safety limits. These changes do not affect initiators of an event that would change the probability of occurrence. The consequences have not been increased. The impact of these changes on the safety analyses has been evaluated. The evaluation results indicate that the safety analyses continue to meet required safety limits.

2) The proposed changes do not create the possibility of a ner or different kind of accident from any previously evaluated. fhe reduced thermal design flow and the revised part power multiplier have been evaluated and found acceptable. The changes affect parameters used in the reactor core design to protect the reactor core from exceeding safety limits assuming an accident or malfunction. There are no new failure modes or mechanisms associated with the proposed revisions. This change does not involve any modification in the physical design of the installed  :

systems, but reflects revised values based on evaluations of existing analysis and reanalysis.

3) The proposed changes do not involve a significant reduction in a margin of safety. These changes do not affect any Technical Specification margin of safety. The changes proposed impact the analysis assumptions and inputs used in the safety analyses and the core power limits used in the reactor design. The affected safety analyses and core power limits have been evaluated and/or reanalyzed as necessary and it has been determined that all applicable safety criteria are met. Therefore, the margin of safety as defined in the bases has not been reduced. Affected Technical Specifications have been revised to reflect the reduced thermal design flow and the revised part power multiplier changes.

Based on the above discussions and those presented in Attachment I, it has been determined that the requested Technical S? edification revisions do not involve a significant increase in the proba]ility or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident over previous evaluations; or involve a significant reduction in a margin of safety. Therefore, the requested license amendment does not involve a significant hazards consideration.

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