ML18113A732

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Requests an Amend to Oper Lic in Form of Changes to Tech Specs Re Emergency Core Cooling Sys & Loss of Coolant Accident. Forwards Check for $4,400
ML18113A732
Person / Time
Site: Surry  
Issue date: 12/26/1978
From: Stallings C
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To: Harold Denton, Schwencer A
Office of Nuclear Reactor Regulation
References
NUDOCS 7812280132
Download: ML18113A732 (54)


Text

December 26~ 1978 Mre Harold R. Denton, Director Office of.. Nuclear Reactor Regulation Attn:

Mr,, Albert Schwencer, Chief Operating Reactors Branch No. l Divisi6n of Reactor Licensing U. S. Nuclear Regulatory Commissi6n Washington, DC

  • 20555

Dear Mr. Denton:

Serial No. 736 FR/MLB:jab Docket Nos.

50-280 50-281 License Nos.

DPR-32 DPR-37 AMENDMENT TO OPERATING LICENSE SURRY POWER STATION UNIT NOS. 1 AND 2

  • PROPOSED.TECHNICAL SPECIFICATIOWCHANGE NO~ 75*

Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company hereby requests an amendment, in the form ~f changes to the Technical Speci-fications, to Operating License Nos. DPR-32 and bPR-37 for the Surry Nuclear Power Station, Units *Nos. l and*2. The proposed changes are attached and have been designated as Change Nao 75.

The proposed amendment is in response to the Exempti.on and Order for Modification of License for Surry Units No. l and 2 (reference your letter of September 13, 1978)e As indicated in our letter of October 11, 1978 (Serial No. 535/091378), a LOCA-ECCS analysis and any required Technical Specifications changes, would be provided to demonstrate compliance to the requirements of 10 CFR 50.46 with a calculational model which fully conforms to the provisions of Appendix K, 10 CFR 50.

This analysis and required Tech-nical Specifications are provided in Attachments l and 2, respectively.

The proposed changes have been determined to be a Class Ill amend-ment for Unit le The amendment involves a single safety issue and does not involve a significant hazards consideration.

  • The proposed change is Class l for Unit 2, si~ce it is duplicated. Accordingly, a check in the amount of

$4400.00 is. attached in payment of the $4006.oo and $400.00 fees.

This proposed change has been reviewed and approved by both the Station Nuclear Safety and bperating Committee, and the System Nuclear Safety and Operating Committee.

It has b~en determined that this request does not involve an unreviewed safety question, as defined in 10 CFR 50.59.

7s122so13c_

p

e VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton Your review and approval of the attached Technical Specifications Change is requested by January'2, 1979.

Should you have questions, we would like to meet with you at your earliest convenience.

Attachments:*

1. Safety Analysis
2.

Proposed Technical Sped f i cations

  • 3. Voucher check no. 8493, in the amount of $4400.00'.

cc:

Mre James*P. O'Reilly, Director Office of Inspections and Enforc*ement Region I I 2

COMMONWEALTH OF VIRGINIA CITY OF RICHMOND

)

) s. s.

)

Before me, a Notary Public, in and for the City and Common-wealth aforsaid, today personally appeared Sam C. Brown, Jr., who being duly sworn, made oath and said (1) that he is Vice President-Power

  • Station Engineering and Construction, of the Virginia Electric and Power Company, (2) that he is duly authorized to execute and file the fore-going Amendment in behalf of that Company, and (3) that the statements in the Amendment are*true to the best of his knowledge and belief.

Given under my hand and notarial seal this 6A1bday of Pc:c:mher:

, 1:2.Zs.*

  • My Commission expires Jar2//q5,Y ZtZ; )'?$/.

~

Notary Public (SEAL)

r--

  • W ATTACHMENT 1

1.0 INTRODUCTION

e e

A reanalysis of the ECCS cooling performance for the postulated large break Loss Of Coolant Accident (LOCA) has been performed which is in compliance with Appendix K to 10 CFR 50.

The results of this reanalysis are presented herein* and are in compliance with 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling was performed with the NRC Systems for Light Water Reactors.

This reanalysis.

(1) approved February, 1978 version of the Westing-house LOCA-ECCS evaluation model.

The analytical techniques used are in full.*

compliance with 10CFR50~ Appendix Kand satisfy the requirements of Reference 2.

As required by Appendix K of 10 CFR 50, certain conservative as-sumptions were made for the LOCA-BCCS analysis.* The assumptions pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur and include such items *as the core.

peaking factors, the containment pressure, and the _performance of the emer-gency core cooling system (ECCS).

All assumptions and initial operating condition input data used in this reanalysis were the same as was used in the previously applicable LOCA-;ECCS analysis(3) except*for l) the limiting value of the heat flux hot channel factor.was increased to 2.05, 2) the*core inlet temperature value was decreased to 534.S°F, 3) the value of initia_l fuel temperature was increased in that the temperature was calculated based on generic values of fuel characteristics rather than as-built values, 4) the change of several ECCS containment parameters. was made to reflect the containment response in a more realistic, but still conservative, manner, and 5) the assumed steam generator tube plugging level was increased to 28 percent.

  • It should be noted that reanalysis of the small break LOCA is not necessary, and therefore, the analysis of this accident submitted by Reference 4 remains applicable.

2.0 DESCRIPTION

OF PO&ATED MAJOR REACTOR C~OLANT. P.RUPT~RE (LOSS OF.

COOLANT ACCinENT LOCA)

A LOCA is the result of a rupture of the Reactor Coolant System (RCS) piping or of any line connected to the system:. The system boundaries 1

f.

AR

. d.

(S) considered 1.n the LOCA ana ys1.s are de 1.ned i.n the FS

.

  • Sens1.t1.v1.ty stu 1.es have indicated that a double-ended cold leg guillotine (DECLG) major break is limiting.

Should a DECLG break occur, rapid depres'surization of the RCS

. occurs.

The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached.

A Safety Injection System (SIS)signal is actuated when the appropriate setpoint is reached and the high head safety injection pumps are activated.* The actuation and subsequent activation of the ECCS, which occurs with the SIS signal, assumes the most limiting single failure event.

These countermeasures will limit. the consequences of the accident in two ways:

1.

Reactor trip and borated water injection complement void*for-

  • mation in causing rapid reduction of power to a residual level correspon~ing to fission product decay heat.* (Ii.~hould be noted, however, that no credit is taken in the analysis for the insertion of co.ntrol rods to shut down the reactor.)
2.

Injection of borated water provides heat transfer. from the core and prevents excessive clad temperatures.

Before the.break occursj the unit is*in an equilibrium condition, i.e., the heat generated in the* core is being removed via *the secondary system.*

During blowdown, heat from decay, ho't internals and the vessel continues to be transferred to the r.eactor-coolant system.

.At the beginning of the blow-down phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling.

Aft~r the b~eak devel~ps, the time to departure fro,m nucleate.boiling is calculated, consistent.with Appendix K of 10CFR50 *. Thereafter, the core heat transfer is based on local conditions with transition boiling arid forced convection of steam as the major heat transfer mechanisms.* During the refill period, it is assumed that rod-to-rod radiation is the only core heat transfer

mechanism.

e.

The heat transfer between the reactor coolant system and the secondary system maybe in either direction depending ori the* relative temperatures.

For the case of continued heat addition to the secondary side, secondary side pressure increases and the main safety valves may actuate to reduce the pressure.

Make-up to the secondary side is automatically provided by the auxiliary.*

feedwater system.

Coincident with the Safety Injection Signal, normal feed-water flow is stopped by closing the main feedwater control valves and tripping the main feedwater pumps.

Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps, The secondary side flow aids*in the reduction of reactor coolant system pressure.

When the reactor.coolant. system depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops.

The.conservative assumption is then made that injected accumulator water bypasses the core and goes out through the break until the termination of bypass.

This conservatism is again consistent with Appe~dix K of 10CFR50.

In addition, the reactor coolant pumps are assumed to be tripped at the initiation of the accident and effects of pump coastdown are included in the 'blowdown analysis.

The water injected by the accumulators cools the core and subsequent operation of the low head safety injection pumps supply water for long term coolin~.

When the RWST is nearly empty, long term cooling of the core *is*

accomplished by switching to the recirculation mode of core coqling, in which the spilled borated water is drawn from the containment sump *by the low head safety injection pumps and returned _to the reactor vessel.

The containment spray system and the recirculation spray system operate to return the containment environment to a subatmospheric pressure.

The large break LOCA transient is divided, for analytical purposes, into three phases:

blowdown,: refill, and reflood.

There are three distinct transients analyzed in each phase, including the thermai-hydr~ulic transient in the RCS, the pressure and temperature transient.within th~ containment,*

i i

I -

~

and the fuel clad temp-tu~e transient of the hottest-el rod in the core.

Based on these considerations, a system of inter-related computer codes has -

been developed for the analysis of the LOCA.

The description of the various aspects of the LOCA analysis methodology 1s_ given in WCAP-8339.(G)

This document describes the major. phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with 10 CFR 50, Appendix K.

The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, *are described in detail in WCAP-8306 (7), WCAP-8326 (S), WCAP-817i<g) and WCAP-8305 (lO), respectively.

These codes are able to assess whether sufficient heat transfer geometry and*

core amenability to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA. - The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown and_

the WREFLOOD computer code is used to calculate the_ transient during the refill and re flood phases of the accident.

The COCO computer code is used to calculate the containment pressure transient during-all thre~_-phases of the LOCA analysis.

Similarly, the LOCTA-IV computer code is used to compare the thermal transient of the hottest fuel rod during the.three phases.

SATAN~vr is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the RCS and steam generator secondary,

  • _as a function of time during the blowdown phase of the LOCA.
  • SATAN-VI also calculates the accumulator mass and pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown.

At the end of the blowdown, the mass and energy release rates during blowdown are transferred to the COCO code for use in the determination of the containment pressure. response during this first phase of the LOCA.

Additional SATAN-VI output data from the end of blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.

\\

With input a the SATAN-VI code, ~FLOO~.s a system thermal-hydraulic model to determine the core flooding rate* (i.e.,.the rate at which coolant enters the bottom of the core), the coolant pressure and temperature,*

and the quench front height during the refill and reflood phases of the LOCA.

WREFLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment.

Sinc.e the mass, flow rate to the containment depends upon the core flooding rate and the local core pressure, which is a function of the containment back.pressure, the WREFLOOD and COCO codes are*

  • interactively.linked.. WREFLOOD is also linked to the LOCTA-IV code in that thermal-hydraulic parameters from WREFLOOD are used by LOCTA-IV in its calculation of the fuel temperature.*

LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel and clad temperature of the hottest rod in the core. :The input to LOCTA-IV consists of appropriate ~hermal-hydraulic output from SATAN-VI and WREFLOOD and conservatively selected initial RCS operating conditions.

These init_ial conditions are summarized in Table* 1 and Figure 1.

(The axial **

  • power shape of Figure.l.assumed for LOCTA-IV is.a cosine curve lmich has been previously verified to be the shape that produces the maximum peak clad"tem-(11) perature

.)

The COCO code, which is also used throughout the LOCA analysis, calculates the containment pressure.

Input to COCO is obtained from the mass and energy flow rates assumed to be vented to the containment as calculated by the SATAN-VI and WREFLOOD codes.

In addition, conservatively chosen initial containment.conditions and an assumed mode of operation.for the containment cooling system are input to coco.* These initial containment conditions and assumed modes of operation are* provided in Table 2.*

I I

3.0 niscussroN OF sAFICANT INPUT Significant changes in the input used in this reanalysis from those used in the currently applicable analyses are delineated in Section 1. 0 and are discussed in more detail below.

The changes made in this analysis reflect the operational conditions and limits necessary to allow full power operation at a steam generator tube plugging level of up to 28%.*

In order to ensure compliance with the _10 CFR 50.46 acceptance criteria, several changes to the operational limits assumed

\\

in the analysis were made.

Specifically, the assumed value of the heat flux

  • hot channel factor was increased from its current value of l.94 in Unit _1 and 1.79 in Unit 2 to a value of 2.05 for both Units 1 and 2.

Changes were also made to the fuel temperature, reactor coolant temperature, and containment

. structural heat sinks.

The previous analysis had been p';?rformed using initial fuel temperatures.

which were derived from as-built fuel parameters.

While this assumption was conservative for the current operating cycles, the possibility did exist. that.*

future reload cycles would have different as-built fuel parameters which could result in slightly nonconservative initial fuel temperature.

In order to*

insure conservative results for all futul:'.e reload cycles, tne initial fuel temperatures for this analysis were calculated based on limiting generic 15xl5.

fuel parameters.

The value assumed for the reactor coolant system core iniet tem-perature for this analysis was changed to be consistent with the overall conservatism*irtherent in the analysis.

Specifically,.a core inlet temperature 0

of 534.5 F was _used in the analysis.

This value is the best-estimate core

This assumption accurately approximates the actual steam generator tube plugging distribution.

The impact of non-symmetric 12 plugging distribution on the LOCA-ECCS analysis has been found to be insignificant

inlet temperature as d-rmined from operational data -

is adequate to encompass the applicable s*team generator tub.e plugging* range.

The amount of the various categories of structural heat. sinks pro-vided in Table 2 were reviewed in detail.

Based on the as-built plant con.tainment, these categories were conservatively revised and credit was taken for carbort steel painted surfaces.. The remainder of the containment initial conditions and pressure*reductian systems parameters were the same as used in the previous analyses.

Finally, this analysis was conducted with the February, 1978 version of the Westinghouse LOCA-ECCS Evaluation Model (lJ,l4,is)

This model version includes a modification to the SATAN VI and LOCTA IV codes to correct this calculation to properly account for the volumetric heat generation due to

06) the metal-water reaction

4.0 RESULTS e

Table 3 p~esents the time sequence of events *and. Table 4 presents the results for the double-ended cold leg guillotine break (DECLG) for the CD= 0.4 discharge coefficient.

The DECLG has been determined to be the limiting.break size and location based on the sensitivity studies reported in Reference 5.

  • Based on all previous LOCA-ECCS submittals for the Surry units, the results* obtained with a c0 = 0.4 discharge coefficient have always been limiting.

The applicability of 'this conclusion (i.e. c0=0.4 is the limiting break size) for this analysis was verified.

Assuming the initial conditions and modes of operation presented in Tables 1 and 2 and.Figure 1, the current analyses resulted in a peak clad temperature of 2172 °F, a maximum local cladding oxidation level of 7.81 *percent, and a total core metal-water reaction of less than 0.3 percent.

The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2 through.18.

5.0 CONCLUSION

S e

)

For breaks up to and including the double-ended severance of a reactor coolant pipe and for the operating conditions specified in Tables

(!

1 and 2, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFRS0.46.

That is:

1.

The calculated peak fuel rod clad temperature.is below the requirement of 2200°F. *

2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.

The localized cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

The core remains amenable to cooling during and after the break.

5.

The core temperature is reduced and the long-term decay heat is removed for an extended period of time.

'... (.*

6.0 REFERENCES

1.

Letter from NRC (J. F. Stolz) to Westinghouse (T. M. Anderson) dated August 29, 1978.

2.

Letter from NRC (A. Schwencer) to Vepco (W. L'. Proffitt) dated September.

13, 1978.

3.

Letter from Vepco (C. M. Stallings) to NRC (E. G. Case) dated May 26, 1978, Serial No. 303.

4.

Letter from Vepco (C. M. Stallings) to NRC (K. R. Goller), Serial No.

5oo~s,dated June 6, 1975.

5.

Buterbaugh, T. L., Johnson, W. J. and Kopelic, S. D., "Westinghuse ECCS-Plant Sensitivity Studies," WCAP-8356, July 1974.

6.

Bordelon, F. M., Massie, H. W., and Zordan, T. A., "Westinghouse ECCS Evaluation Model-Summary" WCAP-8339, July 1974.

7.

Bordelon, F. M., et al., "SATAN-VI Program:

Comprehensive Space-Time.

Dependent Analysis of Loss-of-Coolant,n WCAP-830~, Jurte--1~?.~*

~-'"'**:-*.,

~

  • 8.

Bordelon, F. M. and Murphy, E.T., "Containment Presimre.~nalysis Code (COCO)," WCAP-8326, June, 1974

.. -... ~*--00:-:.. ;.... *......

  • 9. :,Kelly, R. D., et al., ncalculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code)," WCAP-8171, June 1974..
10.

Bordelon, F. M., et al., "LOCTA-IV Program:

Loss-of-Coolant Transient Analysis," WCAP-8305, June, 1974.

11.

Letter from Vepco (C. M. Stallings)* to NRC (E. G. Case) dated February 17, 1978, Serial No. 092.

12.

Letter from Vepco (C. M. Stallings) to NRC (B. C. Rusche), Serial No.**

260/092276 ~* dated October.19, 1976.

13.

"Westinghouse ECCS Evaluation Model - February, 1978 Version'\\ WCAP-9220:

P-A (Proprietary) and WCAP-9221-P-A (Non-Proprietary), February 1~978.

14.. Letter. from Westinghouse (T. M. Anderson) to NRC (J. F.

  • Stolz),.dated November 1, 1978, Serial No. NS-TMA-1981.
15. Letter. from Westinghouse (T. M. Anderson) to NRC (R. Tedesco), dated December 11, 1978, Serial No. NS-TSM-2014.
16. Letter from Westinghouse (C. Eicheldi_nger) to NRC (J. F~ Stolz), dated April 7, 1978, Serial No. NS-CE-1751.

e TABLE 1 INITIAL RCS CONDITIONS Core Power, Mwt, 102% of Peak Linear Power, Kw/ft, 102% of Peaking Factor (FQ)

Accumulator Water Volume, ft 3 Reactor Coolant System Flow, gpm (90% of Thermal Design)

Steam Generator Tube Plugging Level, %

Inlet Temperature, Temperature of the Fluid in the Upper Head Region of the Reactor Vessel Fuel Temperatures Hot Assembly Radial Peaking Factor Hot Rod Radial Peaking Factor Most Limiting Fuel Region Unit 1 Unit 2 Cycle All All

. 2441 12.72 2.05 975 (per accumulator) 28 534.5 100% of TROT Generic lSxlS 1.38 1.45 Region All All

~.

TABLE 2

  • ~AINMENT DATA (DRY CONTAINMENt

. NET FREE VOLUME INITIAL CONDITIONS Pressure Temperatllre RWST Temperature Service Water Temperature

. Outside Temperature SPRAY SYSTEM I CONTAINMENT SPRAY SYSTEM Number of Pumps Operating

. _Runout Flowrate Actuation Time SPRAY SYSTEM II - INSIDE RECiRCULATION SPRAY SUBSYSTEM

  • Number Pumps Operating Runout Flowrate (each)

Actuation Time Heat Exchanger (UA (per pump))

Service Water Flow (per exchanger).

SPRAY SYSTEM II - OUTSIDE-RECIRCULATION SPRAY SUBSYSTEM Number Pumps Operating Runout Flowrate (each)

Actuation Time*

Heat Exchanger (UA (per pump))

Service Water Flow (per exchanger)

1. 863x10 6 F~ 3 9.35 psia 90°F 40°F 32.S°F 9°F 2

3200 gpm

. 52 secs 2

3500*gpm 190 secs 5.18xl06 BTU/HR-°F

  • 6900 gpm.

2 2250 gpm_

410 secs 5.18xl06 BTU/HR-°F 6900 gpm

1--~---~~ ---~--=-=---*

STRUCTURAL HEAT SINKS Type/Thickness (in.)

Concrete 6 Concrete 12 Concrete 18 Concrete 24 Concrete 27 Concrete 36 Carbon Steel 0.375 Concrete 54 Carbon Steel 0.50 Concrete 30 Concrete 26 (Floor)

Carbon Steel 0.239 Stainless Steel 0.306 Aluminum 0.0091 TABLE 2 (Continued) -

Area (ft2), w/uncertainty 8,393 62,271 55,365 11,591 9~404 3,636 46,489*

25,652*

12,110 158,059*

  • 17,519

. 3~ 911

  • Gredit for painted surfaces was taken only-for the nominal surface area

I

-- ~--- ----

e TABLE 3 TIME SEQUENCE OF EVENTS START Reactor Trip S. I. Signal Acc. Injection End of Bypass Pump Injection End of Blowdown Bottom of Core Recovery.

  • Acc. Empty DECLG CD=0.4 (Sec) 0.0 0.538 2.32 15.4 24.45 27032 28.01 36.15 44.67

--*** *~ *...

r.

,.,~... '

e TABLE 4 RESULTS FOR DECLG CD=0.. 4.

0 Peak Clad Temp, F

Peak Clad Location, Ft.

Local Zr/H20 RXN (max),%

Local Zr/H20 Location, Ft.

Total Zr/H20 RXN, %

Hot Rod Burst Time, sec.

Bot Rod Burst Location,. Ft.

2172 7.75 7.81

. 7. 75

<o.3 26.20

  • 6.0

TIME (SEC)

41. 76 50.71 65.21 82.91
102.91 124081 174oll 233.21 312.96 TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLG (CD= 0.4)

TOTAL MASS FLOWRATE (LB/SEC) 36.14 108.20

. 218. 77 258.47 269.19 277.03 291.36 311. 22 330.02 TOTAL ENERGY

-FLOWRATE (105BTU/SEC)

  • 0.467 1.085 1.331 1.371 1.331 1.280 1.171 1.029 0.895

TABLE 6 BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT DECLG, Cn=0.4.

TIME (SEC) 0.0 1.0 3.0 5.0 7.0 lOoO 15o0 20o0 22o0 22.86

  • For energy mass flowrate multiply mass flowrate by a constant of 58. 82.BTU/LBM.

MASS FLOWRATE*

o.o 4205 3590 3185 2892 2566 2189 1934 1857 OoO (LBm/SEC)

10 X 10 TO THE CENTIMETER KEUF"F"EL & ESSER CO *.,,ot IN U.S.A.

10 X 25 ';M*

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ATTACHMENT 2

e TS 3.3-1 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System.

Objective To define those limiting conditions for operation that are necessary to provide,.*

sufficient borated cooling water to remove decay heat from the core in e~ergency situations.

Specifications A.

A reactor shall not be made critical unless the following conditions are met:

1.

The refueling water tank contains not less than 350,000 gal. of borated water with a boron concentration of at least 2000 ppm.

2.

Eachaccumulator system is pressurized to at least 600 psia and con-tains a minimum of 975 ft 3 and a maximum of 989 ft3 of borated water

  • I with a boron concentration of.at least 1950 ppm.
3.

The boron injection tank and isolated portion of the inlet and outlet piping contains no less than 900 gallons of water with a boron concentration equivalent to at least 11.5% to 13% weight boric acid solution at a temperature of at least 145°F.

Additionally, recirculation between a unit's Boron Injection Tank and the Boric Acid Tank(s) assigned to the unit shall be maintained.

e Fq(Z) ~2.05/P x K(Z) for P >.s Fq(Z) _.:: 4.10 x K(Z) for P <.5 F~H ~ 1.55 (1 + 0.2(1-P)) x T(BU)

FN I LOCA

_< 1. 38/P llH Assm.

FN I LOCA

< 1.45/P llH Rod TS 3.12_-4 where Pis the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, Z is the core height location of Fq, and T(BU) is the interim thimble cell rod bow penalty on F~H given in TS Figure 3.12-9.

2.

Prior to exceeding 75% power following each core loading, and during each effective full power month of operation thereafter, power distribution maps using the movable detector system, shall.be made to confirm that the hot channel factor limits of this specification are satisfied.

For the purpose of this confirmation:

a.

The measurement of total peaking factor, F~eas, shall be increased by eight percent to account for manufacturing tolerances, measure-ment error,- and the effects of rod bow..

The measurement of enthalpy rise hot channel factor, *the hot assembly enthalpy rise factor, N ILOCA N I LOCA F t.H Assm., and the hot rod enthalpy n.se factor, F t.H Rod J shall be increased by four percent to account for measurement error. If any measured hot channel factor exceeds its limit specified under 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under 3.12.B.1 are met.

If the hot channel factors cannot be brought to within the limits Fq _<2.05 x K(Z), F~H-<1.55 x T(BU), yN ILOCA<l 45 and u

llH Rod -

F~HI~~~!.~ 1.38 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower 6T and Overtempera-ture llT trip setpoints shall be similarly reduced *.

Deleted

\\

e TS. 3.12-4a.

b.

Fq(Z) shall be evaluated for normal (Condition I) operation of Unit 2 by combining the measured values of, FxyCZ) with the design.

Condition I axial peaking factor values, Fz(Z), as listed in TS Table 3.12-lB.

For the purpose of this specification Fxy(Z) shall be determined between 1.5 feet and 10.5 feet elevations of the core_.

exclusive of grid plane regions located at 25.9 +/-3.2 inches, 52.1 +/-3.2 inches, 78.3 +/-3.2 inches, and 104.5 +/-3.2 inches..

  • The measured values of Fxy(Z) shall be increased by nine percent to account for manufacturing tolerances, measureme:nt error, rod bow,*
    • .xenon redistribution, and any burnup dependent peaking factor increases.

If the results of this evaluation predict that FQ(Z) could potentially violate its limiting values as established in Specification 3.12.B.1, either:

(1) the thermal power _and high neutron flux trip setpoint shall

  • be reduced at least 1% for each 1% of.the potential violation (for the purpose of this specification, this power level shall be called PTHRESHOLD), or (2) movable 4etector surveillance shall be required for operation*

when the reactor thermal power exceeds PTHRESHOLD*

This sur-veillance shall be performed in accordance with the following: - *

(a) The no~rmalized power distribution, !q(Z) I hnM' from.*

thimble j at core elevation Z shall be measured utilizing at least two thimbles of the movable incore flux system for i

1

~

e

.e TS 3.12-5

3.

The reference equilibrium indicated axial flux difference (called..

the target flux difference) at a given power level P0, is that indicated axial flux difference with the core in equilibrium xenon conditions (small or ~6 oscillation) and the control rods more than 190 steps withdrawn.* The target flux difference at any other power level, P, is equal to the target value of P multiplied by the ratio, P/P0

  • The target flux difference shall be measured at least once
  • . per equivalent full"" powel'. quarter.

The target flux difference must be updated during each effective full power month of operation either by actual*measurement, or by linear interpolation using the most recent value and the value predicted for the end of the cycle life.

4. :* Except as modified by. 3.12.B.4.a,,b, c, or d. below/ the. indicated

. *axial flux difference shall be maintained within a +/-5% band about*

the target flux difference (defines the target band on axial flux difference).

a. At a power level greater than 88 percent of rated power, if the.indicated axial flux difference deviates from its target band, within 15 minutes either restore the indicated axial flux difference to within the target band, or reduce the reactor power to less than 88 percent of rated power.
b. At a power level no greater than 88 percent of rated power, (1)

The indicated axial flux difference may deviate from its target band for a maximum of one hour (cumulative) in any 24-hour period provided the flux difference is within the limits shown on Figure 3.12-10.

I

,-----,-------:--------------,,-------------,-------~,cc-.-.-..,

TS 3.12-6 e

e One minute penalty is accumulated for each one minute of operation outside of the target band at power levels equal to or above 50% of rated power~

(2)

If 3.12.B.4.b(l) is violated, then the reactor power shall be reduced to less than 50% power within 30 minutes and*

the high neutron flux setpoint shall be reduced to no greater than 55% power within the next four hours.

(3). A power increase to a level greater than 88 percent of rated power is contingent upon the indicated axial flux difference being within its target band.

(4)

Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Table 4.1-1 provided the indicated AFD is maintained within the limits of Figure 3.12~10.

A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of operation may be accumulated with the AFD outside of the/target band during this testing

. without penalty deviation.

c. At a power level no greater than 50 percent of rated power, (1)

The indicated axial flux difference may deviate from its target band.

(2)

A power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than one hour accumulated penalty during the preceding 24-hour period.

One half minute penalty is accumulated for each one minute of operation outside of the target band at power' levels between 15% and 50% of rated power.

d. The axial flux difference limits of Specifications 3.12.B.4.a, b, and c may be suspended during the performance of physics tests provided:

(1)

The power level is maintained at or below 85% of rated power, and (2)

The limits of Specification 3.12.B.1 are maintainedc The power level shall be determined to be. < 85% of rated power at least once per hour during physics tests. Verification that the limits of Specification 3.12.B.1 are being met shall be demonstrated through in-core flux mapping at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I

TS 3.12-7 e

Alarms shall normally be used to indicate the deviations *from the axial flux difference requirements in 3.12.B.4.a and the.

flux difference time limits in 3.12.B.4.b and c.

  • If the alarms. are out of s.ervice temporarily, the axial flux difference shall be logged, and conformance to the limits assessed,*every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.*.*

The indicated axial flux difference for each excore channel shall be monitored at least once per 7 days when _the alarm is operable and at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the alarm to operable status *

. 5.

The allowable quadrant to average power tilt is 2.0%.

6.

If, except for physics and rod exercise testing, the quadrant to average power tilt exceeds 2%, then:

a.

The hot channel factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the power level adjusted to meet the specification of 3.l2oB.1, or

b.

If the hot channel factors are not determined within two*

hours, the power level and high neutron flux trip setpoint shall be reduced from rated power, 2% for each percent of quadrant tilt.

c.

If the quadrant to average power tilt exceed_s * +/-10%, the power level and high neutron flux trip setpoint will be reduced from rated power, 2% for each percent of quadrant tilt.

e

.TS 3.12-14 Fq(Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F~, Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.

The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad.

Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

F~H' Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power for both LOCA and non-LOCA considerations.

F~Hlx~~!., Hot Assembly Nuclear Enthalpy Rise Factor, is defined as the ratio of the integral *of linear power along the assembly with the highest integrated power to the average assembly power.

It should be noted that the enthalpy rise factors are based on integrals and are used as such in the DNB and LOCA calculations.* Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power ~hapes throughout the core.

Thus the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factors.

The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using an upper bound envelope of 2.05 times the hot channel factor normalized operating envelope given by TS Figure 3.12-8.

e TS *3.12-14a DELETED

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1.

i e

TS 3.12-17 For normal (Condition I) operation of Unit 2, it may be.necessary to perform surveillance to insure that the heat flux hot channel factor, Fq(Z), limit is met.

To determine whether and at what power level surveillance is required, the potential (Condition I) values of Fq(Z) shall be evaluated monthly by com-bining the measured values of Fxy(Z) obtained from the analysis of the monthly incore flux map with the values of the design Condition I axial peaking factors, Fz{Z) e The product of these shall be inc*reased by nine percent to account for measurement uncertainty, manufacturing tolerances, rod bow, radial redistribution of xenon during normal (Condition I) operation, and any burnup dependent peak-ing factor increasesQ PTHRESHOLD is defined as the value.of rated power minus one perc~nt power for each percent of potential FQ_(Z) violation. If the potential values of Fq(Z) for normal (Condition I) operation are greater.than the Fq(Z) limit, then surveillance shall be performe~ at all power levels above PTHRESHOLD*

Movable incore instrumentation thimbles for surveilla:nc:e--are selected so that the measurements are representative of the peak core power density.

By limiting the core average axial power distribution,.the* total.power peaking factor FQ(Z) can '!>e limited since allother components remain relatively fixed.

The remaining part of the total power peaking factor can be derived based on incore measurements, i.e., an effective radial peaking factor, R, can be determined as the ratio of the total peaking

.e TS 3.12-21 power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power. Strict control of the flux difference is not always possible during certain physics tests or during' excore detector calibrations. Therefore, the specifications on power distribution control are less restrictive during physics_ tests and excore*

detector calibrations; this is acceptable due to the low probability ~fa signi-ficant accident occurring during these operations.

In some instances of rapid unit power reduction automatic rod motion will

\\

cause the flux difference to deviate from the target band when the reduced power level is-reached.

This does not necessarily affect the xenon dis-tribution *sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band; however,. to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the bande This ensures

  • that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.

The instan-.

taneous conseqri~nces of being outside the band, provided rod insertion limits are observed, is hot worse than a 10 percent increment in.peaking factor for the allowable flux ~ifference at 88%.power, in the range ;!:13.-5 percent ~10.. 5 percent indicated) where for every 2 percent below rated power, the permissible flux difference boundary is extended by 1 percent.

As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition

~-.

e TS. Table 3.12-lA DELETED.

TS Table 3.12-2 e

DELETED

1.0.

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N H i 0.6 0 z I N

~ 0.4

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e HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION UNIT NOS. 1 AND 2

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8 10 12 CORE HEIGHT (FT.)

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AXIAL FLUX DIFFERENCE LIMITS AS-A FUNCTION OF RATED POWER FLUX DIFFERENCE ~I)%

TS FIGURE 3.12-10