ML19319B519

Attachment 1 to Application to Amend OL for Removal of Burnable Poison Rod & Orifice Rod Assemblies.

ML19319B519
Person / Time
Site:	Davis Besse
Issue date:	05/26/1978
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19319B515	List: ML19319B514 ML19319B519
References
BAW-1489, NUDOCS 8001230637
Download: ML19319B519 (34)

Text

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BAk'-1489 May 1978 Rev. 1 S/26/78 ATTACHMENT 1 TO APPLICATION TO AMEiD OPERATING LICENSE FOR RDiOVAL OF BURSABLE POISON ROD AND ORITICE ROD ASSDiBLIES

- Davis-Besse Nuclear Generating Station, Unit 1-BABCOCK & k'ILCOX Power Generation Group Nuclear Pcuer Generation Division P. O. Box 1260 ,

Lynchburg, Virginia 24505

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Rev. 5/26/78 abcock & wticcx 8M M30

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CONTENTS Page

1. INTRODUCTION AND S'OSIAP.Y . . . . . . . . . . . . . . . . . . . . . 1-1

2. GENERAL DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . 2-1 a

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3. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . . . . . . 3-1 5

3.1. Fuel Assembly Mechanical Design . . . . . . . . . . . . . . 3-1 3.2. Thermal Dasign . . . . . . . . . . . . . . . . . . . . . . . 3-1 j 3.3. Primary Neutron Source Retainer . . . . . . . . . . . ... . 3-1 1

4. NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 t

4 .1. Physics Characteristics . . . . . . . . . . . . . . . . . . 4-1

5. THERM.'.L HYDP.AULIC DESIGN . . . . . . . . . . . . . . . . . . . . . 5-1

6. ACCIDENT. AND TR/.NSIENT ANALYSIS . . . . . . . . . . . . . . . . . 6-1 6.1. General Safety Analysis . . . . . . . . . . . . . . . . . . 6-1 6.2. Rod Withdrawal Acciden

. . . . . . . . . . . . . . . . . . 6-1

6.3. Moderator Dilution Acciden: . . . . . . . . . . . . . . . . 6-2 6.4. Cold *n'ater (Pu=p Startup) Accident . . . . . . . . . . . . 6-3 6.5. Loss of Coolant Flow . . . . . . . . . . . . . . . . . . . 6-3 6.6. S tuch-Ou t , Stuck-In, or Dropped Red Acciden: . . . . . . . 6-4 a 6.7. Loss of Electric Power . . . . . . . . . . . . . . . . . . 6-4 .

i 6.8. Stea= Line Failure . . . . . . . . . . . . . . . . . . . . 6-5 i

6.9. Stea Generator Tube Failure . . . . . . . . . . . . . . . 6-3 6.10. Fuel Handling Acciden: . . . . . . . . . . . . . . . . . . 6-3 6.11. Rod Ejection Accident . . . . . . . . . . . . . . . . . . . 6-5 6.12. Maximum Hypothetical Acciden: . . . . . . . . . . . . . . . 6-6 6.13. Was:e Gas Tank Rupture . . . . . . . . . . . . . . . . . . 6-6 6.14. LOCA Limits . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.15. Loss of Normal Feedwa:er. . . . . . . . . . . . . . . . . . 6-7 l 7. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . . . . . 7-1 S. STAR"JP PROGP.AM - PHYSICS TESTINO . . . . . . . . . . . . . . . . 3-1 S.I. Precri:ical Tests . . . . . . . . . . . . . . . . . . . . . S-1 S.2. Eero Power Physics Tests . . . . . . . . . . . . . . . . . .

8.3. Power Escalation Tests . . . . . . . . . . . . . . . . . . .

S-la E-3 l1 8.4 Procedure if Acceptance Criteria Are Not Me: . . . . .. . . S-6 ,

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! . . REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . .. . . A-1 1

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. List of Trblss f Table .

Page 3-1. Fuel Design Parameters and Dimensions . . . . . . . . . . ...... 3-1 4-1. Davis-Besse 1, Cycle 1 Physics Parameters . . . . . . . . ...... 4-2 4-2. Shutdown ;iargin Calculation for Davis-Besse 1, Modified Cycle 1 .. ..... . . . . . . . . . . . . . ..... . 4-4 5-1. Ther=al Hydraulic Design Parameters . . . . . . . . . . . . ..... 5-2 6-8 l1 6-1. LOCA Limits . . . . .. ... ... . . . . . . . . . ... . . ......

6-2. Comparison of Key Parameters for Accident Analysis. . . . ...... 6-8 7-1. Technical Specification Changes . . . . . . . . . . . . . ...... 7-2 List of Figures Figure 2-1. Davis-Besse 1 Core Modification Plan. . . . . . . . . . . . .. ... 2-2 2-2. Enrichment and Burnup Distribution for Davis-Besse 1, Modified Cycle 1 .. . ... . . . . . . . . . . . . . . ...... 2-3 2-3. Control Rod Core Location and Group Assignments for Modified Cycle 1. . . . . . . . . . . . . . . . . . . ...... 2-4 4-1. Modified Cycle 1 (84 EFPD) Two-Dimensicnal Relative Power Distribution, Full Power, Equilibrium Xenon, Groups 7 and 8 Inserted . . . . . . . . . . . . . . . . . . ..... .

4-5 4-2. Modified Cycle 1 (149 EFPD) Two-Dimensional Relative Power Distribution, Full Power, Equilibrium Xenon, APSRs Inserted. . .. . . . . . . . . . . . . . . . . . . ... . . . . 4-6 4-3 Modified Cycle 1 (485 EEPD) Two-Dimensional Relative Power Distribution, Full Power, Equilibrium Xenon, APSRs Inserted. . . . . . . . . . . . . . . . . . . . . . . ..... 4-7 7-la Reactor Core Safety Limit . . . . . . . . . . . . . . . . ...... 7-2a 7-lb tessure/ Temperature Limits . ... . . . . . . . . . . . ...... 7-2e 1 7-1. Reactor Core Safety Limit . . . . . . . . . . . . . . . . ...... 7-3 7-2. Trip Setpoint for Flux -4 Flux - Flow . . . . . . . . . . ...... 7-4 7-3. Allevable Value for Flux -aFlux - Flow. . . . . . . . . . .. .... 7-5 7-4. Minimum Boric Acid Tank Contained Volume as Function of Stored Boric Acid Concentration - Davis-Besse 1, Cycle 1. . . . . . . 7-6 7-5. Regulating Rod Group Insertion Limits for Operation to 14515 EFFD (Four Pumps) . . . . . . . . . . . . . . . . . .. ... . 7-10 7-6. Regulating Rod Group Insertion Limits for Operation Af ter 14515 EFPL (Four Pumps) .. . .. . . . . . . . . . . . . ...... 7-11 7-7. Regulating Rod Group Insertion Limits for Operation to 14515 EFPD (Three Pumps). . . .. . . . . . . . . . . . . ...... 7-12 7-8. Regulating Rod Group Insertion Limits for Operation After 14515 EFPD (Three Pumps) . . . . . . . . . . . . . ... . . ...... 7-13 7-9. Regulating Rod Group Insertion Limits for Operation To 145i5 EFPD (Two Pumps). . . . .. . . . . . . . . . . . . ...... 7-14 7-10. Regulating Rod Group Insertion Limits for Operation Af t e r 14 5tS EFPD (Two Pump s) . . . . . . . . . . . . . . . . . . . . . . 7-15 7-11. Control Rod Core Locat$on and Group Assignments for Modified Cycle 1 . . . . .. . . . . . . . . . . . . .. .... 7-16 7-lla. Deleted Control Rod Core Location and Group Assignments after 200+10 EFPD . . . . . . .. . . . . . . . . . . . . ...... 7-16a 7-12. Axial Power Imbalance Envelope for Operation to 14515 EFPD (Four Pumps ) . . . . . . . . . . . . . . . . . . . . . . . . . . 7-17 4

- iv - , Rev. 5/26/78

l. INTRODUCTION AD

SUMMARY

This study justifies :ne continued opera:1cn of the first cycle of Davis Besse Nuclear Power Station Uni: 1 at the rated core power of 2772 MWt following re-moval of all 68 burnable poison rod asse=blies (BPRAs) fro = the core and the relocation of e.ight fuel asse=blies wi:hin the core. All orifice rod asse=blies (ORAs) are also being re=oved from the core and two modified ORAs will be located in the two fuel assemblies centnining the pri=ary neutren sources.

These two codified CRAs are described in Sec:1ca 3.3.

The 3PRAs are used to partially cen:rel :he reactivity changes due to fuel burnup and fission product buildup, aug=enting the control provided by soluble boren and centrol rod asse=blies, and are designed to re=ain in the core for the firs: cycle only. 3PRAs are also used :o =ake :he beginning-of-life mod-erator te=perature coefficien: less positive and to flatten the radial pcwer distribution. The ORAs are used to 11=1: bypass flow through fuel asse:blies with e= pry guide tubes. Recen: ano=alous mechanical behavior of the BPRA latching rechanis= in another 36W operating rase:or and the ORAs latching techanism in Davis-Besse Unit i deems it orudent to remove the 3PRAs and the ORAs from the Davis-Besse Unit 1 before completion of the first cycle to preclude the possibility of-1 experiencing any anonalous behavior pf these cocoonents at DB-1.

In additica to the changes described above, the centrol reds will be regrouped and opera:ed differently fres the = ode criginally established for cycle 1.

Addi:ional descriptica of these changes is given in section 2. The analytical

echniques and design bases used to support the operatien of the re=ainder of cycle 1 have been es:ablished in reports tha: have been accepted by the USSRC (see references).

All acciden:s analyced in the FSAR have been reviewed for operation withou:

SPRAs and CRAs. The characteristics for operation of this sedified core are i conservative co= pared to those used in the FSAR analyses, and thus no new acciden: analyses were perforced.

The Technical Specifica:icas have been reviewed, and the modifica: ions required for cycle 1 operation without 3PRAs and CRAs are jus:1fied in this study 3ased Rev. 5/26/78 1-1 Babcock & Wilcox

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. cn tha analyses and considering the postulatsd effects or foal densification and the Final Acceptance Criteria for E=ergency Core Cooling Syste=s, Davis Besse Nuclear Power Station Unit 1 can be operated safely during cycle I without BPRAs and ORAs at the rated core power level of 2772 .W t. 1 4

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l Rev. 5/26/78 1- t a Babcock & Wilcox

3. FUEL SYSTF.M DESICI 3.1. Fuel Asse=bly Mechanical Design The types of fuel asse=blies and pertinent fuel design para =eters for Davis-Besse 1 cycle 1 are listed in Table 3-1. All fuel assemblies are iden:ical in concept and are =echanically interchangeable. The evaluation of core =od-ification indicates that there will be no adverse effect on fuel asse=bly per-for=ance.

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The results eported in the FSAR are applicable for cycle 1 opera-tion without burnable poison rod asse=blies and orifice red asse=blies. 1 3.2. Ther al Desien The removal of the 63 3? ras and 46 of the 48 ORAs frc= the core in:roduces no 1 significant differences in fuel ther=al perfor=ance relative to prior analysis centained in reference 1. Linear heat rate capabilities are based en een:erline fuel =elt and were established using the TAFY-3 code (reference 2) with fuel densification to 96.5P. of theore:ical density.

3.3. Pr1=ary Neutron Source Retainer During initial core opera: ion there are two pri=ary deutron sources located in individual guide tubes of rao fuel asse=blies. Each source is held in a shroud tube which rests on the botto= of a guide tube. A solid stainless rod is placed on top of the source to hold it dcwn against hydraulic lift. To

further insure tha

the source vill not co=e out of the guide tube during 1

l postulated accidents,an Orifice Rod Asse=bly (CRA) is la:ched to :he top of 1

j the fuel asse=bly. The rods of the ORA plug the top of each guide tube including the guide tube containing the source.

i The pri=ary source capturing arrangement described above is being modified to j i prevent the ORA fro: causing wear of the fuel asse=bly end fitting and ec=ing loose. Twelve of the rods in each of the two ORAs that will re=ain in the core are being re=oved, leaving only the rod above the source and the three sy==e:rically located rods. A canned spring pack device (re:ainer) is placed over the hub of the =odified ORA and held down by the reactor internals. The spring is designed to hold the =odified ORA fir =1y agains: the fuel asse=bly

end fitting, taking into account hydraulic lif:, dif ferential ther=al expansion and fuel asse=bly irradiation gro9th. Testing and analyses have shown [>

hydraulic and structural adequacy of the retainer. (A submittal of the results of this testing and analysis will be =ade by 35'n' to the NRC). Rev. 5/26/78 j 3_ t Babcock & Wilcox

Txble' 3-1. Fut1 Design Part=tters c d Di-ensions 3atch 1 3atch 2 Batch 3 Fuel asse=bly type MK B4A MK B4A MK B4 A No. of asse:blies 56 61 60 l Fuel rod OD, in. 0.430 0.430 0.430 Fuel red ID, in. 0.377 0.377 0.377 Flexible spacers, type Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensified active fuel length, in. 143.5 143.5 143.5 Fuel pellet CD (nean specified), in. 0.3675 0.3675 0.3675 Fuel pellet initial density, :T3 96 96 96 Laitial fuel enrichnent, vt : 235 U l.98 2.63 2.96 1

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1 Rev. 5/16/78 3-2 Babcock & Wilcox

5. THERMAL HYDRAULIC DES.GN The re= oval of all burnable poison rod asse=blies (3PRA s) and all but :wo orifice rod asse=blies (ORAs) fro = the core results in an increase in the =axi=u= core bypass flow from 6.04: to 10.75%. However, reactor coolant syste= flow

=easure=ents have de=cns: rated that the actual sys:e= flow rate is at least ,

113% of the design ficw rate. Therefore analyses have been based upon an increased mini =u= ficw rate of 110% of design to de=onstrate that operction with 3?RA s and ORA s re=oved will not result in any loss of ther=al =argin rela:ive to previous licensing submit:als (Ref erences 1,3). Table 5-1 shows significant ther=cl-hydraulic para =e:ers f e sts revised analysis co= pared to those fro = the Fuel Densification Report (Ref erence,3) .

The potential ef fect of fuel rod bow on DN3R was considered in the ini:ial Davis 3 esse I Technical Specifications by providing sufficien: =argin in CN3R -

rela:cd li=its to offse: a 6.0% DN3R penalty, based upon a rod bow proj ectica model sub=itted to NRC by B&W in Septe=ber,1976. In a subsequent sub=i::al (Reference 4), the =argin to offset fuel rod bow was increased (tc 11.2%) by 1 taking credi: for 5% excess RCS ficw. No other-Technical Specifica: ion changes were required to provide the de=onstra:ed =argin.

In evalua:ing :he eff ects of re=oving the ORA's and 3?RA's , the 11.2% I

=argin to offset fuel red bcw has been =aintained by increasing the encess flew credi an additional 5% (up to 110% of desig=) and by adjusting slightly the reactor coolant core outlet pressure and outlet te=perature 11=1 curve (Technical Specification Figure 2.1-1). As shown in Table 5-1. the net effect of these changes 1 is an increase in the Minimum DNER (MDN3R) at design overpcwer, whic'h indicates an i=provement in core ther=al safety =argins relative to the initial core ther=al hydralic analyse:7 l i

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5-1 Rev. 5/26/78

Table 5-1. The:=al Hydraulic Desirn Parameters Modified

. Cycle 1 (ref 3) Cycle 1 1 Rated Core Power, }f4e 2772 2772 4

RCS Flowrate - GPM 352000 387200 RCS Flowrate - L3/HR 131.2x106 143.91x106 Reactor Coolant Inlet Te=perature, F 555.4 557.7 Reactor Coolant Average Te=perature, F 382 582 Reactor Coolant Outlet Te=perature, F 608.7 606.3 Flow Available for Heat Transfer, L3/HR 123.3x106 128.4x106 d

Reference Design Fan . 1.714 1.714 Reference Design Fg 1.50 1.50 '

CHF Correlation BAW-2 BNJ-2 l'DNER @ Design Overpewer 1.77 1.81 i

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6. ACC1LEh7 AND TRANSIENT ANALYSIS 6.1. General Safety Analysis Em'h TSAR 1 accident analysis has,been examined with respect to changes in cycle 1 paraneters due to the =odifications =ade in the core i to determine the effcces of these changes and .co casure tha: thermal perfor=-

ance during hypothe:ical transients is no: degraded.

Core thermal parama:ers used in :he FSAR accident analysis were design operat-ing values based on calculated values plus uncertainties. These parane:ers

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are cor.:en to all the accident analyses presen:ed herein. For each FSAR -

accident ~.

a discussica of the accident and the key part=e:crs are provided.

A comparison of the key para =e:ers (see Table 6-2) from the YSAR and =odified cycle 1 opera: ion is provided with the accident discussions to show tha: :he initial conditions of the transient are bounded by the FSAR analysis. The ef-fects of fuel densifica: ion on the FSAR accident analysis results have been evaluated and are reported in reference 3. Calculational techniques and ne:h-ods for analysis of = edified cycle 1 operation re=ain consis:en: With : hose used for tbc ?SAR.

No new dose calculations were perfor=ed for this report. The dose considers-tions in the FSAR were based on =ax1=um peaking and burnup for all core cy-c1es. '

6_. 2. Rod IJithdrawal Accident This accident is defined as encontrolled reactivity addition to the core due I to vithdrawal of control rods during startup conditions or from rated power conditions. Both types of incidents were analyced in the FSAR.

• The i=por:an:

parameters during a rod withdrawal accident are Doppler coeffi-tien:, =odera:or te=perature coefficient, and the rate at which reactivity is cdded to the core.' Only high-pressure and high-flux trips are accounted for in tha FSAR analysis; the cultiple alarms, interlocks, and trips that nor ally preclude this type of inciden: are ignored.

Rev. 5/26/78 6-1 Babcock t. Wilcox w ,- v , . , - _ - '

For positive reactivity addition indicative of these events, the =oct severe resul:s occur for BOC conditions. The FSAR values of the key parameters for LOC conditions were -1.28 x 10-5 Ak/k *F for the Doppler coefficient, 0.13

= 10-' Ak/k *F for the =oderator te=perature coefficient, and rod group worths up to and including a 10.0% ak/k rod worth. As currently planned, both group 7 and S centrol rods would be fully inserted at the beginning of =odified cycle operation, resul:ing in an EFF =oderator coefficient of -0.3 x 10-"

Ak/k *F and a Doppler coefficient of -1.38 x 10-5 Ak/k *F. At EZP, beginning of codified cycle operation, the moderator coefficient with groups 7 and S in-serted is +0.6 x 10-" a k/k OF. The para =etric studies perfer=ed in the FSAR bound this valuc. Thus, for the red withdrawal transients, the censequences will be no more severe than those presented in the FLAR and the fuel densification report.3 6.3. Moderator Dilutien Accident 3oren (in the for: of boric acid) is used to control excess reactivity. The boren content of the reae:cr ecolan: is periodically reduced to compensate for fuel burnup and transient xenon effects; dilution water is supplied by the makeup and purifica:ica syste=. The moderator dilution transients considered are the pu= ping of water with cero boren concentration frc= :he =akeup tank .

to the reactor coolant syste (RCS) under conditions of full-power operation, het shutdown, and refueling. The key parameters in this analysis are :he ini-tial boron concentratica and the boron reactivity worth, and the =ederator temperature coefficient for power cases.

For positive reactivity addition of this type, the =cs: severe results occur for SOC ccnditions. The FSAR values of the key parcneters for 30C condi: ions were 1407 pp= for the inizial boren cencentra:ica,100 pp=/: ak/k baron reac-

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tivi:y worth . .d +0.13 x 10 ' ak/k *F for the =oderator temperature coef fi-cient. Ce= parable =odified cycle 1 operatica values are 1129 pp= fer the initial beron concentratien,107 pp=/Li ok/k baron reactivity wor:h, and -0.3 x 10-" Ak/k *F for the =odera:or te=perature coef ficient. The FSAR shows tha: the core and RCS are adequately protected during this event. Suf-ficient ti=e for opera:or ac: ion to terminate this transient is also shown in the FSAR, even with =axi=u= dilu: ion and =ini=u= shutdewn margin. The predict-ed para =e:ric values of importance :o =oderator dilution transient during =od-ified cycle 1 operation are conservative with respect to the FSAR design value; th us , the analysis in the FSAR is valid.

6-2 Babcock & \Vilcox

a 6.4. Cold Wate; (Pu=n Startun) Accident The NSS contains no check or isolation valves in the RCS piping; therefore, the classic cold water accident is not poss ible. However, when the reactor is i

opera

ed with one or = ore pu=ps net running, and these pu=ps are then started, the increased flow rate will cause the average core te=perature to decrease.

,1f the moderator temperature coefficient is negative, reactivity will be added to the core and a power increase will occur.

l There are protective interlocks and ad=inistrative procedures to prevent the starting of idle pumps if the reactor power is above 22%. However, these re-1 strictions were not assumed, and two-pump s:artup frc= 60% power was analyzed i as the = cst severe transient.

To =axicice reactivity addition, the FSAR analysis assumed the =ost nega:ive

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(EOC) moderator te=pera:ure coefficient of -3.0 x 10 ' Ak/k *F and the cor-

, responding Doppler coefficien: of -1.45 x 10-5 ak/k *F. The corresponding 1

=cs: negative (EOC) moderator temperature coef ficient and EOC Deppler coeffi-cien: pred'-ted for modified cycle 1 operation are -2.61 x 10-4 ok/k *F and

-1.67 x 10-' Ak/k *F, respectively. Since, for r.odified cycle 1 operation, the predicted moderator te=perature coefficient is less negative and the Dopp-l ler coefficient is =cre nega:ive than the values used in the FSAR, the tran-

! sient results would be less severe than those reported in the FSAR.

6.5. Loss of Coelan: Flow A reduction in reactor coolant flow can be caused by =echanical failure er a loss of electrical power to :he pu=ps. With four independen: pu=ps available, a =echanical failure in one pu=p will not affect the operation of the c:hers.

! Wi:h the reac:or at power, :he effect of loss of coolant flow is a rapid in-1

r crease in coolant te=perature due to reduction of heat re= oval capabili:y.

This increase could result in DN3 if corree:1ve action were net taken i==e di-ately. The key parameters for a four-pu=p coastdown or a locked-ro:or inci-dent are the flow rate, flow coas:down characteristics, Doppler coefficient, i

moderator te=perature coefficien,:, and hot channel DN3 peaking factors. The

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censerva:1ve initial condi:icns assuned for the densification repor: were FSAR

] values of flow and coas:down, -1.28 x 10~5 ak/k *F Doppler coefficient, t0.13 4 _.

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• Ak/k *F =oderator te=perature coefficient, and densified fuel power spike and peaking. The resul:s shewed tha: the DN3R re=ained above 1.32 6-3 _ Babcock & Wilcox

, _ , .,_ __ g-.,p--- w-- -----m- + - - - - w "

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. (BAW-2) for the four-pump coastdown, and the fuel cladding te=perature re-mained below critera limits for the locked-rotor transient.

The predicted para =e:ric values for =odified cycle 1 operation are -1.38 x 10-5 ak/k *F Doppler coef ficient, -0.3 x 10-' Ak/k *F :oderator te=perature coefficient, and peaking factors as shown in table 6-1. Since the BAU-2 CHF

correlation was used, and the predicted modified cycle 1 operation values are bounded by those used in the densification report, the results of that analy-represent the most severe consequences from a loss-of-flow incident.

6.6. Stuck-Out, Stuck-In, or Dropned Control Red Accident if a control red is dropped into the core while operating, a rapid decrease in neutron power would occur, acco=panied by a decrease in the core average cool-an: te=perature. In addi:ica, the power distribution =ay be distorted due to a new control rod pattern. Therefore, under these conditions a return to ra:ed pouer =sy lead to localiced power densities and heat fluxes in excess of design li=itations.

The key parameters for this transien: are =oderator te=percture coefficient, worth of the dropped rod, and local peaking factors. The FSAR analysis was based on 0.65% ik/k rod worth with a coderator temperature coefficient of -3.0

. x 10-" ak/k *F. For-=odified cycle 1 operation, the =ax1=um rod worth at pewer is 0.20% ak/k, and :he cederator temperature coefficient is -2.61 x 10-"

ak/k *F. Since the predicted rod worth is less nd the =oderator te=perature coefficient is more positive, the consequences of this trcnsient are less se-vere than the results presented in the FSAR.

6.7. Loss of Electric Power Two types of power losses were considered in the FSAR: a loss of lead condi-tion caused by separation of the unit frc= the transmission syste=, and a hypothetical condition that results in a comple:e loss of all syste= and uni:

power except the unit ba:teries.

The FSAR analysis evaluated the loss of Icad with and without turbine runback.

When there is no runback, a reactor : rip occurs on high reactor coolant pres-sure or te=pera:ure. This case resul:ed in a non-li=iting acciden:. The largest offsite dose occurs for :he second case, i.e., less of all electrical power except unit ba::eries, and assu=ing operation with failed fuel and 1

6-4 Babcock & Wilecx

steam generator tube leakage. These results are independent of core loading;

therefore, the results of the FSAR are applicable.

6.8. Steam Line Failure A steam line failure is defined as a rupture of any of the steam lines frc=

the stea= generators. Upon ini:iation of the rupture, both steam generators start to blow down, causing a sudden decrease in primary sys c temperature, pressure, and pressurizer level. The temperature reduction leads to positive reactivity insertion, and the reactor trips on high flux cr low RC pressure.

The FSAR has identified a double-ended rupture of :he steam line be: ween the 1

steam generator and ster: stop valve as the worst-case situatien at EOL con-ditions.

The key parameter for the core response is the moderator :e:perature coef fi-cient, which was assu=ed to be -3.0 x 10-4 ak/k *F in the FSAR. The predicted value of modcrator temperature coefficient for modified cycle 1 cperation is

-2.61 x 10-' ik/k *F. This value is bounded by the value used in the FSAR analysis; hence, the results in the FSAR represent the worst situation.

6.9. Steam Generator Tube Failure I A rupture or leak in a steam generator tube allows reactor coolant and asso-cia:ed activity to pass to :he secondary syste:. The FSAR analysis is based en co=plete severence of a steam generator tube. The primary cencern for i

his incident is the poten:ial radiological release, which is independent cf core configuratien. Hence, the FSAR results are applicable to modified cycle 1 operation.

6.10. Fuel Handline Accident The =echanical damage type of acciden: is censidered the =axi=um potential -

source of activity release during fuel handling activities. The primary con-cern is over radiological releases, which are independent of core configura-tion; therefore, the results of the FSAR are applicable to modified cycle 1 operatien.

6.11. Rod Ejection Accident For reactivity to be added to the core = ore rapidly than by uncontrolled rod withdrawal, physical failure of a pressure barrier ce=penent in the control rod drive assembly must occur. Such a failure could cause a pressure differ-ential to act on a control rod assembly and rapidly eject the asse=bly from 6-3 Babcock & Wilecx

~_ __

l 1

the core. This incident represents the =ost rapid reactivity insertion that can I l

be reasonably postulated. The values used in the FSAR and densification report at 30L conditions - -1.28 x 10-5 Ak/k- F Doppler coefficient, +0.13 x 10-' sk/k OF I ~

=oderator te=perature coefficient, and 0.65: 4 k/k ejected rod worth -- represent the

=axi=u= possible transient.

Moderator and Dopplet coef ficients (HFF) for =cdified cycle 1 operation are

-0.3 x 10-4 A k/k- F and -1.38 x 10-5 6 k/k OF, respectively. By specifying rod inser:icn li=its, the maximu= ejected rod worth at HFF is 0.550 s k/k. A: H2F, the nederator coef ficient is -0.6 x 10-46k/k- F, and the beginning of =odified cycle 1 =axi=u= ejected rod worth is less :han 0.35% sk/k as controlled by rod insertion it=1:s. Based on a co=parison of :he HFF parameters to the FSAR and further ce=parison :o H2P FSAR results, the =odified cycle 1 parameters are i

acceptable.

j 6.12. Mhximum Hypotetical Accident There is no postulated =echanism whereby :his accident can occur since it would require a =ultitude of failures in the engineered safeguards. The hypothetical accident is based solely on a gross release of radioactivity to the reactor building.

The consequences of this accident are independent of core loading. Inerefore, the results reported in :he FSAR are applicable for =odified cycle 1 operation.

6.13. Waste Gas Tank Rupture The waste gas tank was assu=ed to contain :he gaseous activity evolved frc= de-gassing all the reactor coolant following operation with ' defective fuel. Rupture of the tank would result in the release of its radioactive contents :o the plan:

ventilatien system and to the at=osphere through the unit vent. The consequences i

of this incident are independent of core loading; therefore, :he results reported in the FSAR are applicable for =odified cycle 1 operation.

4 a

e-6

. , _ - . ._ -. - -_--- - - - _ ~ - - _ _ _ _ _ _ - _ _ _ , - . - - , . . - , - - - , _ _ - . - - , _ _ _ _ - _ _ _ . _ _ , _ _ _ _ ._,x

j 6.14. LOCA Limits

, An ECCS evaluation for Davis-Besse 1 has been perfor=ed and is reported in topical report BAW-10105, Revis ion 1. 5 This ECCS evaluation was further modified by

correspondence between the applicant and the NRC dated February 8,1977 (Serial No.206) and October 21, 1977 (Serial No. 394). Re= oval of the 3PRAs does not affect the results reported in that topical or_ in the later_ sub=ittals._The LOCA limits, given in Table 6-1, ensure that the ECCS will satisfy the require =ents of 10 CFR 50.46.

6.15. Loss of Normal Feed Water A loss of feed water accident results fro either a reduction in or the complete loss of secondary feed water to the steau generators. Wi:h loss or reduction of feed water to the stea= generators, the capability of the secondary syste= to re=ove the heat generated in the reactor coolan: sys:e= is i= paired. The ec=ple:e loss of feed water has been analyzed in the FSAR as the most censervative case.

The i=por: ant parameters for this accident are the 30C :oderator and Doppler co-efficient.

i The FSAR values of the key para =e:ers for the 30C condi: ions were -1.28 x 10-5

/ik/k - F for :he Doppler coefficient and 0.13 x 13-E a k/k- F for the =oderator te=perature coefficient. As currently planned both group 7 and S con:rol rods would

be fully inserted at the beginning of modified cycle operation, resulting in a EF?

modera:or coeffic.ent of -0 3 X 10-'Ak/k- F and a Doppler coef ficient of -1.33 x 10-3 i

ak/k- F. Thus for the loss of feed water acciden:, the consequences will be no

{ more severe than : hose presented in the FSAR.

l 1

s .

3 2

l Table 6-1. LOCA Limits Axial Linear heat

. position, f: rate , kk'/f t 2 16.5 4 17.2 I

6 18.4 9

8 17.5 10 17.0 1

, Table 6-2. Comearisen of Key Parameters for Accident Analysis FSAR and Value predie:cd Parameter EA*a'-1401 value for modified evele'l Doppler coeff, ak/k *F -1.28 x 10-5 (EOC) -1.33 x 10-3 (84 EF?D)

Doppler coeff, ak/k *F

-1.45 x 10-5 (EOC)(a) -1.67 x 10-5 (4E5 EF?D)

Moderator coeff, ak/k *F +0.13 x 10-" (BOC) -0.3 x 10-' (64 EF?D)(b)

Moderator coeff, 'k/k *F -3.0 x 10-' (E00)

-2.61 x 10 " (4S5 EF?D)

All rod group worth, d'd/k 10 7.56 (Banks 1-7)

Initial baron cenc, ppm 1407 1189 (34 EFFD) 3 aron reac:1tity worth 100 107 (EFP), ppt /1?.' ak/k Maximum ejected red wer:h, 0.65 0.55

% ak/k Dropped rod worth (HFP), ak/k 0.65 0.2 3

(*) 1.77 x 10 -5 Ak/k *F was used for the steam line failure analysis.

(b)Centrol red groups 7 and 8 fully inserted.

6-3 Babcock & \Vilcox

~

i t

Table 7-1. Technical Specification' Changes Tech Spec identification (report identification in parentheses) Reason for change 1

Figure 2.1-1, 2.2-1 Bases 2.1 and ORA Removal.

Bases Figures 2.1 (pages 7-2a thru 7-2f)

Figures 2.1-2, 2.2-1, 2.2-2 (Fig- Slightly increased axial peaking be-ures 7-1, 7-2, 7-3) havior in lower third of core due to BPRA removal. Higher flow data on Figure 2.1-2 offsets ORA removal and fuel rod bew effects.

Figure 3.1-1 and page B 3/4.1-2 3PRA removal and standardization of (Figure 7-4 and page 7-7) boric acid volume requirements to minimize future Tech Spec changes.

Secticns 3.1.3.6, 3.1.3.7; Figures Redesignation of control rod group 3.1-2a, -2b; 3.1-3a, -3b, -3c, -3d; assignments and insertion of regulat-3.1-4 (Pages 7-8 and 7-9; Figures ing group 7 into core to maintain 7-5 through 7-11) Delete Figure negative moderator temperature coef-3.1-5 ficient when 3PRAs removed.

Tab le 3. 2- 1 (page 7-24a) Higher ficw offsets ORA removal and 1 fuel rod bow effects.

Figures 3.2-la, -lb; 3.2-2a, -2b; Slightly increased axial peaking be-3.2-3a, -3b and page n 3/4 2-2. havior in lower third of core due to 1 Figures 7-12 through 7-17 and Page EPRA re= oval.

7-22a)

Table 3.2-2 and pag'e B 3/4.2-1 Re-evaluation of individual incore (pages 7-23 and 7-24) -

detector uncertainty and subsequent error propagation into quadrant tilt alarm setpoint.

s Section 5.3.1 BPRA removal.

d 3

e Rev. 5/26/78 7-2 babcock 8. WilCOX

Fig. 7-la Reactor Core Safety Li: sit 2400 _. . . . .-

~ ~ ~ ~

ig RC HIGH PRESSURE TRIP --

619.0 .I.

r_ =--

_ . - . v. _ . _=.p- . - - _ _ - - .m_ - . . - - . - -= M 2355.0 Z

= --

_ RC HIGH TEMPERATURE TRIP ___- g- ~~-

2300

y. . _ . _ _ -

I z i ACCEPTABLE ~~ ' r. ___- = p: -_:T_ z

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==-_._ ~- - =~ __ _ _ _ . -;= y . f_-==_--

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1--  ; ~ r

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2100

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1 RC PRESSURE 5

2000 sbN

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985_. . MF_-f ==E= s=

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c .

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-~

1300

___._=.g..-----.- ~ _.

.._- c

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_ _.__ 4 ___.

~ . . _ . _ . - - . . .i' - - _ . _ . - .

580 590 600 610 620 630 640 Reactar Outitt Temperature, F REACTOR CORE SAFETY LIMIT Tech Spe: Fig . 2.1-1 0

7-2a Rev. 5/26/78 l

. i Y ~

$ TABLE 2.2-1 T

$ REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS U -

.m FUNCTIONAL UNIT TRIP SETP0ini ALLOWABLE VALUES

• i.

l. Manual Reactor Trip Not Applicable Not Applicable

~

2. liigh Flux < 105.5% of RATED TilERMAL POWER < 105.6% of RATED TilERMAL POWER sith four pumps operating sith four pumps operating #

< 80.7% of RATED TilERMAL POWER sith three pumps operating < 80.8% of RATED TilERMAL PgWER With three pumps operating

< 53.0% of RATED TilERMAL POWER with

< 53.1 % of RATED TilERMAL POWER yith y .

one pump operating in each loop one pump operating in each loop r> ~

" g in 3. RC High Temperature < 619 F < 619.08 F

4. Flux - A Flux-Flow (1) Trip Setpoint not to Allowable Values not to exceed ,

exceed the limit line of the limit line of Figure 2.2-2.

. Figure 2.2-1. '

5. RC Low Pressure III > 1985 psig > 1984.0 psig* 1 1976.5 psig**

6. RC liigh Pressure 1 2355 psig < 2356.0 psig* 1 2363.5 psig**

g 7. RC Pressure-Temperature U) > (16.25 T out F - 7873) psig > (16.25 T out F 7873.64 ) psig#

vi R

e a

D r m D vW l 2.1 SAFETY LIMITS I BASES 2.1.1 and 2.1.2 REACTOR CCRE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel cperation to wi:nin the nucleate boiling regime where the hea transfer coefficient is large and the cladding tempera ture. surface temperature is slightly above the c:clant saturation Oceration above the upper boundary of the nuclea:e boiling regime would result in excessive cladding temperatures because of the enset of departure from nucleate boiling (CN5) and the resultan: sharp reduction in heat transfer c: efficient. DNS is not a directly measurable parameter during operatien and therefore THER'GL F0WER and Reacter C clan Temper-ature and pressure have been related to CNS through the 55W-2 CNE correlation. The DNS correlation has been devele;ed :c predict the ON3 flux and the location of DNS for axially uniform and non-uniform heat flux cistributions. The 1:. cal DN5 heat flux ratic, DN5R, defined as the ratio of the heat flux that would cause CNB at a particular core lccr.:icn to the local heat flux, is indicative of the margin :s CNE.

The minimum value of the CNSR during steady state operation, normal operaticnal transients, and an:icipated transients is limited to 1.32.

This value corresponds to a 95 percen: pr:bability at a 99 percent confidence level that DNS will not occur and is chosen as an appr:priate margin :: DN5 for all operating conditicns.

The curve presented in Figure 2.1-1 re:: resents the conditions at which a minimum CNSR of 1.32 is predicted for the maximum possible thermal pcwer 112". when the reactor coolant ficw is 387,200 G?i, which is lic':

design flew rate for four operating reactor coolant pumps. This curve is _

based on the folicwing hot channel factors with p0:ential fuel densifi-cation and fuel red bewing effects:

Fq = 2.56; F =1.M; g h=1.50 The design limit power peaking factors are the most restrictive calculated at full pcwer for the range from all centrol reds fully dithdrawn :: minimum allowable control red withdrawal, and fem the

ore CNSR design basis.

'fAVIS-BE552, UNIT 1 B 2-1 7-2c Rev. 5/26/78

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BASES ror the curve ci-  : s.St s r.igure 2.1, a pressure-tempera ture peint j

abcve and to the lef: cf the curve would result in a CNSR greater than 1.32 or a local cuality at the ;oint of minimum CN5R less Onan -22%

for that particular reac

cr cociant pump situation. The 1.32 CN5R

j curve for fcur pump cperation is acre restrictive than any cther reacter coolant pump situation because any pressure / temperature point above and to the left of the fcur pump curve will be above and Oc the lef t .

of. n.e three pu=p an, :vo pu=p curves. .

1 2.1.3  :: p

. . ..au- n. . . e- n CJ..:. ".+. .t v.. t . .' t. :::

. .s

. ..t. n. .e. .

Th a. r e. s .. #. . #. v r. re #. . 5 i s .< a f e. .y L '. ... '. - - .= r. .s ..W. e. ' r. . =#.: ./-

. . .. - ". . . e React 0r CSciant System from overpressurization and thereby crevents the release of radienuclides con ained in the reac:cr coolan: frcm reaching

• . L.

. . a c.n *.2 :.n....e. n . -cw *.~ee.r .w.e .. ,n..

The reacter pressure vessel and pressuri:er a.re designed Oc Section III of the ASM5 Boiler and Pressure Vessel Ccde which permits a maximum trinsient pressure of 110%, 2750 psig, of design pressure. The Reac:cr Coolant System picing, valves and fi::ings, are designed : pNSI 5 31.7, 1955 Editica, which permits a maximum transient pressure cf 110L, 2750 psig, cf ccacenent design cressure. The Safety Limi of 2750 psig is therafere consisten; with :ne design cri:eria and associated cada

e. u.,.-. ,1 r e.-a -. n . ..2.

t s ,h e e r. * *. r c. R e .a . .. r e.

. . c l a . . *. O./s '. =. ., i a- n' j' '. .-^. '. .= u- *. .= A. .= . .U. . .: .s#-. '..^.:",.

of design pressure, Oc de=cnstrate integrity prior Oc initial coefa,ti:n.

a 1

4 OAVIS-3 ESSE, UNIT I E 2-3 7-Oe Rev. 5/06/78 1

Fig. 7-lb Pressure / Temperature Limi:s 2400 2500 2200 m

=.

d a 2100 M

e

- ... . .. .- _ . .

$ 2000 5

3 1900 --

1800 580 590 600 610 620 630 540 Reactor Outlet ie:perature, F RC FLOW POWER PUMPS OPERAilNG ,

387,200 GPM 1125 F0'JR PUMPS

, PRESSUREc'iEMPERATURE LIMITS AT MAXIMUM ALLOWABLE ?OWER FOR MlN' MUM ONBR Te, :t * . n.t .s e s . . 2.1 7-2f Rev. 5/26/78

'71gu. 7-1. Reacto: Core Safety Li  :

35.0 --120 +20.0 112.0 112.0 0

+30.0 98.0

--100 1

-45.0 . - +20. 0

-55.0 87.2 87.2 85.0 e o I

i

--80

+30.0- 2

-55.0

. 60.0 en v :n 3

-55.0 +30.0 59.5 59.5 ,

E u 5 w e

. o w a o a u c wU c o c w aeo c--40 "a e n e C.

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.c

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u 4 c eA U O u

u .Q LU a eo m --90 s o a c. o -

o u c. u 5 u u o vco e o a o% c m e e u

=

p 4 tt O

, e e 0 e e .

-60 -40 -20 0 t20 v40 w0 Axial Po.ter I= balance, m.,

Curve Reactor Coolant Flow (G?M) .

...em- 1 387,200 2- 290,100 1

. '3 191,000 (Tech 5pec Figure 2.1-2)

. REv. 5/26/78 7-3 Babcock & Wilcox 1

I

TABLE 3.2-1 3

5 DilB MARGift T,

,o, , -

12

. LIMITS Four Reactor Three Reactor One Reactor c.

Coolant Pumps Coolant Pumps Coolant Pump 5

-u Parameter Opera ting Operating Operating in Each loop.

7" i'=

Reactor Coolant.Ilot Leg Temperature T *F

< 611.1 < 611.I II) < 611.1 11 Reactor Coolant Pressure, psig.I ) > 2062.7 > 2058.7 0} > 2091.4 Reactor Coolant Flow Rate, gpm > 395,700 > 396,500 > 195,200 l

%u

'?

I Applicable to the loop with 2 Reactor Coolant Pumps Operating.

k

• I )l.imit not applicable during either a TilERitAL POWER ramp increase in excess of 5% of

, RATED TIIERMAL POWER per minute or a TilERMAL P0tlER step increase of greater than 10%

3 of RATED lilERMAL POWER.

(3)Theseflowsincludeaflowrateuncertaintyof2.2%.

8. STARTUP PR00PJ.M - Pl!? SICS TESTl :G 1

The planned Startup Test Program cssociated with core performance is outlined below. These tests verify that core performance is within the assu=ptiens of the safety analysis and provide confirmation for continued safe operation af-ter re= oval of the lumped burnable poision red asse=blics (L3PRAs). Described below arethe tests to be perfor=cd up to 75% FP; af ter 75" FP testing is co=-

plete, tests vill be conducted in accordance with the ongoing initial Startup Physics Test Program with power Doppler and tc=perature coefficients of reac-tivity =casurancats perfor=ed at the power levcis indicated below.

3 8.1. Precritical Tests - 8.1.1. Control Rod Trio Test Precritical centrol red drop times are recorded for all centrol rods at hot full flow conditions ,efore zero power physics testing begins. Acceptance i criteria state that. the rod drep ti=e from fully withdrawn to 3/4 inserted shall be less than 1.66 seccuds at the conditions above.

It should bc noted that safety analysis calculatiens are based on a red drop l time of 1.40 seconds from fully withdrawn to 2/3 inserted. Since the cos:

accurate position indice tion is obtai ted from the ene reference suitch at the 3/4-inser:cd pusitien, t.:is position is used instead of the 2/3-inserted posi-tion for data gathering. The acceptance criterion of 1.40 seconds corrected to a 3/4-inserted position (by red insertien versus time correlation) is 1.66 -

seconds.

1 I

Rev. 5/26/78 a_, Babcock t. Wilcox

s 8'.1. 2 . RC Flow RC Flow with 4 RCPs running will be measured at hot zero power, steady state l

conditions. Acceptance criteria require that the measured flow be within 1 I

allowable limits.

8.1.3. RC Flow Coastdown l The coastdown of RC flow fro = the tripping of the highest flow RCP from 4 RCPs running will be measured at hot zero power conditions. The coast- 1 down of RC ficw vs. ti=e will then be compared to the required RC flow vs. time to determine if acceptance is met.

i 8.2. Zero Power Physics Tests 8.2.1. Critical Boron Ccacentration ,

Criticality is obtained by deboration at a constant dilution rate. Once criticality is achieved, equilibrius boron is obtained and the critical boron concentration determined. The critical boron concentration is calculated by correcting for any rod withdrawal required in achieving equilibrics boron.

i i

f Rev. 5/26/78 8-la Babcock & \Vilcox i

, m. _

ry ,m, ___-, y -.__ - _ . _ . _ -

9 - . . - , . ,

~

The acceptance critarir , laced on critical boron concer'-ation is that the actual

$oron concentration must be within t100 pp= 3 of predic:ed.

8. 2. 2. Temoerature Reactivity Coefficient The isother=al temperature coefficient is =easured at approxi=ately the group 7 and group 5 rod insertion li=its. The average coolan te=perature is varied by first decreasing then increasing tempera:ure by 507. During the change in te=pera:ure, reactivity feedback is compensated by discrete change in rod =otion; the change in reactivity is then calculated by the summation of reactivi y (obtained from i

reactivity calculation on strip chart recorder) ssociated with the te=perature

, change.

Acceptance criteria state that the measured value shall not dif fer from the predicted

, value by more than 10.4 x 10-0 sk/k OF (predicted value obtained from Physics Tes:

i 1

M_anual curves).

1 f

The moderator coefficient of reactivity is calculated in conjunction with the temperature coef ficient =easurement. After the temperature coefficien: has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to i

4 obtain modera:or coef ficient, This value =ust not be in excess of the acceptance cri:eria limit of +0.9 x 10-4 ak/k OF.

3.2.3. Control Rod Grour Reactivitv Worth Control bank group reac:ivity worths (groups 5,6, and 7) are =easured at hot cero power conditions using the boron / rod swap method. The boron / rod swap =e:hed consists of es:ablishing a deboration rate in the RCS and co=pensating for the reactivi:y J changes of this deboration by inserting centrol rod groups 7, 6, and 5 in incremental i

i steps until the rod insertion 11=1: is reached. The reactivity changes that occur j during :hese measurements are calculated based on reacti=eter data, and differential j rod worths are ob:ained frc= the measured reactivity worth versus the change in rod a

i group position. The differential rod worth of each of the controlling groups are i

i

• hen su=med to obtain in:egral rod group worths.

. The acceptance cri:eria for :he control bank group worths are as fellows:

! 1. Individual bank 5, 6, 7 worth (bank 5 fres 1007. withdrawn to rod insertion lici:):

l oredicted value - =easured value j measured value x 100 s15

2. Sum of groups 50 6 . 7:

, oredicted - measured i measured x 100 510 If acceptance criterion 2 above is not met, the worth of control group 5 from the

rod insertion li=it to f ully inserted and control group 4 are measured, and acceptance criterion 2 is applied to the sus of groups 7, 6, 5, and 4.

8.2.4. Ej ected Control Rod Reactivity Worth 9

Af ter CRA groups 7, 6, and 5 Lave been measured by debora:1on, the ejec:ed rod (a group 6 rod) is borated to 100% wd and the wor:h obtained by adding the incremental changes in reactivity by boration.

Note: If CRA group 4 was deborated to 0; wi:hdrawn, then the eje::ed rod will be =easured after CRA groups 4 and 5 have been berated back to the rod insertion 11=10.

Af ter the ejected rod has been borated to 1007. wd and equilibriu= boron established, the ejected rod is then swapped in versus CRA group 5 and the worth de:er=ined by the change in CRA group 5 position. The baron swap and rod swap values are averaged j, :o determine ej ected rod worth.

Accep;ance criteria for the ejected rod worth test are as follows:

l' oredicted value -measured value O measured value x- 100

2. Measured value (error-adjusted) 21.0T.4k/k The predicted ejected rod worth is given in the Physics Test Manual.

S.3. Power Escalation Tests 3.3.1. Core Power Distribu: ion Verification at4 40 and 75 F? Power Wi:h Nominal Control Rod Group Configuration Core power distribution tests are perfor=ed at 40 and 75: ??. The test at 40 FF is essentially a check on power distribution in the core :o bring attention to any abnor=alities before escalating to the 75 FP plateau. Rod index is es:ablished a:

a no inal full power configura: ion which is where the core power distribution calcu-lations are perfor:ed. APSR position is established to provide a core power i= balance corresponding to the i= balance where the core power distribution calculations are

! perfor=ed.

8-3

. . 1

. 1

. 1 The following acceptance criteria are placed on the 40 FP test:

1. The worst-case maxi =u= linear heat rate must be less than the LOCA 11=1:.

2. The =ini=um DNER nust be grea:er than 1.32.

3. The value obtained fre= the extrapolation of the minimum DNER to the next pow'er plateau overpower trip setpoin: must be greater than 1.32 or fall cutside the RPS power /i= balance trip envelope.

4. The value obtained fro = the extrapolation of the worst case =cxi=u= linear heat rate to the next power plateau overpower trip setpoinc =ust be less then the fuel melt limit or fall outside the R?S power /i= balance trip en-velope.

5. The quadrant power tilt shall no: exceed the li=i:s specified in the Tech-nical Specifications.

6. The highest =easured radial peak and the highes: predicted radial peak shall be within the fellowing li=its:

3 predicted - measured

=easured x 100 < S

7. The highes: =easured totcl peak and the highest predicted Octal peak shall be within the following limits:

Irredicted-=easured)x 100 s ,2

=easured j 1 I:e=s 1, 2, 5, 6, and 7 above are es:ablished for the purpose of verifying core nuclear and ther=al calcula:ional models, thereby verifying the accept-abili:y of da:a frc= these =odels for input to safety evaluations.

Itc=s 3 and 4 es:ablish the criteria whereby escala:1cn to the next power plateau =ay be acec=plished without exceeding any safety limits specified by the safety cnalysis with regard to DNER and linear heat rate.

The power distribu icn test perfor:ed a: 75: power is identical :o the 40: ??

test, except that core equilibriu= xenen is established prior to the 75: FP test. Accordingly, the 75: FP =easured peak acceptance criteria are as folicws:

1. The highest measured radial peak and :he highest predicted radial peak shall be within the following li=i:s:

s_4 Babcock & \Vilcox

m e

fpredicted - =casured)1 x 100 < 5 _

( =casured j

2. The highest measurad total peak and the highes: predicted total peak shall be within the folicwing li=its:

Ioredicted - measuredli

'- x 100 < ,/.5

( =casured j Predicted peaks at 40% and 75% FP are given in the Physics Test Manual.

8.3.2. Incore Vs Excore Detec:c: I= balance Correlation j Verificacion at %40% FP

!= balances are set up in the core by control rod positioning. Imbalances are read si=ultaneously on the incore detectors and execre power range detecters for various i= balances. The i= balances from the execre detectors =ust exceed the i= balances en the incore detectors by a factor of 1.25. If the ratio of excore detector i= balance to incore detector i= balance ia less than 1.25, gain a=plifiers in the encore detec cr signal processing equip =ent is adjusted to i, give the needed gain.

8.3.3. Te=perature Reactivity Coefficien: Above

%90% TP and Below 95% FP The average reacter coolant te=perature is decreased and then increased by about 3F at censtant reactor power. The reactivity associated with each te=-

perature change is obtained frc= the change in the con:rciling rod group pesi-tica. Centrolling red grcup worth is =easured by the fast insert / withdrawal

=ethod. The te=perature reactivity coefficient is calculated from the =ea-sured reactivity change and the =easured temperature change.

Acceptance criteria are that the =cderator te=perature coefficient shall be negative.

8.3.4. Power Doppler Reactivity Coefficien: Above N90% FP and Below 95% FP i

Reactor power is decreased and then increased by about 5% FP. The reactivity 1

change is obtained fro = the change in controlling red group position. Con::ci

rod group worth is =easured using the fast i, sert / withdrawal =e

h do. Reactiv-f-

ity corrections are =ade for changes in xenon and reactor coolant te=perature which occur during the =easure=ent. The power Deppler reactivity coefficien is calcula:ed frc= the =easured reac:ivity change, adjusted as stated above, and the measured power change.

S-5 Babcock & \Vilcox ev----v- -,v+w i-.w*...-g. w 9.-yuy w yv - - - - - - - -

4 g my.r--.r-S. ---e--e-- e--w- ev- 9--'-w -

es , _

. . e The predicted value of the power Doppler reactivi:y coefficient is given in the Physics Test Manual. Acceptance criteria state that the measured value shall be more negative than -0.55 x 10-' k/k-7. FP.

S.4. Procedure If Accectance Criteria Are Not Met An evaluation is performed before the power is escalated in the test progra if acceptance criteria for any test are not ne:. This evaluation is performed by site test personnel, with participation by Babcock & 'Jilcox technical personnel as required.

Further specific actions depend on evalua: ion results. These actions can include test reperformance with more detailed attention to test prerequisites, added tests to search for anc=alles, or detailed analysis of potential safety proble=s because of paraneter deviation by design personnel. The plant is not escala:ed in power until evaluation shows that plant safety will not be cc=pr =ised by escalation.

s l 4 9

, -m-- e . - - , - - . . - . , . , , - - , ,_, - - . , , , , - - - .e-- - ---,e -- ----- - *

.h y W

. r

Attachment 2 REVISION INSTRUCTIONS REMOV" INSERT C010fENT Title Page ~ Title Page Revised p.iii, iv p. iii, iv Revised No change to p. V. i Section 1 Section 1 Revised No change to Section 2 Section 3 Section 3 Revised

--- No change to Section 4 Sections 5 and 6 Sections 5 and 6 Revised No change to p. 7-1 Table 7<-1 (p. 7-2) Table 7-1 (p. 7-2) Revised Figures 7-la (p. 7-2a) New Table 2.2-1 (p. 7-2b) New Bases p. 2-1 (p. 7-2c) New

. Bases p. 2-2 (p. 7-2d) Neu Bases p. 2-3 (p. 7-2e) New 4

Figure 7-lb (p. 7-2f) New Figure 7-1 (p. 7-3) Figure 7-1 (p. 7-3) Revised No change to p. 7-4 through 7-24 Table 3.2-1 (p. 7-24a) New No cht.nge to p. 7-25 Section 8 Spetion 8 Revised e

Rev. 5/26/78

. .- . - . - - . - . . .-