ML20067C095

Application for Amend to License NPF-3,changing Tech Spec 3/4.1.1.3 Re Reactivity Control Sys Moderator Temp Coefficient & 6.9.1.7 Re Core Operating Limits Rept,Per Generic Ltr 88-16

ML20067C095
Person / Time
Site:	Davis Besse
Issue date:	02/06/1991
From:	Myers T, Shelton D CENTERIOR ENERGY
To:
Shared Package
ML20067C093	List: ML20067C091 ML20067C095 ML20067C099
References
GL-88-16, NUDOCS 9102110194
Download: ML20067C095 (9)

Text

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.. Docket Numb 2r NPF-3 Sorial Numbar 1902 Enclosure

, Page'1 APPLICATI0!4 FOR AMENDHENT TO FACII.ITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 Attached ere requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3. Also included is the Safety Assessment and Significant Hazard: Caitside ra tion.

The proposed changes (submitted under cover letter Serial Number 1902) concern:

Technical Specification 3/4.1.1.3, Reactivity Controls System Technical 5pecification Section 6.9.1.7, Core Operating Limits Report.

Fort D. C. anelton, Vice President - Nuclear By: //-

T. J. o , Director - Technical Services Sworn and Subscribed before me this 6th day of February, 1991, ttbn) i fl$J Ilotary Ptrblic, State of Ohio EVEl.vN L tr' ;'q NOTAnt % ;  ; .f ,

Mt Conn;;;;ny;;_;Q,f 9102110194 910206 gDR ADOCK 05000346 PDR

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• . Dockot Number 50-346- i d

., e, Lit;onse Number NPT-3 Serial Number 1902 Enclosure Page 2 I The following information is provided to support issuance of the requested change to the Davis.Besse Nuclear Power Station, Unit Number 1 Operating License Number NPF-3. Appendix A. Technical Specifications. Section 3/4.1.1.3 and 6.9.1.7.

i A. Time Required to implement: This change is to be implemented within 45 days after the NRC issuance of the License Amendment.

B. Reason for Change (License Amendment Request Number 90-0043): ,

, Since a new Negative Hoderator Temperature Coefficient must be calculated for each reload cycle, relocation to the core operating Limits Report vill

. minimite the required number of license amendment applications submitted

to the NRC f1r approval.

C. Safety Assessm e t and Significant Hazards Consideration: See

Attachment 1. .

, . D. Chan3m to the Core. Operating Limits Report See Attachment 2.

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• j Docket Number 50-346 License Number NTP3 Serin 1 Humber 1902 Attachment 1 Page 1 of 9 SAFETY ASSESSHFRr AND SIGNIFICANT llAZARDS CONSIDERATION FOR 1.ICENSE AMENDHEN'T REQUEST NO. 90-0043 DESCRIPTION OF PROPOSED ACTIONS The purpose of this evaluation is to reviev proposed changes to Technical Specification 3.1.1.3c that vill allow a moderator temperature coefficient (HTC) mote negative than the current limit of -3.0 x 10~4 Ak/k/ F. The new negative HTC limit vill be fuel cycle-specific, thus requiring the actual value of the

limit to appear in the Core Operating Limits Report (COLR) for that fuel cycle, as allowed in NRC Generic Letter 88-16. Aside from this change, the votding and

intent of Technical Specification 3.1.1.3c and its associated Surveillance Requirements (4.1.1.3.1 and 4.1.1.3.2) vill not change. Technical Specification 6.9.1.7 vill also be inodified to reflect the revised contents of the COLR.

The introduction of eighteen month fuel cycles, along v'th efforts to reduce the number of assemblies in each teload feed batch, forces overall core burnups to be higher at end of life es,nditions. By increasing core burnup, the core average plutonium concentration also increases, which has the effect of causing moderator temperature coefficients to be more negative, for Cycle 7 (startup ,

July of 1990), the predicted HTC at rated full power condit i lifewasmorenegativethanthecurrentlimitof-3.0x10-),onsatendofcore Ak/k/ 0P. For l future fuel cycles, this problem would be even more severe, and could greatly inhibit future fuel cycle planning and flexibility unless a more negative HTC limit vere justified and implemented.

_ SYSTEMS, COHp0NEffrS, AND ACTIVITIES AFFECTED Reactor Core SAFFTY FUNCTIONS OF Tile AFFECTED SYSTEMS, 00Hp0NENTS, AND ACTIVITIES The nuclcar fuel in the reactor core produces heat through the fissioning of uranium and plutonium. This heat is ultimately used to produce steam which drives the turbine to produce electricity. The safety functions performed by the reactor core and the nuclear fuel are to retain the fuel in an appropriate geometry for heat removal, and to prevent the migration of radioactive fission products away from the fuel pellets by encapsulating the pellets in Zircaloy cladding.

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Doc &tet Nt'mber 50-346 License Number NpF3 Serial Number 1902 Attachment 1 Page 2 of 9 EFFECTS ON SAFETY Analysis The purpose of a negative litnit for inoderator temperature coef ficient is to  !

prevent large, positive reactivity insertions to the reactor core during postulated events that Icad to a tapid cooldown of the Reactor Coolant Fystem (RCS). Six of the events described in the Davis-Besse Updated Safety Analysis Report (USAR) are potentially impacted by a negative HTC at end of life (E3L) conditions. These six events are Event Description USAR Section Dropped Control Rod Assembly 15.2.3 Inactive RCS Pump Startup 15.2.6 Excessive fleat Removal Due To 15.2.10 Feedvater System Malfunction 4

Minor Secondary Pipe Break 15.3.2 Control Rod Assembly (CRA) 15.4.3 Ejection Steam Line Dreak 15.4.4 Each of these events was reevaluated and vill be discussed separately with respect to the proposed change in the negative HTC limit.

Dropped Control Rod Assembly Event The dropped control rod assembly event causes sudden reductions in both neutron and thermal power, resulting in a cooldovn of the RCS by as much as 20 F. This cooldovn overcompensates for the inserted rod vorth, increasing neutron power above the initial conditions. Generically applicab1$ safety analysis calculatiop's have been performed assuming an EOL hot full power (liFp) HTC of

-4.0'x 10- Ak/k/ F, and, although the transient response is slightly more severe than the current USAR analysis, the results of these calculations continue to meet the Safety Evaluation Criteria of USAR Section 15.2.3.2.1.

Docket Number 50-346 1,1 cense Number NPF3

. Serial Number 1902

, Attachment 1 Page 3 of 9 1

Inactive RCS Pump Startup P, vent The inactive RCS pump startup event postulates the startup of two inactive reactor coolant pumps (RCPs) while the reactor is operating at 60 percent of lated full power, thus injecting cooler vater from the inactive RCS loops into the core. Generically applicabic safety analys petformedassuminganE01,ilFPHTCof-4.0x10~{scileglationshavebeen Akn / F, and, although the transient tesponse is slightly more severe than the current USAR analysis, the results of these calculations continue to meet the Safety Evaluation Criteria of USAR Section 15.2.6.2.1. It should be noted that this evaluation is extremely conservative in that power operation with only two RCS pumps is not permitted at Davis-llesse.

Excessive Heat Removal Due To Peedvater System Malfunction 1: vent This event is initiated by either a sudden reduction in feedvater temperature, caused by bypassing the feedvater heaters, or by a sudden increase in feedvater flow, caused by the opening of feedvater control valves. This event may be initiated at either HFP or hot zero power (HZP) conditions, and the previous analysis for this event assumed an HTC of -3.0 x 10"" 6k/k/ F at all conditions.

For the event beginning at HFP conditions, the response of the system and the consequences of the event are bounded by the HFP steam line break event, since they are essentially the same type of transient, with the steam line break being much more severe. A new acceptable MTC for steam line breck at HFP conditions vill be developed belov, and that HTC vill also ensure that the feedvater malfunction event vill yield results that vill continue to meet the Safety Evaluation Criteria of USAR Section 15.2.10.2.1.

A second initial condition for this event is at ilZP conditions with all safety rods (groups 1 through 4) withdrawn from the core, but with the regulating rods (groups 5 through 7) fully inserted. Under these conditions, the event causes neutron power to increase to about 65 percent of rated full power and produces somewhat severe power peaking due to the inserted regulating rod configuration.

This event has not been reevaluated and, therefore, it vill continue to have a limiting negative HTC of -3.0 x 10'4 Ak/k/"F for hot zero power conditions. It vill be jemonstrated below that this HZP MTC value vill be assured and bounded I

by the limiting negative MTC value at HZP conditions for the steam line break event, ensuring that this event continues to meet the Safety Evaluation Criteria of USAR Section 15.2.10.2.1. Therefore, the existing USAR analysis remains valid and bounding for this eveat at HZP conditions.

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Docket Nnmber 50-346 1,1 cense Number NPf3 Serial Numb (r 1902 Attachment 1 Page 4 of 9 Minor Secondary Pipe Break Event Thiu event is essentially a small steam line break and vill be bounded by the steam line break evaluation (see below).

Control Rod Assembly _(CRA)_ Ejection Event The control rod ejection transient is different from the other events in that it is initiated by a positive reactivity insertion not telated to the HTC (ejection of a CRA). Although the ejected CRA event has been evaluated in the USAR vith highly negative HTCs, the most severe consequences occur with more positive HTCs, not with more negative HTCs. Sensitivity studies already incorporated into the U this event has been evaluated at an HTC of

-4.0x10~gARshovthat ok/k/ F, vith results that meet the Safety Evaluation Criteria of USAR Section 15.4.3.2.1.

Steam 1.Inc Break Event This event is unique from the others in that, while it is initiated at hot full power (llFP) conditions, the HTC value is, essentially, of concern only at IIZP conditions or colder. This is because the steam line break event immediately tesults in a reactor trip, and, by design and in accordance with Technical Specification 3.1.1.1, the reactor vill always be at least one percent shutdown at il2P conditions. What is of concern is the value of HTC at HZP conditions and colder, since that value vill determine the total amount of positive reactivity that vill be added to the core during the subsequent cooldown belov HZP cond1tions.

The origingl steam line break analysis for Davis-Besse assumed in HTC of

-3.0 x 10~ Ak/k/"F at all moderator temperatures. A later analysis, also documented in the USAR, was performed using a reactivity-versus-moderator density curve which, while more realistic, was still very conservative. This cutve yielded an average temperature coefficient (combination of Doppler and moderator tempetature coefficients), ovet the range of temperatures between HZP and the migimum RCS temperature during the steam line break event, e'

-3.1 x 10- Ak/k/ F. In other vords, if the-4 transient had been n' fzed using a single temperature coefficient of -3.1 x 10 6k/k/"F for all temperatut es colder than HZP conditions, the same positive reactivity insertion vould be

Docket Number 50-346 License Number NPF3 3 , Setin! Number 1902 Attachment 1 '

Page 5 of 9 l I

produced. Therefore, this average temperature coefficient can, in fact, be shown to be a bounding value for the steam line break event at temperatures colder than llZP.

event (-1.77x10'ghentheDopplercoefficientusedinthesteamlinebreak 6k/k/"F) i equivalent HTC of -2.923 x 10'g subtracted from this Since temperature coefficient, this HTC value is an lessnegativethanthe-3.0x10'g/k/Ffsobtained. 6k/k/ F u.lue assumed at liZP conditions for the feedvater malfunction accident, the HZP temperature coefficient for the stean line break event, with i s associated value of HTC, is a more restrictive limit that bounds the value of . Se llZP HTC for the feedvater malfunction event.

Since the coefficients defined above are derived from an existing USAR analysis, which has not been changed, the steam line areak event vill continue to meet the Safety Evaluation Criteria of USAR Section 15.4.4.2.1.

Application .

Essentially,twojimitshavebeendefinedabove. For ilFP conditions, a negative HTC of -4.0 x 10' ok/k/ F has been shown to be acceptable with respect to the USAR Safety Evaluation Criteria for all of the events sensitive to a HFP HTC limit. Fo

-3.1x10gcondigionsatHZPandcolder,anegativetemperaturecoefficientof Ak/k/ F has been defined for the events which are sensitive to a lizP HTC limit, lloveve t , since Technical Specification 3.1.1.3c refers to a negative HTC limit at rated full pover, the HZP temperature coefficient must be related to a HFP HTC in order to provide a single value for the Technical Specification limit.

A'three-dimensional full core geometry physics model, using the NOODLE analytic nodal code,.vas developed to transform temperature coefficients at HZP conditions, with all control rods inserted except the maximum vorth stuck rod, into HTCs at PPP conditions with all rods withdrawn. It should be noted that the NOODLE code has been successfully benchmarked against measured reactivity coef ficients at both IIFP and HZP conditions and that the NOODLE code-has been topically approved by the NRC. This transformation process, which accounts for the offects of moderator density, soluble boron concentration, control rods, and

, fuel temperature between HZP and ilFP conditions, provides for a fuel cycle-specificTransformationfromthe11m1tiggHZPtemperaturecoefficientfor the steam line break event (-3.1 x 10~4 6k/k/ F) to an equivalent HFP HTC limit for the steam line break event. The new negative HTC limit, which vili appear in the Core Operating Limits Report (COLR), vill either be the fuel cycle specalic steam line break value described above or the -4.0 x 10~f' Ak/k/'nF value used in other events, whichever is least negative.

- -u .-- - . .= - _, _ . - - _. -.

.- .. - _ . . -~ . - - - . - = - - ~ - . - - . . _ . - - . - . _ . . . ~ .

4 Docket Number 50-346 License Number NPF3 Serial Number 1902 Attachment 1 Page 6 of 9 The above-described process has been applied for the Davis-Besse Cycle 7 core.

For Cycle 7 the t

-3.62 x 10-4 6k/k/gansfotmedsteamlinebreakilFPMTChasagalueof F, which is less negative than -4.0 x 10~ Ak/k/ F.

Therefore,J,heneggtivePFPHTClimit for the cycle 7 core is

-3.62 x 10~ Ak/k/ F. This value vill appear in Table 2 of the COLR, which vill be referenced by Technical Specification Limiting Condition for Operation 3.1.1.3c.

The surveillance requirements for HTC (4.1.1.3.1 and 4.1.1.3.2) vill remain unchanged.

SIGNIFICANT IIA?.ARDS CONSIDERATION 1he Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an opetating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would (1) Not involve a significant increase in the probability or consequences of an accident previously evaluated (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. Toledo Edison has revieved tne proposed change and determined that a significant hazards consideration does not exist because operation of Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes would la. Not involve a significant increase in the probability of an accident previously evaluated because there are no accidents whose probabilities of occurrence are related to the value of the HTC.

Ib. Not involve a significant increase in the consequences of an accident previously evaluated because it has been demonstrated that all of the USAR accidents sensitive to-a negative HTC still meet their USAP, Safety Evaluation Criteria under the proposed new limits.

2a. Not create the possibility of a new kind of accident from any accident previously analyzed because a more negative HTC is only a concern during RCS overcooling transients that have already been addressed in the USAR and the value of the HTC cannot create a new accident.

Docket Number $0-346 License Number NPF3

, seital Number 1902 Attachment 1 Page 7 of 9 2b. Not create the possibility of a different kind of accident from any accident previously analyzed because a more negative HTC is only a concern duting RCS overcooling trantients that have already been .

1 addtersed in the USAR and the value of the HTC cannot create a dif(erent accident,

3. Not involve a significant reduction in a margin of safety because all events sensitive to a negative moderator temperature coefficient have been evaluated with tespect to the proposed new limits in a very conservative tashion and have shown no significant change in transient response, and because the proposed change in the negative HTC limit is telatively small compared to the conservatisms in the  !

evaluation. Further, all events sensitive to a negative HTC vill l continue to meet their appropriate USAR Safety Evaluation Criteria under the proposed new limits. ,

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CONCl,USIONS On the basis of the above, Toledo Edison has determined that this License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical  !

Specifications that must be teviewed by the Nuclear Regulatory Commission, i thc License Amendmenc Request does not constitute an unreviewed safety l question. '

A'ITACIIMENTS Attached are the proposed marked-up changes to the Operating. License.

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