3F0389-18, Application for Amend to License DPR-72,consisting of Tech Spec Change Request 166,removing cycle-dependent Core Operating Limits from Tech Specs & Relocating Limits to Core Operating Limits Rept

Application for Amend to License DPR-72,consisting of Tech Spec Change Request 166,removing cycle-dependent Core Operating Limits from Tech Specs & Relocating Limits to Core Operating Limits Rept

ML20246P714
Person / Time
Site:	Crystal River
Issue date:	03/22/1989
From:	Boldt G FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20246P721	List: 3F0389-18, Application for Amend to License DPR-72,consisting of Tech Spec Change Request 166,removing cycle-dependent Core Operating Limits from Tech Specs & Relocating Limits to Core Operating Limits Rept ML20246P724
References
3F0389-18, 3F389-18, GL-88-16, NUDOCS 8903280299
Download: ML20246P714 (14)

Text

1

% l

• j 6 I Florida )

Power COR PO R AT 1ON I March 22, 1989 3F0389-18 Document Control Desk (

U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 '

Technical Specification Change Request No. 166 Core Operating Limits Report

Dear Sir:

I Florida Power Corporation (FPC) hereby submits Technical Specification Change Request No. (TSCRN) 166, requesting amendment to Appendix A of ,

Operating License No. DPR-72. Proposed replacement pages for Appendix  !

A and associated bases are provided. 4 1

This submittal requests that cycle-dependent core operating limits be l removed from Technical Specifications (TS) and relocated to a Core 1 Operating Limits Report. This report will document the specific values of parameter limits resulting from the licensee's calculations including any mid-cycle revisions to such parameter values. l Generic letter 88-16, which was issued by the NRC on October 4, 1988, provides a means for removing the numeric values of cycle specific parameters from the technical specifications. The generic letter-requires that the cycle specific limits be determined using an NRC-approved methodology.

B&W Fuel Company (BWFC) calculates the axial power imbalance ranges, control rod insertion ranges, axial power shaping rod insertion ranges, and quadrant power tilt limits with the methodology described in topical report BAW-10122A, Rev. 1, " Normal Operating Controls." This report received NRC approval in Safety Evaluation Reports (SERs) dated September 14, 1979 and April 20, 1984. The SERs conclude that the procedures and techniques described in BAW-10122A, Rev. 1 are acceptable for establishing limiting conditions for operation for the parameters discussed above.

$ 2 g  :

P POST OFFICE BOX 219 + CRYSTAL RIVER, FLORIDA 326294219 * (904) 563-2943 A Florida Progress Company L

Merch 22, 1989 3F.0389-18 Page 2 BWFC calculates the moderator temperature coefficient with the methodology described in topical reports BAW-10118A, " Core Calculational Techniques and Procedures" and BAW-10152A, " NOODLE, A Multi-Dimensional Two-Group Reactor Simulator." Topical report BAW-10118A . received NRC approval in the Safety Evaluation Report dated September 25, 1979. Topical report BAW-10152A received NRC approval in the SER dated April 24, 1985. The SERs conclude that the procedures-and techniques described in BAW-10118A and BAW-10152A are acceptable for establishing limiting conditions for operation for the moderator temperature coefficient.

The methodology for calculating the programmed rod positions is described in BWFC topical' report , - BAW-10118A, " Core Calculational Techniques and Procedures", and in the attached methodology (attachment

1) provided by BWFC. Topical report BAW-10118A received NRC approval as noted above.- The refuel boron concentration limits are calculated by BWFC using the methodology described in the attached - BWFC report (attachment 1) . It is requested that these methodologies be approved with this change request as they will be utilized in determining the-parameters to be used in the Core Operating Limits Report for Rod Program and Refuel Boron Concentration.

The NRC issued a Safety Evaluation Report on January 26, 1989 to Duke Power Company for their Core Operating Limits Report change request.

Oconee Nuclear Station was the lead plant for the Babcock and Wilcox owners group regarding this submittal.

l FPC requests this amendment become effective 30 days after issuance in order to allow for procedure changes and training.

Very truly yours, 1

7 f '.

l Gary [.Boldt,VicePresident )

Nuclear Production GLB:dlh Attachment xc: Regional Administrator, Region II Senior Resident Inspector i

v i

9

, i STATE OF FLOHIDA COUNTY OF CITHUS i

G.L. Boldt states that he is the Vice President, Nuclear Production for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

I G.L. gplidt, Vice President Nuclear Production l

Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 22nd day of March, 1989.

nso ary Public V

l Notary Public, State of Florida at Large My Commission Expires: June 21, 1991

v

. UNITED STATES OF AMERICA: j NUCLEAR REGULATORY COMMISSION' l

IN THE MATTER ~ )'

) . DOCKET NO. 50-302 FLORIDA POWER CORPORATION )

CERTIFICATE OF SERVICE G.L. Boldt deposes and says that the following has been served on the Designated State Representative and Chief Executive -of Citrus County, Florida, by deposit in the United States' mail, addressed as'follows:  ;

I Chairman. ' Administrator _ _

Board of County Commissioners Radiological;_ Health rSe rvicas ,

of Citrus County Department of Health and-Citrus County Courthouse Rehabilitative Services Inverness, FL 3265G 1323 Winewood Blvd.

Tallahassee, FL 32301 l

l A copy of Technical Specification Change Request No. 166,' Revision 0,-  :

requesting Amendment to Appendix A.of Operating Licensing No. DPR-72.

FLORIDA POWER CORPORATION

~

L XYhW W d%VF 1

G.L.,%1dt, Vice President NuclWar Production SWORN TO AND SUBSCRIBED BEFORE ME THIS 22nd DAY OF March, 1989.

'l l

- J! k SA) Yt No ' dry Public U U

i Notary.Public, State of Florida at Large '

My Commission Expires: June 21, 1991 ,

1-

i.

I

%' f n

FIIRIDik IOGER ERPCRATIN QUSDLL RIVER IMIT. 3 D00ET NO. 50-302/LIQBEiB NO. WR-72 REQUEST NO. 166 CORE WEEUEDG I2MHS REEUtf j LIQ 5EEE D00 MENT INVMRED: 'Dectinical 5t=r ifin'ations ,

RRFIGGS: Index,'Page la Definitions, Page 1-8 3.1.1.3 3.1.3.1 3.1.3.6 Figure 3.1-1 Figure 3.1-2 Figure 3.1-3 Figure 3.1-4 ,

3.1.3.7 .I Figure 3.1-7 3.1.3.9 3.2.1 Figure 3.2-1 Figure 3.2-2 3.2.4 Table 3.2-2 3.9.1 ..

4.1.1.3.1 4.1.3.6 j

4.1.3.9 '

4.2.1 4.2.4 6.9.1.7 IESmIPTICN OF REQUEST: i

'Ihis subnittal requests that cycle-dependent core operating limits be removed fran Technical Specifications (TS) and relocatai to 'a Core operating Limits Report. 'Ibe term, 00RE OPERATING LIMITS REPORP, will be arN1 to the Definitions Section of TS. A new administrative reporting requirausit will be  ;

added to existing reportiny requirements of TS. Individual specifications and l applicable bases will be revised to state ~ that the values of cycle-specific  ;

parameters shall be maintained within the limits identified in Core Operating 1 Limits Report.

REASW MR RBQUEST:

A number of Technical Specifications (TS). address limits associated with.

reactor physics parameters that generally change with each reload requiring the pwcassing of cnanges .to TS to update these limits each fuel cycle. 'Ibese limits are developed using an NRC-approved methodology, therefore, the license aasnisit prma is an unmaan burden on the licensee and the NRC. -An alternative to including the values of these cycle-specific parameters in' individual specifications is responsive to industry and NRC efforts on improvements in TS and is provided in this change request.

v EVAIIRTICE OF REQUESf:

It is essential to safety that the plant is operated within the bounds of cycle-specific parameter limits and that a requirement to maintain the plant within the appropriate bounds must be retained in the TS. However, the i specific values of these limits may be modified by licensees, without affecting nuclear safety, provided that these changes are determined using an NRC-approved methodology. and consistent with all applicable limits of the plant safety analysis that are addIm W in the Final Safety Analysis Report (FSAR).

A Core Operating Limits Report will be subnitted to NRC with the values of these limits. 'Ihis will allow continued trending of this information, even though prior NRC approval of the changes to these limits would not be required.

'Ibe Core Operating Limits Report will elev,=nt the specific values of parameter limits resulting from the licensee's calculations including any mid-cycle revisions to such parameter values. 'Ihe methodology for determining cycle-specific parameter limits (except for Rod Prwtaru and Refuel. Boron Concentration) is dev'=nted in NRC-approved 'Ibpical Reports IRW-10122A, Rev.

1, BAW-10118A, Rev. O, and BAW 10152A, Rev. O. As a consequence, the NRC review of proposed changes to TS for these limits is primarily limited to confirmation that the updated limits are calculated using an NRC-approved methodology and consistent with all applicable limits of the safety analysis. Attachment i describes .the methodology for determinire parameter limits for Rod hWram and Refuel Borun Concentration. It is requested that these methodologies be approved with this change request as they will be utilized in detennining the parameters to be used in the Core Operatiry Limits Report for Rod h^% ram and

, Refuel Boron Coraxultation.

l

'Ibe cycle specific core operating limits calculated in accordance with the approved methodologies will be included in the report, Core Operatirg Limits Report, and provided to the NRC upon issuance as required by the proposed Technical Specification 6.9.1.7. Controlled copies of the Core Operating Limits Report will be maintained at Crystal River Unit-3 (CR-3) and will be revised as required by the future CR-3 cycles. Attachment 2 provides a sartple copy of a Core Operating Limits Report for information purposes only.

SIUIZX EVAIIRTIN OF RB3]EST:

Florida Power Corporation (FPC) pin w the removal of cycle specific core operating limits frun Technical Specifications does not involve a significant hazard consideration. The removal of cycle dep=aiad. variables frcan Teudaical Specifications has no 4W upon plant operation or safety. The Technical Specifications will continue to require operation within the core operational limits for each cycle reload calculated by the approved reload design methodologies. Awwiate actions to be taken if limits are violated will j also remain in Technical Specifications.

FPC concludes this change will not:

1. Involve a significant increase in the probability or consequence of an i accident previously evaluated because the renoval of cycle specific core operating limits from the Crystal River Unit 3 (CR-3) Technical Specifications has no influence or 4W on the probability of a Design Basis Accident (DIR) nmwrence.

The cycle specific core operating limits will be relocated to a CORE OPERATING LIMITS REPORT. The requita-is to operate CR-3 within the limits will continue to be maintained in Technical Specifications.

2. Create the possibility of a new or different kind of accident frcan any  !

accident previously evaluated because the removal of the cycle specific variables has no influence, .bmpact, nor does it contribute to the I probability or ue= sequences of an accident. The cycle specific variables are calculated using NRC approved methodologies, and Technical l Specifications will continue to require the operation within the core operatirg limits. .

3. Involve a significant reduction in the margin of safety because the margin of safety presently provided by current Technical Specifications remains unchanged. This proposed amendment still requires operation within the I core limits as obtained frun the NRC approved methodologies, and appropriate actions to be taken when, or if limits are violated, remain  !

unchanged.

__...__m___._______.m__ _ _ _ . _ . _ _ _ _ _ _ _

v

\

l

)

i l

ATTACHMENT 1 I

METHODOLOGY  !

I e ROD PROGRAM j i

9 REFUEL BORON CONCENTRATION 1

l l

l l

a .. .J 1

I msnmrmy  !

ato Paoans h core operatinJ Limits Report has a figure which provides the Rod Frymu for each cycle. The mod Pr@mu Figure shows the location of each control

'i '

assembly in the core and identifies to which rod group assembly is _ assigned.

The Technical Specification states, "Each control rod assembly (safety, regulating and APSR) shall be sv:pmi. a.3 to operate in the core location and -

rod group specified in the 00RE OPERATING LIMITS RE10RT." The following-diam == ion daaeribes the procedures and methods of analysis used to determine the rods assigned to each group.

Eadi control rod amaemibly in the core can be electrically wui:cted to.any of the eight control rod groups. h only limitation is that each group nust have between 4 and 12 rod amaamblies. The groups are numbered 1 through 8 and are-divided into these categories. Groups 1 through 4 are the safety rods, Groups 5 through 7 are the regulating rods, and Group 8 is the axial power shaping rods (APSR's). The following dacv ription of the methods and guc.=dures for the analysis addramaa= the APSR's first and then the regulating rods and finally the safety rods.

1 N APSR's are unique in their physical characteristics fr m the other cont W rods. They have a shorter poison region and the poison' may 'be a differr -

composition. The latching mechanism for the APSR's is also different from tL other control rods in that these rods cannot be unlatched when the control rods are h cuuusi. The location of the APSR's in the core is symetric by quadrant.

These locations (L12 and N10 in the lower right quadrant) were determined to

w have a minimal impact on radial power peaking while adileving the greatest amount of overall control on core offset (or imbalance). W e locations of the APSR's have not been changed since the first cycle aM are not expected to be changed. l l

'Ihe regulating rods, Groups 5 through 7, are used to control the core power .

1 level. Wese red groups are electrically coupled to be sequentially withdrawn (5,6,7), with an overlap of 25% 5% and sequentially inserted (7,6,5). %e j i

location of the rods in each of the regulating groups is determined beginning with Group 7. Group 7 is usually m M of 8 rod assemblies. We location l

of the Group 7 rod is selected to be symmetrical in the eighth core and have a j i

minimal impact on the radial power distribution. 'Ihe locations are also l l

restricted to positions other than those adjacent to the Group 8 rods, if j i

possible. We Groups 6 and 5 rods are also selected to have eighth core j l

symmetry aM have a minimal impact on the radial power distribution. In l l

addition, the Groups 6 and 5 rods are positioned to ensure an ejected rod will not violate the safety criteria. (See BAW-10118A, " Core Calculational Techniques And Procedures," J.J. Romano, December, 1979, for a d%= ion of ejected red worth analysis.)

Design analyses with Groups 5 through 8 determine the limiting red positions for the operation of each cycle. Wese analyses evaluate the operation of the core at various power levels throughout the cycle and show how normal operating controls on the red groups can be set to ensure safe operation. (See BAW-10122A, Rev. 1, " Normal Operating Controls," G.E. Hanson, April, 1984, for a discussion of rod operation analyses which ensure operating margin with respect to power peaking, shutdown reactivity and ejected rod worths).

3

%e safety rods, Groups 1 through 4, are used to ensure sufficient scram reactivity for a safe shutdown of the reactor core. %ese rod groups are fully withdrawn prior to the core going critical. %ey are withdrawn one group at a time but not twmaat-ily in order. %e core remains ~ shutdown with Groups 1 through 4 out of the core, therefore there is no peaking requirement on the selection of the red groups for Groups 1 through 4. However, there are two conditions which influence the location of the Group 4 rods and the Group 1 rods.

1

% e Group 4 rods may be briefly inserted into the core while the core is in'the j 1

Startup Mode (2) during physics testing. ' mus the location of the Group 4 rod assemblies is chosen such that they are eighth core symmetric. %ey are also positioned with respect to the location of Groups 5 through 7 to have a worth of approximately 1.0%&K/K.

%e Group 1 rods may be withdrawn any time the core is in either the Hot Standby, Hot Shutdown, or Cold Shutdown Modes by. increasing the. boren i concentration above that required by the shutdown margin with all control rods i in. '

% e reason for withdrawing Group 1 when going through a~ heat-up to the Startup Mode (2) is to provide an extra margin of safety by havirg sczne worth for a t

scram should one be r== mary. Sus the location of the Group 1 rods is chosen to be sylunetrical and to have a worth on the order of 1.0% in reactivity. %e results of the Group 1 worth are also ocmpared to the stuck rod worth such that the reactivity difference can be included in the requirunents for.the shutdown boron unceritration if mea 7y.

E e We location of the renainirq safety groups, '2 and 3, are selected to be'in the renaining locations, but may not be eighth core synnetrical. If the Group 1 ruds are towards the interior, the Group 2 rods will be more towards the periphery ard the Group 3 rods will be more towards the interior. If the Group 1 rods are towards the periphery, then the location of the Group. 2 and 3 rods will be reversed. me Group 2 rods will be more towards the interior, and the Group 3 rods towards the periphery.

%e methods and procedures used to analyze the reactivity and power peaking effects of the control rods are diawW in BAW-10118A and BAW-10122A as noted above. In order for these analyses to be valid, the location of the rod groups in the core must be the same as those in the Frapmu Figure. % e verification that the electrical connections of the rod groups do irdeed cotr=:speili to the:

Rod Fregmu Figure is specified in the Surveillance Requirements of Technical Specification 3.1.3.7. mis surveillance requirement provides the link between the design analyses and core operation.

.a f

M !syr r m r q REPUEL B3Di QM3lNBUEIGi h Core Operating Limits Report contains the ' specification for the refuel boron w s iL ation. 'Ihe pv wtires and methods of' analysis to determine the refuel boron saw a rcation are h ibed in the following paragraph.

N calculations are performed with the approved conputer codes, (1) PDQ (BAW-10117A (P), "Rahmr'k & Wilcox's Version of PDQ07--User's Manual, "H.A. Fa-rl, et. al., June, 1976), (2) N00 DIE (BAW-10152A, "A Multi-Dimensional 'IWo-Group Reactor Simulator," 0.W. Mays, et. al. , June,1985), and (3) FIAME (BAW-10124A, "A 'Ihree-Dimensional NODAL Code for Calculating Core Reactivity - and. Power Distributions," C.W. Mays, May, 1976). Technical Specifications set the requirements on the value of the effective neutron multiplication factor (Keff) for the core. To this value and additional 1.0% in reactivity is added to account for uncertainties.

h formula to determine the refuel boron unmakations is:

Refuel Boron (ppn) = Base Case Boron (ppn)

+ Inverse Boron Worth (ppa fy,4 g) x (Base Case reactivity (g) + 6p(Shutdown Feff + 1.0%)

+ 6( (Model Corrections) + 6q (Uncertainty))

+ 6 ppu (Uncertainty) h calculational procedure to determine the values of the terms in the above equation begins with the base case. h base case calculation has 140 F and 14.7 psia, BOL cold, conditions, no u.nikul rods inserted, and uses an estimated BOL cold boron uamd. ration. h results provide the ha= case reactivity. To the base case reactivity three differential reactivities are

added. We first is the differential reactivity frun the critical condition to the shutdown Keff includirg a 1.0% reactivity increase for mdeling uncertainties. We second differential reactivity corrects the base case for l

modeling biases. W ese biases are a result of the methods or approximations in the base case, such a two-diensional calculation. %e third differential reactivity is an adjustment for uncertainties in the core corditions.

Conditions such as the BOL core burnup which is estimated while the previous cycle is operating. Any estimates or uncertainties include a reactivity increase to ensure conservative analyses.

We sum of the base case reactivity plus the second and third differential reactivities produce a total reactivity which is near zero. 'Ihis small total reactivity is converted into a boron concentration by multiplying it by the inverse boren worth. We inverse boron worth is calculated fran two NOODIE cases run at two different boron concentrations, bracketing the boron concentration required for refueling. We change in tin boron concentration due to the total reactivity correction is added to the base case boron concentration to establish the refuel boron concentration.

We refuel boron concentration may be increased by an additional incremental amount to account for uncertainties in the methods and crocedures and provide conservatism. me uncertainties and conservatism would arise from the same type of conditions noted above for the third differential reactivity term (previous cycle length estimates, etc.).

_ _ _ _ - _ _ _ _ _ _ _ - _