ML19276D706

Forwards License Amend Request,Proposed Changes to Tech Specs,App A,Of Operating Licenses DPR-42 & 60 Re Design Features & Limiting Conditions for Operation

ML19276D706
Person / Time
Site:	Prairie Island
Issue date:	12/29/1978
From:	Mayer L NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML19276D707	List: ML19276D706 ML19276D708 ML19276D709 ML19276D711
References
NUDOCS 7901090212
Download: ML19276D706 (26)

Text

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NORTHERN STATES POWER COMPANY M I N N E A PO Li s. M I N N E S OTA 55401 December 29, 1978 V

Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 5 0-282 License No. DPR-42 50-306 DPR-60 License Amendment Request Dated December 29, 1978 Technical Specifications Associated with Use of Exxon Nuclear Company Fuel in Prairie Island Units 1 and 2 Ref e rence : (a) Letter, L 0 Mayer (NSP) to Director of Nuclear Reactor Regulation (NRC) dated September 8,1978 Attached are three originals and 37 conformed copies of a request for a change of Technical Specifications, Appendix A, of Operating Licenses DPR-42 and 60.

Also attached is one copy of the license amendment class determination and a check in the amount of $12700.00 for the amendment fee.

This submittal proposes changes in definition, reactor core design features, and limiting conditions for operation in the areas of minimum conditions for criticality and power distribution limits. These changes will allow operation of Prairie Island Unit 1 and 2 with Exxon Nuclear Company (ENC) Reload Fuel commencing with Cycle 5. The changes in design features and limiting conditions f or operation had been identified previously in reference (a). Supporting documentation for this amendment is included as Exhibits C and D. In accordance with 10CFR 2.790(a)(4), ENC report, XN-NF-78-34, will be transmitted under separate cover due to the proprietary nature of the document. ENC affidavit will be included with that transmittal.

To assure timely startup of the Unit I reactor, issuance of this amendment is requested by no later than April 1,1978.

This revision does not include Technical Specifications implementing ENC calculated exposure dependent F limits as described in Reference (a).

Exxon Nuclear Company has deterbined that there will be no ef fect on Cycle 5 ope rat ion. A separate license amendment is expected to be forwarded in 7 90109 0All h

i i NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 2 Decembe r 29, 1978 February,1979 when Exxon will have completed docuuentation to support exposure dependent F limits for the Prairie Island units. This documentation will be simila9 to that previously submitted to the NRC to support exposure depende nt F limits for Docket No. 50-315. Issue of the forthcoming Technical SpecificatiOnchangedealingwithexposuredependent F limits will not be required until Decenber, 1979. N Yours very truly,

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L 0 Mayer, PE Manager of Nuclear Support Services LOM/ JAG /ak cc: J G Keppler G Charnoff IIPCA - Attn: J W Fe rnan At tachment s

i i Y UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50-282 50-306 REQUEST IUR AMENDMENT TO OPERATING LICENSE NO. DPR-42 & DPR-60 (License Amendment Request Dated December 29, 19 78)

Northern States Power Company, a Minnesota corporation, reques ts authorization for changes to the Technical Specifications as shown on the attachments labeled Exhibit A and Exhibit B. Exhibit A describes the proposed changes along with reasons for the change. Exhibit B is a set of Technical Specification pages incorporating the proposed changes.

Exhibits C and D contain support 4ag documentation for the proposed changes.

This request contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By -C'7 j [ d d L f Wachter Vice President, Power Production

& System Operation On this 29th day of , December , 1978, before me a notary public in and for said County, parsonally appeared L J Wach ter, Vice President, Power Production and System Operation, and being first duly sworn acknowl-edged that he is autt vrized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof and that to the bes t of his knowledge , information and belief, the statements made in it are true and that it is not interposed for delay.

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3 EXHIBIT A PRAIRIE ISLAND NUCLEAR GENERATING PIANT License Amendment Request Dated December 29, 1978 PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS APPENDIX A 0F UPERATING LICENSE DPR-4 2 & DPR-60 Pursuant to 10CFR50.59, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:

1. List of Figures PROPOSED CHANGE Change the title of Figure TS.3.10-6 and add Figure TS.3.10-8 to the List of Figures on page TS-iv.

REASON FOR CHANGE These figures are required based on the power distribution limit method of control PDC 2 specified by ENC, the fuel manufacturer.

SAFETY ANALYSIS These changes have no safety significance and are administrative in nature.

2. Inter.7 Fuel Limits Defin.. ion PROF 0 SED CHANGE Delete definition 1.S. Interim Fuel Limits. Reletter 1.T. Startup Operation and 1.U. Fire Suppression Water System to maintain alphabetical orde r of the definitions.

REASON FOR CHANGE The interim fuel limits definition specifies (1) the power distributions to be used in the loss of coolant accident analyses and (2) the limit on Unit 1 Cycle 1 fuel residence time. Neither of these limits is applicable.

SAFETY ANALYSIS Deletion of the definition of " Interim Fuel Limits" is warranted because the 19 71 Policy Statement and 19 72 Technical Report listed have been superseded by the 10CFR50 Appendix K criteria and the Section 3.10 " Power Distribution Limits" limiting conditions for operation currently in the Technical Specifications. This change will not affect the health and safety of the public since the references stated have been superseded.

3. Minimum Conditions for Criticality PROPOSED CHANGE Change specification 3.1.F.1 and the associated basis to read as shown in 3xhibit B, Pages TS.3 1-17 and 18.

REASON FOR CHANGE Previous safety analyses for the Prairie Island units (Cycles 1 through 4) were conducted assuming a non-positive moderator temperature coef ficient.

In order to obtain better f uel utilization, higher fuel loadings (within Technical Specification limits) may be required, resulting in higher RCS boron concentrations at the beginning of cycle. Under these conditions, a positive moderator temperature meef ficient may exist. However, the core has been designed so that the isothermal temperature coef ficient (the sum of the moderator and fuel temperature coef ficient) is limited to being negative.

This change also deletes ref e re nces that are no longer applicable.

SAFETY EVALUATION Exhibit C is submitted to support this proposed change.

The isothermal coef ficient is a more practical parameter for which to establish a Technical Specification limit because it is empirically dete rmined during s tartup tes ts. The moderator temperature coef ficient, on the other hand, is derived as the dif ference between the isothermal coefficient and the calculated fuel temperature coef ficient at the tes t conditions. The isothermal temperature coef ficient is negative to ensure that there is negative reactivity feedback on a power excursion.

XN-NF-78-35 contains the results of transient analyses that conservatively assume the most limiting isothermal temperature coef ficient. These analyses assumed a more conservative gF (2.32) than that being proposed (2.21) and moderator and f uel temperatore coef ficient values as listed in Table 4.1 (Page 69) of XN-NF-78-35.XN-NF-78-47 (Section 5.1.3) describes the mode rator temperature coef ficient considerations and predicts the hot zero powar all rods out isothermal coef ficient to be negative, which will be confirmed in the startup testing program described later. Other conditions e.g. , higher power or partial iraertion of rods would cause the isothermal coef ficient to have a more negative value.

XN-NF-78-35, indicates MTC will be positive at a low power, approaches zero at about 70% power, and is calculated to be negative at full power.

The trans ient analyses conducted envelope these predicted core conditicas.

Values of the trip setpoints used in the analyses (XN-NF-78-35 Tab le 2.2 Page 10) are as or more conservative than the Technical Specifications or actual plant setpoints.

i The change requested is consistent with previously established NRC accept-ance of allowing plant operation with a positive moderator temperature coef ficient , e.g., Amendment 40 to DPR-36, dated August 18, 1978.

As with any startup af ter a refueling, testing is appropriate to verify design values. The testing program for this reload will include the following as appropriate for refueling with fuel of a dif ferent supplier.

1. Verification of proper core loading.

2. Rod drop times (100% flow, RCS temperature >500 F).

3. Source, and intermediate range calibrations (including functional tes ts, plateau curves, compensating voltage check as appropriate).

4. Rod position indication calibration.

5. Boron endpoint measurements (All rods out, Control Bank D in, Control Banks C and D in, all Control Banks in).

6. Baron worth.

7. Rod worths (Dif ferential and Integral, Control Banks A, B, C, and D) .

8. Isothermal Temperature Coef ficient measurements (All rods out, Control Bank D in, Control Banks C and D in).

9. Flux maps (Hot Zero Power-all rods out, 48%, 100%.)

10. Axial offset power range calibration at 90% after 3 days of fuel preconditioning at 100% (3 flux maps at different power distributions).

11. Plant Heat Balance (35%, 48%, 90%).

These tests are conducted by the Prairie Island Nuclear Engineering staf f using procedures that have been reviewed by the Plant Operations Committee and approved by the Plant Superintendent-Engineering and Radiation Protection.

Descriptions of many of these tests are included in the Unit 1 and 2 startup tes t reports previously forwarded to the USAEC/USNRC. Forty (40) copies of each of these reports were transmitted, as follows:

Unit 1 - L 0 Mayer to E Case, dated October 31, 1974 Unit 2 - L 0 Mayer to A Giambusso, dated May 15, 1975 These tests are consistent with those conducted during a startup after any refueling, regardless of fuel type.. The procedures used and results for startup tests cycles 1 through 4 have been reviewed by Region III I&E personnel. Selection of the tests to be conducted is based on evaluation and consensus by corporate of fice licensing and core management and snalysis personnel and plant nuclear engineering staf f personnel.

In accordance with Technical Specification 6.7.A.1, a startup report will be filed with the NRC within 90 days af ter startup.

4. Specification 3.10.B.1 Power Distribution Limits PROPOSED CilANGE

a. Change the existing section to read as shown on pages IS.3.10-1, lA,

-2, -7A, -8, and -9 and Figure TS.3.10-6.

b. Add Figure TS.3.10-8 of Exhibit B.

REASON FOR CHANGE The Fhvaluesandfootnote l statement were imposed by NRC Order dated May 18, 1978. The F for both Prairie Island units was restricted to 2.2 4 if accumulator mod 9fications were performed and 2.21 if those modifications were not pe rf o rmed .

TheFhandFNillimits are no longe r applicable based on subsequent reanalysis. In addition, ENG has adoptgd Power Distribution Control Phase 2 which imposes restrictions on FQ during target axial flux dif ference determination. ihese changes are required to reflect the different methodology in nuclear design.

SAFETY ANALYSIS Westinghouse LOCA analyses for both units utilized the " February, 1978" model approved by the NRC. The analytical results will be provided under separate cove r.

The ECCS analyses and results are contained in XN-NF-78-46. The analytical results presented in Section 2.4 and Table 2.2 show that the acceptance criteria of 10CFR50.46 would be met for an F of 2.2l for a core loaded with Exxon fuel. The analyses presented are9 applicable for both Prairie Island Units 1 and 2 using the Exxon fuel of the same design as that used in Unit 1, Cycle 5. This approach of applying the same ECCS analyses to both units is consistent with that heretofore employed with Westinghouse fuel and analyses. Containment and RCS parameters are identical for both units.

The ECCS analyses were conducted using the UREM IIA model, described in XN-NF-78-30, previously forwa rded to the NRC by G F Owsley (ENC) to T A Ip po li to (hRC) dated Augus' 15, 1978. In addition, a calculation to evaluate the impact of upper plenum injection is presented. This UPI analysis is based on the interim model developed by the NRC staf f and modified by Westinghouse. The interim UPI model changes and results are described in Section 3. l of XN-NF-78-46. In summa ry, the calculations show that the peak clad temperature for a 0.4 DECLG will be 2198"F (including a 1 F penalty for the interim UPI model).

9 The rod bow penalty has heretofore been established for the Westinghouse fuel. The basis for this penalty not being applicable to the Exxon fuel is described in Section 6.3 of XN-NF-78-47.

The limit Fh < (2.145/P) x (K(Z)/V(Z)) is imposed based on the power distribution control phase 2 method applied to Prairie Island (Section 5.1.1 of XN-NF-78-47). The basis for PDC 2 procedures is described in XN-NF-77-57 " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase 2".

The revised Figure TS.J.10-6 is consistent with the methodology explained in Specification 3.10.B.6.

The q w ffgure (Figure TS.3.10-8) was added to support application of the new FQ limit imposed by Specification 3.10.B.1.b.

The bases were changed to provide the results of the Exxon analyses to the operator responsible for I control.

5. Specification 5.3.A Reactor Core Design Features PROPOSED CHANGES

a. Change subsection 2 as shown in Exhibit B and delete subsection 3.

b. Change the highest enrichment to 3.5 rather than 3.4 weight percent and delete reference (2) of the end of Specification 3.3.A.2.

c. Delete the existing subsection 4 that refers to the burnable poison rods incorporated in the initial core and their construction.

d. Change the existing subsection 5 to become subsection 3.

e. Delete subsection 6 and carry forward the page TS.5.3-2 material.

REASON FOR CEANGE

a. Reference to the three fuel enrichments used in the initial core is no longer applicable. This change maintains consistency by ref erring only to reload cores which is appropriate for Prairie Island.

b. To obtain longer fuel cycles, higher uranium 235 enrichment loadings will be required. In addition, this change would make Section 5.3.A.2 consistent with Section 5.6.A.

c. Burnable poisons have only been used in the Prairie Island units as needed primarily to control the value of the moderator temperature coef ficient at beginning of life.

d. The subsection 4 should be deleted for two reasons:

1. That section is no longer applicable. It refers to the initial core burnable poison rods. Starting with Cycle 5, gadolinium oxide will be used as the burnable poison.

2. Current technical specification regulatory guidance on design features for both PWR's and BWR's contains no reference to burnable poisons.

e. The requirement of 5.3.A.6 was superseded by Section 2.B.(3) of the operating licenses DPR-42 and -60.

SAFETY ANALYSES

a. A reload core typically consists of 4 or more regions of fuel depending on the individual licensee's plan for optimum fuel utilization. The typical burnup history for the 121 fuel assemblies at the beginning of a cycle is:

1 assembly - 3 cycles exposure (~ 27 GWD/MTU) 40 assemblies - 2 cycles exposure (~ 23 GWD/MTU) 40 assemblies - 1 cycle exposure ( - 10 GWD/MTU) 40 assemblies - No exposure Thus the previous description is no longer applicable. This deletion should have no ef fect on the health and safety of the public.

b. For each reload, separate safety analyses are conducted to assure that operation with reload fuel of the enrichment selected will be in compliance with Technical Specifications and NRC regulations. Thus the change from 3.4 to 3.5 w/o should have no ef fect on public health and safety. In addition the 3.5 w/o selected is based on the criticality analyses conducted for the Prairie Island Spent Fuel Pit Expansion Hearings. The 3.5 w/o limit is also specified in Section 5.6.A.

c. Burnable poisons may be incorporated into the relo d core. These may be of the following designs -

1. Burnable poison rod assemblies (BPRA's) consisting of 8,12 or 16 rods of borosilicate glass clad with stainless steel.

2. Gadolinium oxide dispersion in a uranium dioxide matrix.

Safety analyses for Cycle 1 considered use of the BPRA design as described in the plant FSAR. Use of the gadolinia design is described in XN-NF-78-34 and KN-NF-78-47.

The Exxon reload fuel assemblies will be loaded into the outer core loca-tions. Only 64 burnable poison rods will be used, distributed evenly among 8 fuel assemblics as shown in Figures 3.1 and A.3 of XN-NF-78-47. Figure A2 of that document shows that gadolinium burnup should be completed by 5500

!ND/MTU pin cell exposure. This contrasts to the borosilicate glass rods initially loaded in the Prairie Island Units, which did not burn up completely within the first cycle (e-17000 MWD /MTU) and were loaded into all three regions.

Supporting information for use of gadolinia is contained in Sections 5.3, 6.4, and Appendix A of XN-NF-78-47 and Exxon report XN-NF-78-34 on fuel design.

Use of the burnable poison uniformly dispersed in the fuel rod eliminates the disadvantages associated with the BPRA design. The main disadvantage of the BPRA design is that many dif ficulties have been experienced with the tool used to remove these devices f rom fuel assemblies. These problems can have a significant inpact on refueling outage critical path time.

Rep lacement of the discrete burnable poison rods with gadolinium oxide dispersed in the fuel should allow control of the moderator temperature ccef ficient while minimizing localized disturbances in power distribution.

This change also eliminates references that are no longer applicable and corrects a previous reference error.

EXHIBIT B License Amendment Request dated December 29, 1978 Exhibit B consists of revised pages of Appendix A Technical Specifications as listed below:

Pages TS-iv TS.1-6 TS.3.1-17 TS.3.1-18 TS.3.10-1 TS.3.10-1A T S . 3 .10-2 TS.3.10-7A TS.3.10-8 TS.3.10-9 Figure TS.3.10-6 Figure TS.3.10-8 TS.S.3-1 The following Appendix A Technical Specification page would be deleted upon approval of this request:

TS.S.3-2

TS-iv REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Rear ;or Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and opper Content on Shif t of RT NDI Reactor Vessel Steets Exposed to 550 Temperature 3.1-4 Fast Neutron Fluence (E > 1 MeV) as a Function of Full Power Service Life 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3 10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope For F = 2.32 3.10-6 Deviation from Target Flux Dif ference as a Function ofN Thermal Power 3.10-7 Rod Bow Penalty (RBP) Fraction Versus Region Average Burnup 3.10-8 V(Z) as a function of core height g 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Environmenta'.

Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-site Operating Organization 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for on-site Operating Group

TS.1-6 REV

3. Refueling Shutdown A reactor is in the . refueling shutdown condition when a refueling operation is scheduled, the reactor is subcritical by at least 10%

Ak/k and the reactor coolant average temperature is less than 140 F.

Q. Thermal Power Thermal power of a unit is the total heat transferred from the reactor core to the coolant.

R. Physics Tests ,

Physics tests are those conducted to measure fundamental characteristics of the core and related instrumentation. Physics tests are conducted such th at the core power is sufficiently reduced to allow for the perturbation due to the test cnd therefore avoid exceeding power distribution limits in Specification 3.10.B.

Low power physics tests are run at reactor powers less than 5% of rated power.

S. Startup Operation The process of heating up a reactor above 200 F, making it critical, and bringing it up to power operation.

T. Fire Suppression Water System l The fire suppression water system consists of : Water sources; pumps; and distribution piping with associated sectionalizing isolation valves.

Such valves include yard hydrant valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser.

TS.3.1-17 REV F. MINIMUM CONDITIONS FOR CRITICALITY Specification

1. The reactor shall be made critical only at or above the coolant temperature at which the following reactivity coef ficient is negative and remains negative for any coolant temperature increase (except ,

during low power physics tests): i (a) Moderator temperature coefficient for a reactor loaded with I

Westinghouse fuel only.

(b) Isothermal temperature coefficient for a reactor either full or partially loaded with Exxon fuel.

2. The reactor shall not be brought to a critical condition until the reactor coolant temperature is higher than that defined by the criti-eality limit line shown in Figure TS.3.1-1.

3. When the reactor coolant temperature is below the minimum temperature as specified in 1. above, the reactor shall be subcritical by an amount equal t' or greater than the potential reactivity insertion due to reactor cu slant depressurization.

Basis At the beginning of a fuel cycle the moderator temperature coef ficient has its most positive or least negative value. As the boron concentration is reduced throughout the fuel cycle, the moderator temperature coef ficient becomes more negative. The safety analyses conducted for Prairie Island i units with Westinghouse fue assumed a non positive moderator temperature coefficient. The isothermal temperature coef ficient is defined as the reactivity change associatud with a unit change in the moderator and fuel temperatures. Essentially, the isothermal temperature coef ficient is the sum of the moderator and f uel temperature coef ficients. This coefficient is measured directly during startup physics tes ting, whereas the moderator temperature coefficient is an inferred parameter determined by subtracting the predicted f uel temperature coef ficient from the experimentally deter-mined isothermal temperature coefficient.

TS.3.1-18 REV For extended optimum f uel burnup it is necessary to either load the reactor with burnable poisons or increase the boron concentration in the reactor coolant system. If the latter approach is emphasized, it is possible that a positive moderator temperature coef ficient could exist at beginning of cycle (BOC). For cycles with Exxon fuel, safety analyses are conducted assuming a positive moderator temperature coef ficient. These analyses predict the isothermal coef ficient to be negative at an all rods out, hot zero power condition. Other conditions, e.g. , higher power or partial rod insertio" would cause the isothermal coef ficient to have a more negative value. These analyses demanstrate that applicable criteria in the NRC Standard Review Plan (NUREG 75/087) are met.

Physics measurements and analyses are conducted during the reload startup test program to (1) verify that the plant will operate within safety analyses assumptions and (2) establish operational procedures to ensure safety anal /ses assumptions are met. The 3.1.F.1 requirements are waived during low power physics tests to permit measurement of reactor temperature coef ficient and other physics design parameters of interest. Special operating precautions will be taken during these pg{yics tests. In addition, the strong negative Doppler coef ficient and the small integrated A k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical except as specified in Figure TS.3.1-1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility temperature range. Heatup to this tempera-ture will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters. The pressurizer heater and associated power cables have been sized for continuous operation at full heater power. The shutdown margin in Specification 3.10 precludes the possibility of accidental criticality as a result ofan{ycreaseofmoderatortemperatureora I decrease of coolant pressure.

References:

(1) FSAR Figure 3.2-10 (2) FS AR Table 3.2-1

4 a T S . 3 .10-1 REV 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations.

Obiective To assure 1) core subcriticality af ter reacto .' trip, 2) acceptable core power distributions during power operation, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

Specification A. Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in F: gure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, including ef fects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at hot shutdown conditions if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon, boron, or part-length rod position.

B. Power Distribution Limits

1. a. At all times except during low power physics tests, the hot l channel factors defined in the basis must meet the following limits F 1 (2.145/P) x K(2.) for P > 0.5 Q

F" < (4.29/P) x K(Z) f o r P _< 0.5 q _

1 1.55 (1 + 0.2(1-P))(1-RBP(BU))

b. In addition, F the target flux dif ference onc0ach e(Z) shall be effective measured full at power quarter and must meet the following limit for a unit with Exxon fuel:

(Z) 1 (2.145/P) x (K(Z)/V(Z)) for P > 0.5 where:

1. The (1-RBP(BU)) multiplier is only applicable for Westinghouse Fuel.

TS.3.10-1 A REV (1) P is the f raction of full power at which the core is operating (2) K(Z) is the function given in Figuge TS.3.10-5 (3) Z is the core height location of F (4) RBP(BU) istheRodBowPenaltyasSfunctionofregion average burnup as shown in Figure TS.3.10-7 (5) Region is defined as those asserobif as with the same loading date (6) V(Z) is the function given in Figure TS.3.10-8

2. a. Following initial loading and at regular ef fective full pow r monthly l intervals thereaf ter, power distribution maps, using the moi,,ble detector system, shall be made to confirm that the hot channel f actor limits of this specification are satisfied. For the purpose of this compa rison,

1. The measured peaking f actor, FQ, shall be increased by l five percent to account for measurement error.

2. The measurement of enthalpy rise hot channel f actor, [A H' shall be increased by four percent to account for measurement error.

b. If either measured hot channel factor exceeds its limit specified under 3.10.B.l.a,the reactor power and high neutron flux trip setpoint shall be reduced so as ngt to gxceed a fraction of rated power equal to the limit to ratiooftheFQorF,hsequent measured value, whichever is less. If su in-core mapping cannot, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, demonstrate that the hot channel factors are met, the reactor shall be brought to a hot shutdown condition with return to power authorized up to 50% power for the purpose of physics testing. Identify and correct the cause of the out of limit condition prior to increasing thermal power above 50%

power, thermal power may then be increased provided F (Z) is demonstrated through in-core mapping to be within its limits.

c. If the measured hot channelfactorFhexceeds its limit as specified under 3.10.B.1.b, then either of the following two actions shall be taken:

1. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> place the reactor in a configuration for which specification 3.10.B.1.b is satisfied; or

2. Rgduce thermal power by 1% for each percent that the measured Fy exceeds the limit specified in 3.10.B.1.b. Thermal pgwer may be increased to a power such that the associated Fy would comply with 3.10.B.1.b.

c .

TS.3.10-2 REV

3. The reference equilibrium indicated axial flux dif ference for each excore channel as a function of power level (called the target flux dif ference) shall be measured at least once per equive' at full power quarter. The target dif ferences must be updated monthly. This may be done either by using the measured value for that month or by linear interpolation using the mos t recent measured value and a value of 0 percent at the end of the cycle life.

4. Except during physics tests, and except as provided by Item 5 through 8 below, the indicated axial flux difference for at least the number of operable excore channels required by TS.3.5 shall be e_ intained within a +5% band about their target flux differences (defines the target band on axial flux dif ference).

5. At a power icvel greater than 90 percent of rated power, if the indicated axial flux difference of two operable excore channels deviates f rom its target band, either such deviation shall be elimi-nated, or the reactor power shall be reduced to a level no greater than 90 percent of rated power.

6. At a power level no greater than 90 percent of rated power,

a. The indicated axial flux dif ference may deviate from its + 5%

target band for a maximum of one* hour (cumulative) in any 24-hour period provided that the dif ference between the indicated axial flux dif ference and the target flux dif ference does not exceed an envelope bounded by -10 percent and +10 percent at 90% power and increasing linearly to -25 percent and +25 percent at 50 percent power as shown in Figure TS.3.10-6.

b. If 6.a is violated for two operable excore channels then the raactor power shall be reduced to no greater than 50% power and the high neutron flux setpoint reduced to no greater than 55 percent of rated values.

• May be extended to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> during incore/excore calibration.

TS.3.10-7A REV F Dependent Heat Flux Hot Channel Factor, is defined as the m3x(Z),

imum Height local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux,allowingformanufacturing erances on fuel pellets and rods. F is the product of F and Q

FE , Engineering Heat Flux Hot Channel Factor, is defined as the aklowance on heat flux required for manufacturing tolerances. The engineer-ing f actor allows for local variations in enrichment, pellet density and diameter, surf ace area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net ef fect is a factor of 103 to be applied to fuel rod surf ace heat flux.

N F Nuclear Hot Channel Factor, is defined as the maximum local l nOu, tron flux in the core divided by the average neutron flux in the core.

N FoH, u ar a se Hot Gannd Factor, is denned as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

N It should be noted that FAH is based on an integral and is used as such in the DNB calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizcatal (x-y) power shapes throughout the core. Thus th<

horizontal power shap at the point of maximum heat flux is not necessarily directly relatei t AH' An upper bound envelope for F of 2.145 times the normalized peaking l factor axial dependence of Fikure TS.3 10-5 has been deterrdned f rom extensive analyses considering all operating maneuvers consistent with the technical specifications on power distribution control as given in Section 3.10. The results of the loss of coolant accident analyses based on this upper bound envelope indicate an adequate peak clad temperature margin to the 2200 F limit.

When an nF measurement is taken, both experimental error and manufacturing tolerance %ust be allowed for. Five percent is the appropriate allowance for experimental error for a full core map taken with the movable incore I detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

TS.3.10-8 REV N

In the specified limit of F there is an 8 percent allowance for uncertaintigs which means t$!t normal operation of the core is expected to result in F * "#E*# ""C"# ""Y AH I

• * *

• 8" '"

in this case is that (a)abnormagperturbationsintheradialpowershape (e.g. rod misalignment) affect F g, in mos t cases without necessarily affecting F , (b) the operator has a direct influence on Fg through movement of rogs, and can limit it to the desired value, he has no direct control over F and (c) an error in the predictions for radial power shape, wkkch may be detected during startup physics tests can be gmpensated for in gF by tighter axial control, but cogensation for F is less readily available. When a measurement of F 3

taken, experimental error must be allowed for and 4 percentA H '"isthe appro-priate allowance for a full core map taken with the movabfe incore detector flux mapping system. The penalties applied to F to account for rod bow of Westinghouse fuel as a functice o/hurnupare l consistent with those described in the NRC safety evaluation report,

" Interim Safety Evaluation Report an Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors," Revision 1, February 1977.

The rod bow pe nalties are not applicable for Exxon fuel based on independent measurements cf rod-to-rod spacings for interior and peripheral rod bows of spent fuel similar in design to that used in the Prairie Island core.

The plant technical specification includes a total nuclear peaking augmen-tation factor of 1.0815 (product of 1.03x1.05) in the calculation of ECCS safety limits. This factor is adequace to accommodate nuclear augmentation due to rod bow in a limiting assembly to 28,150 MWD /MTU. Fuel assembly exposures above this are expected to be operating well below the LOCA limits due to reduction of assembly reactivity. Thus no additional penalty due to rod bow needs to be applied to calculation of the LOCA limits.

Measurements of the hot channe t factors are required as part of startup physics tes ts , at least once each full power month of operation, and I whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel f actors. The incore map taken following int *ial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise af fect these bases.

For normal operation, it is not necessary to measure these quantities.

Ins tead it has been determined that, provided certain conditions are observed, the hot channel f actor limits will be met; these conditions are as follows:

1. Control rods in a single bank move together with no individual rod insertion dif fering by more than 15 inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalignment greater than 15 inches with consid-eration of maximum instrumentation error.

2. Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10.

TS.3.10-9 REV

3. The control bank insertion limits are not violated.

4. Ine part length control rods are not inserted.

5. Axial power distribution control procedures, which are given in terms of flux dif ference control and control bank insertion limits are observed. Flux dif ference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux dif ference is a measure of the axial of fset which is defined as the dif ference in normalized power between the top and bot tom halves of the core.

The permitted relaxation in F$H and F allows for radial power shape changes with rod insertioO to the insertion limits. It has been determined that provided the above conditions 1 through 5 are obserygd, these hot channel f actor limits are met. In specification 3.10 f is arbitrarily limited for P 4 0.5 (except for low power physich tests) .

The procedures for axial power distribution control referred to above l are designed to minimize the ef fects of xenon redistribution on the axial power distribution during load-follow maneuvers. Basically control of flux difference is required to limit the difference between the current value of Flux Dif ference ( ti I) and a reference value which corresponds to the full power equilibrium value of Axial Of fset (Axial Of fset = /k I/ fractional power). The reference value of flux dif ference varies with power level an burnup d

but expressed as axial offset it varies only with burnup.

The tgchnical specifications on power distribution control assure that the F upper bound envelope of 2.145 times Figure TS.3.10-5 is l not eOceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.

The target (or reference) value of flux dif ference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux dif ference is noted with part length rods withdrawn f rom the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e. , normal f ull pow'r operating position appropriate for the time in life, usually withdrawn f arther as burnup proceeds). This value, divided by the fraction of full power at which the core was opers .ing is the full power value of the targe t flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power.

Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated deviation of +5 percent Al are permitted f rom the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux dif ference every month. For this reason, the specification provides two methods for updating the target flux dif ference. Figure TS.3.10-6 shows the allowed deviation f rom t arget flux difference as a function of thermal power.

REV 1

100 __

Unacceptable Unacceptable 90 I Operation Operation 80 -

70 _

60 _

50 __

Acceptable Operation 40 _

30 _ g @

g a.

20 _ T &

o x g -

10 _

I I I I I I 0

-50 -40 -30 -20 -10 0 10 20 30 40 50 Deviation From Target Flux Difference Figure TS.3.10-6 DEVIATION FROM TARGET FLUX DIFFERENCE AS A FUNCTION OF TilERMAL POWER

2 l 1

)

5 6 i 1

1

, 0 0 l 1 t

2 h 1 g

( i e

i H

)

1 e 1 r

. o 1 C

, I 8 5 f 2 o 9 n

( o i i t

c n

u f

I 6 a

s a

)

Z

(

V 1 4 8 0

1 3

S T

e r

i 2 u g

i F

./ i 0

4 2 0 8 6 4 2 0 8 6 9 0 0 1 1 1 1 1 0 0 .

1 1 1 1 1 1 1 1 1 1

)

Z

(

v

. *

• TS.5.3-1 REV 5.3 REACTOR A. Reactor Core

1. The reactor core contains approximately 48 metric tons of uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is cade up(gj 121 f uel assemblies. Each fuel assembly contains 179 fuel rods

2. The average enrichment of the reload ccr is a nominal 2.90 weight per cent of U-235. The highest enrichment is a nominal 3.50 weight per cent of U-235.

I

3. In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inck2}ength of silver-indium-cadmium alloy clad with stainless steel i

B. Reactor Coolant System

1. The design of the ryggtor coolant systc5 complies with all applicable code requirements.

2. All high pressure piping, components of the reactor coolant system and their supporting structures are designed to Class I requirements, and have been designed to withstand:

a. The design seismic ground acceleration, 0.06g, acting in the horizontal and 0.04g acting in the vertical planes simultaneously, with stresses maintained within code allowable working s tresses.

b. The maximum potential seismic ground accele ration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.

3. The r.ocinal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.

C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of neutron flux as measured by the excore nuclear instruments.yg9nge of The system is intended to triy4yhe reactor upon the abnormal dropping I of more than one ccc. trol rod If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10.

References (1) FSAR, Section 3.2.3 (3) FS AR, Table 4.1-9 (2) FS AR, Sections 3.2.1 and 3.2.3 (4) FS AR, Section 7 l

EXHIBIT C License Amendment Request dated December 29, 1978 Exhibit C consists of the following Exxon Nuclear Company reports:

XN-NF-78-35 " Plant Transient Analysis for the Prairie Island Nuclear Power Plant Units 1 and 2".

XN-NF-78-46 "ECCS Large Break Spectrum Analysis for Prairie Island Unit I using ENC WREM-IIA PWR Evaluation Model".

XN-NF-78-47 " Prairie Island Unit 1 Nuclear Plant Cycle 5 Safety Analysis Report".

EXHIBIT D License Amendment Reques t dated December 29, 1978 Exhibit D contains the list of errata for the Exxon Documents included n Exnibit C.

Document Page Comment XN-NF-78-35 10 Delete the line referring to "b) 3.32 f t steam generator level and 25 sec".

69 The reference cycle values for moderator temperature coefficient should be corrected to read:

BOC EOC Inactive Loop Startup -400 Excessive Feedwater 0 and -400 Excessive Load Increase 6 0 and -400 where the units are (af /F) x 10 73 Document number (Upper Right Hand Corner) should read XN-NF-78-35.

XN-NF-78-47 25 2nd paragr3ph, line 4 should read ". . 3 28.15 x 10 }ND/MTU . . ." not "28.5 x 10 KWD/MTU".