ML18088B274

License No. DPR-67 Application for Amendment: Appendix a Technical Specification Proposed Change Concerning Deleting Requirements for Part Length Control Element Assemblies from Unit 1 Technical Specifications

ML18088B274
Person / Time
Site:	Saint Lucie
Issue date:	04/12/1978
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
Download: ML18088B274 (23)

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

DISTRIBUTION FOR INCOMING MATERIAL 50-335 REC: STELLQ V ORG: UHRIG R E DOCDATE: 04/12/78 NRC FL PWR 5 LIGHT DATE RCVD: 04/14/78 DOCTYPE: LETTER NOTARIZED: YES COPIES RECEIVED

SUBJECT:

LTR 1 ENCL 40 FQRWARDING LIC NO DPR-67 APPL FQR AMEND: APPENDIX A TECH SPEC PROPOSE CHANGE CONCERNING DELETING REQUIREMENTS FOR PART LENGTH CONTROL ELEMENT ASSEMBLIES FROM UNIT 1 TECH SPECS... NOTARIZED 04li3/78... WlATT LIC FEES.

PLANT NAME: ST LUCIE Ni REVIEWER INITIAL: XJM DISTRIBUTOR INITIAL:PLr DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTFR ISSUANCE OF OPERATING LICENSE.

(DISTRIBUTION CODE A001>

FOR ACTION: BR W/7 ENCL INTERNAL: 'EG FILE+~ NRC PDR4eW/ENCL

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P.O. BOX 013100, MIAMI, FL 33101

~gg FLORIDA POWER & LIGHT COMPANY gggg)ypyq~(y .].,g,37~a O CQ e

Office of Nuclear Reactor Regulation C

Ch C'Cl Attention: Mr. Victor Stello, Director P'(

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+O Division of Operating Reactors t33C C CC/3 JW,, Dl na U. S. Nuclear Regulatory Commission CA~CA Washington, D. C. 20555~ mrnn M

Dear Mr. Stello:

n O Cll Re: St. Lucie Unit No. 1, Docket No. 50-335 Proposed Amendment to Facility Operating License No. DPR-67 ln accordance with 10 CFR 50.30, Florida Power and Light Company (FPL) submits herewith three (3) signed originals and forty (40) copies of a request to amend Appendix A of Facility Operating License.DPR-67.

The purpose of this proposed amendment is to delete the require-ments for Part Length Control Element Assemblies (PLCEA's) from the Technical Specifications for St. Lucie Unit No. 1 because of tne CEA guide tube wear problem and the requirement to maintain the PLCEA's fully withdrawn and non-scrammable. Plant operation at power is not allowed with the PLCEA's in the core. FPL plans to remove the PLCEA's during the present refueling outage. CEA guide tube plugs will be installed into the locations previously occupied by the PLCEA's to preserve the current dynamic operating characteristics of the reactor. This request must be approved to support startup following refueling.

The proposed amendment is described below and shown on the accom-panying Technical Specification pages bearing the date of this letter in the lower right hand corner.

~Pa e 1-3 Reference to PLCEA's in Definition 1.13 is deleted.

Pa e 3/4 1-23 Specification 3/4.1.3.2 requiring PLCEA withdrawal and surveillance is deleted.

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Mr .. Victor Stello April 12, 1978

'age Two L-78-125 Reference to "part length" in Specification 3.1.3.3 and ACTION statement a are deleted. ACTION statement b is renumbered to reflect the above deletion.

Pa e 3/4 1-25 ACTION statement. c and d are renumbered to reflect the deletion of ACTION statement a.

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Reference to PLCEA's in Specification 4.2.2.3 is deleted.

Specification 3.10.1.b is deleted. References to the PLCEA's in ACTION stat'ement a and Specification 4.10.1.1 are deleted. Specification 4.10.1.3 is deleted.

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Reference to the PLCEA's in BASES 3/4.1.3 is deleted.

Pacae 5-5 Reference to the PLCEA's in Specification 5.3.2 is deleted.

In accordance with the criteria stated in 10 CFR 170.22, FPL has

,determined tnat this is a CLASS III amendment. A check in the amount. of $ 4,000 to cover the requisite amendment, fee is enclosed.

Deletion of the PLCEA's has been requested by a number of plants, including Millstone 2, Calvert Cliffs, and Indian Point, 2. The St. Lucie Plant Facility Review Group and the Florida. Power and Light Company Nuclear Review Board have reviewed the proposed amendment and concluded that it. does not involve an unreviewed safety question. A safety evaluation is attached.

Very truly yours, pA Robert E. Uhrig

,Vice President REU:MAS:sl Attachment cc: Mr. Peter B. Erickson Mr. James P. O'Reilly, Region II Harold F. Reis, Esquire

STATE OF FLORIDA )

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COUNTY OF DADF )

A. D. Schmidt , be'ng first duly sworn, deposes and says:

That he is Vice President of Florida Power. a Light Company, the licensee herein'hat ne has executed the foregoing document; that the state-ments made in this said document are true and correct to the best of his knowledge," in ormation, and bel"'ef, and that he is authorized to execute the document, on behalf of said L'censee.

A. D Sc 1 Subscribed and sworn to before me this day of 19'OTARY PUBLIC, in and for the County of Dade, State of Florida NOTARY PUBUC STATE OF FLORIDA at LARGE MY COMMISSION EXPIRES AUGUST 24, 1981 i4ly commission expires: @>No& mmu tIATNARo QDNDINQ AGENcY

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DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION

.1.12 CORE ALTERATION shall be th movement or manipulation of any-component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTDOMN MARGIN 1.13 SHUTDOtAN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or, would be subcritical from its present condition

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assuming all full length conttol element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest r activity worth which is assumed to be fu'l ly withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a.. Leakage (except CONTROLLED LEAKAGE) into closed systems', such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

b. leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system.

ST. I UCIE - UNIT 1 1-3 4/12/78

THIS PAGE INTENTIONALLY LEFT BLANK ST. LUCIE - UNIT 1 3/4 1-23 4/12/78

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I REACTIYITY CONTROL SYSTEMS POSITION INOICATOR CHANNELS LIMITING CONOITION FOR OPERATION 3.1.3.3 All shutdown and regulating CEA reed switch posi-tion indicator channels and CEA pulse counting position indicator chan-nels shall be OPERABLE and capable of determining the absolute CEA positions within + 2.25 inches.

APPLICABILITY: MOOES 1 and 2.

ACTION:

a. With a maximum of one reed switch posit'ion indicator channeI per group or one (except as permitted by ACTION item d. belo>>)

pulse counting position indicator channel per group inoperable and the CEA(s} with the inoperable position indicator channel partially inserted, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

Restore the inope'rable position indicator channel to OPERABLE status, or

2. Be in HOT STANOBY, or

3. Reduce THERMAL POWER to < 70". of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. 'Operation at or below this reduced THERMAL POWER level may continue provided that within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a) The CEA group(s) with the inoperable position indi-cator is fully withdrawn while maintaining the withdrawal sequence required by Specification 3.1.3.6 and when this CEA group reaches its fully withdrawn position, the "Full Out" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully withdrawn. Subsequent to fully withdrawing this CEA group(s), the THEfU1AL POllER level may be returned to a level consistent with all other applicable specifications; or ST. LUCIE - UNIT 1 3/4 1-24 4/12/78

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REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS Continued LIMITING CONDITION FOR OPERATION b) The CEA group(s) with the inoperable position indi-cator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawa1 sequence and THERMAL POWER leve1 required by Speci-fication 3.1.3.6 and when this CEA group reaches its fully inserted position, the "Full In" limit of the CEA with the inoperable position ic:dicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6.

b. With a maximum of one reed switch position indicator channel per group or one pu1se counting position indicator channel per.

group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter by i ts "Full In" or "Full Out" limit (as applicable),

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2. The fully inserted CEA group(s) containing the inoperable position indicator channel is subsequently maintained fully inserted, and

3. Subsequent operation is within the limits of Specifica-tion 3.1.3.6.

With one or more pulse counting position indicator channels inoperable, operation in NOOES 1 and 2 may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided all of the reed. switch position indicator

'hannels are OPERAS E.

SURVEILLANCE RE UIRENENTS 4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 4.5 inches at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals'when the Oeviation circuit is inoperable, then compare the pulse counting position indicator and reed switch position indicator channels at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ST. LUCIE - UNIT 1 3/4 1-25 4/12/78

POWER DISTRIBUTION LIMITS SURYEILLAHCE RE UIRB1ENTS Continued

c. Within four hours if the AZIhUTHAL POWEP, TILT (T ) is )'O.G2.

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4.2.2.3 F shall be determined each time a calculation of FT is required

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by using tEe incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term. Steady State Insertion Limit for the existing Reactor CooTant Pump combination. This determination shall be limiied xo core planes be-tween 15/ and 85K of full core height and shall exclude regions influenced by grid effects.,

4.2.2.4 T shall be determined each,tiye a calculati'on of F is required'nd the value of T used to determi'ne Fshall be the measure value of T .

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ST. LUCIE - UNIT 1 3/4 2-7 4/12/78

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTOOWH iQRGIH requirement of -Specification 3.1.1.1 may be for measurement of CEA worth and shutdown margin provided 'uspended reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODE 2.

ACTION:

With the reactor critical (K > 1.0) and with less than the above reactivity equivaient IvfiTable for trip insertion,.

immediat ly initiate and continue boration at > 40 gpm,of 1720 ppm boron or equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With the reactor subcritical (K < 1.0) by less than the above reactivity equivalent, irk(3iatsly initiate and continue boration a't > 40 gpm of 1720 ppm boron or equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is

• restored.

SURVEILLANCE REQUIREMENTS I

4.10.1,1 The position of each full length .

CEA either partially or fu'fly withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each CEA.not fully inserted shall be demonstrated OPEPABLE by verifying its CEA drop time to be < 3.3 seconds within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of. Specification 3.1.1.1.,

ST. LUCI E - UNIT 1 3/4 10-1 4/12/78

REACTIYITY CONTROL SYSTEHS BASES 3/4.1.3 MO~/ABLE CONTROL ASSEHBLIES Continued The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the. core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES 1 and 2,T'n'ethe full"leng-.'h CEAs be nearly fully withdrawn. amount of CEA insertion permitted by the Long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the un-rodded burnup assumption but will still provide sufficient reactivi ty control. The Power Dependent Insertion Limits of Soecification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUT00'llN MARGIN is maintain d, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configuration.

ST. LUCIE - UNIT 1 B 3/4 1-5 4/12/78

DESIGN FEATURES COHTROL ELEllEHT ASSEt'1BLIES 5.3.2 The reactor core shall contain 73 full length I control element assemblies. The control element assemblies shall be designed and maintained in accordance with the original 'design provisions contained in Section 4.2.3.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AHD TEi~1PERATURE 5.4 ~ 1 The reactor coolant system is designed and shall be maintained:

a 0 In accordance with the code requirements specified in Section 5.2 of the FSAR with allo, ance for normal degradation pursuart,

. to the applicable Surveillance Requirements,

b. For a pressure of 2485 ps',g, and C. For a temperature of 650'F, except for the pressurizer which is 700'F.

'lOLUNE 5.4.2 The total water and steam volume of the reactor coolant sys em is 11,100 + 180 cubic feet at a nominal T of 567'F.

avg 5.5 Ei~i RGEHCY CORE COOLIHG SYSTEhS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with th original design provisions contained in .Section 6.3 of .he FSAR with allowance for normal degradation pur-suant to the applicable Surveillance Requirements.

5.6 FUEL STORAGE CRITICALITY 5.6.1 The new fuel storage racks are designed and shall be maintained with a center-to-center distance of not less than 21 inches between assemblies placed in the storage racks. The spen't fuel storage racks are designed and shall be maintained with a center-to-center spacing of not ST. LVCIE - UNIT 1 4/12/78

4/12/78 SAFETY EVALUATION

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RENOVAL OF PART LENGTH CONTROL ELEMENT ASSEi~lBLIES FROth ST. LUCIE UNIT 1 I. INTRODUCTION This report provides information to justify plant operatioji following the removal of the part-length control element assemblies (PLCEAs).

Plant operation at power is currently not allowed with PLCEAs in the core.

CEA guide tule plugs will be installed into the locations previously occupied by the PLCEAs. These plugs are being installed to preserve the current dynamic operating characteristic of the reactor, i.e.

pressure drops, coolant flow rates, etc., which could be affected if just removal of the PLCEAs was performed.

II. CEA Pl UG HECHNICAL DESIGN The CEA guide tube plugs, which will 'be inserted into locations pre-viously occupied by PLCEA, have essentially the same configuration as the casting of standard four arm CEA spiders with elements extending down from the end of each spider arm and from the center. Also in the manner of a standard CEA, there are springs and retainers around

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The only significant differences between the plugs and standard CEAs are the following items:

1. The fingers of the plug are very short and extend only approximately five (5) inches into the top of the fuel assembly, and are made from solid 304 stainless steel. Each finger has a leaf spring tyoe attachment which positively positions the finger such that it will not vibrate against the wall of the upper end fitting post. The fingers of Standard CEAs, by comparision, are made of Inconel 625 tubing and extend approximately 150 inches into the fuel assembly.

In view of the vast differences between the dynamic characteristics of the plug elements and CEA elements, it is concluded that the plugs will not b'e susceptible to the same mechanism which has pro-duced guide tube wear. Moreover, since the plug does not extend into 'the Zircaloy portion of the assembly whatever vibration may occur will not result in guide tube wear.

2. The plug incorporates design features which engage the lower 'ends of the flow scupper in the upper guide structure and compress a spring which holds the plug against the under side of the scupper.

This preload spring is loacted in approximately the same location as the CEA spider soring on standard CEAs, but is somewhat longer in order to accoranodate the effects on differential expansion, irradiation induced Zircaloy growth, and component tolerances.

3. 'Hinor hardware modifications will be made to make the plug com-patible with existing fuel and CEA handling equipment. In order to preclude inadvertent pickup of a plug, the PLCEA extension shafts will'be removed during plant operation. ,

It should be noted that the basic design of the CEA guide. tube plugs are similar to that of plugs which have been used successfully in the Fort Calhoun reactor during all operating cycles and has been revised for use in 2560 tNl class plants only to reflect minor dimensional variations between the two classes of plant.

III. THERMAL HYDRAULIC EFFECTS A. Thermal Effects Physics analysis indicates that there will be no adverse effect .

of the plug assemblies on the core power distribution. Since the plugged fuel assemblies have no adverse effect on the design core flow distribution, calculated core thermal margin will be 'unaffected.,

B. Hydraulic Effects The following hydraulic aspects were considered with respect to the installation of the plug assemblies in the part length CEA locations.

Hydraulic uplift force on the plug assembly.

The hydraulic uplift force on the plug assembly is calculated to be 90 lbs. This uplift force is offset by a hold-down spring design force of at least 150 lbs plus the wet weight of the plug assembly. Because of .the large margin, no uplift.

problem is anticipated with the plug assembly.

. 2. Hydraulic uplift force on the fuel assembly.=

The uplift force on the fuel assembly increases by less than 5 lbs. due to the installation of the plug assembly. This represents an insignificant change in fuel assembly uplift margin.'.

Core Bypass Flow Rate.

The core bypass flow. rate increases very slightly, 0.02~<. of vessel flow rate, due to the reduced hydraulic 'resistance of the plug assembly relative to that for the part-length CEAs.

The resulting total core bypass flow, 3.4Ã at the end of the second cycle, remains below the thermal design bypass flow rate of 3.7X.

4. Guide Tube Flow Rates and Crossflow Velocities on Fuel Rods.

The flow rate into a guide tube increases by 225 when replacing the part length CEA with the plug assembly. The lateral

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velocities across the fuel rods in the immediate vicinity of the guide tube cooling holes are calculated to increase from 2.5 fps to 3.1 fps. Results from hydraulic tests on 14xl4 fuel assemblies, in which the lateral velocities were more adverse, showed no fuel rod fretting. Therefore, no problems with fuel rod fretting are expected with the installation of the plug assembly.

5. Flow-Induced Vibration Characteristics The CEA fir.gers are known to vibrate within the guide tubes due to a f'tow-induced mechanism. That mechanism will roost likely be present when a plug assembly is installed in a fuel assembly. The expected response o.'he plug assembly has been covered ab ove in Section II.

IV; NEUTRONICS EFFECTS

'he removal of part length rods has no impact on any physics information generated in the past for St. Lucie Unit l. The use of part length CEAs has been prohibited by Technical Specifications and they have been locked in the full out 'position during operation. The inbidllativn of CEA plug assemblies as described in Sections II and III will have no influence on the physics characteristics of the reactor.Section II states that the lowest portion of the plug assemblies will not be within several, inches of the top of the fuel. Therefore, operation with installed plugs will not invalidate any of the physics parameters con-tained in the Cycle 2 reload submittal.

V. ACCIDENT AN0 TRANSIENT ANALYSES A. Impact on Analyses of Design Basis Events A list of design basis events (DBEs) which the cui.rent plants are required to accommodate is presented in Table l. An assessment has been made to determine the impact of replacing the PLCEAs with "plug" CEAs on safety r'elated input data used in analyses of these

-DBE. It has been determined that none of the safety related input ,

data -is worse, than corresponding input data used in the Cycle 2 analyses when the PLCEAs are removed. Hence, operational thermal margins are not reduced below the design values, nor are consequences introduced which are more adverse than those previously ~eported in the'St. Lucie Unit 1, Cycle 2, license submittal.

B. Impact on Probability of Occurence A potential safety concern is that the probability of some event previously analyzed can be increased due to the replacement of PLCEAs with "plug" CEAs. No information exists which suggests that

the replacement of PLCEAs with "plug" CEAs increases the probability of any event previously analyzed.

C. Other Mal functions Hot Previously Analyzed No information exists which suggests that the replacement of PLCEAs with "plug" CEAs introduces a possibility for an accident or any malfunction of a different type than those previously analyzed.

Hence, it is concluded that the replacement of PLCEAs with plugs does not introduce the possibility of events not, previously analyzed.

D. Margin of Safety It is evaluated that the consequences of replacing the PLCEAs with "plug" CEAs does not reduce the margin of safety, as defined in the bases for applicable technical specifications.

E. -

Sugary The probability of occurrence of events has not increased and the consequences of these events remain within those reported in pre-vious analyses. The possibility of other types of accidents or malfunctions has not increased. Hence, the information presented in this report leads to the conclusion that Cycle 2 operation of St.

Lucie Unit 1, with the "plug" CEAs instead of PLCEAs, do s not pre-sent any danger to the health and safety of the public.:i any new technical iriformation is obtained which would change the con-clusions, such information will be reported in a timely manner.

TABLE I

. DESIGN BASIS EVENTS Control Element Assembly Withdrawal I

Boron Dilution Startup nf an Inactive Reactor Coolant Pump Excess Load Loss of Load Loss of Feedwater Flow Excess Heat Removal due to Feedwater Malfunction Reactor Coolant System Depressurization

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Loss of Coolant FIow Loss of AC Power Full Length CEA Drop Transients Resulting form Malfunction of One Steam Generator Part Length CEA Drop Part Length CEA Malpositioning

VI. TECHNICAL SPECIFICATION CHARGES As a result of removal of PLCEAs, the following Unit 1 technical specifications must be revised as indicated on the attached sheets.

l. Section 3.1.2.3 5.

6.

Bases 3/4.1.3 Definition 1.13

2. Section 3.1.3.3

3. Section 3.10.1 7: Section 4.2.2.3

4. Section 5.3.2 VII. COHCLUSIOi)S Based on the considerations discussed above, we have concluded that removal of the PLCEAs does not invalidate any of the safety analyses provided in the reload submittal. In addition, removal of the PLCEAs assures no CEA to guide tube interaction can take place in these fuel assemblies.

The public health and safety will not be endanagered by removal of the PLCEAs.

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