L-78-263, Submit Proposed Amendment to Facility Operating Licenses to Delete Requirements for Part-Length Rod Cluster Control Assembles from Tech Spec

Submit Proposed Amendment to Facility Operating Licenses to Delete Requirements for Part-Length Rod Cluster Control Assembles from Tech Spec

ML18227B284
Person / Time
Site:	Turkey Point
Issue date:	08/09/1978
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-78-263
Download: ML18227B284 (32)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS DISTRIBUTION FOR INCOMING MATERIAL -2 l251 REC: STELLO V ORG: UHRIG R E DOCD ATE: 08/09/78 NRC FL PWR 5 LIGHT DATE RCVD: 08/17/78 DOCTYPE: LETTER NOTARIZED: YES COPIES RECEIVED

SUBJECT:

LTR 3 ENCL 40 FORWARDING LIC NOS DPR-31 5 41 APPL FOR AMEND: TECH SPEC PROPOSED CHANGE i

CONCERNING 'REMOVAL OF THE PART-LENGTH ROD CLUSTER. CONTROL ASSEMBLIES AND INSTALLATION OF THIMBLE PLUG ASSEMBLIES IN LOCATIONS PREVIOUSLY OCCUPIED BY THE PLRCCA "S... NOTARIZED 08/09/+Q e

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PLANT NAME: TURKEY PT 03 REVIEWER INITIAL: XJM TURKEY PT N4 DISTRIBUTER INITIAL:,g+OJ DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.

(DISTRIBUTION CODE A001)

FOR ACTION:, BR CHIEF ORBN1 BC++W/7 ENCL'NTERNAL:

REG FILE~~W/ENCL NRC PDR>>W/ENCL OELD4+LTR ONLY kiANAUER++W/ENCL CORE PERFORMANCE BRw+W/ENCL AD FOR SYS Sc PROJ++W/ENCL ENGINEERING BR+<M/ENCL REACTOR SAFETY- BR~>W/ENCL PLANT SYSTEMS BR++W/ENCL EEB~~W/ENCL EFFLUENT TREAT SYSwwW/ENCL J. MCGOUGH+~W/ENCL EXTERNAL: ,, LPDR S MIAMI > FL++W/ENCL TERA>+W/ENCL'S I C44 W/ENCL ACRS CAT B'++W/16 ENCL ffff@4fSS@f@@ff@444f0%%%44444f4%444f%44fffOSSfff@40 4 CHECK NBR: 54'54 AMOUNT: 44> 000. 00 f CHECK AND COPY OF TRANSMITTAL 'LTR ADVANCED f

f TO W. MILLER (LFMB) (08li7l78) UPON,RECIEPT

'444%$ 4440444040444f@f%@f@0$ f 0f@40$ $ 440f 404404%4f @4 DISTRIBUTION: LTR 40 ENCL 39 CONTROL NBR: 782190320 SIZE: 2P+13P

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'gg[gg5 ÃIXEiI'K I A POWER & LIGHT COMPANY August 9, 1978 L-78-263'irector of Nuclear Reactor Regulation Attention: Mr. Victor Stello, Jr., Director Division of Operating Reactors U. S. Nucl'ear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stello:

Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-.41 In accordance with 10 CFR 50.30, Florida Power 8 Light Company:( FPL) submits herewith three (3) signed originals and forty (40) copies of a request to amend Appendix A of Facility Operating Licenses DPR-31 and DPR-41.

The purpose of this proposed amendment is to delete the requirements for Part-Length Rod Cluster Control Assemblies (PLRCCA's) from the Technical Specifications for Turkey Point Units 3 and 4. Because of the requirement to maintain the PLRCCA's fully withdrawn and non-scrammable, plant operation at full power is not allowed with the PLRCCA's in the core. FPL plans to remove the PLRCCA's at Turkey Point Units 3 and 4 during their respective refueling outages. Thimble Plug Assemblies will. be installed in to the locations previously occupied 'by the PLRCCA's 'to preserve the current dynamic operating characteristics of the Reactor. We request that our proposal be approved to support startup following the upcoming refueling of Turkey Point Unit 4.

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

A footnote, specifying that any reference to part-length rods no longer applies after the part-length rods are removed from the reactor,, has been added to the following pages:

Page 3.2-1

'Page 3.2-2 Page 5.2-1 Page 5.2-2 Page B2.1-2 782190320 Page B3.2-la Page B3.2-2 Page B3.2-6 Page B3.2-7 l'EOPLE SERVING PEOPLE

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I Director of Nuclear Reactor Regulation Page Two In accordance with the cri'ter'ia stated in 10 CFR 170.22, FPL has determined that 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 PLRCCA's has been requested by a number of Westinghouse plants.

The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 8 Light Company Nuclear Review Board.

They have concluded that it does not involve an .unreviewed safety question.

In addition, removal of part length rods has been approved for several operating reactors. A Safety Evaluation is attached.

Very truly yours, Robert E. Uhrig Vice 'President REU/thAS/RL/cpc Attachment cc: Nr. James P. O'eilly, Region II Robert Lowenstein, Esquire

41 L

CONTROL AND POWER DISTRIBUTION LIMITS distribution limits.

~Ob ective: To ensure (1) core subcriticality after a reactor trip, (2) a limit on potential reactivity insertions from a hypo-thetical control rod ejection, and (3) an acceptable core power distribution during power operation.

1. CONTROL ROD INSERTION LIMITS

a. Whenever the reactor is critical, except for physics tests and control rod exercises, the shutdown control rods shall be fully withdrawn.

.b. For Unit 4, whenever the reactor is, critical, except for physics tests and control rod exercises, the control group rods shall be no further inserted than the limits shown on Figure 3.2-1 for three loop operation and on Figure 3.2-1(a) for two loop operation.

c. For Unit 3, whenever the reactor is critical, ex-cept for physics tests and control rod exercises, the control group rods shall be no further in-serted than the limits shown on Figure 3.2-1(b)'or three loop operation and on Figure 3.2-1(c) for two loop operation.

d. The Unit 4 consol rod insertion limits shown on Figure 3.2-1 and the Unit 3 control rod insertion limits shown on Figure 3.2-1(b).may be revised on the basis of physics calculations and physics data obtained during startup and subsequent operation.

e.* Part length rods shall not be permitted in the core except for low power physics tests and for axial offset calibration tests performed below 75% of rated power.

Any reference to part-length z'ods no longer applies after the part-length.

rods are removed from the reactor.

3.2-1

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f; Except for low power physics tests, the shutdown margin with all'owance for a stuck control rod shall'xceed the applicable value shown on Figure 3.2-2 under all steady-state operating conditions from zero to full power, including effects of axial .

power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (540 F) if all control rods were tripped, assuming that the highest worth control rod remained fully withdrawn,

~ and assuming no changes in xenon, boron con-centration or part-length rod position.

g. During physics tests and control rod exercises, the insertion limits need not be met, but the wequired shutdown margin, Figure 3.2-2 mast be maintained or exceeded.

'2o, HXSALXGNEDi'CONTROL ROD If a part length. or full length control rod's more than 15 inches out of. alignment with its. bank, and is not corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> power shall be reduced so

as not to exceed 75Z of interim power for 3'oop or 45%

of interim power for two loop operation, unless the hot channel factors are shown to be no greater than allowed

, by Section 6a of Specification 3.2.

3' ROD DROP'XME

. ~ '..The dro P time of each control rod shall be no g reater than 1.8 seconds. at full flow and operating temperature r'

,"from the beginning of rod motion:to dashpot entry.

4~ ~ INOPERABLE CONTROL RODS a., No-more than one inoperable control rod shall be permitted during sus'tained power operation, except it shall not be permitted if,the rod .has a .potenti'al Any reference to part-length rods no longer applies after the part-length rods are removed from the reactor.

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5. 2 REACTOR REACTOR CORE The reactor core contains approximately 71 metric

.tons of uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encap-sulated in Zircaloy 4 tubing to form fue1 rods The reactor cork, is made up of 157 fuel assemblies.

.Each fuel assembly contains 204 fuel rods.

2. .The average enrichment of the initial core is a

. mominal 2.50 weight per cent of U-235. Three fuel enrichments are used in the initial core. The ihighest enrichment is a nominal 3 10 weight per cent

. of U-235.

.3. Reload'fuel will be similar in design to the initial core. The enrichment of reload fue1 wiD be no more

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.. Chan 3.5 weight per cent of U-235.

13urnable poison rods are incorporated in the initia1 core. There are 816 poison rods in the form of 12-

.mod c1usters, which are located in vacant rod cluster control guide tubes. The burnable poison rods

~onsist of borated pyrex glass clad with stainless steel.

There are 45 Lull-. length.RCC assemb1ies .and 8 partial-

.length RCC assemblies in the reactor core. The Xull-

• after. the part-length Any reference to part-length rods no longer. applies rods are removed from. the reactor.

5.2-1

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-D length RCC assemblies contain a 144 inch 1ength of

. silver-indium-cadmium alloy clad with the stainless steel. The partial-length RCC assemblies contain a 36 inch length of silver-indium-cadmium alloy with remainder of the stainless steel sheath filled

'he

. with F1203.

6. Up to 10 grams of enriched fissionable material may be used either in the core, or available on the site, in the form of fabricated neutron flux detectors for

.the purposes of monitoring core neutroa.flux.

REACTOR COOLANT SYST&f 3..'e design of the Reactor Coolant System complies

. with the code requirements.

2. All piping, components and supportiag structures of the Reactor Coolant System are designed to Class I requirements, and have been designed to withstand:

a. The design seismic ground acceleration, 0.05g acting in the horizontal and 0.033g actiag in the

-.wertical planes simultaneously, with stresses

'aaaiatained within code allowable working stresses

b. The maximum potential seismic ground acceleration 9 I5g, acting ia the horizontal and 0.10g acting M the vertical directions simultaneously with .

mo loss of function.

.3. The nominal liquid volume of the Reactor Coolant

.: System, at rated operating coaditions, is 9088 cubic

-Eeet.

Any reference to part-length rode'o longer applies after the part-length'ods are removed from the reactor'.

5&2 2

il Use of these factors results in more conservative curves than the Interim Limits.were used.

these limiting hot channel factors are higher than those ca3.

culated at full power for the range from all control rods fully withdrawn to maximum allowable control xod insertion. The con-trol rod insertion limits are covered by Specification 3.2.

'.Slightly higher hot channel factors could occur at lower power 1evels because additional control rods are in the core. How-ever~ the control xod insertion limits dictated by Figure 3'-1 ensure that the DNBR is sways greater at partial power

.&an at full power.

Me hot channel factors are also sufficiently large to account for the degree oi malpositioning of part-length rods that is allowed before the reactor trip set points axe reduced and

.rod withdrawal Mock and 'load runback may be required. ( ) Rod withdrawal block and load runback occur before reactor trip setpoints are reached.

Zhe Reactor Control and Protection Sy'tem is designed to prevent any anticipated combination of transient conditions that would result in a DNBR of less than 1.30. (2)

Feference (1) ,FSAR 3.2.2 (2) CESAR u .1.1

. Any reference to part-length rods. no longer applies after th'e part-length rods are removed'rom the. reactor.

Z2. 1-2

The overlap. between successive control banks is allowed because the control rod worth is lower near the top and bottom of the core than in the center.

Positioning of the part-length rods is governed by the requirement to main-tain the axial power shape within specified limits or to accept an automatic cutback of. the overpower bT and overtemperature bT set points (see Specification 2.3). Thus, there is no need for imposing a limit on the physical positioning of the part-length rods.

Any reference to part-length'ods no longer applies after the part-length rods are. removed from the reactor.

B3.2-1a

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The various bank, that is, cobol rod banks are each to be m~d as a

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with all rods& the bank within one step (5/8&eh) of the bank position.

The control system is designed to permit individual rod movement for test purposes. Position indication is provided by two methods: a digital count of actuating pulses which shows the demand position of the banks and a

&near position indicator (LVDT) whi'ch indicates the actual rod position.

The relative accuracy of the linear position indicator (LVDT) is such that,

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with the most adverse error, an alarm will be actuated if any two rods

-within a bank deviate by more than 15 inches In the .event that an LVDT is not in service, the effects of a malpositioned control xod are observable

'.on nuclear and process information displayed in the contxol room and $ y

-.core. thermocouples and in-core movable detectors. Complete xod misalignment

~: (part-length or full-length control rod 12 feet out of alignment with its

-.hank).does not result in exeeding core limits in steady-state operation at

=

rated power. If the condition cannot be readily corrected, the specified

-xeduction in power to 75% (3 loop) or 45% (2 loop) will insuxe that design 0 ~,l anargins .to core limits will be maintained under both steady-state and an-le

'Mcipated transient. condi.ti.ons. The 8-hour permissible limit on xod mis-m1ignment is short with respect to the probability of an'independent ac-

.-cident. The 24-hour period ensures that no signi.ficant buxnup effects

~ould be caused by the inserted xod.

Zhe specified 'rod drop time is consistent with safety, analyses that have

.been perf ormed. (1)

'The In-Core Instrumentation has five drives with detectors each of which

' has -.

~ .ten thimbles assigned. (3) This provides broad capability for'detailed flux swapping.

'The ion chambers lo'cated outside the reactor vessel measure flux distribution.

=at t:he top and bottom of the core. Core traverses in a few of the in-core

'nstrument paths will establish that the fixed flux measurement equipment

.Cs properly calibrated.

.-Operating experience has established that the flux measurement system is of

-a reliable design, and that the 10% load reduction, in the event of xe-

-.calibration delay, is ultra cons rvative compensation.

References:

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(1) FSAR Section 14

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(2) FSAR Section 7.2 (3) FSAR Section 7.t' Any reference to part-length rods no longer applies after the part-length rods are: removed from the reactor.

B3. 2-2

Flux Differen~(Ag) and a reference value whi orresponds to the full 0

design power e~ilibrium value of Axial Offset vial Offset ~ Af(fractional power). The reference value of flux difference varies with power level and .

burnup but expressed as axial offset it varies only with burnup.

The technical specifications on power distribution control assure that the F upper bound envelope of 2-22~times Figure 3.2-3 is not exceeded and xen'on q

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 difference is, determined as follows.

At any time that equilibrium xenon conditions have been established, the in-dicated flux difference is noted with part length+ rods withdrawn from the core and with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal rated power operating position appropriate for the time in life.

Control rods are usually withdrawn farther as burnup proceeds). This value, divided by the fraction of design power at which the core was operating is the design power value of the target-flux difference. Values for all other core power 1'cvels are obtained by multiplying tne design power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of

+5% AI are permitted from 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 difference every rated power month. For this reason, methods are permitted by Item 6c of Section 3.2 for updating the target flux differences. Figure B3.2-1 shows a typical construction of the target flux difference band at BOL and Figure B3.2-2 shows the typical variation of the full power value with burnup.

Strict co'ntrol of the flux difference (and rod position) is not as necessary during part power operation. This. is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict con-trol at part power. Strict control of tne flux difference is not possible during certain physics tests or during the required, periodic excore calibra-

• For steam generator tube plugging in excess of 10%, this value becomes 2.20.

+

Any reference to part-length rods no longer applies after the part-length rods are removed from, the reactor..

B3.2-6

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tions which require larger flux differences than permitted. Therefore, the specifications on power distribution control are not applied during physics- tests or excore calibration. This is acceptable due to the extremely low probability of a significant accident occurring during these operations.

In some instances, of rapid plant power reduction automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distribu-tion sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power'ithin the target band. However, to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band. This ensures that the resulting xenon, distributions are not significantly different from those resulting from operation within the target band. The instantaneous conse-quences of being outside the band, provided rod insertion limits are observed, is not worse than a 10 percent 'increment in'peaking factor for flux difference in t'e range +14% to .14% (+13.% to -11% indicated) increas-ing by, + 1% for each 2% decrease in rated power. Therefore, while the deviation exists, the power level is limited to 90% of design power or lower depending on the indicated flux difference.

If, for any reason, flux difference is not controlled within the +5% band for as long a period as one hour, then xenon distributions may be significantl ican y changed and operation at 50% of design power is required to protect against potentially more. severe consequences of some accidents.

I As discussed above, the essence of the procedure is to maintain the xenon

'istribution in the core a's close.to the equilibrium full power con'dition as possible. This can be accomplished without part length rods by using the boron system to position the full length control rods to produce the required indicated flux difference.

Any reference to part-length rods no longer applies after the part-length rods are removed from the reactor.

B3.2-7

SAFETY EVALUATION REMOVAL OF PART LENGTH CONTROL ELEMENT ASSEMBLIES

'FROM TURKEY 'POINT'NITS NO. 3 'AND 4 Z. 'NTRODUCTZON This report provides information to justify plant operation following the removal of the part-length rod cluster control assemblies (PLRCCA's). Plant operation 'at power is currently not allowed with PLRCCA's in the "core.

Thimble plug assemblies will be installed into the locations previously occupied by the PLRCCA's. 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 removal of the PLRCCA's was performed.

if just ZI. THIMBLE PLUG 'ASSEMBLY MECHANICAL DESZGN The Thimble Plug Assembly, which will be inserted into locations previously occupied by PLRCCA's, consists of a flat base plate with short rods suspended from the bottom surface and a spring pack assembly. The twenty short rods, called thimble plugs, project into the upper ends of the guide thimbles to red'uce the bypass. flow area. Fuel

-..assemblies without control rods, burnable poison rods, or source rods use identical devices. Similar short rods are also used on the source assemblies and fuel assembly guide thimbles. At installation in the coze, the thimble plug assemblies interface with both the upper core plate and with the fuel assembly top nozzles by resting on the adapter plate. The spring pack is compressed by the upper core plate when the upper internals assembly is lowered into place.

Each thimble plug is permanently attached to the base plate by,a nut which is locked to the threaded end of the plug by a pin welded to the nut.

All components in the thimble plug assembly, except for'the springs, are constructed from type 304 stainless steel.

The springs are wound from Inconel x-750 for corrosion resistance and high strength.

These thimble plugs will effectively limit bypass flow through the rod cluster control guide thimbles in the fuel assembl'ies from which the PLRCCA's have been removed, just as they currently limit bypass flow in those assemblies which

~ Og do not contain control rods, source rods, or burnable poison rods.

III. THERMAL HYDRAULIC EFFECTS A. Thermal Effects Physics analysis, as well as incore monitoring, indicates that there, will be no adverse effect of the plug assemblies on the core power distri-bution. Since the plugged fuel assemblies have no adverse effect on the design core flow distri<<

bution, calculated core thermal margin will be unaffected.

B. Hydraulic Effects Hydraulic .aspects were considered with respect to the installation of the thimble plug assemblies.

Since the pl'ug assemblies are already extensively used in existing fuel assemblies with no adverse effects, it can be concluded that there will be no adver'se effects from the installation additional thimble plugs. of'hese IV. 'EUTRONICS EFFECTS The removal of part length rods has no impact on any physics information generated in the past for Turkey Point No. 3 & 4..

The use of part length RCCA's has been prohibited by Technical 'Specifications and they have been locked in the full out position during asoperation. The installation of described in Sections II and III thimble .plug assemblies will .have no influence on the physics characteristics of will the reactor. The lowest portion of the plug assemblies not be within several inches of the top of the fuel. There-fore, operation with installed plugs will not invalidate any of the physics parameters.

V. ACCIDENT AND TRANSIENT ANALYSES Based on foregoing discussion, the following conclusions relating to accident and transient analyses can be reached.

A. Impact on Probability of Occurrence A potential safety concern is that the probability of'ome event previously analyzed can be increased due to the replacement of PLRCCA's with thimble plug assemblies. No information exists. which suggests .that the replacement of PLRCCA's with thimble plug assemblies increases the probability of any event previously analyzed.

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.0 B. Other Malfunctions, Not Previously Analyzed No- information exists which suggests that the-replacement of PLRCCA's with thimble plug assemblies introduces 'a possibility for, an accident or any malfunction of a different type than those prev'iously analyzed. Hence, it is concluded that the replacement of PLRCCA's,

'with plugs .does: not introduce the possibility of events not previously'nalyzed.

C. Margin of Safety It is evaluated that the consequences of replacing the PLRCCA's with thimble plug assemblies does not'reduce the margin of safety, as defined in the bases for applicable technical specificati'ons.

D. Summary The probability of occurrence of events has not increased and the consequences of these events remain within those reported in previous 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 operation of Turkey Point No. 3 6 4 with the thimble plug assemblies instead of PLRCCA-'s does not present any danger to the health and safety of the public.

VI. TECHNICAL SPECIFICATION CHANGES's a result of removal of PLRCCA's, the following Unit 3 a 4 technical specifications must be revised as indicated on the attached sheets.

1. Section 3.2.1

2. Section 3'.2.2

3. Section 5.2

4. Bases 2.-1

5. Bases 3.2

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STATE OF FLORIDA )

) ss COUNTY OF DADE )

Robert E. Uhrig, being first duly sworn, deposes and says:

That he is a Vice President of Florida Power & Light Company, the Applicant herein:

. That he has executed the foregoing document; that the statements made in this said document are true and correct to the best of his knowledge, information and belief, and that he is authorized.

to execute the document on behalf of said Applicant.

Robert E. Uhra.g Subscribed and sworn to before me this g~ day of j ~K 19 7V NOTARY-."PUBLI , in and for the County of Dade, State of Flor da NOTAI ABLIC STATG OF FLOfViOA =E tAAGG MY COMNSSIOM WPihES MAACH 27, 1~

gy COmmj.SSj.On eXPj.reS: enVOrO ~W> V~'~AoO er'iO"lG AAFb~

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