ML18227C687

Letter Transmitting Proposed Amendment to Facility Operating Licenses, Changing to Tech Spec Consisting of Revisions to OL Regarding Reactor Coolant System Pressure Temperature Limit

ML18227C687
Person / Time
Site:	Turkey Point
Issue date:	05/21/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-76-199
Download: ML18227C687 (46)

Text

U.S. NVCLI!AAAGGVLATOII,IMMISSION NIIC F<."AM 195 (2 76I I

NRC DISTRIBUTION rOII PART 50 b~VCI(ET MATERIAL OOCKGTN: I 5

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Miami, Flor'dia R.E. Uhrig OATE OF OOCVMENT 5-21-76

'ATE AECEIVEO 5-26-76 I LI LE TTEn 8OAICINAL QCOPV 1NOTOAIZEO UNC LASS IF I E 0 P AOP.

INPUT FOAM NUMOEA OF COPIES AECEIVGO

'i 40 I OESCAIPTION Ltr. notarized 5-25-76, Re. our 2-17-'76 ltr. Tran the following.......

ENCLOSUAE

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Proposed Amdt. to Facili.ty Operating Licenses Chanqe'.to Tech;'ppc.. consist'i'ng of revisions to OL'egarding-the i eactor coolant system pressure temperature limits'...W/Written

'.. Safety'Evaluation..;...

(3 Signed 5 37 Carbon Cys". Received)

ACKNOWI.

PLAIQ NAME:

Turkey Pt.

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SAFETY ASSIGNED AD:

BRANCH CIIXEF:

ear FOR ACTION/INFO RftiIATION ASSXGNEi;D AD:

BRANCII CIIXEF

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COLI.INS IIOUSTON PETERSON HILTZ IIELTEMES SKOVIIOLT LI'Dl 1 BAll I,'XC NSX AS I.13 ACAS J5 Mmzml~l.it f INTERNAL0 I;II~S 7

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IPPGLXTO OPERATING REACTORS STELLO.

OPERATXNG TECH XSENIIUT IIAO AER SCI RII'.NCI'.R RItlES ST.TI'. SAFETY Ec I'.Ng~

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6 JAnl{ Se<<ttnn Doekee C4ik FLORIDA POWER 8E LIGHT COMPANY May 21, 1976 L-76-199 Director of Nuclear Reactor Regulation Attention:

Mr. Victor Stello, Director Division of Operating Reactors U.

S. Nuclear Regulatory Commission Washington DC 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 s Light Company submits herewith three (3) signed originals and forty (40) conformed copies of a request to amend Appendix A of Facility Operating Licenses DPR-31 and DPR-41.

On February 17,

1976, by letter from George Lear to Robert E. Uhrig, your office requested that we review the reactor coolant system pressure-temperature limits in the Turkey Point Technical Specifi-cations to determine if they are in full compliance with 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements."

We have conducted such a review and find that appropriate Technical Spe-cification changes are necessary.

We are also proposing changes in response to an April 14, 1975 letter from George Lear to R.

E. Uhrig which requested that heatup and cooldown rates in the Turkey Point Technical Specifications be expressed as "'F in any one hour" instead of "'Fihr."

The intent is to increase operational flexibilityand decrease the possibility of a Technical Specification violation by more clearly specifying the time interval over which a temperature change is to be averaged.

The proposed changes are as described below and as shown on the accompanying Technical Specification pages bearing the date of this letter in the lower right, hand corner.

Page V

The List of Figures is revised to include new Figures 3.1-la through 3.l-ld, 3.1-2a through 3.1-2d, B3.1-1, and B3.1-2.

Pages 3.1-2 through 3.1-4 These pages are revised such that Specification 3.1.2.1 applies to Unit 3 only and'Specification 3.1.2.2 applies to Unit 4 only.

The heatup and cooldown rates are expressed in terms of "'F in any one HELPING BUILD FLORIDA

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Mr. Victor Stello page 2

hour" instead of "'F/hr."

The heatup and cooldown rate limits are based on test results obtained from the surveillance capsules most recently removed from the reactors.

Page 3.1-4 is revised such that the final part of Specification 3.1.2 no longer appears on that page.

Fi ures 3.l-la throu h 3.1-ld and 3.1-2a throu h 3.1-2d New reactor coolant system pressure-temperature limit curves are provided for Units 3 and 4.

The n'w curves are based on analysis of data obtained from the most recent reactor vessel surveillance capsules.

Figure 3.1-2 is deleted.

Pa es B3.1-1 throu h B3.1-4 Revised bases pages are included in support of the proposed amendment.

Fi ur'es B3.1-1 and B3.1-2 Two new figures are added to Bases Section B3.1.

The proposed changes meet the requirements of 10 CFR Part 50, Appendix G.

The proposed amendment has been reviewed and the conclusion reached that it does not involve a significant haz'ar'ds consideration; therefore, prenoticing pursuant to 10 CFR 2.105 should not be required.

A written safety evaluation is attached.

Very truly yours, Robert E. Uhrig Vice President REU:tg Attachment cc:

Mr. Norman C. Moseley Jack R.

Newman, Esquire

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STATE OF FLORIDA

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SS COUNTY OF DADE

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

That he is a Vice President of Florida Power 6 Light. Company, the Licensee 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 Licensee.

Robert E. Uhrig Subscribed and sworn to before me This dayof'976 Notar Public, in an for the County of

Dade, State of Florida N~TA~ PlTF)ll: 'KTW%'QF'FI,ORIDA AT LAROE SLY 'COMMISSION EXPIRES JIB

'34 WH 0 THRU, GENER4L'IIQNWtÃUNDERWRITSa My commission expires

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LIST OF FIGURES

~Pi use Title TECHNICAL SPECIFICATIONS

2. 1-1 Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation 2 ~ 1 2

Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation 3,l-la thru-ld Reactor Coolant System Pressure Limits:

Unit 3 Heatup and Cooldown 30 1 2a thru-2d Reactor Coolant System Pressure Limits:

Unit 4 Heatup and Cooldown 3 ~ 2 1 Control Group Insertion Limits for Unit 4, Three Loop Operation 3.2-1a Control Group Inser'tion Limits for Unit 4, Two Loop Operation 3 2 lb Control Group Insertion Limits for Unit 3, Three Loop Operation 3.2-lc Control Group Insertion Limits for Unit 3, Two Loop Operation 3 ~ 2 2 30 2-3 Required Shutdown Margin Hot Channel Factor Normalized Operating Envelope

3. 2-4 4.12-1 6.1-1 6.1-2 Maximum Allowable Local KW/PT Sampling Locations Offsite Orgaixization Chart Plant Organization Chart B3.1-1 Effect of Fluence and Copper Content on Shift of RT for Reactor Vessel Steels Exposed to 550'F Temperature B3.1-2 Fast Neutron Fluence (E>1MEV) as a function of Effective Full Power Years Target Band on Indicated Flux Difference as a Function of Operating Power Level B3.2-2 Permissible Operating Band on Indicated Plux Difference as a

Function of Burnup 5/21/76

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UNIT 3 2.

'RESSURE-TEMPERATURE LIMXTS l.

Unit 3 The Reactor Coolant System (except for the pressuriz'er) pressure and temperature shall be limited during heatup, cooldown, criticality (except for low power physics tests),

and inservice leak and hydrostatic testing in accordance with the limit lines shown on Figures 3.1-la through 3.l-ld.. Allowable pressure-temperature combinations are BELOW AND TO THE RXGEIT of the lines on the Figures.

Heatup and cooldown rate limits are:

a.

A maximum heatup rate of 100 'F in any one hour.

b.

A maximum cooldown rate of 100 'P,in any one hour.

c.

A maximum temperature change of >

5 'F in any one hour during hydrostatic testing operation above system design pressure.

The pressurizer pressure.and temperature shall be limited in accordance with the following:

d.

The pressurizer shall be--limited to a max'mum heatup or cooldown rate of 200 'P in any one hour.

e.

The pressurizer shall be limited to a maximum Reactor Coolant System spray water temperature differential of 320 'P.

With any of the above limits exceeded, restore the temperature and/or pressure within the limits within 30 minutes; determine that the RCS or pressurizer remains acceptable for continued operations or, if at

power, be in at least Hot Shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The reactor shall not be made critical unless the moderator temperature coefficient is zero or negative.

When the coefficient is greater than zero, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization.

These moderator temperature coefficient conditions do not apply to low power physics tests.

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!! ~fj", ! !iii ~ I ~ ~ I 'tj jao .roti o'1 ~ el!i es ~ iii:: ii!i 0 ~ ~ ~ 559 ~ 50 100 150 200 250 300 350 400 450 500 jNOl CATED TEMPERATURE ('F j FIGURE 3.1-2dTURKEY POPPET UNIT 4 REACTOR COOLANT COOLDOWN LIMITATIONSAPPLICABLEFOR PERIODS FRG))f 5 TO 10 EFFECTIVE FULL'POWER YEARS. I BASES FOR LIHITING CONDITIONS FOR OPERATION, REACTOR COOLANT SYSTEH 1. 0 erational Com onents The specification requires 'that a sufficient number of reactor coolant pumps be operating to provide coast down 'ore'ooling in the event that a loss of flow occurs. The flow provided will keep DNBR well above 1.30. When the boron concentration of the Reactor Coolant System is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Hixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if, at least one reactor coolant pump or one residual heat removal pump is running while the change i's taking place. The residual heat removal pump vill circulate the reactor coolant system volume in approximately one half hour. Each of the pressurizer safety valves is designed to relieve 293,330 lbs. per hr. of saturated steam at. the valve set (1) lt point. Below 350 F and 450 psig in the Reactor Coolant

System, the Residual 11eat Removal System can remove decay heat and thereby control system temperature and pressure.

If no residual heat were removed by any of the means available / the amouht of steam which could be generated at safety valve 1'ifting pressure would be less than the capacity of a single valve. Also, two safety valves have capacity greater than the (2) maximum surge rate resulting from complete loss of load. 2. Pressure/Tem erature Limits All components in the Reactor Coolant Syst: em are designed to withstand the effects of cyclic loads due to syst: em temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.;'7hz various categories of load cycles used for design purposes are provided in B3.1-1 5/21/76 f Section 4.1.5 of the FSAR. During startup and shutdown, the rates of temperature and. pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for prevention of brittle fracture. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location. The heatup analysis also covers the determination of pressure-temperature 1'mitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses axe additive to the pressure induced tensile stresses which axe already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined,. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling

location, each heatup-rate of int'crest= must be analyzed on an individual basis.

33.1-2 II The heatup limit curves are composite curves prepared by determining the most conservative

case, with either the inside or outside wall controlling, for any heatup rate up to 100'F per hour.

The cooldown limit curves are composite curves which were prepared based upon the same type analysis with the exception that the controlling location 'is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period. The reactor vessel materials have been tested to determine thei" initial RT ,, ~ Adjusted reference temperatures, based upon the fluence and copper content of the. material in question, are then determined. The heatup and cooldown limit curves include the shift in RT T at the end of the service period shown on the heatup and cooldown curves. 3 The actual shift in NDTT of the vessel material will be -'stablished'eriodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiat'ion surveillance specimens installed near the inside wall of'he reactor. vessel in the core area. Since the neutron spectra at the irradiation samples has a definite relationship to the spectra at the vessel inside

radius, the measured transition shift for a sample can be related with confidence to the adjacent section of the reactor vessel.

The heatup and cooldown curves must be ]33. 1-3 5/21/76 'I recalculated when the ARTNDT determined from the surveillance capsule is different irom the calculated ARTNDT for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.2-1 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements 133. 1-4 5/21/7C) 0.201, COPPER CASE. 0.255 MELO LtD OYER 0.25$ COPPER

SASE, 0.20$

MELO Q.205 COPPER BASE. O. I5$ -ME ~ ~ O.lSS COPPER AASE. O.[05 MELO ~ 0 f0/ COPPER BASE. 0'.05$ MELO e y ~ 'I q g S ~O~O Effect of Fluence'and Copper Content on Slrift of RTN>T for Reactor Vessel Steels Exposed to 550 'P Temperature Pigure 83.1-1 0 s ( ~ ~ ~ ~ ~ ~ ~ ~ \\ I ( I ( Suc ace ~. ':=.='.i. ~ ~ I 1/4 T ~ ( s s s I s ( ~ Its I I I I I! s ~ ' s I I s s I l ( s ~ ~ ~ 1 s s ~ 1 s ~ s ~ s I s I I I s I s I I I ' 1 I s 1 t I s ~ ~ ~ s I ~ ~ s I I ~ ~ s 1 i 1 ~ ~ s ( ~ s ~ I i I 1!1'I s I I I!,! I I I I! I I ~ ! I! I >> s L ~ ~ ~ ~ ~ O ~, I I ~, I, ( ~ ( ~ ( ~ ~ ( s s s s ( s, ( s s ~ ~ s ~ ~ I ~ I 1 s ~ s ( s ~ 1 ~ 1 I I ( I s t +i I I I s J s i+/ I s s s ~ I ~ s s s ( ~ I I I !'I I I I I ( I (. III( I !i! II ~ I I I I I I I I I I! I l lsl i I I I ':! I ~' s ~ s t ~ ~t ~ I s ~ ~ s s 1 ~ I s s ~ t ~~~( I ~ s ( ( ( ~ s 1 ~ s ~ I ~ t ' s ~ \\ ( ~ 1 ~ ~ ( ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ Q st~1 s s I 1 s ~ s I 1017 I ISi l.g. L !! I! !r! I I I 5 10 15 20 25 30 Past Neutron Pluence (H > l MEV) as a Punction of Effective Pull Power Years Pigure D3.1-2 I SAFETY EVALUATION Introduction Limits for reactor coolant temperature and pressure are given in Turkey Point Technical Specifications 3.l.2. This submittal proposes changes to the specification to bring it into compliance with 10 CFR 50, Appendix G. The changes are based on test results recently obtained'rom the latest reactor vessel surveillance capsules from Unit 3 and 4. Discussion Since the physical properties of reactor vessel materials can be affected by neutron irradiation, these materials are subject to a surveillance program. Reactor vessels are designed to accommodate surveillance capsules which contain materials used in the manufacture of the vessels. Irradiated surveillance materials are periodically analyzed and the data used to revise operating pressure-temperature limits. Turkey Point FSAR Section 4.4 conta'ns detailed infor-mation on the reactor vessel surveillance program. Current Technical Specification 3.l.2 provides a method for revising heatup and cooldown curves based on the radiation exposure of surveillance specimens. 3Iowever, the Turkey Point reactor vessels were manufactured before the publication of 10 CFR 50, Appendix G and ASME Bo'ler and Pressu e Vessel

Code, Sectioii IIX, Appendix G, both of which provide information on reactor vessel design and calculation techniques based on surveillance specimen data.

Now i=hat Turkey Point surveillance capsule data have been reviewed and analyzed in accordance with Appendix G and the ASNE Code, we find it necessary to revise the pressure-temperature curves in the Tecnnical Specifications. The new limit curves for normal heatup and cooldown of the primary reactor coolant system were calculated using i=he methods discussed in the proposed Bases pages. Conclusion Based on these considerations, (l) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and does not reduce the'argin of safety as defined in the basis for any technical specification, therefore, the change does not involve a significant hazards consideration, (2) there is reasonable assurance thai. the health and safety of the public will not be endangered by operation in the proposed

manner, and (3) such activities will be conducted in com-pliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

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