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Request to Amend Appendix a of Facility Operating Licenses DPR-31 & DPR-41. Proposes Changes in Response to 04/14/1975 Letter from George Lear Which Requested That Heatup & Cooldown Rates in Technical Specifications .
	ML18227D854
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: ML18227D854 (42)
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INI'VT I'OIIM NUMOEA OF COI'IES AECEIVCO 40 I OESCAIPTION ILtr. notarized 5-25-76, Re. our 2-17-76 ltr. Tran ithe following.......

LNCLOSVAC Proposed Amdt. to Facility Operating'icenses Chanqe to Tech.

Sppc.. consisting of revisions to OL regarding the reactor caolant system pressure temperature limits...w/Written Safety'Evaluation..;...

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~ F.O. ROX 013100, FhLAML, FLORIDA 33101 FLORIDA POWER 8I 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:

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~ISSIOH MelI Se440 lj'e:

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 ln accordance with 10 CFR 50.30, Florida Power

& Light Company submits herewith three (3) signed originals and forty (40) conformed

'opies 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'etermine 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're 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 "'F/hr."

The intent is to increase operational flexibilityand decrease the possibility of a Technical Specification violation by more clearly specifying the time interval over wh'ich 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.

3

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

Pa es 3.1-2 throu h 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 86 HELPING BUILD FLORIDA

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 3.1-la throu h 3.1-1d 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 new curves are based'n 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 ures 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. hazards 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 r

cc:

Mr. Norman C. Moseley, Jack R.

Newman, Esquire

STMZ OP FLORIDA

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SS COUNTY OP DK)E

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Robert Ee Uh&gg bGLng erst 4uly QUorng deposes and says:

That he ia a Vice President of Flora.da Power 5 Xight Company, Q 'the Licensee horeint That he haa executed the foregoing documents that the;.~

statements made in thea said document are true and correct to tho beat of hio knowledge, information and belief, and that ho Xs authorised to execute the document on behalf of said Xicensee.

Robart E. Uhrig Subscribed and avorn to before mo Thea A.

day oE 1976 Notary ubl c n anO Nor t o County oR DadeP tata << ~~o<<da NdTAA ~"<<<<;,'IvF of fII'".'PA n" I~"~

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LIST OF FIGURES

~Pi use Title TECHNXCAL SPECIFXCATIONS 2 ~ 1 1 Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation

2. 1-2 Reactor Core Thermal and Hydraulic Safety Limits, Operation Two Loop 3.1-la thru-ld Reactor Coolant System.Pressure Limits:

Unit 3 Heatup and Cooldown 3\\ 1 2a thru-2d E

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-la Control Group Insertion Limits for Unit 4, Two Loop Operation

3. 2-3.b Control Group Xns rtion Li...its for Unit 3, Three Loop Oper"tion 3.2-1c Control Group Insertion Limits for Unit 3, Two Loop Operation 302-2 3 ~ 2 3

3. 2-4 Required Shutdown Margin Hot Channel Factor Normalized Operating Envelope Maximum Allowable Local Kl</FT 4.12-1 Sampling Locations

6. 1-1

6. 1-2 Offsite Organization 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>lMEV) as a function of Effective Full Power Years B3. 2-1 Target Band on Indicated Flux-Difference as a Function of Operating Power Level B3.2-2 Permissible Operating Band on Indicated Flux Difference as a

Function of Burnup 5/21/76

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UNIT 3 2.

PRESSURE-TEMPERATURE LIMITS 1.

Unit 3 The Reactor Coolant System (except for the pressurizer) 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.l-la through 3.l-ld.

Allowable pressure-temperature combinations are BELOW AND TO THE RIGHT of the lines on the Figures.

Heatup and cooldown rate limits are:

a.

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

b.

A maximum cooldown'ate. of 100 'F in any one hour.

c

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A maximum temperature change of >

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

The pressurizer pressure and temperature shallbe limited in accordance with,the following:

d.

The pressur'zer, shall be limited +o a maximum heatup or cooldown rate of 200 'F in any one hour.

e.

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

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 f:o low power physics tests.

3.1-2 5/21/76

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UNXT 4 2.

Unit 4 The Reactor Coolant System (except for the pressurizer) pressure and temperature shall be limited during heatup, cooldown, criticality (except for'ow power physics tests),

and inservice leak and hydrostatic testing in accordance with the limit lines shown on Figures 3.1-2a through 3.1-2d.

Allowable pressure-temperature combinations are BELOW AND TO THE RXGHT 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 'F in any one hour.

c.

A maximum temperature change of a 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 maximum heatU'p-or cooldown rate of 2QQ 'F in an>> one hour.

e.

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

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Cold Shutdown. within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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.

3.1-3 5/21/".6

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LEAKAGE a

Any reactor coolant system leakage indication in excess

.of 1 gpm shall be the,subject of an investigation and

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evaluation initiated within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months " of the. indication (ex.

I water inventory changes, radiation level increases, visual or audible indication).

A leak shall be assu'med to exist I

tilit is determined that no unsafe condition exists

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. end that the indicated leak cannot be sub -tantiated.

Leakage of reactor coolant through reactor,.pump seals end system valves.to connecting closed systems from which

-coolant can be returned to the reactor. coolant system sh~

lnot be -considered as leakage except that such losses sha11

aot exceed 30Igpm.

Xf a reactor coolant system leakage indication is

-.proven real, and. is not evaluated as safe, or ex-ceeds 10 gpm, reactor shutdown shall be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the initial indication.

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Xf reactor coolant leakage exists through a fault in the system boundary that cannot be isolated (ex. vessels, giping, valve bodies) the reactor shall be shutdown and cool dotm to cold shutdown shall be initiated within

'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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jc The safety evaluation shall consider the source I

4 end magnitude of the leak, rates of change'.of

. detection variables, and if shutdown is required

this evaluation shall be used to determine shut-

'down rates and conditions.

A written log of the

~ection taken shall be made as soon as practicable.

.The evaluation shall assure that no potent9.al gross leak is developing and that: potential re-3.case of activity will be within the guidelines

'of 10CFR20.

3.1-4 5/21/76

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~ ~ Ei: jf ~II '3?i ittj ~4 l}ia lhi IIII ~ ~ ~ ~ ~ ~ ~ ~ ie 200 250 300 350 400 450 '00 )NDj CATED TEMPERATURE - ('F ) FIGURE3.1"2c TURKEY POPiET UNIT 4 REACTOR COOLANT HEATUP LEMITATIOiNSAPPLICABLE FOR PERIODS FROM 5 TO 10 EFFECTIVE FULL POWER YEARS. 100 150 ~ ~ ",'I: 550 4 ~ ~ 4 ~ 4 ~ ~ s> 4
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~ ~ ~ ~ I ~ eel I ~ ~ ti"sfft); ~ ~ ~ e e a! I 200 ~~ ~ ~ - et ~ ~ s ~ h ~. .")if e -I ttt: tfI) ij I ~if)i Iijii:if! llge00 !.I! jfii iit 1!t'-)atIje'f ~ 1!!l ft')I'ji I
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et Il.t Pl! t-;fi ,t 11a) ll i atajlj gÃbTfj ~ ~ ~ ~ II::i IN! 0 ~ 1:i 550 50 100 150 200 250 300 350 400 450 500 }jh)DlCATEO TEMPERATURE ( F FIGURE 3.1-2!ITURKEY POBsjT UNIT 4 REACTOR COOLANT COOLDOWN LIMITATIONSAPPLICABLE FOR PERIODS FROM e5 TO 10 EFFECTIVE FULLPOWER YEARS. 0 e ~ ~ c a ~ ' e ~ ~
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BASES FOR LIMITING CONDITIONS FOR OPERATION, REACTOR COOLANT SYSTEM 1. 0 erational Com onents The specification requires that a sufficient number of reactor coolant pumps be operating to provide coast down core cooling in the event that a loss oi 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. Mixing 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 is taking place. The residual heat removal pump will circulate the reactor coolant system volume in approximately one half hour. Hach of the pressurizer safety valves is designed to relieve 293,330 lbs. per hr. of saturated steam at the valve set
point, Below 350 F and 450 psig in the Reactor Coolant
. (1) Syst: em, the Residual lieat 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 amount oi steam which could be generated at safety valve lifting 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 irom complete loss of load.
2. Pressure/Tem erature Limits All components in the Reactor. Coolant System are designed I to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are pr6vided in B3.1-1 5/21/76 1
Section 4.1.5 of the PSAR. 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 i.nduced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of a11 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 limitations 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 are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the raj:e 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 oi the vessel becomes the stress controlling
location, each heatup rate of interest must be analyzed on aIi individual basis.
33.1-2 5/21/76
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'P '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 tempexature at the end of the service period. Thc reactor vessel materials have been tested to determine their 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 at the end of the service period shown on the heatup and cooldown curves. The actual shift in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accoxdance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples has a definite xelationship,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 83.1-3 5/21/76
recalculated when the bRT D, determined from the surveillance capsule is different from the calculated ARTIST 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 CPR 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 CPR Part 50. The limitations imposed on pressurizer heatup and cooldown and spray water temperature diiferential are provided to assure that the pressurizer is operated within the design criteria assumed ior the fatigue analysis performed in accordance with the ASMH Code requirements B3.1-4 5/21/76
200 0.30% COPPER QASE. Q.25$ MELO A%0'OVER 0.25$ COPPER aasE. 0.20$ uELO Q.P.0$ COPPER BISE. O.le WE g~n V f00 o. gQ 60
0. t S8 OOP P ER 3LSE. 0, )OX VE'LO 0-lOg COPPER BASE. 0.05$
tfELO tO'6 Effect of Pluence and Copper Content on Shift of RTNDT for Reactor Vessel Steels Exposed to 550 '1'emperature Figure 33.1-1 5/21/76
1 s s 1 'I 1 ~ y ~ ~ I ~ S ~~ ~ 1/4 T.=- y ~ ~ 1 s 1 ~ I ' 1 I y ~ I I y 1 I I ~ s y ~ 1 1 ~ ~ ~ s s 1 I 1 I I I 1 I 1 ~ ~ I ss I 1 I s s I 1 I I ~ Illy I 1 ~ ~ I ~ ~ ~ I I l ~ ~ - ~ --} I I ~ I ) II I 1 I I }. } I I I I I I I I ~ I 1 I ~ ' I y <<+I( y 1 ~ J' I 1 ~ 0 I y 1 ~ I' s 1 I y s s ~ I I ~ 1 1 ~ ~ ~ 1 11 I I I ~ s ~ 1 1 1 ~ 1 s y I y 1 1 1 ~ ~ ~ I 1 ~ 1 y ~ ~ ~ ~ 1 s ~ ~ I I I }i}L I I I I s I Is} } } } I I } } .1 I ~ 1 I ~ ~ 1 ~ ~ ~ I ~ ~ ~ I ~ I ~ sty 1 ~ I I ~ 1 1 ~ ~ 1 ~ ~ 1 ~ I I I I ~ I ~ 1 i I 1 I ~ ~ ~ 1 1 I I ~ 1 s ~ s y 1 I ~ I 1 ~ ~ 1 I s ~ ~ 1 1017 'e Ere!t I~JJ I zieci veyIF" I }'1"i'>c;erl~ee I I I I gll }} I I 1 I I I 1 ~ 5 10 15 20 25 30 Past Neutron Fluence (E > 1'HEV) as a Function of Effective Pull Power Years Figure B3.1-2
'a
SAFETY EVALUATXON Xntroduction Limits for reactor coolant temperature and pressure are given in Turkey Point Technical Specifications 3.1.2. This submittal proposes changes to the specification to bring it into compliance with 10 CFR'0, Appendix G. The changes are based on test results recently obtained from the latest reactor vessel surveillance capsules 'from Unit 3 and 4. P 4 ~ Discussion Since the physical properties of reactor vessel mater'i'als-,can. be affected by neutron irradiation, these materials. are subjec'4".to a surveillance program. Reactor vessels are designed'o accommodate surveillance capsules which contain materials used in ""the rpanuiacture of the vessels. Xrradiated surveillance materials are p'erRodically analyzed and the data used to revise operating pressure-temperature. limits. Turkey Point FSAR Section 4.4 contains detailed infor-mation on the reactor vessel surveillance program. Current Technical Specification 3.1.2 provides a method'or revising heatup and cooldown curves based on the radiation exposure of surveillance specimens. However, the Turkey Point, reactor vessels were manufactured before the publication of 10 CFR 50, Appendix G and ASHE Boiler and Pressure Vessel
Code, Section Xi~', appendix G:
both of which provide information on reactor vessel design and calculation techniques based on surveillance specimen data. Now that Turkey Point surveillance capsule data have been reviewed and analyzed in accordance with Appendix G and the ASb1E Code, we find it necessary to revise the pressure-temperature curves in the Technical Specifications. The new limit curves for normal heatup and cooldown of the primary reactor coolant system were calculated using the methods discussed in the proposed Bases pages. Conclusion Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and does not reduce the margin of safety as defined in the basis for any technical specification, therefor'e, the change does not. involve a significant hazards consideration, (2) there is reasonable assurance that 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.
l I <<gp4 ' 'f t l}}

L-76-199, Request to Amend Appendix a of Facility Operating Licenses DPR-31 & DPR-41. Proposes Changes in Response to 04/14/1975 Letter from George Lear Which Requested That Heatup & Cooldown Rates in Technical Specifications .

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Dear Mr. Stello:

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L-76-199, Request to Amend Appendix a of Facility Operating Licenses DPR-31 & DPR-41. Proposes Changes in Response to 04/14/1975 Letter from George Lear Which Requested That Heatup & Cooldown Rates in Technical Specifications .

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Dear Mr. Stello:

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