Text

Proposed Tech Specs,Providing New RCS pressure-temp Limit Curves Applicable Up to 21 EFPYs
	ML20080G275
Person / Time
Site:	Davis Besse
Issue date:	01/30/1995
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20080G269	List: ML20080G266; ML20080G275;
References
	2269, NUDOCS 9502070082
	Download: ML20080G275 (16)
	v • d • e

_ _ _ _ _ _ _ __

Docket Number 50-346 License Number NPF-3

' Serial Number 2269 P

Attachment uH5w u

0RINFORMATION ONLY

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REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOVN LIMITING CONDITICt! FOR OPERATION 3.4.2 Decay Hest Removal System relief val @e DH-4849fshall be OPERABLE vith a lift setting of < 330 psig* and isolation valv.es T4-11 and DH-12 open and

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control power to their valve operators removed.

APPLICABILITY: MODES 4 and 5.

ACTION:

Vith DH-4849 not OPERABLE:

A.

1.

Make the valve OPERABLE vithin eight. hours; or-Vithin next one hour, disable the capability of both high pressure 2.

a..

injection (HPI) pumps to inject water into the reactor coolant system; and b.

Vithin next eight hours 1.

Disable the automatic transfer of makeup pump suction to the borated water storage tank on lov makeup tank level; and l

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2.

Reduce makeup tank level to < 73 inches and reduce reactor coolant system pressure and pressurizer level within the acceptable region on Figures 3.4-2a (in MODE 4) and 3.4-2b (in MODE 5).

B.

Vith DH-11 or DH-12 closed, open DH-21 and DH-23 within one hour.

C.

Vith the control power not removed from DH-11 and DH-12, remove the power to the valve operators at the Motor Control Centers within one hour.

SURVEILLANCE REQUIREMENTS 4.4.2 Decay Heat Removal System relief valve DB-4849 shall be determined OPERABLEt per the surveillance requirements of Specification 4.0.5.

a.

b.

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying either:

1.

isolation valves DB-11 and DH-12 open with control power removed from their valve operators; or 2.

valves DH-21 and DB-23 open.

The lift setting pressure shall correspond to ambient conditions of the 3

valve at nominal operating temperature and pressure.

t /

DAVIS-BESSE, UNIT 1 3/4 4-3 Amendment No. 57,779,135 9502070082 950130 PDR ADOCK 05000346 P

PDR

Docic2t Nunb:r 50-346 Licensa Number NPF-3 S2 rial Number 2269 Attachment a

FORINFORMAIl0N DNLY REACTOR C001).NT SYSTEM SAFETY VALVES AND PILOT OPERATED RELIEF VALVE - OPERATING l

LIMITING CONDITION FOR OPERATION

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3.4.3 All pressurizer code saf,ety valves shall-[be OPERABLE vith a lift setting

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of < 2525 psig.* vhen not isolated, the pressurizer' pilot operated relief valve l shall have a trip setpoint of 1 2435 psig and an allovable value of 1 2435 psig.**

APPLICABILITY: HODES 1, 2'and 3.

ACTION:

Vith'one pressurizer code sa'fety valve inoperable, either restore the inoperable valve to OPERABLE. status within 15 minutes or be in BOT SHUTDOVN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i SURVEILLANCE REOUIREMENTS forthe'pressurizercodesafetyvalves,therearenoadditional 4.4.3 Surveillance Requirements other than those required by Specification 4.0:5.

For the pressurizer pilot operated relief valve a CHANNEL CALIBRATION check shall be l

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performed every 18 months.

The lift setting pressure shall corresoond.to ambient conditions of the f

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valve at nominal operating temperature and pressure.

Allovable value for CHANNEL CALIBRATION check.

l DAVIS-BESSE, UNIT 1 3/4 4 4 Amendment No. 77, /90/77F,135 1

Dock,et Number 50-346

.. Li6inse Number NPP-3 Serial Numbar 2269 Figure 3.4-2a l

i Attachment age 0 Reactor coolant. System Pressure - Pressurizer Level Limits for inoperable Decay He'ac Removal System Relief Valve in MODE 4 400 t

UNACCEPTABLE REGION 350

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. NOTE: NOT CORRECTED FOR INSTRUMENT ERROR l

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40 80 120 160 200 240 Initial Preesurizer Level (Inches)

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FORINFORMEl0N ONLY DAVIS-BESSE, UNIT 1 3/4 4-4a Amendment No. ff, 116

Docket Number 50-346 rieure 3.4-2b l

'

License NL.mber NPF-3 11 umber 2269 Reactor coolant System Pressure - Pressurizer Level p,gg 33 -

Limits for inoperable Decay Heat Removal System Relief Value in MODE 5 400 l

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k NOTE: NOTCORRECTEDFORINSIRUMENTERROR' 350 G

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250 UNACCEPTABLE REGION Y

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ACCEPTABLE RECION' N

0 40 80 120 160 200 240 Initial Pressuricer level (Inches)

IIIIS PAGE PROVIDED FOR NFORE110N ONU DAVIS-BESSE, UNIT 1 3/4 4-4b Anendment No. $/,116

Docket Number 50-346

, Licehse Number NPF-3 Serial Number 2269 Attachment Page 12.

THIS PAGE PROVIDED

""" ~"""

3/4.4.9 PRESSURE / TEMPERATURE LIMITS FORINFORE00N ONN LIMITING CONDITION FOR OPERATION ~

3.4.9.1 TheReactorCoolantsystem(exceptthhpressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 during heatup, cooldown, crit.icality, and-inservice leak and hydrostatic' testing with:

a.

A maximum heatup of 50"F in any one hour period, and l

b.

A maximum cooldown of 1,00'T in any one hour period -with cold leg temperature > 270*F and a maximum cooldown of 50*F in any one hour period with cold leg temperature <270*F.

APPLICABILITY: At all times.

ACTION:

With"any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering

~

evaluation to determine the effects of the out-of-limit condition on the integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUIDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system bestup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens representative of the vessel materials shall be removed and examined, to determine changes in material properties, at the intervals defined in BAV.

l 1543A. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.

l DAVIS-BESSE, UNIT 1 3/4 4-24 Amendment No. 82,116

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m : m e-n n n n.F nno-

.n Floure 3.4-3

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,n Reactor Coolant System Pressure-Temperature Limits -

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for Cooldown for the First 21 EFPY

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1. Allowable cooldown rate above 270 F as 100 F/hr Ramp). Emited

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emu a 15 F step change followed by an 9 minute hold.

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r. ----- r!---- r- ----i ------1

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3. A maximum step temperature change of 15 F is anowable when removing!

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i c-1800 all RC pumps from operation with the DHR system operating. The step l----- --j-----j-.----!...

Point

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g temperature change is defined as RC temp minus the DHR retum temp j j

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A 70 156 i

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.o to the reactor coolant sistem prior to stopping a!! RC pumps.

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4. When the decay host removal system (DH) is operating without any i

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170 477 i

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reactor vessel shaR be used.

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6. Instrument error is not accounted for in these limits.

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nn am-o x-Reactor Coolant System Pressure-Temperature Heatup and J E E *+

Cooldown Limits for Inservice Leak and Hydrostatic Tests "5m*z 2600 -

for the First 21 EFPY E E if E i

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S 2000 a 15 F step change fortowed by an a minute hoid.

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3. Aliowable cooldown rate below 270 F is 50 F/hr (Ramp), limited by e

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1600

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oE all RC pumps from operation with the DHR system opera 6ng. The step i

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temperatum change is defined as RC temp minus the DHR retum temp i

i Point Temc Press ':

1400 tg the reactor emiant system prior 2 stopping an RC pps.

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Q reactor vessel shall be used.

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250 1242

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Indicated Reactor Coolant Inlet Temperature, F J

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vocKet numoer au-acu ulcense Number NPF-3 3

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' ' Serial Number 2269 f

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g Attachment I Es sw Page 19 REA'CTOR COOLANT SYSTEM BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVAIINT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation during one continuous time interval with specific activity levels exceeding 1.0 pci/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T to < 530*F prevents the release of activity should a steam generator fu6e rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessiye specific activity levels in the primary coolant will be detected in suf-ficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary are established in accordance with t,he requirements of Appendix G to 10 CFR 50 and with the thermal and loading cycles used for design purposes.

The limitations prevent non-ductile failure during normal operation, including anticipated operational occurrences and system hydrostatic The limits also prevent exceeding stress limits during cyclic tests.

operation. The loading conditions of interest include:

1.

Normal operations, including heatup and cooldown, 2.

Inservice leak and hydrostatic tests, and l

3.

Reactor core operation.

l, The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50.

The closure head reactor vessel outlet nozzles and the beltline region have been

region, identified to be the only regions of the reactor vessel, and consequently

~

i of the reactor coo: wt pressure boundary, that determine the pressure-temperature limitatiou. concerning non-ductile failure.

l i

I DAVIS-BESSE, UNIT 1 B 3/4 4-o Amendment No. 104 1

(

Docket Number 50-34'.

i

., License Wumber NPF-3 Serial Number 2269 Attachment Page 20 REACTOR COOLANT SYSTEM BASES The closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt pre-load).

This region largely controls the pressure-temperature limitations of the first several service periods. The outlet nozzles of the reactor vessel also affect the pressure-tem)erature limit curves of the first several service periods.

This is due to t1e high local stresses at thu side corner of the nozzle which can be two to three times the membrane stresses of the shell.

After the first several years of neutron radiation exposure, the RTum temperature of the beltline region materials will be high enough so that the beltline region of the reactor vessel will start to control the pressure-temperature limitations l

of the reactor coolant pressure boundary.

For the service period for which I

the limit curves, are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point com]arison of the limits imposed by the closure head region, outlet nozzles, and )eltline region. The maximum allowable the three calculated pressures. pressure is taken to be the lower pressure of The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The limit curves were prepared based upon the most limiting adjusted reference temperature of all the beltline region materials at the end of the tenth twentysone ef fective full power years.

The actual shift in RT of the beltline region material will be established um periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, re6ctor 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 and vessel inside the radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The limit curves must be recalculated when the ART determined from the surveillance capsule is different from the calculated $km for the equivalent capsule radiation exposure.

l l

4 DAVIS-BESSE. UNIT 1 B 3/4 4-10 Amendment II6 c

L

- - - ' - ^ - - - ^

'~

i Docke,t. Number 50-346

. hicchse Number NPF-3 Serial Number 2269

-Attachment Page 21 REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline region materials, required by A)pendices G and H to 10 CFR 50 were determined for those materials for whic1 sufficient amounts of material were available.

The adjusted reference temperatures are calculated by adding the predicted radiation-induced ART and the unirradiated RT The procedures described c

e1 in Regulatory Guide 1.69. Rev. 2 were used for predicting the radiation induced ART r as a function of the material's copper and nickel content and e

neutron fluence.

I Figure 3.4-2 presents the pressure-temperature limit curve for normal heatup.

i This figure also 3 resents the core criticality limits as required by Appendix G to 10 CFR 50.

rigure 3.4-3 presents the pressure-temperature limit curve for normal cooldown.

Figure 3.4-4 presents the 3ressure-temperature limit curves for heatup and cooldown for inservice leac and hydrostatic testing.

i All pressure-temperature limit curve are applicable up to the tenth tWentyj, one effective full power years.

The protection against non-ductile failure is assured by maintaining the codlant pressure below the upper limits of Figures 3.4-2. 3.4-3 and 3.4-4.

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DAVIS-BESSE. UNIT 1 B 3/4 4-11 Amendment No. II6 l

_ pocket Nurber 50-346

'Licen::2 Nunbcr NPF-3 Serial Number 2269

="r IfilS PAGE PROVIDED

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FORINFORMATION DN REACTOR COOLANT SYSTEM

_S 1

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in BAW 1543A. The withdrawal schedule is based on four considerations:

l (a) uncover possible technical anomalies as early in life as they can be detected (end of first fuel cycle), (b) define the material properties needed to perform the analysis required by Appendix G to 10 CFR 50, (c) reserve two capsules for evaluation of the effectiveness of themal annealing in the event inplace annealing becomes necessary, (d) provide material property data corresponding to the reactor vessel beltline conditions at the end of service. This withdrawal schedule is specified to assure compliance with the requirements of Appendix H to 10 CFR 50.

Appendix H references the requirements of AST!i E185 for surveillance Program criteria.

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DAVIS-EESSE, UNIT 1 B 3/4 4-12 Amendment No. EJ 116

/I Dockat Nurber 50-346

.,. lLicenso Nuxbar NPF-3

.S2 rial'Nunber 2269 Attachment,

?

Page 23 License Condition

.2.C(3)(d) Prior to operation beyond 21 Effective Full Power Years, the Toledo Edison Company shall provide to the NRC a reanalysis and proposed modifications, as'necessary, to ensure continued means of-protection against lov temperature reactor coolant system overpressure events.

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ML20080G275

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