ML20058N288

Proposed Tech Specs for Pressure Vs Temp Operating Limit Curves

ML20058N288
Person / Time
Site:	Cooper
Issue date:	12/10/1993
From:	NEBRASKA PUBLIC POWER DISTRICT
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ML20058N283	List: ML20058N279 ML20058N288
References
NUDOCS 9312210316
Download: ML20058N288 (9)

Text

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LIMITINd CONDITTONS FOR OPEPATION 91'RVETT T ANCE PEOUTPEMENTS

-3.6 Primary System Boundarv A.6 Primarv System Boundarv t'

l Apolicability ADolicabilitv-l l

Applies to the operating status of Applies to the periodic examination the reactor coolant system.

and testing ' requirements for the l

reactor cooling system.

Obiective:

Obiective:

To assure the integrity and safe To determine the condition of the operation of the reactor coolant reactor coola'nt system and the system.

operation of the safety '. devices related to it.

Soecification:

Specification:

A.

Thernal and Pressurization A.

Thermal and Pressuritation.

Limitations Limitations 1.

The average rate of reactor coolant 1.

During heatups and ; cooldowns, the -

temperature change during normal following. temperatures.shall. be heatup or cooldown shall not exceed permanently logged. at. least every 100'F/hr when averaged over

.a

15. minutes until.the difference one-hour period.

between any two readings taken over

.a 45. minute' period is less than 50*F.

a.

Bottom head drain.

b.

Recirculation loops A.and B, 2.huringoperationwhere the core isl 2.

Reactor

. vessel, temperature ' and critical. <2 l during heatup by reactor coolant - pressure shall be I

nonnuclear means or-cooldown,

permanently logged at least every following shutdown. f the reactor 15 minutes whenever the shell j.g vessel metal and fluid temperatures temperature is below 220*F and the "5 "U shall be at or above the reactor vessel is not vented-.

temperatures shown on the limiting 7his SMCa4 ton cdces not off$ h ]

curves of Figures 3.6.1.a or 3.6.1.b.* This specification applies he.ofcp or coolcleun for +4porpoSE i

when the reactor vessel head is e 4 perferrnW_ iniern.e., hydrosMC tensioned.f

- ( or C%V. ie69 a

3.

The reactor vessel metal 3.

Test specimens of the reactor vessel temperatures for the botto[r head base, veld and heat affected tone region and beltline region shall be metal subj ected to the' highest at'or above the temperatures shown fluence of; greater than 1 Mev on the limiting curves of neutrons shall:be installed in the Figure 3.6.2 during inservice reactor vessel adj acent to-the i

hydrostatic or leak testing.

The vessel' wall at the core : midplane Adjusted Reference Temperature (ART) level.

The specimens and sample for the beltline region must be program shall-conform to ASTP.

approgriate E 185-73 to the degree possible.

determined from the beltline curve (H,.13. = 2%EFPY) depending on the current accumulated L 15 % El or 3 3

number of effective full power years (EFPY).

m l

l 9312210316 931210 W;m pR ADOCK0500g8 j.

132.

%1RMiV

3.6..A & 4.6.A BASES (cont'd) usAQ As de' scribed in the saf:t :n:lv:i r gert, detaile'd stress analyses have

^

been made on the reacto'r ves'sel for both steady-state and transient conditions with respect to material fatigue.

Thec results of these analyses are compared to allowable stress limits. -Requiring the coolant-temperature in an idle recirculation loop to be within 50'F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.

The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the' vessel when there is nc-recirculation flow.

This colder water. is forced up when : recirculation pumps, are '

started.

This will not result in stresses which exceed ASME Boiler and Pressure Vessel Code,Section III limits when the temperature differential

_ is not greater than 145'F.

e first surveillance capsule was removed at 6.8 EFPY of operation' b

metal, weld metal and HAZ specimens were tested.

In additio ux wire were tested to experimentally determine the integrated n ran flux (fluen at the surveillance capsule location.

The tes results are presente n General Electric Report MDE-103-0986.

M ured shifts 11n 4

RTm of the ase metal and weld metal were compared predicted values' per Regulato Guide 1.99, Revision 1 which was effect at that time.

The measured va s were higher than predicted o the 1.99 methods were modified to reflec

• he surveillance data, e test results for.the flux-wires were used with a 1 tically determi lead factors ro determine the ence at the T Vessel wall dep6th.-pevalue peakend-of-life {EOL) is 1.5 x 10 n/cm.

corresponding to 0 years eration 2 EFPY) j Subsequent to this evaluatio e NRC issued Regulatory Guide 1.99, Revision 2.

This revision-r es that two surveillance capsules be.

tested before.the test res s are crored -into the adjusted reference temperature (ART) shift edictions..

e adjusted ~ reference temperature -

nitial RT y from the RTthe.-i$ltial-lus due of a beltline materia s defined as th velope7 to irradiation.

erefore, the. curves surveillance caps e testing were re-evaluat in - accordance - with-the guidance provid in Regulatory Guide 1.99 Revisi 2.

Based strictly: on and Revision 2,iven the chemist actors provided in Regulatory Guide consideri each beltline material chemistry and peak f at a g

EFPY, pressure-temperature curves in Fipres 3.6.1.a an

.l.b.

whic eflect a beltline ART of 110*F, were determined to be vali

.r 21 PY.

Figure 3.6.2, the pressure test curve, was re-evaluated inflik nner and includes curves for 13. 18 and 21 EFPY ' to provide more.

flexibility in pressure testing.JAFigure A b.Z also has a separace' curve for the bottom neaa region, inelbottom head, curve does notl shift with p [e increased operation; therefore,- the bottom head temperature can be l

monitored against lower temperature requirements than the beltline during pressure testing.

The surveillance capsule withdrawal schedule for the remaining specimens is located in Section IV.2.7 of the CNS USAR.

l B.

Coolant ChemistII Materials in the primary system are primarily Type-304 stainless steel and Ziracioy cladding. The reactor water chemistry limits are established to provide an environment favorable to these materials. Limits are placed on 2

conductivity and chloride concentrations. Conductivity is limited because it can be continuously and reliably measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant; Chloride limits are specified to prevent stress corrosion cracking of stainless steel.

Several' investigations have shown that in neutral solutions some oxygen is required to cause stress corrosion cracking of stainless steel, while in the absence of oxygen no cracking occurs. One of these is the chloride-oxygen relationship of Williams, where it is shown that at high chloride l

concentration little oxygen is required to cause stress corrosion cracking of stainless steel, and at high oxygen concentration little' chloride is required to cause cracking. These measurements were determined in a wetting and drying siruation using alkaline-phosphate-treated boiler water and therefore,.are of limited significance to BWR conditions.- They are.,-

however, a qualitative indication of trends.

2W. L. Williams, Corrosion 13, 1957, p. 539t.

-147-10/13/92

M&W

~

The second surveillance capsule was removed from the vessel at the end of Fuel Cycle 14, following 11.2 EFPY of operation. Based on the analysis of flux wires l

contained in the second capsule combined with analytically determined lead factors, the peak end-of-life (40 years of operation, or 32 EFPY) fluence at the 1/4 T vessel wall depth is predicted to be 1.1 X 10 n/cm Based on the 28 2

testing and analysis of the base metal, weld metal, and' HAZ charpy specimens I

removed from both the first and second surveillance capsules, new pressure-temperature curves were generated. This testing and analysis is documented in General Electric Report No. GE-NE-523-159-1292, dated February, 1993.

The i

testing and analysis performed following the removal of the first surveillance capsule was reported in General Electric Report No.

MDE-103 0986, dated i

May, 1987.

l Figures 3.6.1.a.

3.6.1.b, and 3.6.2 reflect the results of this testing and analysis, which includes consideration of the test results obtained from the specimens removed with the first surveillance capsule.

Figures 3.6.1.a and a calculated

3. 6.1.b have been calculated through end-of-life, and ' reflect Adjusted Reference Temperature (ART) of 128'F at 32 EFPY. Ff.gure 3.6.2 includes curves corresponding to 15, 18, 21, 24, and 32 EFPY which correspond with ARTS of 89, 96, 101, 108, and 128'F respectively, in order to provide' greater flexibility during pressure testing. e gg q g I

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l APPENDIX B 1

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l LIMITING' CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS 3.6 Primary System Boundarv 4.6 Primary System Boundary Applicability:

Applicability:

Applies to the operating status of Applies to the periodic examination the reactor coolant system.

and testing requirements for the reactor cooling system.

Obiective:

Obiective:

To assure the integrity and safe To determine the condition of the l

operation of the reactor coolant reactor coolant system and the system.

operation of the safety devices j

related to it.

Specification:

Specification:

A.

Thermal and Pressurization A.

Thermal and Pressurization-Limitations Limitations 1.

The average rate of reactor coolant 1.

During heatups and cooldowns, the j

temperature change during normal following temperatures shall be heatup or cooldown shall not exceed permanently logged at least every 100*F/hr when averaged over a

15 minutes until the difference i

one-hour period, between any two readings taken over a 45 minute period is less ' than

]

2.

During heatup by nonnuclear means or 50*F.

cooldown following shutdown, or a.

Bottom head drain.

during operation where the core is I

critical, the reactor vessel metal b.

Recirculation loops A and B.

and fluid temperatures shall be at or above the temperatures shown on 2.

Reactor vessel. temperature and the limiting curves of reactor coolant pressure shall be l

Figures 3.6.1.a or 3.6.1.b as permanently logged at least every applicable.

This specification 15 minutes whenever the shell applies when the reactor vessel head temperature is below 220*F and the j

is tensioned.

This specification reactor vessel is not vented.

does not apply to heatup or cooldown for the purpose of performing inservice hydrostatic or leak l

testing.

3.

The reactor vessel metal l

temperatures for the bottom head 3.

Test specimens of the reactor vessel region and beltline region shall be base, veld and heat affected zone at or above the temperatures shown metal subj ected to the highest on the limiting curves of fluence of greater than 1 Mev Figure 3.6.2 during inservice neutrons shall be installed in the hydrostatic or leak testing.

The reactor vessel adj acent. to the Adjusted Reference Temperature (ART) vessel wall at the core midplane l

for the beltline region must be level.

The specimens and sample determined from the appropriate program shall conform to ASTM beltline curve (15, 18, 21, 24, or E 185-73 to the degree possible.

32 EFPY) depending on the current accumulated number of effective full I

power years (EFPY).

i

-132-j

l 3.6.A & 4.6.A BASES (cont'd)

As dcscribed in the USAR, detailed stress analyses have been made on the-l rer.ctor vessel for both steady-state and transient conditions with respect to ' material fatigue.

The results of these analyses are compared to t

allowable stress limits.

Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature-l befcre a recirculation pump is started assures that the changes in coolant t

temperature at the reactor vessel nozzles and bottom head region are acceptable.

[

The coolant in the bottom of the vessel is at a lower temperature than l

that in the upper regions of the vessel when there is no recirculation i

flow.

This colder water ~is forced up when recirculation pumps are started.

This will not result in stresses which' exceed ASME Boiler and Pressure Vessel Code,Section III limits when the temperature differential is not greater than 145'F.

The second surveillance capsule was removed from the vessel at the end of Fuel Cycle 14, following 11.2 EFPY of operation. Based on the analysis of flux wires contained in the second capsule combined with analytically determined lead factors, the peak end-of-life (40 years of operation, or i

32 EFPY)18 fluence at the 1/4 T vessel wall depth is predicted - to be 2

1.1 X 10 n/cm.

Based on the testing and analysis of the base metal, weld metal, and HAZ charpy specimens removed from both the first ' and second surveillance capsules, new pressure-temperature curves were j

generated.

This testing and analysis is documented in General Electric Report No. GE-NE-523-159-1292, dated February, 1993.

The testing and l

analysis performed following removal of the first surveillance capsule was i

reported in General Electric Report No. MDE-103-0986, dated May, 1987.

i Figures 3.6.1.a.

3.6.1.b, and 3.6.2 reflect the results of this testing and analysis, which includes consideration of the test results obtained t

from the specimens removed with the first. surveillance capsule.

l Figures 3.6.1.a and 3.6.1.b have been calculated, through end-of-life, and reflect a calculated Adjusted Reference - Temperature (ART) of 128'F at t

32 EFPY.

Figure 3.6.2 includes curves correspondin 18, 21, 24, and 32 EFPY which correspond with ARTS of 89, 96, g to 15, 101, 108, and 128'F respectively, in order to provide greater flexibility during prussure testing.

Figure 3.6.2 also has a separate curve for the bottom head region.

The bottom head curve does not shift with increased operation; therefore, the bottom head temperature can be monitored against lower temperature requirements than the beltline during pressure testing. The surveillance capsule withdrawal schedule for the remaining specimens is located in Section IV.2.7 of the CNS USAR.

l B.

Coolant Chemistrv i

t Materials in the primary system are primarily Type-304 stainless steel and l

Ziracloy cladding. The reactor water chemistry limits are established to i

provide an environment favorable to these materials. Limits are placed on f

conductivity and chloride concentrations. Conductivity is limited because it can be continuously and reliably measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant.

I Chloride limits are specified to prevent stress corrosion cracking of stainless steel, i

Several investigations have shown that in neutral solutions some oxygen is required to cause stress corrosion cracking of stainless-steel, while in the absence of oxygen no cracking occurs. One of these is the chloride-1 oxygen relationship of Williams, where it is shown that at high chloride-concentration little oxygen is required to cause stress corrosion cracking of stainless steel, and at high oxygen concentration little chloride is required to cause cracking. These measurements were determined in a wetting and drying situation using alkaline-phosphate-treated boiler water and therefore, are of limited significance to BWR conditions. They are, i

however, a qualitative indication of trends, 1

IV. L. Williams, Corrosion 13, 1957, p. 539t.

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