ML20116N833
| ML20116N833 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/30/1985 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20116N821 | List: |
| References | |
| NUDOCS 8505070369 | |
| Download: ML20116N833 (14) | |
Text
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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION
' 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure; shall be limited in accordance with the limit lines .shown on Figures 3.4-2 and 3.4-3 during heatup, . cooldown, criticality. and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 60 F in any one hour period.
- b. A maximum cooldown of 100 F in-any one hour period.
' APPLICABILITY: At all. times.
ACTION: 1 With any of the - above limits exceeded, restore the temperature and/or pres-sure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out--of-limit condition on the structural integrity of. the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for ; continued operations or be .in at least HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to'less than 200 F and 500 psig, respectively, within the folfoE ng 30 hours. 1 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--
heatup, cooldown,-and inservice leak and hydrostatic testing operations. l 4.4.9.1. 2 The reactor vessel material irradiation surveillance specimens l shall'be removed and examined, to determine changes in material-properties,'at i the intervals shown in Table 4.4-5.. The results of-these' examinations shall i be used^to update Figures 3.4-2'and 3.4-3. l i l l NORTH ANNA . UNIT 1 '3/4.4 l law agry.o
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Figure 3.4.3 North Anna Power Station I Reactor Coolant System Cooldown Limitations Valid up to 10 EFPY
- NORTH ANN /. - UNIT 1 3/4 4-28 i
L:
REACTOR ~ COOLANT SYSTEM
' OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least = one of the following overpressure protection systems shall i be OPERABLE: '
- a. Two power operated relief valves (PORVs) with a lift setting of:
- 1) .less than or equal to 420 psig whenever any RCS cold- leg temperature is less than or equal to-375 F, and 2) less than or equal to 350 psig whenever any.RCS~ cold leg temperature is less than 185 F, or
- b. A reactor - coolant system vent . of greater than or . equal to 2.07 square inches, or
- c. A maximum pressurizer water volume of 457 cubic feet with all RCS cold leg temperatures greater than or equal to 320 F.
APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 375 F, except when the reactor vessel head is removed. ACTION:
- a. With one PORV inoperable,- either restore the inoperable PORV to OPERABLE status within . 7 days . or depressurize' and vent the RCS through 2.07 square inch vent (s) within the next 8 hours; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status,
- b. With,both PORVs inoperable, depressurize and vent the RCS through a 2.07 square . inch vent (s) within 8 hours; maintain ' the RCS in a vented condition until both PORVs have been restored ' to OPERABLE status. ,
- c. In the event either J the PORVs or the ?RCS: vent (s) - are used to mitigate a- RCS ' pressure transient, a Special Report-~shall be prepared and ' submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe ' the circumstances initiating the transient, the effect of ' the PORVs or vent (s) ~ on the transient and~ any corrective action necessary: -to- prevent recurrence. -
'd . The provisions of Specification 3.0.4 are not; applicable.
i NORTH ANNA'- UNITjl. 3/4 4-31 g , E
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l l l REACTOR COOLANT SYSTEM l E BASES
- The hea' up analysis also covers the determination of - pressure-temperature
- limitations' for
- the case in'which the outer wall of the vessel becomes the 1
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 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. Consequently, for the' cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on-an individual basis. f The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining'the most conservative case, with either'the inside or outside wa11' controlling, for any huatup' rate up to 60*F per hour. The cooldown'11mit
. curves of Figure 3.4-3 are composite curves which were prepared based upon the
'same type analysis with the exception that the controlling location is always j the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing corpressive 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-10 EFPY. The adjusted.
? reference temperature was- calculated using ' results from a capsule removed after the first cycle. The' results are documented in' Babcock and .Wilcox j Reports BAW-1638, May 1981 and BAW-1872, April 1985. The reactor vessel materials have been tested to determine their - initial RT . The results of these tests are 'shown in Table B 3/4 4-1.' Reactor opbItion and resultant fast neutron (E>1 Mev) irradiation will cause . an increase in the RT Therefore, an adjusted reference temperature ' based upon the fluence' EN.. copper content of the material 'in question, can . be predicted using Figures B 3/4.4-1 and B. 3/4.4-2. The heatup. and cooldown limit curves (Figure 3.4-2 and 3.4-3) include predicted adjustments for this -. I shif t :in RT at the end of 10 EFPY, as well as ' adjustments for possible ! ' errors in thhress;ure .:and temperature sensing' instruments. l i The actual shift in ' RT - of the vessel material will be established periodically during operabn by. removing and evaluating, in'accordance with ASTM E185-70,o reactor' vessel material ' irradiation surveillance specimens installed near.the inside wall of the reactor' vessel in the= core area. Since p the neutron spectra at the. irradiation samples and NORTH ANNA'- UNIT 1 B 3/4 4-8 ' P _ . . - ,. .a e J__ __ a.i 4 - - . . . , _
;,-. .. . . - . - - _ , ~ . . -. - . . .. . -.. - . . - -
e u p REACTOR COOLANT SYSTEM , , BASES vessel.inside 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 heatup and cooldown curves must be recalculated when the ART determined from the surveillance - capsule is different from the calb$atedARTg for the equivalent capsule radiation exposure.
.The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided l to assure compliance with the minimum temperature requirements of. Appendix G f '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 4.4-3.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 fatique analysis performed in accordance with the ASME Code requirements. 3 l The - OPERABILITY of two PORVs or an RCS vent opening of . greater than 2.07 square inches. ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the 'RCS - cold ' legs are less than or _ equal to' 375*F. Either PORV has l 4 adequate relieving capability _to. protect the RCS from overpressurization when the transient 'is . limited to either (1) the start of an idle- RCP .with the secondary water temperature of the steam generator'less than or equal to 50*F above the RCS cold leg temperatures. or' (2) the start of ~a charging pump and its injection'into a water. solid RCS. When the temperature of the RCS cold legs is - between 320*F and 375'F, . i ~~ overpressure protection can also be provided by a bubble in the pressurizer. In such a case, a maximum pressurizer water ; volume Lof -~457 cu. ft. has been selected to provide at least 10 minutes for' operator response in the event of a malfunction resulting in maximum flow from one charging pump. i: i 4 i .. f-I < 4 ! . NORTH ANNA - UNIT 1 .B 3/4 4-11: I' .
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8 .. ATTACHMENT'2 PROPOSED TECHNICAL SPECIFICATION CHANGE UNIT 2
-I
REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 60 F in any one hour period.
- b. A maximum cooldown of 100 F in any one hour period.
- c. A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. 1 ACTION: With any of the above limits exceeded, restore the temperature and/or pres-sure to within the limit within 30 minutes; perform an~ engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at ' least HOT STANDBY vithin the next 6 hours and reduce the RCS T and pressure to less than 200 F and 500 psig, respectively, within the foll'o%ng 30 hours. 4 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 heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor . vessel = material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these-examinations shall-be used to update Figures 3.4-2 and 3.4-3. NORTH ANFA - UNIT 2 3/4 4-26
4 . 1
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m , .i
' REACTOR COOLANT SYSTEM-OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION
, 3.4.9.3 At:least one of the following overpressure protection systems shall
- be OPERABLE:
- a. Two power operated relief valves -(PORVs)~ with a lift setting of:
- 1) less than -or equal to 520 psig whenever any RCS cold leg temperature is less than or equal to 340 F, and 2) less than or equal to 375 psig whenever any RCS cold leg temperature is less than 190 F, or
-b. A reactor - coolant : system vent of greater than or equal to 2'.07 square inches, or
- c. A maximum preesurizer water volume of 457 cubic fact with all RCS cold leg temperatures greater than or equal t'o 320 F.
APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 340*F, 'except when the reactor vessel head is removed. ACTION:
- a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent . the RCS through 2.07 square inch vent (s) within the next 8 hours; maintain the RCS in a vented' condition until both PORVs have.been restored to OPERABLE status,
- b. With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) ~ ' within 8 - hours; maintain' the RCS in a vented condition until' both PORVs have been. restored' to OPERABLE-status.-
- c. In the event either the PORVs or the RCS vent (s) are u' sed to mitigate' a RCS pressure . transient, 'a. Special Report-~shall be prepared and submitted ' to the Commission pursuant to Specification 6.9.2 within 30 days. The ' report shall describe the circumstances initiating'the transient, the~effect of the PORVs or vent (s) on the transient .and any . corrective- action necessary .to prevent recurrence.
- d. .The provisions of Specification 3.0.4 are not applicable..
s NORTH ANNA - UNIT 2 3/4 4-30 m.
c .
-REACTOR COOLANT SYSTEM BASES Heatup and cooldown.-limit curves are calculated using the most limiting value
'of the nil-ductility reference temperature, RTNDT, at the end of 10 effective
-full power years of service life. The 10 EFPY service life period is chosen such. that the limiting RT a e a n n e c re regi n is NDT greater than the RT NDT f the limiting unirradiated material. The selection.
of such a limiting RT assures that all components in the Reactor Coolant NDT System will be operated conservatively in accordance with applicable Code ! requirements... The reactor vessel materials have been tested to determine their initial RT NDT; the results of these tests are shown in Table B 3/4.4-1. Reactor [ operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause'an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2. The adjusted reference temperature was calculated using results from a capsule removed af ter the first core cycle. The results are documented in Babcock -and Wilcox Reports BAW-1794, October 1983 and BAW-1872, April, 1985. The heatup and cooldown
~1imit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT at the end of 10 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.
Values of ARTNDT determined in this manner may be used until the results from -l the material surveillance - program, evaluated - according to~ ASTM E185, are available. . The first capsule was removed at the end of.the first core cycle. l Successive . capsules will - be removed in accordance with the requirements of ASTM E185-73 and the latest revision of 10 CFR'50, Appendix H. The heatup and l > cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the' calculated ART f r the equivalent capsule NDT radiation exposure. f r NORTH ANNA - UNIT 2 B 3/4 4-11 o _ _ -_-_-__ _ _ _ -_ _ ___}}