ML20217B827
| ML20217B827 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/15/1998 |
| From: | IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | |
| Shared Package | |
| ML20217B803 | List: |
| References | |
| NUDOCS 9804230136 | |
| Download: ML20217B827 (9) | |
Text
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RTS-300 to NG-98-0493 Page1ofI PROPOSED CHANGE RTS-300 TO THE DUANE ARNOLD ENERGY CENTER IMPROVED TECIINICAL SPECIFICATIONS The holders oflicense DPR-49 for the Duane Arnold Energy Center propose to amend the Improved Technical Specifications by deleting the referenced pages and replacing them with the enclosed new pages.
t
SUMMARY
OF CHANGES:
Pace Description of Chances 3.4-21 Revises SR 3.4.9.1.b to state "RCS heatup and cooldown rates are s 20 F in any I hour period during Inservice Leak and Hydrostatic testing (Curve A)."
3.4-21 Adds SR 3.4.9.1.c,"RCS heatup and cooldown rates are s 100 F in any I hour period during Non-Nuclear Heating (Curve B) and Nuclear Heating (Curve C)."
3.4-24 Replaces Figure 3.4.9-1," Pressure Versus Minimum Temperature Valid to Sixteen Full Power Years, per Appendix G of 10CFR50" with updated curves labeled " Pressure Versus Minimum l
Temperature Valid to Thirty-Two Full Power Years, per Appendix G of 10CFR50."
l B 3.4-51 Revises LCO " element a" to reflect the revisions to SRs 3.4.9.1.b and 3.4.9.1.c.
B 3.4-54 and Revises SR 3.4.9.1 to insert " initiated and" and " starting and" in l
B 3.4-55 the sentence, " Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be initiated and discontinued when the criteria given in the relevant plant procedure for starting and ending the activity are satisfied."
B 3.4-58 Revises Reference 1 to state "10 CFR 50, Appendix G, December 1995."
B 3.4-58 Updates Reference 7 to reflect this proposed amendment.
I 9804230136 980415 PDR ADOCK 05000331 l
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1 RCS P/T Limits 3.4,9 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C.
NOTE---------
C1 Initiate action to Immediately Re uired Action C.2 restore parameter (s) sh 11 be completed if to within limits.
this Condition is entered.
AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LC0 not met in other operation.
or 3.
than MODES 1. 2, and 3.
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SP 3.4.9.1
NOTE--------------------
Only required to be performed during RCS i
heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
Verify:
30 minutes
{
a.
.RCS pressure and RCS temperature ara within the applicable limits in Figure l
3.4.9-1.
i b.
RCS heatup and cooldown rates are
@WF in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> periodxdor,3 1as<. eda.o Lask. 2d 4 Jees+s&h. faKq t e.wus A ).
3 C,.
'itC5 ke.slup 2ad coolkwa r2I*-5 sre.
(continued)
.Alco*f o*a 99 1 hear p<.ried durky goa-9ve,lesr Hes4.s (curve.W) w/ yde.sr HesQ (curg., ch.
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DAEC 3.4-21 Amendment 223
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RCS P/T Limits 3.4.9 Curve A(EFPY) 18.20.24.28.32 B
C 1400 A - System Hydrotest Lim:t with Fuel in l
l Vessel (20*F/hr heatup/cooldown rate).
/
B - Non-Nuclear Heating Limit, Valid to 32
/
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EFPY(100'F/hr heatup/cooldown rate)!
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C - Nuclear (Core Critical) Limit, Valid to 32 i
1200 EFPY(100*F/hr heatup/cooldown rate)[
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50 100 150 200 250 300 350 Minimum Reactor Vessel Metal Temperature (F) i Figure 3.4.9-1 (page 1 of 1)
Pressure Versus Minimum Temperature Valid to mieen 71"rb - Lo Full Power Years, per Appendix G of 10CFR50 DAEC 3.4-24 Amendment 223 l
RCS P/T Limits B 3.4.9 BASES APPLICABLE limits related to the P/T limits.
Rather, the P/T limits SAFETY ANALYSES are acceptance limits themselves since they preclude (continued) operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
t LCO The elements of this LCO are:
a.
RCS pressure and temperatures are within the limits Figure 3.4.9-1 and heatup or cooldown rates are s 100 F/hr during RCS heatup,bJ cooldown, and inservice j
leak and hydrostatic testing;
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b.
The tem)erature difference between the reactor vessel bottom lead coolant and the Reactor Pressure Vessel (RPV) coolant is s 145 F during recirculation pump startup:
c.
The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is s 50 F during recirculation pump startup:
d.
RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-1 prior to achieving criticality; and e.
The temperatures at the reactor vessel head flange and the shcl1 adjacent to the head flange are 2 74 F when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as in)uts for calculating the heatup, cooldown, and inservice leacage and hydrostatic testing P/T limit curves.
Thus, the LC0 for the (continued)
DAEC B 3.4-51 Amendment 223
RCS P/T Limits B 3.4.9 BASES ACTIONS C.1 and C.2 (continued)
Operation outside the P/T limits in other than MODES 1. 2.
and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Recuired Action must be initiated without delay and continuec until the limits are restored.
Besides restoring the P/T limit parameters to within limits.
an evaluation is required to determine if RCS caeration is allowed.
This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 212 F.
Several methods may be used. including comparison with pre-analyzed transients, new analyses, or inspection of the components.
ASME Code.Section XI. A)pendix E (Ref. 6). may be used to support the evaluation:
lowever, its use is restricted to evaluation of the beltline.
Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered.
The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within limits is required l
every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments. 30 minutes permits a reasonable time for assessment and correction of minor deviations.
The limits of Figure 3.4.9-1 are met when operation is to the right of the applicable limit curve.
i Surveillance for heatup, cooldown, or inservice leakage and i
hydrostatic testing may be continued when the criteria u5850 (continued)
DAEC B 3.4-54 Amendment 223
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l RCS P/T Limits B 3.4.9 BASES i
SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS M8d'd *j given in the relevant plant procedure for ending the activity are satisfied. During heatu)s and cooldowns, the temperatures at the reactor vessel siell adjacent to the shell flange, the reactor vessel bottom drain, recirculation loops A and B. and the reactor vessel bottom head shall be monitored.
During inservice hydrostatic or leak testing, the reactor vessel metal temperatures at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to the shell flange shall be monitored.
This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
The limits of Figure 3.4.9-1 are met when operation is to the right of the applicable limit curve.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time l
of the control rod withdrawal.
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Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design l
allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation pump (Ref. 8) are satisfied.
(continued)
DAEC B 3.4-55 Amendment 223
i i
RCS P/T Limits B 3.4.9 BASES (continued)
REFERENCES 1.
10 CFR 50. Appendix G. May-4983 bec e.mbex 199K 2.
ASME, Boiler and Pressure Vessel Code.Section III.
Appendix G.
3.
ASTM E 185-82 July 1982.
4.
10 CFR 50. Appendix H.
5.
Regulatory Guide 1.99 Revision 2. May 1988.
6.
ASME, Boiler and Pressure Vessel Code.Section XI.
Appendix E.
7.
C--Shi rak i- ( NRC&to-L-td u-(-I ELP-)--TS-Amendment-No.
If 2-tofa ci-lity-Operati ng-ti cense-NorDPR-49 Mated August 42. 1991. ('A m ~ /, % + re 4 a A 4. r " 4 4 /
L.1 unc.h, 8.
UFSAR, Section 15.4T5.
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l DAEC B 3.4-58 Amendment 223 t
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RTS-300 to NG-98-0493 Page1of1 SAFETY ASSESSMENT By letter dated April 15,1998, IES Utilities Inc. submitted a request for revision of the Improved Technical Specifications (ITS) for the Duane Arnold Energy Center (DAEC).
The proposed amendment would update the reactor pressure vessel pressure-temperature limit curves to include operation up to 32 effective full power years (EFPY), as well as intermediate curves for 18,20,24, and 28 EFPY. The proposed amendment also revises surveillance requirements (SRs) and the Bases to reflect operation with the new curves.
Evaluation:
The amendment will update the pressure-temperature curves (and associated surveillance requirements and Bases) already existing in the pending Improved Technical Specifications to encompass 32 EFPY of operation. The curves were derived from analysis of test specimens removed from the DAEC reactor vessel to the requirements of 10 CFR 50, Appendix G. The results of that analyses are within the limits of Regulatory Guide 1.99 Revision 2. The values of adjusted reference temperature and upper shelf energy determined as a result of the 10 CFR 50, Appendices G and H analysis are expected to remain within the limits of Regulatory Guide 1.99, Revision 2 and Appendix G of 10 CFR 50 (less than 200 F and greater than 50 ft-lbs respectively) for at least 32 EFPY of operation. Therefore, based on the above, we have concluded that the proposed revision to the DAEC ITS is acceptable.
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RTS-300 to NG-98-0493 Page1ofI l
ENVIRONMENTAL CONSIDERATION 10 CFR Section 51.22(c)(9) identifies certain licensing and regulatory actions which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and (3) result in a significant increase in individual or cumulative occupational radiation exposure. IES Utilities Inc. has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical l
exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR Section 51.22(b),
no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:
Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9) for the following reasons:
l 1.
As demonstrated in Attachment I to this letter, the proposed amendment does not l
involve a significant hazards consideration.
j 2.
The proposed change updates pressure-temperature limits that were not derived from Design Basis Accident Analyses. They are prescribed by the ASME B&PV Code and 10 CFR 50 Appendix G and 11 as restrictions on normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile l
failure of the reactor coolant pressure boundary. Therefore, this change will not l
result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
1 3.
The proposed change will not appreciably change the way the plant or its systems are operated. The change will merely update the pressure-temperature curves (and l
associated surveillance requirements and Bases) already existing in the plant l
Technical Specifications to provide limits from 18 to 32 EFPY of operation, which are based upon evaluation and analysis of actual in-vessel material specimens, per 10 CFR 50, Appendices G and II. The pressure-temperature curves are established to the requirements of 10 CFR 50, Appendix G to assure that brittle fracture of the reactor vessel is prevented. There will be no significant increase in eitber individual or cumulative occupational radiation exposure.
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