ML17326A698
| ML17326A698 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/09/1980 |
| From: | Dolan J INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17318A747 | List: |
| References | |
| AEP:NRC:00121, AEP:NRC:121, NUDOCS 8005130234 | |
| Download: ML17326A698 (14) | |
Text
P 1
REGULATORY I RMATION DISTRIBUTION SYST (RIDS)
'UBJECT:
Submits Tech Specs Change 1 re extension to surveillance test interval re ice condenser lower inlet door testing 8
Change 2 re reactor surveillance capsule withdrawal schedule. Forwards proposed changes to Tech Specs 8 fee.
DISTRIBUTION CODE:
AOOIS COPIES RECEIVED:LTR Q ENCL Q SIZE: Q2 i TITLE: General Distribution fo'r after Issuance of Operating Lic NOTES: Q~JCPIE~~4+gdd ~ ~V.'
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RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LlTTR ENCL ID CODE/NAME LTTR ENCL 7
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'ACTION:
05 BC 48 0 l
'NTERNAL 0
G F ILE 12 II 17 ENGR BR 19 PI ANT SYS BR 21 EFLT TRT SYS OELD 02 NRC PDR 15 CORE PERF BR 18 REAC SFTY BR 20 EEB EPB DOR STS GROUP LEADR 1
1 2
2 1
1 1
1 1
1 1
0 ACCESSION NBR08005130234 DOCSDATE! 80/05/09 NOTARIZED+
NO DOCKET FACIL:50-315 Donald C,
Cook Nuclear Power Planti Unit ii Indiana 8
05000315 50-316 Donald C,
Cook Nuclear Power Pl anti Unit 2i Indian'a 8
05000316 AUTH,NAME AUTHOR AFFILIATION DOLANgJED Indiana 8 Michigan Electric Co, REC IP, NAME RECIPIENT AFFILIATION DENTONEHRR ~
Office of Nuclear Reactor Regulation EXTERNAL: 03 LPDR 23 ACRS 1
1 0<i NSIC 16 16 TOTAL NUMBER OF COPIES REQUIRED:
LTTR 38.
ENCI 37
r E(
I II
8005180 <9+
P INDIANA L MICHIGAN ELECTRIC COMPANY P. 0.
BOX 18 BOWI ING GREEN STATION NEW YORK, N. Y. 10004 May 9, 1980 AEP: NRC:00121:
Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 License Nos..
DPR-58 and DPR-74 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
This letter requests changes to the Appendix 'A'echnical Specifications for the Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2.
Attachment 'A'o this letter contains the description and review of each change.
Attachment 'B'ontains the appropriate revised pages.
The proposed Technical Specification changes contained herein have been reviewed and approved by the Plant Nuclear Safety Review Committee and will be reviewed shortly by the AEPSC Nuclear Safety and Design-Review Committee.
The PNSRC review indicates that in no instance will the proposed Technical Specification change adversely affect the health and safety of the public.
Change No.
1 is considered to be a Class II Amendment as per the provisions of 10 CFR 170.22.
Change No.
2 is considered to be a Class II Amendment and a Class I Amendment.
Accordingly, attached is a check in the amount of $2,800.00.
Your prompt attention to this matter is requested.
JED/emc cc:,
(Attached)
Very truly yours, John E.
Dolan Vice President pooi i'/c A@<~p:
g y85O.oo
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Mr. Harold R. Denton, Director AEP: NRC: 00121 cc:
R.
C. Callen G. Charnoff R. S. Hunter R.
W. Jurgensen D. V.'haller - Bridgman
ATTACHMENT 'A'O AEP:NRC:00121
.CHANGE NO.
1 - SPECIFICATION 4.6.5.3.1 This change ser'ves to request a 'one-time'xtension to the surveillance test interval of Unit No.2 Technical Specification No.
4.6.5.3.1 regarding
.Ice Condenser Lower Inlet Door Testing.
The present surveillance interval, including the
'Grace Period'llowed by Specification No. 4.0.2, will end on June 8, 1980.
As previously discussed with the NRC Staff, IN1ECo. intends to bring Unit No.
2 off line during the forthcoming Unit No.
1 refueling outage to complete modifications to the Auxiliary Feedwater System.
This Unit No.
2 outage is tentatively scheduled to begin in late June or ear ly July, 1980.
It is requested that an extension to the surveillance test interval of Specification 4.6.5.3.1 be granted until July 20, 1980.
This extension will avoid an unnecessary unit shutdown with the corresponding challenge to safety systems and fuel thermal stresses.
The required lower inlet door surveillance would be completed prior to startup from the aforementioned Unit No.
2 outage.
Extension of the lower inlet door surveillance test interval will have no adverse effect on the ability of the Cook Plant to safely mitigate the consequences of a hypothetical accident.
No significant icing has been experienced with the Unit No.
2 lower inlet doors and there is no reason to expect any -icing during the extended surveillance interval.
Previous survei'llance testing of the lower inlet doors has clearly demonstrated that the doors would indeed open as designed during a
hypotheti'cal accident.
A summary of the previous tests is shown in Table 1 below.
TABLE 1
UNIT NO.
2 LOWER INLET DOOR SURVEILLANCE HISTORY TEST DATE COMMENTS March 1978 May 1978 July 1978 November 1978 May 1979 October 1979 All-doors found acceptable All doors found acceptable All doors found accepts'ble One door out of forty eight found unacceptable
(*)
All doors found acceptable All doors found acceptable The right door in Bay 9 was found to require a greater opening force than allowed by the Technical Specifications.
All succeeding sur-veillance tests on this door have been successful.
CHANGE NO.
2 - TABLE 4.4-5 UNIT NOS.
1 AND 2 Technical Specification Tables4.4-5 of Units 1 and 2 have been revised to reflect recent changes in-the reactor vessel surveillance capsule withdrawal schedule which will allow for the collection of data from high lead factor capsules in a more meaningful time frame of reactor service life.
The proposed revised withdrawal schedule has been recommended to us by Westinghouse Electric Corporation and by Southwest Research Institute, our inservice inspection consul-tant.
,The recalculated lead factors as supplied to us by Westinghouse are higher than the previously reported factors also generated by Westinghouse.
The changes marked on pages B 3/4 4-11 (Unit 1) and B 3/4 4-10 (Unit 2) correct a previous typographical error.
ATTACHMENT "B'O AEP:NRC:00121
TABLE 4.4 5
REACTOR VESSEL HhTERIAL IRRADIATION SURVEILLANCE SCHEDULE
- l. Capsule T
3 SPEC INEN t
- 2. Capsule X
REHOVAL INTERVAL 1.25 EFPY 3
- 3. Capsule Y
- 4. Capsule U
- 5. Capsule S
- 5. Capsules V>
M>
5 EFPY 9
e
~
0 REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTN
, the results of these tests are shown in Table 8 3/4.4-1.
Reactor opelZtion and resultant fast neutron (E>l Nev) irradiation wi11 cause an increase in the RTN Therefore, an adjusted reference temperature, based upon the inhuence and copper content of the material in question, can be predicted using Figures 8 3/4.4-1 and 8 3/4.4-2.
he heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted adjustments for this shift in RTR at the end of 12 EFPY, as uelI as adjustments for possible errors the pressure and temperature sensing instruments.
The actual shift in RT T of the vessel material will be established periodically during operatii by removing and eialuating, in accordance 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 and 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 LRT, determined from the surveillance capsule is different from the calNfated aRTRR 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 to assure compliance with the minimum temperature require-ments of Appendix 6 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-$ to assure compliance with the requirenents 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 fatigue analysis performed in accordance with the ASME Code requirements.
3/4.4.10 STRUCTURAL INTEGRITY The r equired inspection programs for the Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level through'out the life of the plant.
To the extent applicable, the inspection program for the Reactor Coolant System components is in compliance with Section XI of 0,
C.
COOK-UNIT 1 8 3/4 4-11
TABE L) tV C7 0A A
0O 1 ~
CAPsuLR 7
2~
CAPSULE Y
Gar SULK X
.4i CAPSULE U
5o CAPSULE s
6.
CAPSULES V, M, Z 1
EFPY EFPY 5
EFPY 9
$2 EFPY STANOSY
~
~
. *4
\\s 1
v O.
REACTOR COOLANT SYSTEM BASES s
7 The actual shift in RTN T of the vessel material will be established periodical'ly during operatioI2 by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance speci-mens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and 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 calcu1a h aRT<pT 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 to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and the frequen~ies for re'moving and testing these specimens are provided in Table 4.4-5 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 fatigue analysis performed in accordance with the ASME Code requirements.
3/4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1,
2 and 3
components ensure that the. structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
D. C.
COOK - UNIT 2 B 3/4 4-10
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