ML20087L250
| ML20087L250 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/21/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.G.1, TASK-TM NUDOCS 8403270103 | |
| Download: ML20087L250 (5) | |
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s DUKE POWER GOMPANY P.O. DOX 33180 CifARLOTTE, N.O. 28242 HALB.TL*CKEH meum
.J"Z ""E March 21, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Cnief Licensing Branch No. 4 Re: Catawba Nuclear Station, Unit 1 Docket No. 50-413 Proof and Review Technical Specifications
Dear Mr. Denton:
In response to NUREG-0737 Item I.G.1, " Natural Circulation Testing," Duke Power Company has committed to conduct a natural circulation verification test at Catawba (FSAR Table 14.2.12-2, pages 35 and 36, Attachment 1).
Various Technical Specification test exemptions will be required in order to perform this test. The exemptions as shown on Attachment 2 were included in the McGuire Unit 1 Technical Specifications and will need to be included in the Catawba Unit 1-Technical Specifications.
If you have any questions concerning this proposal, please contact Mr..R. W. Quellette at (704) 373-7530.
Very truly yours, ULLd%M Hal B. Tucker RWO/php Attachments cc: Mr. James P. O'Reilly, Regional Administrator I
U. S. Nuclear Regulatory Comission J300 l
Region II lf{
101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 l
NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 1
8403270103 840321
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' Mr.THarold R. Denton, Director March 21, 1984 Page 2-cc': Palmetto Alliance 21351 Devine Street Columbia, South Carolina 29205 Mr. Jesse L. Riley Carolina Environmental Study Grouo
~854 Henley-Place.
' Charlotte, North Carolina 28207 l
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', page 1 Table 14.2.12-2 (Page 35)
NATURAL CIRCULATION VERIFICATION TEST Abstract Purpose To demonstrate the capability of the NSSS to remove sensible heat by natural circulation flow in the primary loop.
To verify that p essurizer pressure and level control systems can respond automatically to a loss of for:ed circulation and can maintain reactor coolant pressure within acceptable limits.
To verify that steam generator level and feedwater flow can be maintained under natural circulation conditions in order to maintain effective heat transfer from the reactor coolant system.
To verify, to the extent possible, the adequacy of station operating procedures and to provide operator training to satisfy NUREG 0737 requirements.
Prerequisites The reactor is critical at a power level of approximately 3% full power with all reactor coolant pumps in operation.
Rod control is in manual with Bank D positioned to maintain a slightly negative isothermal temperatere coefficient.
Pressurizer pressure and level control are in automatic.
Steam dump control is in the pressure control mode.
Steam generator tevel is being maintained through use of the auxiliary feedwater header.
The intermediate and power range (low setpoint) high level reactor trips have been reduced to approximately 7% rated thermal power.
UHI isolation valves have been gagged.
All automatic safety injection (Ss) functions except re-actor trip have been blocked.
Manual initiation of safety injection functions and automatic alarms indicating channel trips remain available te the cM mtor.
The steam generM or low-low level reactor trip setpoints have been reduced i.o 5% of span.
Overtemperature and overpower AT reactor trip signals have been blocked.
Safety injection initiation on low steamline pressure has been blocked.
Various Technical Specificatio m test exemptions are required for the conduct of this test.
These special test exemptions are provided in Technical Spec-ifications.
Special operator action guidelines are provided by the test pro-cedure to compensate for the blockina of various safety injection functions and reactor D ips.
The test is required to be performed at core burnups which ensure that no signifigant core decay heat levels are present.
Test Method The test will be initiated by tripping all operating reactor coolant pumps.
The establishment of natural circulation will be verified by observing the response of wide range hot and cold leg temperatures as well as cor< exit thermocouples. The response of pressurizer level and pressure will be ob-served.
Steam generator level and pressure response will be conitored.
Dur-l ing the performance of this test on Catawba Unit 1 only, the test will be re-peated for each operating shift at Catawba or suitable simulator facility, for the purpose of initial operator training.
Each R0 and SR0 will observe or participate in the ini".iation, detection and maintenance of astural circula-tion conditions during at least one of the test runs.
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Attachm:nt 1, page 2 Table 14.2.12-2 (Psge 36)
NATURAL CIRCULATION VERIFICATION TEST Abstract Acceptance triteria Core exit temperatures, loop ATs, and loop average temperatures do not exceed values specified by the NSSS vendor. ATs as determined by the hot leg wide range and core exit temperature indications when compared to the w!de range cold leg temperatures stabilize and do not exceed limits supplied by the NSSS vendor.
If data taken during first performance of test demonstrates accept-able performance, data does not need to be taken during subsequent operator training.
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Q AL FSAE. Tabl4. W.2. 12. r345 35 -A 3(a Special Test procram 5
For conducting the special low power test program as described in R tien I.C.1 Wh: E:
"e:! ear. fety E;; h: tion R ?:rt re hted t: th; :per: tion of the id4Mn Stati:n, Suppl::ent "umter i d:ted January 13&l, the Technical specifications may be excepted as follows: -(T:st number 3 ref r to th::e speeMied i., umia: 0.- P:rkcr's idhr of October 1k1531 t: Marcid-R. Cente,./-
Ap?:ndix C "";Cuire " cieer Steti:n - Re: pen:: te T"! conce 4 a
Technical Specificathns 2.1.1 Core Safety Limits X
2.2.1 Various Reactor Trips Overtemperature AT X
Overpower AT X
Steam Generator Level X
3.3.1 Various Reactor Trips Overtemperature AT X
Overpower AT
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X Steam Generator Level X
i 3.3.2 Safety Infection - All 1,
X automatic functions Auxiliary feedwater automatic start X
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" 0_: = Unit 1 7-1
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