ML19250C445
| ML19250C445 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 11/21/1979 |
| From: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-04-01.A, TASK-4-1.A, TASK-RR NUDOCS 7911260134 | |
| Download: ML19250C445 (9) | |
Text
.o CONNECTICUT YANKEE ATOMIC POWER COMPANY BERLIN. CO N N ECTIC U T P. O. BOX 2 70 H ARTFORD. CONN ECTICUT 06101 T a t s.n oss a 203-666-6911 November 21, 1979 Docket No. 50-213 Director of Nuclear Reactor Regulation Attn:
Mr. D. L. Ziemann, Chief Operating Reactors Branch #2 U. S. Nuclear Regulatory Commission Washington, D. C.
20555
References:
(1)
W. G. Counsil letter to D. L. Ziemann dated September 22, 1978.
(2)
W. G. Counsil letter to D. L. Ziemann dated October 20, 1978.
(3)
D. L. Ziemann letter to W. G. Counsil dated October 24, 1978.
Centlemen:
Haddam Neck Plant N-1 Loop Operation During the past several months, our respective Staffs have discussed the topic of three-loop operation at the Haddam Neck Plant, focusing on t6 technical and analytical basis justifying operation in this mode. A synopsis of the justification is provided in Attachment 1, N-1 Loop Operation at the Haddam Neck Plent.
As noted in the Attachment, either a specific analysis or an engineering evalua-tion has been completed for each incident addressed in Chapter 10 of the FDSA with particular emphasis placed on steamline break analyses. The results of these analyses are summarized in Table 3.
The methodology employed for all steamline break analysis, both three-loop and four-loop cases, discussed in the Attachment is identical to that provided in References (1) and (2).
It is noted that these methods were reviewed and accepted by the NRC Staff in Reference (3). The results indicate that for the limiting case, three-loop operation at 33% power, the minimum shutdown margin is 1.30%oK. The values presented in Table 3 for operation at reduced power represent a refinement of the information presented in Reference (2).
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, Based upon the information presented in the Attachment, Connecticut Yankee Atomic Power Company (CYAPCO) continues to conclude that three-loop operation at the Haddam Neck Plant is justified.
Very truly yours, CONNECTICUT YA?'SE ATOMIC POWER COMPANY s
W. G. Coansil Vice President Attachment bi w, l
DOCKET NO. 50-213 ATTACHMENT tiADb.*M NECK PLANT N-1 LOO.' OPERATION l i ef 10 4
\\J,r Ia NOVEMBER, 1979
N-1 LOOP OPERATION IT THE HADDAM NECK PLANT The technical justification for three-loop operation at the Haddam Neck Plant for non-LOCA transients is based on two efforts: recent steamline break analyses and previously performed accident analyses. These analyses combined with the docketed steady-state and LOCA analyses provide a complete engineering basis for three-loop operation at the Haddam Neck Plant.
The previously performed set of accident analyses were conducted in order to support a proposed power uprating to 86% power for three-loop operation.
The effort involved a reanalysis of the Chapter 10 FDSA non-LOCA transients. The accidents specifically reanalyzed at the uprated power level, assuming similar input parameters as those in the FDSA were: Loss of Flow, Isolated Loop Startup, Loss of Feedwater, Rod Withdrawal, and Steamline Break.
Table 1 presents a comparison of the results between four-loop 100% power transients and three-loop 86% power transients. The three-loop results shown would conservatively bound these accidents initiating at the current license limit on power level for three-loop operation (i.e., 65% power).
All other FDSA non-LOCA transients (i.e., Excess Load, Boron Dilution, Excess Feedwater Flow, Control Rod Ejection, Dropped Rod, Steam Generator Tube Rupture, Loss of Load, Fuel Handling Incident, Waste Gas Incident and Hypothetical Accident) not specifically analyzed, were determined not to be limiting in the three-loop configuration, based upon an evaluation of the parameters of interest for the above incident.
In general, for those accidents not explicitly re-analyzed, in each case, it was determined that either four-loop operation was inherently more severe than three-loop operation or that the accidents were completely independent of reactor operating conditions.
Due to the modification, incorporated in Cycle 8, of the charging pump automatic starting logic following a safety injection signal, it was necessary to reanalyze the Steamline Break analysis. CYAPCO calculated the minimum approach to criticality following a reactor trip during a steamline break for four-loop HFP, HZP, and intermediate power cases (docketed September 22 and October 20, 1978), and subsequently performed three-loop HZP and 65% cases. Table 2 shows the various key parameters used in both the three and four-loop cases.
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, Table 3 summarizes the results of ther,e analyses (Cycle 8) and also the results of the similar calculations performcd for Cycle 9.
It reveals that the limiting configuration (33% power, three-loop) still maintained a minimum shutdown margin at the end of cycle greater than or equal to 1.3%AK.
Please note that a specific analysis was not performed for three-loop opera-tion in Cycle 9 because the Cycle 8 analysis showed that the three-loop configuration was only slightly more severe than the four-loop, which was demonstrated to have adequate shutdown.
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TABLE 1 Incident Limitino Parameter
_% Power Level Loops Result Loss of Flow DNBR 100 4
1.30*
75 3
1.80*
86 3
1.32 Isolated Loop DNBR 75 3
1.87*
Startup 8C 3
1.52 Loss of Feedwater S/G Drain Time 100 4
14.5* min.
Flow 75 3
15.3 min.
86 3
12.4 min.
Rod Withdrawal DNBR 100 4
1.32*
86 3
1.72 Steanline Break DNBR 100 4
1.65*
(full power) 86 3
1.62
- FDSA Analysis O ] k
?!'] b
TABLE 2 PARAMETERS USED Ifl SLB AtlALYSIS INITIAL RX.
RX. TRIP R0D WORTH, &
STEAf1 FLOW EACH FEEDWATER STEAM GEN LIQ.
CASE
- POWER, %
T F SETPOINT IllIT./AFTER C00LDOWN DEFICIT STEAM GEN.
LB/SEC TEf1P.,
F INVENTORY. LB.
in.
4 LOOP-0PA 100 529 109%
8.18/7.387 547 432 38245 4 LOOP-CLOP 100 529 t=0 8.18/7.387 547 432 38245 4 LOOP-0PA 0
535 34%
6.90/6.07 1.9 90 72377 4 LOOP-CLOP 0
535 t=0 6.90/6.07 1.9 90 72377 3 LOOP-0PA 65 517 85%
7.352/6.804 449 388 39403 3 LOOP-CLOP 65 517 t=o 7.352/6.804 449 388 39403 3 LOOP-0PA 0
535 34%
7.62/6.07 1.9 1.9 72377 3 LOOP-CLOP 0
535 t=0 7.62/6.07 1.9 1.9 72377
- 0PA - Offsite Power Available CLOP - Coincident Loss of Offsite Power
,I sa
TABLE 2 (CONT)
PARAMETERS USED IN SLB ANALYSIS INITIAL RX.
TIME BORON STEAM GEN. FEED STEAM GEN. STEAM CASE POWER, %
REACHES CORE FLOW RESPONSE FLOW RESPONSE 4 LOOP-0PA 100 55 0-8 sec. (N); 8-11 sec. (N+F) 0-8 sec. (N); 8-11 sec. (N+D) 11-13 sec. (F); 13-16 sec. (F+0) 11-13 sec. (D); 13-16 sec. (D40) 4 LOOP-CLOP 100 76 0-20 sec. (R+0) 0-3 sec. (N+0); 3-13 sec. (D) 13-16 sec. (D+0) 4 LOOP-0PA 0
61 constant constant 4 LOOP-CLOP 0
99 constant constant 3 LOOP-0PA 65 59 O-11.5 sec. (N); 11.5-18.5 sec. (N+0) 0-11.5 sec. (N); 11.5-14.5 sec. (N+0)
(full feed until Tave < 535) 3 LOOP-CLOP 65 61 0-20 sec. (N40) 0-3 sec. (R+0) 3 LOOP-0PA 0
61 constant constant 3 l.00P-CLOP 0
93 constant constant N - Normal Operation of Valves at Initial Reactor Power D - Steam Dump Valves Fully Open F - Feedwater Control Valves Fully Open 0 - Valves Completely Closed C
4 Dm.
<r CD
TABLE 3 SLB MARGIN TO CRITICALITY AT E0C (% Ak)
CYCLE 8 CYCLE 9 CASES ANALYZED (RELAP 4 M00. 5) 4 LOOP 3 LOOP 4 LOOP 1
I l
HZP 1.67 1.53 1.64 3
33%
1.58 1.30I 1
65%
1.58 HFP 1.18 1.48 10ffsite Power Available 2Values Interpolated at the Point of Minimum Available Scram Worth (w/ Stuck Rod) 3.69% Ak was Calculated for this case in the docketed 0
letter of October 20, 1978, using very conservative assumptions as described in the submittal.
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