ML20054A938
| ML20054A938 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/07/1982 |
| From: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-15-17, TASK-RR 810480, NUDOCS 8204160228 | |
| Download: ML20054A938 (17) | |
Text
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CONNECTICUT YANKEE ATOMIC POWER COMPANY BERLIN. CONNECTICUT P O BOX 270 H A RTFORD. CON NECTICUT C6101 v.<._,
April 7, 1982 g
C Docket No. 50-213 g y;g g
B10480 P-APR 151982>
c um wm tvm -10 Director of Nuclear Reactor Regulation
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Attn:
Mr. Dennis M. Crutchfield, Chief ff Operating Reactors Branch #5
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U. S. Nuclear Regulatory Commission N
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Washington, D.C. 20555
References:
(1)
W. G. Counsil letter to D. G. Eisenhut, dated July 29, 1981.
(2)
W. G. Counsil letter to D. M. Crutchfield, dated September 30, 1981.
Gentlemen:
Haddam Neck Plant SEP Topic XV-17, Steam Generator Tube Failure In Reference (1), Connecticut Yankee Atomic Power Company (CYAPCO) committed to sobmit the Safety Assessment Reports (SARs) for the l
Section XV Design Basis Events for the Haddam Neck Plant. In Refer-l ence (2), CYAPCO provided Safety Assessment Reports for all Section
'YV topics with the exception of Topic XV-17, Steam Generator Tube Failure.
It was noted in Reference (2) that CYAPCO was in the pro-cess of reanalyzing this event since the original analysis did not assume a loss of offsite power or iodine spiking. The purpose of this submittal is to forward the results of this reanalysis.
In accordance with the Reference (1) and (2) commitments, CYAPCO is providing the Safety Assessment Report for SEP Topic XV-17, Steam Generator Tube Failure, which is included as Attachment 1.
We trust the Staff will appropriately use this information to develop a Safety Evaluation Report for this SEP topic.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY f'
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l W. G. Counsil Senior Vice President 8204160228 820407 PDR ADOCK 03000213 p
Docket No. 50-213 l
ATTAC11 MENT 1 SAFETY ASSESSMENT REPORT SEP TOPIC XV-17, STEAM GENERATOR TUBE FAILURE l
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i April, 1982 I
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1.0 Introduction The safety objective of this topic is to assure that the releases from a steam generator tube rupture will not result in exposures in excess of the established guidelines.
In the event of a steam generator tube rupture, primary coolant will be discharged to the secondary side until the primary system pressure is reduced to less than the pressure in the steam generators. This discharge of primary coolant will result in a radioactive release to the environment.
2.0 Criteria The review criteria for evaluation of the steam generator tube rupture event are presented in Standard Review Plan section 15.6.3.
3.0 Discussion The original analysis of the Steam Generator Tube Rupture event for the Haddam Neck Plant is contained in the Facility Description and Safety Analysis (FDSA).
For the Systematic Evaluation Program, this event has been reanalyzed using Standard Review Plan criteria.
3.1 System Response The Steam Generator Tube Rupture event has been analyzed utilizing the RETRAN2 computer code. The break was assumed to occur at the bottom tube sheet on the cold leg side of the steam generator. A loss of offsite power was assumed to occur concurrent with the tube rupture. The loss of offsite power causes a reactor trip, reactor coolant pump trip, and feedwater pump trip. The analysis also included the following assumptions:
1.
102% initial power level with end-of-life (EOL) core conditions 2.
2 x ANS decay heat curve (includes allowances for primary metal heat and analytical conservatisms)
I 3.
Highest worth control rod was assumed stuck in the fully withdrawn position with the most limit-ing E0L scram reactivity 4.
Single failure of the only Atmospheric Dump Valve l
5.
No letdown flow was assumed 6.
One of two auxiliary feedwater pumps wae modeled j
7.
Thirty minutes was assumed for operator action to isolate the affected steam generator
The tube rupture, loss of offsite power, reactor trip, reactor coolant pump trip, and feedwater pump trip are assumed to occur simultaneously at t=2.0 seconds. The charging pumps are not automatically sequenced onto the diesel generators when offsite power is not avail-able. Therefore, it was assumed that they were manually started by the operator at t=60.0 seconds.
It was assumed that the first signal to automatically initiate auxiliary feedwater (i.e. opening of main feedwater pump breakers) failed and the system was initiated on low steam generator level at 350.0 seconds.
Isolation of the affected steam generator and cooldown initiation occur at 1800.0 seconds.
The Safety Injection Signal and starting of the SI pumps occurs at 1908.0 seconds. The flow from the broken tube was assumed to continue to t=7500.0 seconds, at which time the Icak is isolated.
The following table summarizes the sequence of events:
Sequence of Events Event initiation 2.0 sec.
LOOP initiation 2.0 sec.
Reactor trip 2.0 sec.
R. C. pump trip 2.0 sec.
F. W. pump trip 2.0 sec.
Charging initiation 60.0 sec.
Aux. F. W.
initiation 350.0 sec.
Affected S. G.
isolation 1800.0 sec.
Cooldown initiation 1800.0 sec.
SIS trip actuation 1908.0 sec.
Leak flow isolation 7500.0 sec. (estimated)
The attached Figures 1 through 8 show primary and secondary system responses for the event. As can be determined from these figures, no core damage is expected to occur. Thus, the radiological consequences are limited to the release of primary and secondary coolant activity.
3.2 Radiological Consequences l
The potential radiological consequences of a steam generator tube rupture event have been evaluated in accordance with Standard Review Plan 15.6.3.
The event was evaluated both with and without a pre-incident iodine spike and with and without a coincident loss of offsite power.
The iodine concentration in the primary coolant was assumed
[
to be 1.0 uCi/gm Dose Equivalent Iodine 131 (DEQ I-131) for l
the cases which do not assume a pre-incident spike.
Iodine l
spiking caused by the tube rupture was modeled as a 500-fold increase in the iodine release rate from the fuel. The iodine concentration in the primary coolant was assumed to be 60.0 uCi/gm DEQ I-131 for cases which assumed a pre-incident spike. No additional fuel was assumed to fail.
1 P
In the loss of offsite power case, the release paths for activity were the steam generator safety relief valves and atmospheric steam dump via the terry turbine exhaust.
The case with offsite power available assumed that the release paths for activity were the terry turbine exhaust and the Steam Jet Air Ejector.
Decontamination factors in the steam generator were computed in a manner consistent with NUREG-0409, " Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident".
]
The doses were calculated using the semi-infinite cloud model and the dose conversion factors in Regulatory Guide 1.109, Revision 1.
Meteorological dispersion coefficients were based on the 95% worst case meteorology over five years, as discussed in j
Section 7.2 of Reference (2). A list of assumptions is included as Table 1.
l The calculated thyroid and whole body doses at the Exclusion Area boundary and at the low population zone are shown in Table 2.
The resultant doses are below the limits of 10CFR100 and are, therefore, accept-able.
4.0 Conclusions l
The steam generator tube rupture accident at the lladdam Neck Plant has been evaluated using current licensing criteria. Results show that no fuel damage is expected to occur and the radiological consequences are well below 10CFR100 guidelines. Therefore, no further action is required.
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1, TABLE 1
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4 Assumptions For The Radiological Evaluation of a SGTRA at Haddam Neck Assumption Basis 1.
Iodine Concentration in primary 1 uCi/gm DEQ I-131, equilibrium case Westinghouse t
Coolant
=
i 60 uCi/gm DEQ I-131, pre-accident spike case Standard Tech Specs
=
2.
Iodine Concentration in Secondary Coolant (including DWST inventory) 0.10 uCi/gm DEQ I-131 Tech Specs
=
3.
Iodine Partition Factor in Steam Generator = 100 S.R.P. 15.6.3 4
4.
Iodine Spiking Factor = 500 S.R.P. 15.6.3 6
5.
Iodine Dose Conversion Factor = 1.49 x 10 Rem Reg. Guide 1.109, Ci-DEQ I-131 Rev. 1 l
6.
Breathing Rate = 3.47 x 10 4 3/sec Reg. Guide 1.4 3
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7.
Decontamination Factor in Condenser /SJAE = 133 Demonstration of Compliance with 10CFR50, Appendix I Document 8.
Total Length of Release = 7500 sec Conservative extra-polation 5 lbm Conservative Calcula -
9.
Total Mass of Primary Coolant Release = 1.94 x 10 tion based on 7500 sec.
i duration of release 3
10.
Atmospheric Dispersion Coefficients (sec/m )
(0-2 hr) EAB = 1.08 x 10-3 t
(0-8 hr) LPZ = 2.91 x 10-5 OFor a discussion of atmospheric dispersion' coefficients, see Section 7.2 of Reference (2).
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TABLE 2-r Radiological Consequences of a Steam Generator Tube Rupture Accident Doses, Rem No Loss of Offsite AC Power Loss of Offsite AC Power Equilibrium & Spike Pre-Accident Spike Equilibrium & Spike Pre-Accident Spike Thyroid Whole Body Thyroid Whole Body Thyroid Whole Body Thyroid Whole Body EAB 0.247 1.150 0.807 1.150 8.648 2.384 18.05 2.384 LPZ 0.0067 0.0310 0.0217 0.0310 0.2330 0.0642 0.4863 0.0642 1
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