ML19352A605
| ML19352A605 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/03/1975 |
| From: | Switzer D NORTHEAST NUCLEAR ENERGY CO. |
| To: | Anthony Giambusso Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18025B195 | List:
|
| References | |
| FOIA-80-587 NUDOCS 8104170265 | |
| Download: ML19352A605 (6) | |
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P o Bcx 27C HARTFO Ao CoNNEOTICuT C6tC1 NORTHLAST NUCLEAR ENERGY COMPANY
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2C3-666-6911 A NORTHEAST UTIUTIES CoVPANf
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February 3,1975 Mr. A. Giambusso Docket No. 50-336 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555
References:
(1) Technical Report on ATWS for Water Cooled Power Reactors: WASH-1270, September 1973 (2)
D. C. Switzer (NNECO) letter to A. Giambusso, (AEC),
dated September 30, 1974 (3)
D. C. Switzer (NNECO) letter to A. Giambusso, dated January 3, 1975 (4)
F. M. Stern (CE) letter to 0. D. Parr, dated September 27, 1974
Dear Sir:
Millstone Nuclear Power Station, Unit No. 2 Submittal of ATUS Evaluation in Reference (2), it was stated that NNECO intended to submit the re-sults of the ATWS evaluation required by Reference (1) follewing review of the Combustion Ly.gineering (CE) generic report, CENPD-158, by NNECO, Northeast Utilities Service Co'epany (NUSCO), and the Bechtel Power Corporation. The CE report has been submitted to the NRC Staff for review by Reference 4.
The CE generic report has been reviewed and is considered to be gener-ally applicable to the Millstone Unit No. 2 system design. This report described the results of the analyses of ten separate transients. The I
results indicated that in only three of these transients, namely, Loss of External Load, Loss of Normal Electrical Power, and Complete Loss of Normal Feedwater Flow were the limits of the criteria, as stated in Section 1.2 of CENPD-lS8, approached.
For the remaining transients, all of the criteria specified in WASH-1270 are met, i.e., the consequences
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of the postulated ATWS are acceptable.
With respect to the applicability of the CE generic report, it should be noted that there are minor differences between t'he specific Millstone Unit No. 2 design and the CE model; these differences are enumerated in.
It is felt, however, that the differences do'not signifi-cantly affect the results of the analyses except as noted in subsequent F "" {p j5" discussion.
In addition, the initial conditions used in the specific Millstone Unit No. 2 ATUS evaluations are tabulated in Attachment 2.
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Two of the aforementioned transients, namely the Loss of Electric Power 1 and Loss of Normal Feedwater, as analyzed, are not deemed to be antici y.-
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@ n the specific Millstone Unit No. 2 design. For the Loss g e of Power transients, two types of common mode failures (CMF) have been
- g postulated as causing a failure to scram. The first is a CMF in the
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Reactor Protection System (RPS); however, recent studies (Reference 3) p',,, 7 J'.J '
have indicated that the Mil? stone Unit No. 2_RPS is not Darticularly f
vulnerahla to CMF.
Even assuming a CMF, the system is designed f ail- (,*/
' safe in that power is always lost to the Control Rod Drive Mechanism (CRDM) and therefore, the Control Element Assemblies (CEA) will fall by gravity into the core. The results of this transient would be less' severe than those of accidents analyzed in Section 14.0 of the Millstone Unit No. 2 FSAR.
The second type of CMF postulated is that in which the CRDMs experience a mechanical failure such that no CEA is inserted into the core when.
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power is removed from the CRDM holding coils.
It is felt that this type
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of CMF has an exceedingly low probability of occurrence cue to the CRDM 7
magnetic jack design and frequent exercising of the CEA's during normal plant operation. Therefore, any development of conditions which could affect CEA insertion would be detected and corrected before a signifi-cant number of CEA's were affected.
It is our understanding that CE has been discussing this matter with the NRC Staff and plans to document their analyses shortly.
With respect to the complete Loss of Feedwater transient, failure mode A
evaluations of the Millstone Unit No. 2 system as described in FSAR y
Section 10.4.5 have indicated that no single active failure will cause a e
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complete loss of feedwater, except in one specific instance. During certain operating modes, e.g., one feedwater pump out of service, it is
[ ~,f possible that a single factor may result in a complete loss of normal
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i feedwater flow. However, it is felt that the combined probability of an
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anticipated transient in conjunction with operating in an abnormal M
operating mode with one feed pump out of service is exceedingly low and, j
therefore, acceptable.
Based on the above discussion and the redundancy of the Millstone Unit No. 2 Main Feedwater System, partial loss of feedwater flow, as dis-i cussed in.Section 2.6.2.2.1 of CENPD-158, is considered to be the limiting 'anticipaten transient that can be caused by a single failure in l
the Main Feedwater System. The peak pressure for this transient is less than 2750 psia.
For the Loss of Load transient, preliminary calculations based on the N
general assumptions in Section 1.5 of CENPD-158 and Attachment 2 have
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dN indicated that the maximum Reaccor Coolant System (RCS) pressure is 3542 psia. This value is slightly above the calculated emergency stres I ' '
level for the limiting RCS component in this plant.
I.aative to the criteria for the maximum allowable RCS pressure (Ecer-gency Conditions) in Appendix A of WASH-1270 it is felt that it is unduly stringent compared to the probability of an ATWS event.
Such an
_3-Y eventchouldbeclassifiedasConditionIV,asdescribe b.a ANSI-N.18.2-1973.
According to the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, Faulted Conditions
/ f/ 'p '.{ 3 (Limiting Faults) apply to Condition IV events.
It should be noted that the NRC Staff has not yet completed review of
.i the n.ethods used in CDiPD-158. Therefore, the methods are presently un-approved and subject to modification.
For example, the de ree of con-
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g servatism of the pres _surizer safety valve discha.xgg_.f, low-model has.npr.
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yendet__erminedasyet and itD6pTn' ion thatJarts_of_tAis_,jnodel
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are unnecessarily ~ conservative. An experimental program sponsored by
-T the ElecG Ic' Tower Research Inaritur. r m T)
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i deriti Q f D M valve discharge flows under ATWS co'n'dition (see y,4 Riference 2 of CENPD-158L It is 4xpe'cted
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thaf~the flow rates predicted!
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by th'e'model (homogenous, ther=al-equilibrium, isentropic dxpansion Jj.i g'
i flow theory) will be shown to be overly conservative, when data become
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available.
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) In conclusion, in view of the lack of calculational model approval and
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' the continuing discussions between CE and the NRC as well as the uncer-t,.
M' tainty with respect to the specific AT.'9 t74eav43, it is not considered i.',f appropriate to propose any plant design modifications at this time.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY O C..C 'm D. C. Switzer
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Applicability of CENPD-158 to Millstone Unit No. 2 Section Comment 2.1 Initiation of boration via charging pumps instead of
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safety injection pumps.
(Applies to other transients analyzed).
2.6.1.1 Auxiliary feedwater contains 2-motor dri,ven and 1-turbine driven pump (s) instead of vice versa.
2.6.2.2.2 Closure of MSIV will not isolate steam generator from the atmospheric dump system.
2.6.3 Containment design pressure is 54 psig instead of 50 psig.
2.7.1.2 Delete last sentence of the second paragraph.
2.8.1.2 Delete Item 1, second paragraph.
2.8.2.2 Delete the sentence in the second paragraph starting
" Auxiliary feedwater is initiated...."
i 2.10.2.2 Delete "which also isolates..." from third sentence, second paragraph.
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Initial Condition for Millstone Unit No.,2 ATWS Evaluation ANS-N661 Parameter Value Basis Classification
~b Moderator BOC -0.2X10 ag/ F Least negative calcu-d Temperature lated valve for Coefficient Millstone 2 (without' uncertainties)
Doppler Figure 3.4-1 Calculated value for d
Coefficient Millstone 2 FSAR beginning of cycle (without uncertainties)
Core Power 2560 Mwt Design power c
(without uncertainties)
Inlet 542 F Design inlet tempera-c Temperature ture at design power (without uncertainties)
Reactor 381,100 GPM Minimum calculated d
Vessel Flow value (without un-certainties)
Decay Heat ANS Decay Heat Reference 15 of d
Function Standard CENPD-158 Pressurizer 764 ft (total Normal pressurizer e
Water Volume pressurizer volume =
vater level 1500 ft3 (without uncertainties)
Pressurizer 2250 psia Normal pressurizer e
Pressure pressure (without uncertainties)
Steam 815 psia Normal steam generator 4
c Generator pressure (without un-Pressure certainties)
Steam 137,800 lbs.
Design steam generator c
Generator mass (without un-Mass certainties)
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3 ANS-N661 Parameter Value Basis Classification Feedwater 415 Btu /lb Design feedwater c
Enthalpy enthalpy (without uncertainties) 2 Pressurizer
.058 ft 2 Relief valves at d
Relieving
.0113 ft 2 each Area plus 2 Safety Valves at.0177 ft 2 each
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