ML19323B999
| ML19323B999 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/09/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Rubenstein L Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8005140500 | |
| Download: ML19323B999 (6) | |
Text
8006140 000 f
400 Chestnut Street Tower II May 9, 1980 s
Director of Nuclear Reactor Regulation Attention:
Mr. L. S. Rubenstein, Acting Chief Light Water Reactors Branch No. 4 Division of Project Management U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Rubenstein:
In the Matter of the Application of
)
Docket No. 50-327 Tennessee Valley Authority
)
Reference:
Letter from L. M. Mills to L. S. Rubenstein dated April 9, 1980 Enclosed for your review is additional information regarding the safety evaluation of the special test program for Sequoyah Nuclear Plant (SNP). This information should be incorporated into section 4.0 of the master document for the SNP special test program that was transmitted to you in the referenced letter.
This information summarizes the analytical effort made by Westinghouse and provides details on the analysis methods and assumptions used and the basis for the conclusions reached in the safety evaluation document.
Very truly yours, TENNESSEE VALLEY AUTHORITY
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L. M. Mills, Manager Nuclear Regulation and Safety Enclosure (10)
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%D <-> t THis DOCUMENT CONTAINS P00R QUAUTY PAGES
s ENCLOSURE TABLE I SUiMARY OF SAFETY EVALUATION, SECTI0ll 4.0*
Section Transient Test:
1 2
3 4
5 6
7 8
9A 93 4.1
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1.1 RCCA Bank With., Suberit 2,4 2,4 2,4 2,4 2,4 1
2,4 2
1 4
1.2 RCCA Bank With., at Power 4
4 4
4 4
1 4
4 1
4 1.3 RCCA Misalignment 1
1 1
1 1
1 1
1 1
1 1.4 Doron Dilution 1
1 1
1 1
1 1
1 1
1 1.5 Partial Loss of Flow 1
1 1
1 1
1 1
1 1
1 1.6 Start Inactive Loop 1
1 1
1 1
1 1
1 1
1 1.7 L ss of Load 1
1 1
1 1
1 1
1 1
1 1.8 Loss of Fecdwater 1
1 1
1 1
1 3
1 1
1 1.9 Loss Offsite Powc.
1 1
1 1
1 1
3 1
1 1
1.10 Excessive Feedwater 2
2 2
2 2
1 2
2 1
2 1.11 Excessive Load 2
2 2~
2 2
1
,.2 2
1 2
1.12 RCS Depressurization 1
1 4
1 4
1 1
1 1
1 1.13 Steam Depressurization 1
1 1
1 1
1 1
1 1
1 1.14 Spurious Safety Injection 1
1 1
1 1
1 1
1 1
1 2.1 Small LOCA 3
3 3
3 3
3 3-~
3 3
3 2.2 Small Secondary Breaks 2
2.
2 2
2 1
2.
2 1
2 2.3 Single RCCA Withdrawal 4
4 4
4 4
1
'4 4
1 4
2.4 Misloaded Fuel Assembly 1
1 1
1 1
1 1
1 1
1 Complete Loss of Flow 1
1 1
1 1
1 1
1 1
1 Waste Gas Decay Tank Brk.
I 1
1 1
1 1
1 1
1 1
3.1 Major LOCA 3
3 3
3 3
3 3
3 3
3 3.2 Major Secondary Break 2,3 2,3 2,3 2,3 2,3 1
2,3 2,3 1
2,3 3.3 S/G Tube Rupture 1
1-1 1
1 1
1 1
1 1
3.4 RCP Locked Rotor 1
1 1
1 1
1 1
1 1
1 3.5 Fuel Handl.ing 1
1 1
1 1
1 l'
1 1
1 3.6 Ruptured CRDM 3,5 3,5 3,5 3,5 3,5 1
3,5 3,5 1
3,5
- Dases of Evaluation 1.
Bounded by FSAR analysis results si 2.
Rea.nalysis shows fuel clad integrity is maintained
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3.
Operator action is required for protection 4.
Probability of occurrance reduced by restrictions on operating conditions 5.
Probability of occurrance reduced by short testing period solely
y Tests 6 and 9A Cooldown Ca20bility of the Chargino and Letdown f_oyced Circulation Cooldown For these tests all transients.with the exception of small LOCA and major LOCA are bounded by FSAR analyses.
For both small and mcjor LOCAs, operator action is required to mitigate the consequences of the transient.
For small LOCA the operator action (SI initiation) is required within s1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
For a major LOCA assuming cold leg accumulator injection, the operator action (SI initiation with eventual transfer to long term cooling) is required within sl.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to prevent core uncuvering.
Tests 1, 2 and 4 Natural Circulation, Loss of Offsite AC Power and Steam' Generator Isolation Loss of coolant considerations for these tests are the same as those for Tests 6 and 9A as discussed above.
The test procedures require that only a limited amount of excess reactivity is available via the control rods.
This.in combin-ation with the fact that automatic rod control is not used reduces the prob-ability of an extended rod seithdrawal.
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Analyses have 'shown that for cases similar to those presented in the FSAR, i.e.
continuous insertion rate, a reactor trip will occur on the low setting of the high flux trip. Because of the difficulty of calculating core hydraulic be-havior under test conditions, the potential for DNB renulting from inadvertant rod withdrawal has not been precluded by analysis. Once reactor trip is obtained 6 rod assumed to be in DNB would have adequate cooling.
Figure 4.2.4 of the Safety Evaluation shows the time of reactor trip as a function of insertion rate.
This figure represents-the maximum amount of time a rod could ba in Di;B as a result of the rod withdrawal.
As a consequence of limiting the excess reactivity and the fact that many norcal trips are bypassed, there is another condition not considered in the FSAR. This condition -is that a continuous rod withdrawal will not result in sufficient in-crease in power (105 assuming setpoint errors) to cause a reactor trip..-Because of the difficulty of predicting DHD under the test conditions, it has not been absolutely " demonstrated that DNB uill not result with an ensuing clad tempera-ture excursion until manual reactor trip.is initiated. Again, due to analysis limitations, a reasonable low bound on required operator action ti: c has not been demonstrated. Therefore, major reliance is placed on the low probability of inadvertant rod withdrawal events occurring during the limited test duration.
An analysis of the single uncontrolled RCCA withdrawal has not been perforced at these test conditions. The probability of an uncontrolled single RCCA with--
~. :r drawal has been reduced by a restriction on manual operation and lack of a' requirement for extended rod bank withdrawal as a part of the test procedure.
-For_steamline breaks outside containment, automatic DNB protection is provided by the flux trip. = Since the NIS detectors and associated wiring are not totally environmentally ' qualified for steamlinc~ break conditions, a break inside con-tainment may be more severe. Thiscis due to additional environmental errors ~ on
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the flux channel which could delay or possibly prevent the flux trip. -In such
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N a case operator action would be necessary to terminate the transient. Analyses have not been performed for this case.
Due to the low probability of occurrance and the short operat.'ng time at low flow, the potential for a reactivity transient induced by a postulated rupture of a~ control rod drive mechanism housing for low reactor coolant flow condition is reduced.
The subscquent consequences of the rod cjection, i.e., a small LOCA are covered by the above LOCA discussions.
Test 7 Simulated Loss of All Onsite and Offsite Power The above discussion of tests 1, 2 and 4 apply equally to Test 7.
During this test the motor-driven auxiliary feedwater ptmips will not start on loss of' off-site power or loss of feedwater. Operator action will be required to provide power to the motor-driven auxiliary feedwater pumps in th? event that the turbine-driven auxiliary feedwater pump fails to provide adequate feedwater flow. The operator has at least.30 minutes in which to accomplish this action to prevent steam generator dryout and filling the pressurizer.
Tests 3 and 5 flatural Circulation with Loss of Pressurizer Heaters and at Reduced Pressure The previous discussions of tests 1, 2 and 4 apply equally to tests 3 and 5, with the exception that automatic protection against inadvertant RCS depres-surization (i.e. reactor trip on low pressurizer pressure safety injection
. signal) will be blocked for these tests during low pressure operation. The pressurizer power operated relief valves will also be blocked, which reduces the probability of occurrance. The other credible cause of depressurization, short of LOCA, is auxiliary pressurizer spray which has no automatic actuation.
The depressurization rate with auxiliary spray is small compared to the rate produced by an assumed stuck open PORV. ilo seccific analyses have been performed to determine whether or not DiB would occur and hence possible clad damage has not been precluded.
Test 8 Establishment of ?!atural Circulation from Staonant Conditiens The previous discussion of LOCA apply to test 8.
Uncontrolled rod withdrawal or rod ejection from any power level during the time in Test 8 when coolant flow lags behind the steady-state natural circulation flow for the prevailing power level could be more severe because of this lower flow than for any cther tests except for 98.
The conclusions discussed previously. for Tests 1, 2 and 4 apply except that the probability of an inadvertant rod withdrawal event may be increased for this-test due to t.,he test procedures requiring rod withdrawal.
Tent 98 Boron liixing and Cooldown for this test the following transients do not fall within the bounds of FSAR analyses or reanalyses do not assure that clad integrity is maintained:
RCCA Dank Withdrawal from Subcritical,' RCCA Bank Uithdrawal at Power, Single'RCCA Withdrawal, and Ruptured CRDM. ~The small_ LOCA and major LOCA, transients are the same as described for the other tests. This test presents unt sual circum-i stances for uncontrolled' rod or bank movement because of the need for long tenn
p-
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control rod movement to remain at power while per forming the boron. mixing portion of the test, and to respond to reactivity changes brougnt about by the changing moderator temperature coefficient with decreasing temperature.
The test procedures should be revised to instruct the operator to borate at a sniall rate, i.e. at less than 5 gpm of 12L boric acid blended with charging flow, such that a need for large control rod movements are reduced.
Very little data exists for mixing at these low flow rates. The consequences can be bounded by assuming either perfect mixing of the charge Icop with the ~other loops such that the boron concentration is uniform in the core, or no mixing where the charged loop does not mix and causes a non-uniform boron concentration in the These two extremes result in different effects on the power distribution core.
and may indicate to the operator different rod withdrawal rate requirements because the reactivity ccmputer is connected to only one excore detector. To help the operator receive a better indication of core power level, e.g. radial tilt and quadrant power, it is recommended that the test procedures instruct the operator to place the incore detectors in the lower half of the core and uonitor them continuously during this portion of the test.
During the second phase of the test the operator will have to first run the control rods in until the transition of the isothermal moderator coefficient from a negative to a positive value, at which point the operater will withdraw the control rods with decreasing temperature. This coupled with the increasing uncertainty in actual core power due to the increasing error in excore detector
. power levels requires that the operator. decrease the system temperature in a very orderly fashion.
In the event an uncontrolled rod withdrawal at specified operating power occurs during the boration pilase, the discussion of consequences for Tests 1, 2, 3, 4, 5 and 7 would apply.
Should this event occur during the ecoldo..n phase, the " shadowing" of NIS detectors by cold downcemer water.would increase th:-
actual power at which high ncutron flux trip would occur, and thus the time taken to reach trip would be greater than in the other tests.
L'hile the consequences of an uncontrolled RCCA bank withdrawal fro.n a subc'ritical condition in test SB is reduced by operating restrictions against any rod with-draul in that condition, should it occur during the boration phase, the can'-
sequences could be more severe than for the other tests because of the decper rod insertion and greater available reactivity. Analysis was not perfcreed for this condition.
Applicability of Analysis Methods The analyses presented in Section 4.~2 of the Safety Evaluation attcapt to Lound the consequences of some of the postulated faults,-using standard FSAR bounding
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assumptions as applicable to the test conditions and standard FSAR analysis methods.
Because of the far off-normal conditions existing 'during the tests, strict applicability of those methods to transients originating from test' con-ditions has not. bcen established.
Some o.f these aspects are here: discussed.
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,3 System Codes __(LOFTRNI)
The LOFTRAN model is a nodal code, as such there is the concern of the treat-ment of non-unifonn conditions, e.g. different boron concentrations within a node.
For the standard FSAR cases with high flow this is not a concern, however, for ~these tests it cannot be stated that the effects of fronts (boron, cold water slugs) are correctly treated. liixing within the core is not modelled and mixing in the vessel is not verified with experimental data. This could have an affect on the no mixing extremes of the baron mixing and steam generator cooldown and isolation tests.
LOFTRAN has'been verified for several natural circulation tests in operating Westinghouse reactors including at least one four loop reactor of size and configuration similar to Sequoyah.
Core Thermal-Hydraulic Code (TilINC-IV)
The conditions of natural circulation are within the scope of n:odelling in the TillHC-IV computer code. The experimental verification included in UCAP-7956 (THIHC-IV. An Improved Program for Thermal-Hydraulic Analysis of Rnd Cundle Cores) discusses the analysis of plant natural circulation testing with THlUC-IV.
However, some accident analyses for this study results in some boiling in the core and convergence was difficult to obtain. These cases were assumed to go into DMB as part of the safety evaluation performed in order to bound the consequences of the accidents under study.
Fuel Temperature (FACTRAM)
If DNB is assumed to occur, both. local flow conditions and a heat transfer cor-
-relation applicable to those conditions are required to calculate clad temper-i
.atures.
For the analysis of P.CCA bank withdrawal from a suberitical condition for Test 8, an extremely low value of 2 Btu /hr ft.2 F was used to bound the temperature transient. This value is appropriate to a core unccvering condi-tion with steam cooling and is extremely conservative for the case with the RCS filled with water.
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