ML20140F677
| ML20140F677 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/28/1983 |
| From: | Brandon R, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20140F668 | List: |
| References | |
| NEDO-30060, NUDOCS 8604010128 | |
| Download: ML20140F677 (29) | |
Text
NEDO-30060 DRF B31-00074 83NED017 Class I February 1983 VERMONT YANKEE NUCLEAR POWER STATION SINGLE LOOP OPERATION Approved: _
Approved: 3[/7![
R.L/ Gridley( Managdr R.J.B/[ don, Manager Nuclear Services Engineering Fuel and Services Licensing Operation NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL $ ELECTRIC B604010128 860327 PDR ADOCK 05000271 p PDR L
- /
NEDO-30060 IMPCRTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General Electric Ccr:pany respecting infomation in this document are contained in the Generat Electric Company Single Loop Operation Analysis Proposal No. 424-TY817-NB1 dated July 22, 1992, and nothing contained in this doc:c ent shall be construed as changing the contract. The use of this information by anyone other than Yemont Yankee Nuclear Power Corporation, for any purpose other than that for which it is intended, is not authorized; and uith respect to any un-authorized use, General Electric Comany makes no representation or carranty, and assic':es no liability as to the completeness, accuracy, or usefulness of the information contained in this doeur:ent.
h .
NEDO-30060 CONTENTS Page
- 1. INTRODUCTION AND
SUMMARY
l-1
- 2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 2.1 Core Flow Uncertainty 2-1 2.2 TIP Reading Uncertainty 2-4
- 3. MCPR OPERATING LIMIT 3-1 3.1 Core-Wide Transients 3-1 3.2 Rod Withdrawal Error 3-2 3.3 Operating MCPR finit 3-4
- 4. STABILITY ANALYSIS 4-1
- 5. ACCIDENT ANALYSES 5-1 5.1 Loss-of-Coolant Analysis 5-1 5.2 One Pump Seizure Accident 5-3
- 6. REFERENCES 6-1 iii/iv L_
I . .
NED0-30060 l
i TABLES Table Title Page 5-1 Limiting MAPLHGR Reduction Factors 5 .4 ILLUSTRATIONS Figure. Title Page r
2-1 Illustration of Single Recirculation Loop Operation Flows 2-5 3-1 Main Turbine Trip with Bypass Manual Flow Control 3-5 4-1 Typical Decay Ratio versus Power Curve for Two Loop and Single loop Operation 4-2 5-1 Vermont Yankee Core Uncovery Time versus Suction Break Area '5-5 5-2 Vermont Yankee Core Reflooding Time versus Suction
- Break Area 5-6
[ 5-3 Vermont Yankee Core Total Uncovered Time versus Suction Break Area 5-7
]
1:
5-4 Vermont Yankee Core Uncovery Time versus Discharge Break Area 5-8 l-5-5' Vermont Yankee Core Reflooding Time versus Discharge
, Break Area 5-9 6 Vermont Yankee Core Total Uncovery Time versus Discharge Break Area 5-10 v/vi L
- 1. INTRODUCTION AND
SUMMARY
The purpose of this report is to provide the justification for the operation of the Vermont Yankee Nuclear Power Station with one recirculation loop out of service. The current technical specifications limit single loop operation to a maximum 24-hour period (Technical Specification 3.6.G.1). The analysis described herein provides the' basis for removing the 24-hour time constraint.
The capability of operating at reduced power with'a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event that maintenance of a recirculation pump or other component renders one loop inoperative. To justify single loop operation, the safety analyses documented in the Final Safety Analysis Report and Reference 1 were reviewed for one pump operation. Increased uncertainties in the core total flow and Traversing In-Core Probe (T1P) readings can resu~ t in a 0.01 incremental increase in the Minimum Critical Power Ratio (MCPR) fuel cladding integrity safety limit during single loop operation. This 0.01 increase is reflected in the MCPR operating limit. No other increase in this limit is required as core-wide transients are bounded by the rated power / flow analyses performsd for each cycle. Recommendations are provided to adjust the recirculation flow-rate dependent rod block and scram setpcint equations given in the technical speci-fications for one pump operation. The least stable power / flow condition, the natural circulation point achieved by tripping both recirculation pumps, is not affected by one pump operation.
Under single loop operation, the flow control should be in master manual, since control oscillations may occur in the recirculation flow control system under these conditions.
The derived MAPLHGR reduction factor is 0.83 for the following fuel types:
8x8, 8x8R and P8x8R.
1-1 L. ,
b NED0-30060 The analyses were performed assuming the equalizer valve was closed. The discharge valve in the idle recirculation loop is normally closed, but if its closure is prevented, the suction valve in the loop should be closed to pre-vent the loss of Low Pressure Coolant Itijection (LPCI) flow out of a postulated break in the idle suction line.
1-2
\, .
- 2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT The basis for the MCPR fuel cladding integrity safety limit was derived independent of whether the coolant flow to the core is provided by one or two recirculation pumps. Suf ficient margin was included in this basis to account for uncertainties in monitoring the reactor core operating state. The two core operating parameters that might be af fected during single loop operation are the total. core flow measurements and the TIP readings. The possible ef fects on these parameters during single loop operation are discussed in this section.
Uncertainties used in the two loop operation analysis are documented in Table S.2-1 of Reference 1 for reloads. A 6% core flow measurement uncertainty has been established for single loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one stand-ard deviation (one sigma) accuracy of the core flow measurement system docu-mented in Reference 2. The random noise component of the TIP reading uncer-tainty was revised for single recirculation loop operation to reflect the operating plant test results given in Subsection 2.2 below. This revision resulted in a single loop operation process computer uncertainty of 9.1% for reload cores. A comparable two loop process computer uncertainty value is 8.7% for reload cores. The net effect of the revised core flow and TIP uncer-tainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.
2.1 CORE FLOW UNCERTAINTY 2.1.1 Core Flow Measurement During Single Loop Operation The jet pump core flow measurement system is calibrated to measure core flow when both sets of jet pumps are in forward flow; total core flow is the sum of the ' indicated loop flows. For single loop operation, however, the inactive jet pumps may be flowing in reverse (backflowing). Therefore, to obtain the total. core flow the measured flow in the backflowing jet pumps must be subtracted fror the measured flow in the active loop. In addition, the jet pump flow coef fic ient is different for reverse flow than for forward flow, and the indicated reverse flow must be modified to account for this dif ference.
2-1 L
NED0-30060 For single loop operation, the total core flow is derived by the following formula:
Total Core Active Loop Inactive Loop
-C Flow =
Indicated Flow . .
Indicated Flow.
where C (=0.95) is defined as the ratio of " Inactive Loop True Flow" to
" Inactive Loop Indicated Flow." " Loop Indicated Flow" is the flow indicated by the jet pump " single-tap" loop flow summers and indicators, which are set to indicate forward flow correctly. The 0.95 factor is the result of a conserva-tive analysis to appropriately modify the single-tap flow coefficient for reverse flow. The expected value of this factor is s0.88. Therefore, if a more exact (or less conservative) core flow measurenent is required, special in-reactor calibration tests would have to be made. Such calibration tests would involve calibration of core support plate 'P versus core flow during two pump operation along the 100% flow control line and during one pump operation along the 100% flow control line. The correct value of C can then be deter-mined based on the core flow derived from the core support plate AP and the loop flow indicator readings.
2.1.2 Core Flow Uncertainty Analysis The analysis procedure used to establish the core flow uncertainty for one pump operation is essentially the same as for two pump operation, except for some extensions. The core flow uncertainty analysis is described in Reference 2. The analysis of one pump core flow uncertainty is summarized below.
l l For single loop operation, the total core flow can be expressed as follows (refer to Figure 2-1):
W ~w ~w C A I i
2-2
where W = total core flow; W = active 1 P flow; and A
W = inactive loop (true) flow.
7 By applying the "propagat' ion of errors" method to the above equation, the variance esf the total- flow uncertainty can be approximated by:
2 2 I l 32 #2 la h f2 2)
U w"CW +1 *1 # # '
'Y' \al/
1 Wg rand k 1-a
/
l l k
W
. rand C
)
where c """"#t*i"*I f t t*1 * ** f1 "I WC "
og,y, = uncertainty systematic to both loops; egA
= random uncertainty of active loop only; rand cyg = random uncertainty of inactive loop only;
= uncertainty f "C" coefficient; and C
a = ratio of inactive loop flow (Wy) to active loop flow (WA)*
Resulting from an uncertainty analysis, the conservative, bounding values of 7W sys' W * " are 1.6%, 2.6%, 3.5%, and 2.8%, respectively.
I rand C Arand Based on the above uncertainties and a bounding value of 0.38 for "a",
the variance of the total flow uncertainty is approximatly:
2-3 C
. ~ - - - . ._ - - _ _ - _ _ . . _ _ . _
.. s NEDO-30060
- + + (* +(* "(*I C"(*
(l-0 38) (1 0 8 / - -
'When the effect of 4.1% core bypass flow uncertainty at 12% (bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncertainty is:
' active
" ( +
10 2 coolant which is less than the 6% core flow uncertainty assumed in the statistical f
analysis.
In summary, core flow during one pump operation is determined in a con-servative way, and its uncertainty has been conservatively evaluated.
J .
2.2 TIP READING UNCERTAINTY. '
'To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR, The test was performed r
at a power level 59.3% of rated with a single recirculation pump in operation (core flow 46.3% of rated). A rotationally symmetric control rod pattern existed prior to the test.
Five consecutive traverses were made with each of five TIP machines, giving a total of 25 traverses. Analysis of this data resulted in a nodal TIP noise of 2.85%. Use of this TIP noise value as a component of the process
> computer total uncertainty results in a one-sigma process computer total uncertainty value for single loop operation of 9.1% for reload cores.
2-4
_U
P
- NEDO-30060 N CORE a
/ L 1
J i C W g A
WC = TOTAL CORE FLOW ACTIVE LOOP Flee , j W, =
l 1
1 Ffgure 2-1. Illustration of Single Recirculation Loop Operation Flows j
\
2-5/2-6 m
I
\ u.
NED0-30060
- 3. MCPR OPERATING LIMIT 3.1 CORE-WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which is approximately 20% to 40% below that which is attainable for two pump.
operation. Therefore. the consequences of abnormal operational transients from one loop operation will be considerably less severe than those analyzed from a two loop operational mode. For pressurization, flow decrease, and cold water increase transients, previously transmitted Reload and Final Safety Analysis Report (FSAR) results bound both the thermal and overpressure conse-quences of one loop operation.
Figure 3-1 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be seen, the consequences of the transient during one loop operation are considerably lass severe because of the associated reduction in operating power level.
The consequences of flow decrease transients are also bounded by the full power analysis. A single pump trip from one loop operation is less severe than a two pump trip from full power because.of the reduced initial power level.
Cold water increase transients can result from either pump speed-up or introduction of colder water into the reactor vessel by events such as loss of the feedwater heater. For the former, the flow-adjustment factor, Kg is derived by assuming that both recirculation loops increase speed to the maxi-mum permitted by the M-G set scoop tube position. This condition produces the maximum possible power increase and hence maximum ACPR for transients initiated from less than rated power and flow. When operating with only one recircv'.ation loop, the flow and power increase associated with the increased speed on only one M-G set will be less than that associated with both pumps increasing speed; therefore, the Kf factors derived with the two pump assump-tion a're conservative for single loop operation. For the latter, the loss of 3-1
- 4 NED0-30060 feedwater heater event is generally the most severe cold water increase event with respect to the increase in core power. This event is caused by posit'ive reactivity insertion due to the increased core flow inlet subcooling; there-fore, the event is independent of two pump or one pump operation. The severity of the event is primarily dependent on the initial power level. The higher the initial power level, the greater the CPR change during the tran-sient. Since. the initial power level during one pump operation will be sig-n'ificantly lower, the one pump cold water increase case is conservatively bounded by the full power (two pump) analysis.
From the above discussions, it can be concluded that the transient con-sequences during one loop operation are bounded by those established by previously performed full power analysis. The maximum power level that can be attained in one loop operation is only restricted by the MCPR and over-pressure limits established from a full power analysis.
3.2 ROD WITHDRAWAL ERROR Analysis of the rod withdrawal error (RWE) at rated power is given in the FSAR for the initial core and in cycle-dependent reload supplemental sub-mittals. These analyses are performed to demonstrate that, even if the operator ignores all instrument indications and the alarms which could occur during the course of the transient, the Rod Block Monitor (RBM) system will stop rod withdrawal at a minimum critical power ratio which is higher than the fuel cladding integrity safety limit.
However, one pump operation can result in backflow through 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps. Because of the potential for backflow through the inactive jet pumps, the direct active-loop flow measurement may not indicate actual ,
flow. Therefore, it is necessary to adjust the flow-biased RBM equation for single loop operation to account for the backflow which is expected in the inactive loop jet pucaps, as the active loop jet pump drive flow is increased.
This adjustment accounts for the discrepancy between actual flow and indicated 3-2 J
NEDO-30060 flow in the active loop and preserves the original relationship between rod block and actual effective drive flow when operating with a single recircula-tion loop.
A procedure was developed for correcting the rod block monitor equation
.for single loop operation and is described below:
The two pump rod block monitor equation is:
RB = mW + [RB 100 - m( 00)]
The one pump equation becomes:
RB = mW + [RB 100 - m( 00)] - maw where AW = difference, determined by utility, between two loop and single loop effective drive flow at the same core flow; RB = power at rod block in %;
m = flow reference slope for the RBM; W = drive flow in % of rated; and 1
RB 100 " t p level rod block at 100% flow.
If the rod block monitor setpoint (RB100) is changed, the equation must be recalculated using the new value.
The APRM trip _ settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip setting discussed above.
3-3 11
. c NEDO-30060 The combination of the flow-biased corrections described above and the lower reactor power attainable in single loop operation provides assurance that the MCPR safety limit will not be violated during the postulated RWE.
3.3 OPERATING MCPR LIMIT For single loop operation, the rated condition steady-state MCPR limit is increased by 0.01 to account for the increase in the fuel cladding integrity safety limit (Section 2). At lower flows, the steady-state operating MCPR limit is conservatively established by multiplying the rated flow steady-state limit by the flow adjustment- Kg factor. This ensures that the 99.9% statisti-cal limit requirement is always satisfied for any postulated abnormal opera-tional occurrence.
3-4
. _ . __ .- - d
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NEDO-30060 1160 1140 -
0 1170 -
200 $<
c E
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R ANGE OF E XPECTED + 4 MAXIMUM 1 LOOP POWER OPERATION I
%0 O 20 40 60 80 100 ipo 143 POWER LEVELik NUCLEAR 80tLER RATED)
Figure 3-1. Main Turbine Trip with Bypass Manual Flow Control 3-5/3-6
i s -
NED0-30060
- 4. STABILITY ANALYSIS The least' stable power / flow condition attainable under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This condition may be reached following the trip of both recirculation
-pumps. As shown in Figure 4-1 for a typical BWR, operating along the minimum forced recirculation line with one pump running at minimum speed results in a smaller decay ratio, and is therefore more stable than operation with natural circulation flow only. However, it is slightly less stable than operation with both pumps operating at minimum speed. Un'er d single loop operation, the flow control should be in master manual, since control oscillations may occur in the recirculation flow control system under single loop conditions.
a C
l 4-1 i
l i
t.
W
NEDO-30060 1.0
- == == == SINGLE LOOP, PUMP MINIVVV $PE E O
- ===== 80Tw LOOPS PUMPS MINivuv SPE E D S&
C z
4 NATURAL R ATED FLOW-
$ CIR CUL ATION ! CONTROL
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LIN E / LINE
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// ATT AIN AB LE FOR $1NGLE
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, POWEH '%I Figure 4-1. Typical Decay Ratio versus Power Curve for Two-Loop and Single Loop Operation 4-2 J
- 5. ACCIDENT ANALYSES The broad spectrum of postulated accidents is covered by six categories of design basis events. These events are the loss-of-coolant, recirculation pump . seizure, control rod drop, main steam line break, and refueling and fuel assembly loading accidents. The analytical results for loss-of-coolant and re:irculation pump seizure accidents with one recirculation pump operating are given below. The results of the two loop analysis for the last four events conservatively bound those for one pump operation.
5.1 LOSS-OF-COOLANT ANALYSIS A single loop operation analysis utilizing the models and assurptions documented in References-1 and 3 was performed for Vermont Yankee. Using this method. SAFE /REFLOOD code evaluations were made for a full spectrum of break sizes for both the suction and discharge side breaks. The core reflooding minus core uncovery time for the single loop analysis is similar to the twe loop analysis. The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) curves currently applied to Vermont Yankee were modified by derived reduction factors for use during one recirculation pump operation.
5.1.1 Break-Spectrum Analysis A break-spectrum analysis for single loop operation was performed for Vermont Yankee using the model, assumptions and procedure documented in Reference 3. This analysis is necessary because :he coastdown flow from the unbroken recirculation loop (which would occur during two loop operation) would not be available during a postulated large break LOCA in the active recirculation loop during single loop operation. This could cause an earlier boiling transition to occur in the core which would increase 'the calculated Peak Cladding Temperature (PCT). Thus, single loop LOCA calculations are performed assuming a bounding (early) boiling transition time (Reference 3) which leads to revised MAPLHGR limits.
5-1 I
L -- l
WED0-30060 The suction break spectrum core uncovery and reflooding times for single ioop operation are compared to the previously performed two loop operation in Figures 5-1 and 5-2, respectively. The total core uncovered time (reflooding time minus uncovery time) for the suction break is compared in Figure 5-3.
The most limiting break size for both two loop and single loop operation is the 100% DBA break of the discharge line. The break spectrums depicting both single loop and two loop core uncovery and reflooding times corresponding to the discharge line break are shown in Figures 5-4 and 5-5, respectively.
The total core uncovered time is shown in Figure 5-6 for the discharge break spectrums.
5.1.2 Single Loop MAPLHCR Determination The small dif ferences in uncovered time and reflooding time f or the limit-ing break size would result in a small difference in the calculated peak cladding temperature. Therefore, as noted in Reference 3, the one and two loop SAFE /REFLOOD results can be . considered similar and the generic alterna-tive procedure described in Section II.A.7.4 of this reference was used to calculate the MAPLHGR reduction factors for single loop operation.
J MAPLHGR reduction factors were determined for the 100% DBA discharge break. The most limiting reduction factors for each fuel type are shown in Table 5-1. One loop operation MAPLHCR values-are derived by multiplying the current two loop operation MAPLHGR values by the reduction factor for that feel type. As discussed in Reference 3, single recirculation loop MAPLHCR values are conservative when calculated in this manner.
5.1.3 Small-Break Peak Cladding Temperature Section II.A.7.4.4.2 of Reference 3 discusses the small sensitivity of the calculated Peak Cladding temperature (PCT) to the duration of nucleate b,iling and to the assumptions used in the one pump operation analysis.
While there is a potential for a slight increase (s50*F) in PCT, this increase 5-2
I' i .
NEDO-30060 is more than of fset by the decreased MAPLHGR (equivalent to 300* to 350'T PCT) for one pump operation. .Therefore, the calculated PCT values for small breaks are significantly below the 2200'F cladding temperature limit specified in 10CFR50.46 and are not limiting.
5.2 ONE PUMP SEIZURE ACCIDENT The pump seizure event is a very mild accident in relation to other accidents such as the Loss-of-Coolant Accident (LOCA). This has been demon-strated by analyses in Reference 2 for the case of two pump operation, and
~
that it is also true for the case of one pump operation is easily veritied by consideration of the two events. In both accidents, the recirculation driving loop flow is lost rapidly. In the case of the pump seizure, stoppage of the pump occurs; for the LOCA, the severance of the line has a similar, but more rapid and severe influence. Following a pump seizure event, natural circulation flow continues, water level is maintained, the core remains sub-merged, and the combined effects provide a continuous core cooling mechanism.
However,' for the LOCA, complete flow stoppage cccurs and the water level decreases as a result of loss-of-coolant, resulting in uncovery of the reactor core and subsequent heatup of the fuel rod claddir.g. In addition, for the pump seizure accident, reactor pressure does not decrease, whereas complete depressurization occurs for the LOCA. Clearly, the increased temperature of' the cladding and reduced reactor pressure for the LOCA both combine to yield a much more severe stress and potential for cladding perforation for the LOCA than for the pump seizure. Therefore, it can be concluded that the sotential effects of the hypothetical pump seizure accident are very conservatively bounded by the effects of a LOCA.
5-3
(- .- -. - _ - _ - . .. -_- _- __ _ _ _ _ .
8 0 NED0-30060 Table 5-1 LIMITING MAPLHGR REDUCTION FACTORS Fuel Type Reduction Factor 8x8 0.83 8x8R 0.83 P8x8R 0.83 5-4
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- 6. REFERENCES
- 1. " General Electric Standard Application for Reactor Fuel," General Electric Company, January 1982 (NEDE-240ll-P-A-4), and " United States Supplement,"
General Electric Company January 1982 (NEDE-240ll-P-A-4-US) .
- 2. " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application," General Electric Company, January 1977 (NEDO-10958-A).
- 3. -"General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K Amendment No. 2 - One Recirculation Loop Out-of-Service," General Electric Company, Revision 1. July 1978 (NEDO-20566-2).
6-1/6-2