ML20205T101
| ML20205T101 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/16/1999 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1920-99-20184, NUDOCS 9904270145 | |
| Download: ML20205T101 (9) | |
Text
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s GPU Nuclear, Inc.
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Route 441 South NUCLEAR Post Office Box 480
%ddletown. TA 17057-0480 Tel717-944 7621 April 16,1999 1920-99-20184 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Ladies and Gentlemen:
I
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating L,icense No. DPR-50 Docket No. 50-289 Additional Information - Technical Specification Change Request No. 279 Core Protection Safety Limit i
This letter provides the additional information as discussed on April 6,1999, regarding TMI-l
/
Technical Specification Change Request No. 279 previously submitted to the NRC in GPU Nuclear letter dated December 3,1998 (1920-98-20669).
If any additional information is needed, please contact Mr. David J. Distel, Nuclear Licensing and Regulatory Affairs at (973) 316-7955.
Sincerely, i
James W. Langen ach Vice President and Director, TMI
/DJD Attachment cc: Administrator, Region I TMI-l Senior Project Manager
(,15 TMI-l Senior Resident Inspector File No. 98195 9904270145 990416
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PDR ADOCK 05000289 P
PDR l.
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- Attachment
~. g 1920-99-20184
. Page.1 of 8 ATTACHMENT 1.
Confirm that the instrument uncertainties have been accounted for in the conservative -
direction for the following parameters:
Loss of Coolant Flow - initial RCS pressure LOCA - low RCS pressure reactor trip setpoint, high pressure injection ESAS setpoint, low pressure injection ESAS setpoint, core flood tank pressure, core flood fank volume, ECCS injection flow rate.
Response
Loss of Coolant Flow The DNB calculation is contervatively performed by applying a pressure penalty of--65 psi to the nominal system pressure. A lower initial pressure will result in a lower initial DNBR.
Since VIPRE-01 code input requires initial core outlet pressure, the --65 psi was used to account for ICS control band, instrument error, and uncertainty associated with the calculation of the delta P between the core outlet and the hot leg measurement. This is consistent with GPUN Topical Report No. 087-A, "TMI-l Core Thermal-Ilydraulic Methodology", using the VIPRE-01 Computer Code, which was reviewed and approved by i
the NRC in SER dated December 19,1996.
LOCA I
The above parameters used in the reanalyses supporting TMI-l 20% tube plugging limits include instrument urtertainty in the conservative direction. GPU Nuclear uses standard industry practices to calculate the instrument string uncertainty, including calibration tolerance, instrument accuracy, temperature effect, power supply effect, drift, test equipment accuracy, and process effects.
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- . Attachment 1920-99-20184 Page.2 of 8 2.
Confirm that the following parameters have a negligible effect on the corresponding analyses:
Loss of Feedwater - core inlet temperature and initial pressure
.l LOCA - value of 133.9E6 lbm/hr total RCS flow instead of 'ae DNB value of133.8E6 lbm/hr.
i Resoonse Loss of Feedwater The core inlet temperature and initial pressure assumed in the Loss of Feedwater (LOFW) analysis were nominal values for normal operation at full power. These are consistent with i
the analyses of the Design Basis Accidents in Chapter 14 of the TMI-1 FSAR. Original non-LOCA analyses on B&W plants were performed using nominal initial pressure, temperature and pressurizer level conditions.
A small change in the core inlet temperature will have an insignificant change in the system -
volumetric expansion and consequently have a negligible effect on peak pressure. A decrease in the initial RCS pressure will result in more time at power before reactor trip. A delayed reactor trip will result in greater energy addition to the reactor coolant. Increased energy addition will result in larger peak pressurizer level and peak RCS pressure, which may not be considered a negligible effect. An increase in the initial RCS pressure will result in a shorter time at power before reactor trip. An earlier reactor trip will result in less energy addition to the reactor coolant. Less energy addition will result in a lower peak pressurizer level and a lower peak RCS pressure.
An overall conservative result is ensured by the conservative application of reactor trip setpoints, safety valve lift setpoints, reactivity coefficients, tripped rod worth, and steam generator mass inventories. A conservative pressure response is ensured because measurement uncertainty is included in the reactor protection system trip setpoint.
LOCA Framatome Technologies, Inc. (FTI) calculated the total reactor coolant system (RCS) flow value (133.9E6 lbm/hr) for the LOCA analysis based on 1.02 times the "RCS Functional Specification". This value is based on a slightly different density for the conversion from gpm to Ibm /hr than used for establishing the total RCS flow value of 133.8E6 lbm/hr used for DNB analysis. This flow difference will have a negligible effect'on the LOCA analyses.
-1
' '. Attachment
~ 1920-99-20184i Page.3 of 8-3.
Confirm that core burnup was selected to yield the most limiting combination of moderator temperature coefficient, void coefficient, Doppler coefficient, power profile, and radial power distribution; and that the most reactive rod was assumed to not insert after the reactor trip in the Loss Of Coolant Flow, Loss Of Feedwater, and Loss Of Electric Power analyses.
Response
j For the Loss of Coolant Flow, Loss of Feedwater, and Loss of Electric Power analyses a moderator temperature coefficient (MTC) of 0.0 was used. This corresponds to a bounding beginning-of-cycle core condition, and is conservative for these events since MTC becomes more negative towards the end-of-cycle. A negative MTC would provide negative reactivity feedback due to the core average temperature increase for these events.
A Doppler coefficient of-1.17 E-5 dk/k/F is used for the Loss of Feedwater and Loss of Electric Power events. A value of 0.0 was conservatively used for the Loss of Coolant Flow event. These values correspond to the bounding beginning-of-cycle (BOC) core condition, and are conservative for these events. The BOC Doppler coefficient represents the most positive value during core life, thus providing less negative reactivity feedback than later points in core life.
Void coefficient is not used for these analyses.
A bounding 1.65 Cosine axial power profile was used in the VIPRE analysis of DNB and is consistent with GPUN Topical Report No. 087-A, "TMI-l Core Thermal-Hydraulic ~
Methodology", using the VIPRE-01 Computer Code, which was reviewed and approved by the NRC in SER dated December 19,1996.
1 A conservative radial power distribution in the hot channel was used in the VIPRE analysis of DNB and is consistent with GPUN Topical Report No. 087-A.
A minimum tripped rod worth (MTRW) of 2.36% dk/k was used for these analyses. This MTRW corresponds to a required minimum shutdown margin of 1% dk/k and provides for the maximum worth stuck rod.
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Attachment 1920-99-20184 Page.4 of 8-L 4.
Provide ajustification for the following assumed values:
a.
initial SG inventory of 39,000 pounds b.
nominal RCS temperature and pressurizer level for LOCA '
a delay of 35 seconds for ECCS injection on the high pressure injection c.
d.
a delay of the greater of 35 seconds after the high pressure injection signal or 10
)
seconds after the low pressure injection signal for ECCS on low pressure injection
- e. -
decay heat of 1979 plus 1-sigma (should use 2 sigma)
- Response a.
The development ofinitial OTSG mass is described in FTI Document 51-5002595-00.
The methodology calculates a total OTSG mass of about 40,700 lbm for CR-3 at an operating range level of 50%. Normal operating range level for CR-3 and TMI-I is 60%. This mass was calculated with 20% SG tube plugging at a power level of 2620 MWt, and RCS T,, of 579"F. Both TMI-l and CR-3 are 177FA-LL plants, and the initial conditions and geometries are sufficiently similar that the CR-3 SG secondary i
steady-state data are applicable to TMI-1. For additional conservatism, a lower value of 39,000 lbm was used in the Loss of Feedwater (LOFW) analysis. This is consistent with the value used by FTI for the CR-3 LOFW re-analysis with 20% SGTP.
l b.
Nominal RCS temperature and pressurizer level for LOCA analyses are utilized as j
specified in the NRC approved FTI Topical Report BAW-10192P-A, Rev. O, June
~1998, "BWNT LOCA Evaluation Model for OTSG Plants".
c.
A delay of 35 seconds for ECCS injection on the High Pressure Injection is consistent with existing TMI-l design basis accident analysis assumptions. The breakdown of the 35 second delay for HPI is as follows (time from reaching the HPI actuation pressure):
Instrument response time 1sec Emergency power source start 10 see HPl pump coastup time 10 sec i
HPI valve response time 24 sec (concurrent with 10 sec pump coastup)
Total ECCS Delay 35 sec
- d.
LPI is started on both the HPI (1600 psig nominal) and LPI (500 psig nominal) actuation signals._ The 20% LOCA analysis assumed the addition of system logic to j
actuate LPI on the later of the HPI signal plus 35 seconds and LPI signal plus 10 seconds. Therefore, the analysis is conservative with respect to the current plant configuration. The analysis was performed in this manner in order to bound the effects of a potential future plant modification to remove the LPI pump stat signal from the HPI actuation. Although, it is noted that this modification is not planned at this time. The new 10 second ECCS delay would be applied in the plant so that it accounts for both instrument response time and LPI pump coastup.
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. Attachment -
.1920-99-20184 Page.5 of 8-The breakdown of the 35 second delay for LPI is as follows (time from reaching the -
' HPI actuation pressure):
Instrument response time -
1 see Emergency power source start 10 see LPI pump coastup time ~
10 sec
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LPI valve response time 24 sec (concurrent with 10 sec pump coastup) 1 Total ECCS Delay 35 sec j
If the LPI pump start signal was removed from the HPI actuation then the ECCS time delays would be as follows:
(time from reaching the HPI actuation pressure):
1 Instrument response time Isec Emergency power source start 10 sec
]
LPI valve response time 24 see Total ECCS Delay 35 sec (time from reaching LPI actuation pressure):
Instrument response time 1 see LPI pump coastup time '
9 see Total ECCS Delay 10 sec e.
In the analysis submitted to the NRC in GPU Nuclear letter dated March 26,1999 (1920-99-20088), the 1979 ANS 5.1 Decay Heat Standard was used with a 1.05 multiplier for the Loss of Feedwater (LOFW) and Station Blackout analyses. This was based on the assumption of decay heat from U "only, together with the 2
conservative assumption of an infinite operating period. This was considered sufficient to account for other fissioning isotopes and for two sigma uncertainty since 2
a significant fraction of fissions occur in Pu ", which generates less decay heat than 2
2 U " and more than offsets the increased decay heat (relative to U ") from the much 23 smaller fraction of fissions occurring in U,
In response to NRC concerns regarding adequacy of the two sigma uncertainty used,
. the Loss of Feedwater (LOFW) analysis was re-analyzed. The re-analysis was performed by specifying the fraction of decay heat from each fissile isotope, U u 2
238 2
U, and Pu ", with the total decay heat being the sum of these three contributions.
The decay heat was assumed to be at equilibrium conditions and the fissile isotope split was obtained by using plant specific values for a typical two-year cycle (U "/ U /Pu "i 0.718/0.072/0.210). In addition, a two sigma uncertainty of 4%,
2 23 2
20%, and 10% was applied to each decay heat curve for U ", U, and Pu 2
23 2n respectively. This _ uncertainty was applied to each isotope contribution before they were summed. The results of the re-analysis were similar to the previous results, and all the acceptance criteria were met. Revised tables for LOFW analysis input values and sequence of events are attached. The Station Blackout'was not re-analyzed since small changes in decay heat will have insignificant effects on the conclusion that adequate natural circulation exists for this event, given the large margins that currently exist.
' Attachment 1920-99-20184 Page.6 of 8
SUMMARY
OF LOSS OF FEEDWATER ANALYSIS INPUT VALUES PARAMETER ANALYSIS VALUE HFP BOC Moderator Temperature 0.0 Coefficient, pcmf'F HFP BOC Doppler Coerricient, pcm/'F
-1.17 HFP Delayed Neutron Fraction, b 0.007 Spray Capacity, gpm 190 PORV Capacity, Ibm /hr/ valve 100,000 PSV Capacity, Ibm /hr/ valve 297,846 PSV Setpoint Drift, %
3 Initial Core Power, MWt 2619.36 (102% of 2568)
Decay Heat model ANS 5.1 1979 (2o uncertainty)
(3 isotopes)
High Flux Trip, percent full power 112 High Flux Trip Delay Time, sec 0.4 High Pressure Trip, psia 2402 i
High Pressure Trip Delay Time, sec 0.6 5
RCSInletTemperature, F 555.7 Initial RCS Pressure, psia 2170 RCS Flow Rate, Mlbm/hr 133.8 Total EFW Flow Rate to both SGs, gpm 550 EFW Temperature,"F 120 l
W L
. '. ' Attachment 1920-99-20184 Page 7 of 8 l
l Sequence of Events LOFW with PORV and Spray Event Tinie, seconds Main feedwater control valve closure initiated 0.0 Main feedwater flow reaches zero 7.0
]
Pressurizer spray on 10.71 RCS high pressure trip setpoint reached 16.98 i
Turbine trip 17.48 PORV lin (first) 17.61 Peak RCS pressure reached (2645.35 psia) 20.0 PSV lift 20.02 OTSG low level setpoint reached 50.22 EFW flowinitiated 93.22 PORV lift (last) 127.85 Peak RCS temperature reached (614.00 F) 255.0 Pressurizer spray off 402.60 j
Peak pressurizer level reached (39.44 n) 403.0 End of transient 800.0 t
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Attachment '
1920-99-20184
, Page 8 of 8 s
Sequenee of Events LOFW without PORY or Spray.
Event Time, seconds Main feedwater control valve closure initiated 0.0 Main feedwater flow reaches zero 7.0 RCS high pressure trip setpoint reached 16.92 Turbine trip 17.42 PSV lift (first).
19.45 Peak RCS pressure (2670.07 psia) 20.0 OTSG low level setpoint reached 50.22
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EFW flowinitiated 93.22 PSV lif1(last) 245.96 Peak RCS temperature (614.03 *F) 246.0 Peak pressurizer level (36.44 ft) 262.0 End of transient 800.0 v:..