ML20246L208
| ML20246L208 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/31/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89532, NUDOCS 8909060279 | |
| Download: ML20246L208 (29) | |
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MM F7 Log. # TXX-89532-
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915 lilELECTRIC August 31, 1989' WHussa J. Cahul, Jr. -
Enecutive Vke President U..S. Nuclear Regulatory Commission Attn:. Document Control Desk Washington, D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 m.
ADVANCE FSAR CHANGE SUBMITTAL REVISION OF RCS FLOW RATE FOR LOSS OF FLOW EVENTS Gentlemen:
The enclosure to this letter provides an advance submittal-of FSAR changes I
related to use of a revised reactor coolant loop flow rate in reanalysis of
.both the Partial and Complete Loss of Forced Reactor Coolant Flow events, along with related supporting documentation. These events were reanalyzed to i
provide more realistic acceptance criteria for use'during the Initial Startup l
Test Program. These changes will be included in a future FSAR amendment l
In order to facilitate NRC staff review of these changes, supporting i
information related to the FSAR changes is organized as follows:
i l
1.
Draft revised FSAR pages, with changed portions indicated by a revision bar in the margin (denoted as " DRAFT"), as they are to appear in a future amendment.
2.
Line-by-line description / justification for each revised FSAR item
]
together with the group and classification designation, as well as 1
an indication of whether the change impacts the SER/SSER.
3.
A copy of related SER/SSER sections.
4.
An index page containing the title " bullets" which consolidate and categorize similar individual FSAR changes by subject and related l
SER section.
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S.
A discusion of each " bullet" which includes:
- The line-by-line description / justification for each FSAR item 1
related to the " bullet" which has been screened as a group 1 or 2 item or a group 3 or 4 item that impacts the existing SER/SSERs.
(The discussion of these groups is contained in TV Electric letter TXX-88467 dated June 1, 1988.)
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8909060279 890831 KN PDR ADOCK 05000445 A
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TXX-89532-
,j' August 31, 1989-Page 2'of 2
- The bold / overstrike version of the revised FSAR pages referenced:
by the description / justification for'each item identified above; The bold overstrike version facilitates review of the revisions by highlighting with a slash (/) the portion of the text that is deleted. 'In. some cases, where the bold / overstrike version is '
unavailable, a hand marked-up version will be provided.;
TU Electric requests that the NRC perform an expedited review of this FSAR'
. change'and inform us as to its acceptability.
Sincerely, it'!
J )
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William J. Cahill, Jr.
By:
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Rog'er(/L Walker Manager Nuclear Licensing RLA/vid Enclosures c -~ Mr. R. D. Martin, Region IV Resident' Inspectors, CPSES (3) m___-__
m._.____-m.
fttach;er}t to TXX-89532 i.
August 31., 1989
, Page 1 cf 27 l
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Enclosure to TXX-89532 August 31,- 1989 Advance FSAR Change Related to Revision of RCS Loop Flow Rate for Loss of Flow Events SUBJECT East Item 1 Draft revised FSAR pages 2
Item 2 Description / Justification for all FSAR changes 12' Item 3 Related SER/SSER pages 14 Item 4 Index Page for Bullets 15 Item 5 Description / Justification for Bullets 16 and Associated Bold / overstrike Pages
F-
'Ittschment to TXX-89532 L
August 31, 1989 CPSES/FSAR
' Page 2 of 27 TABLE'15.3-1 (Sheet 1 of 2)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE-Accident Event Time (seconds)
Four-loop 76 operation 76 Partial loss of forced reactor coolant flow Four loops operating, one pump coasting down Coastdown begins 0.0 Low flow reactor trip 1.3 DRAFT Rods begin to drop 2.3 DRAFT Minimum DNBR occurs 3.6-DRAFT 76 Complete loss of forced reactor coolant flow l 76 All operating pumps 0.0 lose power and begin coasting down Reactor coolant pump 0.0 76 undervoltage trip I
point reached Rods begin to drop 1.5 76 Miniaum DNBR occurs 3.6 DRAFT l
CPSES/FSAR
'/
- Attachi::ht to TXX-89532 August 31. 1989 TABLE 15.0-3 j
Paga 3 of 27
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NOMINAL VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN THE ACCIDENT ANALYSESa Thermal output of hSSS (MWt)
See Table 15.0 2 Core inlet temperature (OF) 559.6 41 Vessel average temperature (OF) 589.2 Reactor Coolant System pressure (psia) 2250 Reactor coolant flow per loop (gpm)b 94,400 70 Steam flow from NSSS (lb/hr) 15,140,000 Steam pressure at steam generator outlet (psia) 1000 Maximum steam moisture content (1) 0.25 Assumed feedwater temperature at steam 440 generator inlet (OF) 2 Average core heat flux (8tu/hr-ft )
189,800 a
Steady state errors discussed in Section 15.0.3 are added to these values to obtain initial conditions for transient analyses.
b 95,700 gpm is assumed for analysis of all Section 15.2 events, DRAFT Steam Generator Tube Rupture Analyses (Section 15.6.3), and Partial and Complete Loss of Forced Reactor Coolant Flow Analyses (Sections 15.3.1 and 15.3.2).
' Attach;ent to TXX-89532 l.
. August 31, 1989 Page 4 of 27 1.2 l
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' Attach =ht to TXX-89532 l
August 31, 1989 Page 6'of 27 1.2 5
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FIGURE 15.3-4 l
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Rt.tachment to TXX-89532 August 31o 1989 Page 8 of 27 I
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FIGURE 15.311 294'5 11.12
' Attach::.ht to TXX-89532 August 31, 1989~
'Page 11 of 27 2.2
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- Attachment to TXX-89532 CPSES FSAR AMENDMENT August 31, 1989 DETAILED DESCRIPTION Page 12 of 27 FSAR Page (as amenced)
Groun Description Table 15.0-3 2
Modifies footnote b (reactor coolant flow per loop) to indicate that the 95,700 gpa value is also used in the analysis of both the Partial and Complete Loss of Forced Reactor Coolant Flow events.
Revision:
The revised flow rate represents the value used in the current Westinghouse methodology for analysis of both the Partial and Complete Loss of Forced Reactor Coolant Flow events.
FSAR Change Request Number: 89-519.1 Commitment Register Number: NL-4079 Related SER Section: 15.2.2 SER/SSER Impact: No Table 15.3-1 2
See Sheet No(s):1 Changes the time sequence of events resulting from reanalysis of the Partial Loss of Forced Reactor Coolant Flow event.
Revision:
These changes reflect the use of as-shipped CPSES reactor coolant pump performance data.
FSAR Change Request Number: 89-519.2 Related SER Section: 15.2.2 SER/SSER Impact: No Table 15.3-1 2
See Sheet No(s):1 Changes the time sequence of eyents resulting from reanalysis of the Complete Loss of Forced Reactor Coolant Flow event.
Revision:
j The change reflects the use of as-shipped CPSES reactor coolant pump performance date.
FSAR Change Request Number: 89-519.4 Related SER Section: none I
SER/SSER Impact: No Figure 15.3-1, 2, 3, 4 2 Revises the figure; depicting certain core dynamic response characteristics for the Partial Loss of Forced Reactor Coolant Flow event.
Revision:
The revised figures represent the current analysis results for the Partial Loss of Forced Reactor Ceolant Flow event.
FSAR Change Request Number: 89 519.3 Related SER Section: 15.2.2 SER/SSER Impact: No l
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I-Attachmint to TXX-89532-August 31. 1989 CPSES FSAR AMENDMENT Page-13 of 27 DETAILED DESCRIPTION FSAR Page (as amended)
Groun Description -
1 15.3 9, 10, 11 2
See Sheet No(s): Figure 15.3-12 Figure Revises the figures depicting certain core dynamic response characteristics for the Complete Loss of Forced Reactor Coolant flow event.
Revision:
The revised figures represent the current analysis results for the Complete Loss of Forced Reactor Coolant Flow event.
FSAR Change Request Number: 89-519.5-Related SER Section: none SER/SSER Impact: No
Attachtsnt to TXX-89532 August 31. 1989
- Page 14 of 27-(5) startup of an inactive reactor coolant pump at an incorrect temperature None of these transients are lietting; the most severe in tems of departure from nucleate boiling ratio and system pressure are the excessive load increase events.
Only slight charges in primary system pressure were calculated, and the departure from nucleate boiling ratio did not fall below 1.4.
The staff 1
finds these results acceptable because thsy do not violate the appropriate limits.
15.2.2 D(creased Coolfng Transients The appifcant has analyzed the following events which produced decreased primary system cooling:
1 i
(1) loss of external electrical load (2) turbine trip (3) inadvertent closure of main steam isolation valves (4) loss of condenser vacuum and other events resulting in turbine trip (5) loss of nonemergency ac power to the station auxiliaries (6) loss of nomal feedwater flow (7) partial loss of forced reactor coolant flow None of these transients are limiting; the most severe in tems of primary syst overpressurization is the turbine tr' p transient, which results in a peak RCS pressure of approximately 2550 psia.
Because this peak pressure is much lower than 115 of the RCS design pressure, the-staff finds these results acceptable.
15.2.3 Increased Core Reactivity Transients 15.2.3.1 Boron Dilution Events The principal means of causing an inadvertent boron dilution are the opening of the primary water makeup control valve and failure of the blend system, either by the controller or mechanical failure.
The chemical volume and control system (CVCS) is designed to limit, even under various postulated failure modes, the dilution rate to values which, will allow sufficient time for automatic or operater response (depending on the mode of awration) to teminate the dilu-tion before the shutdown margin is exhausted.
Tiis dilution rate is indicated by instrumentation.
The applicant has analyzed the boron dilution event for all modes of operation.
Dilution During Refueling Uncontrolled boron dilution cannot occur during the refueling mode because all sources of unborated water are isolated in this mode.
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. August 31. 1989?.
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The:FSAR has been changed to' reflect.results of'recent' analyses of partial.and completelloss.of flow events..
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August 31 1989 Pagt 16'of 27 15.1.2 ' Decreased Coolina Transients l
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Table 15.0-3 2
Modifies footnote b (reactor coolant flow per loop) to indicate that the 95,700 gpm value'is also used in the' analysis of both the Partial and Complete Loss of Forced Reactor Coolant Flow events.
Revision:
The revised flow rate represents the value used in the current Westinghouse methodology for analysis of both the Partial and Complete Loss of Forced Reactor Coolant Flow events.
FSAR Change Request Number: 89,519.1 Commits <ent Register Number: NL-4079 Related SER Section: 15.2.2 SER/SSER Impact: No i
Table 15.3-1 2
See Sheet No(s):1 Changes the time sequence of events resulting from reanalysis of the Partial Loss of Forced Reactor Coolant Flow event.
Revision:
These changes reflect the use of as-shipped CPSES reactor coolant pump performance data.
FSAR Change Request Number: 89-519.2 Related SER Section: 15.2.2 SER/SSER Impact: No Table 15.3-1 2
See Sheet No(s):1 Changes the time sequence of events resulting from reanalysis of the Complete Loss of Forced Reactor Coolant Flow event.
Revision:
The change reflects the use of as-shipped CPSES reactor coolant pump performance data.
FSAR Change Request Number: 89-519.4 Related SER Section: none SER/SSER Impact: No
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Attach;;nt to'TXX-89532 L
.. 'Page 17 of 27' August' 31 1989 x
I Figure 15.3:1. 2. 3. 4. 2 Revises the figures depicting certain core dynamic response characteristics for the Partial Loss of Forced Reactor Coolant Flow event.
Revision:
The revised figures represent the current analysis results for the Partial Loss of Forced Reactor Coolant Flow event.
FSAR Change Request Number: 89-519.3 Related SER Section: 15.2.2 SER/SSER Impact: No i
Figure 15.3:9. 10, 11 2
See Sheet No(s): Figure 15.3 12 Revises the figures depicting certain core dynamic response characteristics for the Complete Loss of Forced Reactor Coolant Flow event.
Revision:
The revised figures represent the current analysis results for the Complete Loss of Forced Reactor Coolant Flow event.
FSAR Change Request Number: 89-519.5 Related SER Section: none SER/SSER Impact: No
' / Attach::it to TXX-89532 CPSES/FSAR
' August 31, 1989 Page 18 of.27 TA8LE 15.3-1 (Sheet 1 of 2)-
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE Accident Event Time (seconds)
Four-loop 76 operation 76 Partial loss of forced reactor coolant flow-Four loops operating, one pump coasting down Coastdown begins 0.0 Low flow reactor trip 1.3 1/5 Rods begin to drop 2.3 2/5 Minimum DN8R occurs 3.6 3/3 76 i
Complete loss of forced reactor coolar.t flow All operating pumps 0.0 76 lose power and begin coasting down Reactor coolant pump 0.0 76 undervoltage trip point reached Rods begin to drop 1.5 76 Minimum DN8R occurs 3.6 3/5 76
1 Attachiint to TXX-89532 CFSES/FSAR I
August 31, 1989 TABLE 15.0-3 Page 19 of 27 MQMIMAL VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN THE ACCIDENT ANALYSESa Thermal output of NSSS (MWt)
See Table 15.0-2 Core inlet temperature (OF) 559.6 41 Vessel average temperature (OF) 589.2 Reactor Coolant System pressure (psia) 2250 Reactor coolant flow per loop (gpm)b 94.400 70 Steam flow from NSSS (lb/hr) 15,140,000 Steam pressure at steam generator outlet (psia) 1000 Maximum steam moisture content (%)
0.25 Assumed feedwater temperature at steam 440 generator inlet (OF) 2 Average core heat flux (Stu/hr-ft )
189,800 a
Steady state errors discussed in Section 15.0.3 are added to these values to obtata initial conditions for transient analyses, b
95.700 gpa is assumed for analysis of all Section 15.2 events. U d 70 Steae Generator Tube Rupture Analyses (Section 15.6.3), and Partial and Complete Loss of Forced Reactor Coolant Flow Analyses (Sections 15.3.1 and 15.3.2).
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