View source for LD-91-019, Forwards Response to NRC 910131 Request for Addl Info Re C-E Std SAR - Design Certification

{{Adams
| number = ML20077C710
| issue date = 05/06/1991
| title = Forwards Response to NRC 910131 Request for Addl Info Re C-E Std SAR - Design Certification
| author name = Kennedy E
| author affiliation = ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05200002
| license number = 
| contact person = 
| case reference number = PROJECT-675A
| document report number = LD-91-019, LD-91-19, NUDOCS 9105230164
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 9
}}

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A D ED P%KBE0 AM A OFOWN BOWN May 6, 1991 LD-91-019 Docket No. 52-002         (Project No. 675)
U. S. Nuclear Regulatory Commission Attn:      Document Control Desk Washington, DC 20555

==Subject:==
Response to NRC Requests for Additional Information

==Reference:==
NRC Letter, Reactor Systems Branch RAIs, T. V.
Wambach (NRC) to E. H. Kennedy (C-E), dated January 31, 1991

==Dear Sirs:==

The reference requested additional information for the NRC staff review of the Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC). Enclosure I to this letter provides our responses. Enclosure II contains a proposed revision to CESSAR-DC. A response to RAI 440.35 will be provided separately.
Should you have any questions on the enclosed material, please contact me or Mr. Ritterbusch of my staff at (203) 285-5206.
Very truly yours, COMBUSTION ENGINEERING, INC.
E. H. Kennedy Manager Nuclear Systems Licensing EHK:BF

==Enclosures:==
As Stated cc:   P. Lang (DOE - Germantown)                                                                                      f J. Trotter (EPRI)
T. Wambach (NRC)
ABB Combustion Engineering Nuclear Power
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Enclosure to LD-91-019 RESPONSE TO'NRC REQUESTS FOR ADDITIONAL INFORMATION, REACTOR SYSTEMS BRANCH                           I l
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                             ' Question 440.32 Discuss why an initial top peaked ASI of -0.3 is conservative for the rod withdrawal events since this would appear to result in a more rapid negative reactivity insertion on scram.
Response 440,32 The -0.3 ASI identified in CESSAR-DC as being used in the CEA withdrawal analyses was used as the limiting axial power shape for only DNBR calculations. A bottom peaked axial power shape was used to model scram reactivity int rtion.
For the low-power and full-power CEA withdrawal cases, a
                                          +0.3 ASI power shape was used for scram reactivity insertion.
l Question 440.33 Why is the maximum assumed reactivity rate at the maximum CEA withdrawal rate only 1.5 X 10 4 delta rho /sec compared to 2.5 X 10 4 delta rho /sec for System 80?
Response 440.33 The dffference in the maximum reactivity rate between the System =80+ and System 80 designs is that the System 80 l
design utilizes a-5 bank CEA regulating system whereas System 80+ utilizes a 3 bank CEA regulating system. In addition, the reactivity worth of each bank is different between the two_ designs.                                                 The combination of these differences accounts for the two different reactivity rates.
The System 80+ design results in reduced peaking factors in the core during critical, HZP operating conditions. Please see CESSAR-DC Sections 4.3.2.4 and 4.3.2.5 for additional information.
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Qngstion 440.34 Table 15.4.s-2 gives 0.1 sec. as the time for a drcpped CEA to be fully inserted. Since the event is analyzed from full power, the core should essentially be unrodded and a CEA drop over the entire core would taka several seconds.
Please justify the 0.1 sec, time interval used.
Epsnonse 440.34 It is true that a CEA drop would take several seconds; 1.e.,
the technical specifications state that the individual CEA drop time, from a fully withdrawn position, shall be less than or equal to 4 seconds from electrical power interruption to 90% insertion. However, the objective of the single CEA drop analysis is to calculate the minimum DNBR that occurs during this event.
The methodology used for the. single CEA drop event uses static power peaking factors such that the drop time of the CEA does not impact the results of the DNBR calculation. An assumed 0.1 sec. drop time merely produces a faster initial transient response than would actually be expected. The DNBR calculation is only sensitive to "where" the CEA is dropped in the core as opposed to "how fast" the CEA is dropped in the core. Since one of the initial conditions used to calculate DNBR concerns radial peaking factors, the maximum radial peaking factor is detsrmined by the locat19D of the dropped CEA and this.value in turn is used in the DNBR calculation.
It should be noted that the reactor does not trip for this event and that the minimum DNBR is reached at 105 seconds into the transient, well after the time of CEA drop.

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Question 440.36 Discuss the adequacy of a high neutron flux alarm to indicate a boron dilution event in sufficient time during Modes 3, 4 or 5.
Resppnse 440.36 The high neutron flux alarm is activated whnn the SRM (Source Range Monitoring) ratio exceeds its setpoint.                               The SRM ratio is defined as follows:
source aanse signat u )
SRM ratio   =
Source range signal at start of dilution For Mode 3, 4 and 5 operation, time is calculated from event initiation to loss of shutdown margin. From this time 30 minutes is subtracted to determine the latest allowable time for alarm actuation. In all above modes, it was calculated that at 30 minutes prior to loss of shutdown the SRM ratio will have exceeded its setpoint. Therefore, an operator response time of at least 30 minutes is demonstrated.
i It should be noted that the high neutron flux alarm is used.
when at least one reactor coolant pump is operating during these modes. For cases where the reactor. coolant pumps could be idle (Modes 4 and 5), the reactor makeup water flow alarm would provide indication of any boron dilution event, ensuring the 30 minute operator response time before shutdown margin is lost.
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        'Ouestion 440 22 Standard Review Plan 15.4.6 requires redundancy of alarms that alert the operator to an unplanned boron dilution event. Describe the redundant alarms available in each operating mode.
Response 440.37 The Standard Review Plan recommends that 15 minutes exist in Modes-1 through 5 and 30 minutes in Mode 6 between the time that the operator is made aware of an ongoing boron dilution    j and the time of loss of shutdown margin. However, a 30      l minute interval was used as a goal for Modes 1 through 5 as     l well as the acceptance criterion for Mode 6 for the CESSAR-DC analyses.
The following pre-trip alarms are available for operational Modes 1 and 2: a high power or, under certain conditions, a high pressurizer pressure pre-trip alarm in Mode 1 or a high logarithmic power pre-trip alarm in Mode 2. Furthermore, a high RCS temperature alarm may also occur prior to trip. In operational Modes 3 through 6, either a boron dilution alarm or a reactor makeup water flow alarm will alert the operator I
to an unplanned boron dilution event. In Modes 3, 4 and 5 with the Reactor Coolant System (RCS) full and at least one Reactor Coolant Pump (RCP) operating, a high neutron flux (boron dilution) alarm will provide indication of a boron dilution event. In Modes 4 and 5 with the RCS full and all RCPs idle or for Mode 5 with the RCS partially drained for system maintenance, deboration is prohibited. Therefore, the reactor makeup water flow alarm will provide indication of any boron dilution event. In Mode 6, the boron concentration is at least 2200 ppm before entering this mode and deboration is prohibited. Therefora, the reactor makeup water flow alarm will provide indication of a boron dilution event.
Depending upon the mode of operation, there are a number of alarms available to. alert an operator of a boron dilution event. In addition to the above mentioned alarms, there are also sampling and boronometer indications which wculd provide information in the case of a boron dilution event.
To address EPRI guidance to reduce the number of alarms presented to operators, Combustion Engineering is currently performing a confirmatory analysis to show that 30 minutes is available for operator action time if a boron dilution alarm is used in place of the reactor makeup water flow alarm in modes other than Mode 6. The results of this analysis will in included in a future amendment to CESSAR-DC.                                                           ;

                             ' Question 440.38 The first paregraph describing the results-of the CEA ejection analysis should state that the_ radial averaged fuel enthalpy is less than 280 cal /gm "at the hottest axial location of the hot fuel pin".
Response 440.38 Combustion Engineering agrees, and the phrase "at the hottest axial location of the hot fuel pin" will be added as            -
suggested. The_ full description of the radial averaged fuel enthalpy for the CEA ejection analysis will then read, "The results show that'the radial averaged fuel enthalpy is less than 280 cal /g at the hottest axial location of the hot fuel pin." This correction to the "Results" section of CESSAR DC Section 15.4.8.3 will be incorporated in a future revision of CESSAR-DC.
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Enclosure II to LD-91-019 PROPOSED REVISION TO THE COMBUSTION ENGINEERING STANDARD SAFETY ANALYSIS REPORT - DESIGN CERTIFICATION

C ES S A R l'!a%uum
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: c. Itesultn l                  A opect um of        initial staten were connidered.                                                                                              The cano initiated from hot full power Ull'P) initial conditionn in expected to result                                      in the creatent po t e nt. i a l Ior atinito done connequencen                (i.e., the cane renulting in the largent nusaber of pontulated fuel failuren). The renultu show that,~
the radial averaged Inel enthalpy in lenn than 280 ca l /f.Y                                                                                                            )
The following paragraphn describe thin event,                                                                                                   in detail.~
Refer to Table         15.4.8-2 for the                                                                              i n i t. i a l condit ionn and acuumptionn uned f or thin analyni:..
Table 15.4.0-1 cont.a i nu the nequence of events that occut during a CI:A 1jection initiated from hot full power initial conditionn.
l'iquren 15.4.H-1 through 15.4.8-S nhow the reactor power, heat flux, and clad and iuel temperature'. during ihe nignificant portion of trannient.
1-jection of a CEA caunen the core power t.o increane rapidly due    to   the    almont innlantaneoun addition of                                                                                                           ponitive   I reactivity,       floweeet , the rapid increano in core power in
(.
terminated by a combination of Doppler ic ".>ack and delayed neutron offactn.                  Thin increane in power resultn in a high power trip and ihe reactor power beginn io decreare an the Cl:Au enter the core.                                                       Itoactivity ef f ect '. are shown i n l'igure 15.4.8-6.
In the hot channel, the increane in heat ! lux in such that.
DNH in calculat ed t o occur, t enult ing in:
4
: 1. A   rapid     dectcane                                                                               in  the  nurtace             hi at                irannfer coefficient.
: 2. A rapid decreano in heat ilux.
: 3.    /. rapid incroac.e in clad temperature The rapid increane in clad t errporatu re in quificient to override t~ e decreased nurface heat tiannier coe f f icient ,
resulting in a necond peak in the h o t. channel heat flux.                                                                                                         At thin time the Cl:Au are nearly fully innerted, renul t iny in a rapid reduction in the core power                                                                                      level.           The heat ilux continue <. to decrease ior the remainder of the i rannient .
Initial   p r e n n u r i ;'.o r prenuure in                                                                      1900 pnia.                The itCS and steam generator prennure for tnin cane in shown in l'i q u r e n 15.4.8-7 through 15.4.8-12.
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Amendment 11 15.4-23                        Augunt 31, 1990
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