ML20034C893
| ML20034C893 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/16/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034C892 | List: |
| References | |
| NUDOCS 9005210064 | |
| Download: ML20034C893 (5) | |
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NUCLEAR REGULATORY COMMISSION a
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WASmNGTON, D. C. 206S5
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION-RELATED TO AMENDMENT NO.148-TO FACILITY' OPERATING LICENSE'NO. NPF-31 TOLEDO EDISON COMPANY.
AND THE CLEVELAND ELECTRIC' ILLUMINATING COMPANY DAVIS-BESSE NULLEAR POWER STATION; UNIT NO. I-DOCKET-NO. 50-346
1.0 INTRODUCTION
By' letter dated February 21,L1989, as supplemented by letters' dated July 19 and September 1,1989, Toledo Edison Company requested.an-amendment.to the
'i Davis-Besse Technical: Specifications (TS)..Specifically, the TS change 7 quest t
a is to increase-from.451' seconds to.631 seconds the response' time requirement i
for the "High Flux / Number of Reactor Coolant-Pumps On" trip in Table 3;3-2 of TS 3/4.3.1.1, Reactor Protection System. Instrumentation.. The present response-
'I time of.451 seconds is 'close to the physical limit of the-system,'and the increased response. time may prevent _ unnecessary plant outage time. This proposed change is based on an analysis performed with the VIPRE-01 code. The staff '
evaluation of these changes follows.
.p 2.0 DJSCUSSION Duringpoweroperation,thestatusoftheireactorcoolantpumps(RCP)is.
monitored. The "High Flux / Number of RCP On" trip, or Power / Pumps. trip, circuit utilizes the pump status information to determine the trip setpoint, and initiates a trip signal when the reactor. power exceeds the setpoint. The safety 1 function of the Power / Pumps trip is to provide protection against-1 departure from nucleate boiling (DNB) for loss of forced. reactor-coolant flow-q transientsincluding(1)multipleRCPcoastdowns,(2)singleRCP'coastdowns i
from partial pump-operation, or (3) RCP coastdowns resulting in;the loss of both pumps in either loop.
For a coastdown of a single RCP from four.RCP initial condition, the Power / Pumps trip is'not necessary because the flux / delta
- f. lux / flow-trip ensures that there is sufficient flow and heat removal capability while the reactor is automatically running back in power..
3.0 EVALUATION,
The most limiting transient for the loss of forced reactor coolant flow that relies on the Power / Pumps trip is the four-pump coastdown transient. The-t 9005210064 90051s p
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four-pump coastdown transient had previously been analyzed using' a closed channel thermal-hydraulic code RADAR and the B&W-2 critical heat flux cor-relation. Using a total RPS response time of.620 seconds,-the resulting-calculated minimum DNBR'was-higher than 1.80 compared to the minimum DNBR
. limit of 1.30. _Therefore, no DNB is, anticipated with 95 percent probability at a 95 percent confidence level if the reactor is tripped within.620 seconds of the' time the reactor power exceeding the trip setpoint. Since the RPS response time specified in the TS is defined as the time. interval'from the-l time the monitored parameter exceeds:its trip setpoint at the channel sensor to the power interruption at the control' rod drive-(CRD) breaker (i.e., the:
sum of the sensor and RPS' delay and'CRD breaker delay times), the RPS response time of.451' seconds was obtained by subtracting.125 seconds for the CRD-release delay and.044 seconds for a-dedicated margin for uncertainty from the total response time of.620 seconds used in th'e analysis.
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In support of the proposed response time.of.631 seconds, the four-pump coastdown transient-was reanalyzed using the VIPRE-01 thermal-hydraulic code 1 and the B&W-2 CHF correlation.- Using a total response time of.800 seconds, 1.
the calculated minimum DNBR is 1.885, significantly. higher than the DNBR~ limit 1
of 1,3 for the B&W-2 correlation. Subtracting the same.CRD release delay and uncertainty margin previously used (a total of.169:secontis), the resulting-response time is.631 seconds for-the "High Flux / Number of RCPs On" trip in TS Table 3.3-2.
The VIPRE-01 code is an open-lattice subchannel core thermal-hydraulic code developed by Battelle Pacific Northwest Laboratories under the sponsorship of j
Electric Power Research Institute.- In the open-lattice analysis, the reactor core or fuel bundles is divided into a-number of quasi-one-dimensional channels 1
that communicate laterally.by' diversion crossflow and turbulent mixing.- This approach more realistically considers the flow redistribution effects in the open-lattice core of a pressurized water' reactor.and results in less= severe hot channel thermal hydraulic conditions than that obtained from the closed-channel approach used-in RADAR. The VIPRE-01 code has been approved-for licensing applications with conditions requiring the licensee to document its' intended-use of-VIPRE-01 and justification for its specific modeling assumptions,-
choices of particular models and. correlations, and input values of plant specific data.
In addition, since the DNBR limit of 1.3 for the B&W-2 CHF correlation was developed with another thermal-hydraulic code, it is necessary for the licensee to demonstrate that use of this DNBR limit in VIPRE-01 can predict its data base of the DNB" occurrence with at least a 95 percent probability at a 95 percent confidence level, or to' increase-the DNBR limit.
In using the VIPRE-01-code,.the licensee has developed an 11-channel core model for the thermal-hydraulic analysis of the complete loss of RC flow transient.
The 11-channel model characteristics, including the' geometry, power
~3 distribution, flow and heat transfer correlations, mixing models and CHF-correlation, are sumarized in Appendix A to the February 21, 1989 letter.
Since the B&W-2 correlation with the LYNXT code has been approved with DNBR limit of 1.3, the licensee also provided a VIPRE-01/B&W-2 benchmark against LYNXT/B&W-2. The results of analyses for the locked rotor transient showed a DNBR difference of 0.04 between VIPRE-01/B&W-2 and LYNXT/B&W-2.
A 1
1; At-the staff request, the licensee, in letters'of July 19 and September-1,.
- 1989,'provided sensitivity studies on.VIPRE-01 input parameters, heat transfer correlations solution scheme, radial and' axial.noding, hot channel-location,=andpowerdIstribution. The results show that the 11-channel-model is appropriate for the analysis of the loss of flow transient.. No analysis was performed to show the ap)ropriateness of the VIPRE-01/B&W-2 DNBR limit of 1.30. However; since tie four-pump coastdown transient
- analyzed with VIPRE-01/B&W-2 has a minimum DNBR of 1.89, there is_ ample margin to account'for uncertainty for the DNBR limit of 1,3, and there is reasonable-assurance that DNB will be avoided; This-is-further assured by-the.small difference (0.04) in:the DNBRs. calculated by LYNXT/B&W-2 and VIPRE-01/B&W-2 for the locked rotor transient. Therefore, the staff concludes that the-four-pump (coastdown' analysis using VIPRE-01/B&W-02 -is acceptable to show no DNB occurrence with a total: response time of.800 seconds even though no effort is made to justify the:DNBR limit of 1.30. The licensee has indicated-in the September 1, 1989 letter that this is alone-time specific use of VIPRE-01
= and further justification will be provided to use VIPRE-01-in other applications.
The staff has reviewed the licensee's request for-a TS change to increase the "High Flux / Number of RCP-ON" trip response time from.451 seconds to.631 seconds, and has found that the four-pump coastdown transient analysis using_
j VIPRE-01/B&W-2-and'the-TS change are' acceptable; However : further =
J justif t::ation will be needed to use VIPRE-01/B&W-2 with a:DNBR limit of*1.3 -
i for other applications.
2 4.0 MVIRONMENTALCONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51'.35,-an. environmental. assessment and finding of no significant impact has been pre Federal Register on=May 9,1990(55FR19374)paredandpublishedinthe
' Accordingly, based on the environmental assessment, the Connission has determined that the issuance of this amendment will not.have a significant,effect on the.
quality of the human environment.
5.0 CONCLUSION
The sta#f has concluded, based on the considerations discussed above, that:
(1)tha e is reasonable assurance that the health and> safety of the public will not be endangered by operation in.the proposed manner, and (2).such activities will be conducted in compliance with the Connission's regulations,- and the 9
issuance of this amendment will not be; inimical to the connon defense and l
security. or to the health and safety of the public.
Principal Contributor:
Y. Hsii, NRR/SRXB Dated: May 16, 1990 l
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UNITED STATES NUCLEAR-REGULATORY COMMISSION
.g THE~ CLEVELAND ELECTRICAL ILLUMINATING COMPANY 'ET:AL.
DOCKET NO. 50-346 NOTICE OF ISSUANCE OF AMENDMENT TO~
FACILITY OPERATING LICENSE The U.S. Nuclear Regulatory Comission (Comission) has-issued' Amendment-No.148 to Facility Operating License No. NPF-3, Lissued to Toledo Edison' Company;and Cleveland. Electric Illuminating Company, which revised the Technical Specifications for operation of.the Davis-Besse Nuclear Power Station located?in Ottawa County, Ohio.- The amendment was effective as of thel ate-d l.
of issuance ~.
The amendment modified the Technica1' Specifications to increase the..
allowable response time for.the 'igh. Flux / Number of Reactor Coolant _ Pumps On--
H (power / pumps) trip function of Table 3.3-2 of the Reactor Protection System from'451' milliseconds to 631 milliseconds.-
In addition, a' change li_s made to a footnote of Table 3.3-2 to clarify the. identification of the pump' monitor.
The application for the amendment complies with the standardsLand requirements of the Atomic Energy Act of 1954, as amended (the Act). and 4
the Comission's rules and regulations. The-Comission has;made appropriate t
findings as required by the Act and the Comission's rules and. regulations; in 10 CFR Chapter I, which are set forth in the license amendment, a
Notice of Consideration of Issuance.of Amendments and Opportunity for Hearing in; connection with this action was published in the FEDERAL REGISTER-on October 26-1989(54FR43636).
No request for a hearing or petition for leave to intervene was filed following this notice. '
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The Commission has prepared an Environmental Assessment related to the-'
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- action and-has determined not to prepare an environmental-impact statement..
1 Based upon the environmental assessment, the Commission has concluded that the c
1ssuance of this amendment will not have a significant effect on the quality i
of the human. environment.
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For further details with respect to the action see (1) the application for' amendment dated February 21,.1989, and supplemented-July:19 and September 1,1989,(2) Amendment No.148 to License No,' NPF-3.(3) the Commission's related Safety Evaluation dated May 16,' 1990 -
and.
N (4)theEnvironmentalAssessmentdatedMay 1, 1990 (55 FR'19374). ~All of these items are available for public-inspection at the.' Commission's Public Document
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Room, Gelman Building, 2120 L Street N.W., and at the University of Toledo i
Library, Documents Department, 2801 Bancroft Avenue, Toledo Ohio' 43606. - A 1
copy of items (2), (3) and-(4) may be obtained upon request addressed to-the U.S. Nuclear Regulatory Commission, Washington',. D.C. 20555, Attention: Director.
i Division of Reactor Projects III,'IV, V and Special Projects..
Dated at Rockville, Maryland this 16th day of.
May.
1990.
FOR THE NUCLEAR REGULATORY COMMISSION
{
4 ear A John N. Hannon,' Director Project Directorate.III-3 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear _ Reactor Regulation
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