ML20154C546
| ML20154C546 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/11/1988 |
| From: | Eury L CAROLINA POWER & LIGHT CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| NLS-88-119, NUDOCS 8805180217 | |
| Download: ML20154C546 (8) | |
Text
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CaroHna Power & Ught Cr;-7 P. o Ben 1551
- Roegh, N C. 27602 M AY 1 1 1988 LYNN W EURY Semot Vice Prescent OPat. ne se SERIAL:
NLS 119 10CFR50.62 United States Nuclear Regulatory Commission ATTENTION: Dr. Thomas E. Murley Director, Nuclear Reactor Regulation Washin6 ton, D. C.
20555 BRUNSWICK STEAM ELECTRIC PIRIT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR 71 & DPR-62 COMPLIANCE WITH THE TECHNICAL AND SCHEDULE REQUIREMENTS OF THE ATWS RULE
Dear Dr. Murley:
The Nuclear Regulatory Commission issued a Safety Evaluation Report (SER) for the Brunswick Steam Electric Plant Unit 2 on April 8, 1988 regarding recirculation pump trip instrumentation.
In the SER, the Staff concluded that the type of signal conditioning (Rosemount analo o
transmitter / trip units) provided for the Brunswick ATWS design does not meat the diversity requirements of 10CFR50.62 in that diversity, to the extent reasonable and practicable, has not been provided.
The NRC granted a one cycle extension for Unit 2 to allow the Company time to achieve compliance with the An'S Rule.
The Company disagrees with the conclusions of the Staff for the reasons stated in Enclosure 1 and requests reconsideration of the Staff position. An identical system is planned to be installed during the upcoming Unit 1 outage scheduled to commence in November, 1988.
Therefore, the Company requests to delay implementation of the ATWS rule for one additional operating cycle for Unit 1 if this appeal is denied.
Justification for this extension is provided in Enclosure 2.
It is requested that your response to this appeal, and the one cycle extension if necessary, be provided no later than June 15, 1988 to minimize the impact on the upcoming Unit 1 outage.
In accordance with the requirements of 10CFR170.12, a check for $150 is also enclosed.
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8805180217 880511 PDR ADOCK 05000324 P
Documsnt Control Desk NLS 88119 / Page 2 Please refer any questions regarding this submittal to Mr. Stephen D.
Floyd at (919) 836 6901.
Yours very truly, d'3C L. W. Eury LWE/SDF(\\cor\\)
Enclosures cc:
Dr. J. Nelson Grace Mr. W. H. Ruland Mr. E. D. Sylvester NRC Document Control Desk R. F. Janecek (BWR Owner's Group)
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Docum2nt Control Dask NLS 88 119 / Page 4 ENCLOSURE 1 Carolina Power and Light Company believes the design of the ATWS system installed in Unit 2 and planned to be installed in Unit 1 meets the requirements of the ATWS rule (10CFR50.62) for the reasons stated below.
The ATUS rule requires an alternate rod injection (ARI) system that is diverse (from the reactor trip system) from sensor output to the final actuation device.
The Staff has concluded that the Brunswick Plant design does not meet this requirement in that Rosemount analog trip units are used in both the reactor trip system and the ARI system.
Appendix 1 to the April 8, 1988 SER states that an "... acceptable level of component / hardware diversity can be achieved in accordance with combinations of allowable methods such as energization states, AC versus DC power, functional capability, and the use of components from different manufacturers." The Company contends that diversity does exist between the Brunswick Plant reactor trip and ARI systems based on the following:
1.
The reactor trip system de energizes to actuate whereas the ARI system energizes to actuate.
2.
The reactor trip system is DC powered whereas the ARI system is AC powered.
3.
The reactor trip system utilizes a one out of two twice logic scheme whereas the ARI system utilizes a two out of two logic scheme.
4 Rosemount analog trip units are only used in the reactor pressure and reactor level trip channels of the reactor trip system.
The following reactor trip channels are available to trip the reactor and do n21 employ a Rosemount analog trip unit.
- a. High Neutron F1 tx
- b. Turbine Stop Valve Closure
- d. Main Staam Line Isclation
- e. Scram Disenarge Volume High Water Level
- f. Main Steam Line High Radiation
- g. Manual Scram
- h. Peactor System Mode Selector Switch in SHUTDOWN
Document Control Desk NLS 88-119 / Page 5 Therefore, equipment diversity is achieved in the majority of the reactor trip system channels. The NRC concern centers on the potential for a common mode failure of the Rosemount analog trip units such that the reactor pressure and reactor icvel trip channels would not be available in either the RPS or ARI system.
Table 1 (Reference NEDC 30844) shows that even if a common mode failure of the Rosemount analog trip units in the ARI system and the RPS were to occur, a scram signal would be initiated by at least one remaining channel in the RPS that does not utilire Rosemount analog trip units. The common mode failure of concern to the Staff would not prevent a reactor scram because there is equipment diversity to provide a scram signal for the postulated ATVS initiating events, even if the common mode failure were to occur.
As shown above, the ARI system is functionally and logically diverse from all RPS channels.
In addition, as discussed above, the ARI system is equipment diverse from a sufficient portion of the RPS channels to provide a scram signal if a common mode failure of the Rosemount analog trip utits were to occur.
In essence, the NRC is requiring additional equipment diversity.
The Statements of Consideration to the ATWS rule state that "Equipment diversity to the extent reasonable and practicable to minimize the potential for common cause failures is required from the sensors to and including the components used to interrupt control rod power or vent the scram air header." CP&L has concluded that it is not "reasonable and practicable" to provide additional equipment diversity for the following reasons.
1.
The common mode failure of concern to the Staff does not prevent a reactor scram for the postulated ATWS initiating events.
2.
The Company has recently completed a Level 1 PRA for the Brunswick Plant.
The significance of the common mode failure concern has been assessed by determining the change in core damage frequency (CDF) for two cases.
Case 1 assumed an ARI system beta factor (common cause factor, which includes common mode failure) of 0.2 from NEDC 30844 The selection of this factor is conservative since the actual contribution of common mode failure would be less than the total common cause failure potential. Case 2 assumed a beta fsetor for the ARI system of zero (i.e., ali common cause and common mode failure is eliminated). This factor is also conservative since merely providing equipment diversity in the analog trip units would not remove all common mode failure or common cause failure potential.
The combined conservatism of the two cases serves to exaggerate the perceived benefits of removing common mode failure potential by replacing the analog trip units.
Even so, the following results demonstrate the insignificance of removing
Documsnt Control Dask NLS 88-119 / Page 6 the potential for common mode failure of the Rosemount analog trip units.
CASE Gpf Case 1 Beta - 0.2 2.38E 5 Case 2 Beta - 0.0 2.36E 5 The difference between the two cases is 2.0E 7.
Given the negligible improvement in CDF, we do not consider it reasonable to provide additionai equipment diversity.
The fault tree for the above assessment is depicted in Figure 1.
3.
The Company has also eva12ated the cost of providing additional equipment diversity.
It would cost the Company approximately $675,000 per unit for engineering, procurement and installation of an analog trip unit of a different manufacturer in order to reduce the potencial for common mode failure.
Since we believe the ARI system is diverse from the RPS for the reasons stated above, and since additional equipment diversity provides negligible improvement in CDF, we do not believe it is reasonable or practicable to incur the cost of additional diversity.
In summary, the Company believes the ARI system is diverse from the RPS and, therefore, meets the requireaents of 10CFR50.62.
The Staff position that additional equipment diversity is needed does not meet the test of "reasonable and practicable" discussed in the Statements of Consideration to the ATWS Rule because of the high cost incurred in achieving essentially no safety improvement and the fact that the common mode failure of concern would not prevent a reactor scram for the postulated ATVS intitiating events.
4 4
TABLE 1 SENSOR DIVERSITY FOR INITIATING EVENTS SCIAM $T.NSC13 FOR INITIA;;NC IYDC3 Pressure'or Differential Pressure
- Pressure, Neutros Tranesittore Position Flux or and Trip or Micro Switch Radiation Unite Contact Opening Secaore
.1 3
JT e
A*
21 2
4 1
2"
- a g *t 1
4 E
12 t*
4 t
J.
Je di 4"
e3 w-w-
32 32 Si' 3:
's "2 :
=
^
t 33 J
t-tt!
22 5
24 E
2.
Av 45 42e Ev 4
24 Initiating Ivente nsIV cleeure I
I I
I Turb Trip (with bypsee)
I I
I I
Cenerator Trip (with bypese)
I I
I Preneure Regdater Tailure I
I I
I (Primary pressure decrease)
(MSIY closure trip)
Pressure Regulator Ta11ure I
I I
(Primary pressure decrease)
(Level 8 trip)
Pressure Resdatot Ta11ure I
I (Primary pressure increase)
Teedvater now Control I
I I
Ta11ure (Kish reactor water level)
Teedvetor now Coattol I
I I
I Ta11ure (Low reactor water level)
Lees of Condenser Vacuun I
F.
I I
Imee of AC Power (14ee I
I I
I of transforsst) 14ee of AC Power (lose I
I I
I I
I of grid connections)
2.-
.....n.. ----.u
.r.. u. a-- : 2-u. -. _
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=
FIGURE 1 C: (((CE
((CE
RPS ELECTRICAL
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RPS PECHANICAL l
g
. gpg
( cM )
SCRAM C0tt10N CAUSE FAILURE RPS ELECT Arc AR! j l
(CE
on NHAL C0Pt10N CAUSE FAILURE SCRAM APS ELECT Ate ARI i en 1 R
l i
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l C0rt10N CAUSE FAILORE RPS ELECT AND ARI
( CC )
- AR!
CE = 2.0 E-5 RM gCBICAL ASSUME ARI = 2.0 E-3 AA! Fr.LS MS = 5.0 E-2 CM = 1.0 E ~,
CASE 1 BETA = 2.0 E-1 CC = 0.4 E-5 C = 1.02 E-5 CASE 2 APS ELECTRICAL ARI BETA = 0 FAILt$E FAILS CC = 0
( (E )
( ARI )
C = 1.0 E-5
Documsnt Control Desk NIS 88119 / Page 7 ENCLOSURE 2 The ATVS Rule requires justification if the schedule calls for final implementation later than the second refueling outage after July 26, 1984.
A final schadule can then be mutually agreed upon by the Cenaission and licensee. Carolina Power and Light Company believes justification exists for mutually agreeing to an implementation date coincident with startup for Unit 1 Reload 7, currently scheduled for the Sprits of 1990, based on the following:
1.
The Company was not notified of the NRC decision that the ARI system design did not meet the ATVS Rule requirements until April 8, 1988, 2.
The next rcfueling outage for Unit 1 is currently scheduled to commence November 11, 1988. We have estimated it will take approximately 7 months to redesign the system to utilize analog trip units from a different manufacturer, 4 months for final desi6n review and acceptance, and 6 months for procurement.
Inaufficient time remains to accomplish these tasks before the next outage if our appeal for acceptance of the current desiEn is rejected.
3.
The ATWS rule acknowledged the time it takes to design, procure, and install ATWS modifications by allowing two refueling intervals to accomplish the work. The above schedule is consistent with this time interval.
4.
The Company intends to install the as designed system during the next refueling outage.
The as designed system provides substantial compliance with the ATWS rule. As discussed in, the NPC desired changes to the as designed system provide only negligible improvements in safety.
Therefore, allowing CP6L an additional outage to perform the modifications to the ARI system does not result in a significant risk to the public health and safety.
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