ML20093M458
| ML20093M458 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 10/11/1984 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8410220132 | |
| Download: ML20093M458 (8) | |
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.,,'N Commonwealth Edison A
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N) One First National PI'n, Chicago, lltinois t
Address Reply to: Post Offica Box 767
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/ Chicago, lifinois 60690 October 11, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulations U.S. Nuclear Regulatory Commission Washington, D.C.
20005
SUBJECT:
Byron Generating Stations Units 1 and 2, Braidwood Generating Stations Units 1 and 2 Technical Specifications NRC Docket Nos. 50-454, -455, -456, and 457
Reference:
(a) July 19,1984 letter from B.J. Youngblood to D.L. Farrar (b) July 26, 1984 letter from E.D. Swartz to H.R. Denton
Dear Mr. Denton:
This letter provides supplemental responses to NRC questions regarding the proposed technical specifications for Byron Station.
NRC review of these responses is needed to resolve the remaining concerns of the Reactor Systems Branch reviewers.
In reference (b), Commonwealth Edison's responses were provided to the NRC's questions transmitted in reference (a). Questions 1, 4, 10, and-11 raised issues which appear to be generic to Westinghouse PWR's, so we comitted to resolve those issues through the efforts of the Westinghouse Owners' Group (WOG). After review of those responses with the NRC staff, we recognize the need for a comittment to a plant-specific resolution of these issues in the event that the WOG' efforts do not satisfactorily address the NRC's concerns. We propose to do just that. If the WOG is unable to resolve the generic technical specification issues relating to operability of pressurizer relief valves, main steam isolation valves (in mode 4), power range neutron
-monitor flux trips (in modes 3 and 4), and the residual heat removal 8410220132 841011 PDR ADOCK 05000454 A
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4 pump'.(in mode 4), then Comonwealth Edison will undertake to resolve these issues specifically for the Byron and Braidwood plants. This comittment is also made for one ~of the other technical specification-
- issues, operability of the containment pressure high-1 instrumentation in mode 4.
Enclosed with this letter are the supplemental responses to NRC questions 1, 4, 5b, 10, and 11 and the additional containment pressure instrumentation topic. Please address further questions regarding,.his
, matter to this office.
Very truly yours,
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Thomas R.' Tram Nuclear Licensing Administrator cc: NRC Senior Resident Inspector - Byron Lenney 01shan
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l QUESTION 1:
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Belief Valves (Section 3.4.4, page 3/4 4-10)
It is the staff's understanding that your steam generator tube rupture analysis presented in Chapter 15 of yout FSAR telied on the availability and operability of the pressurizet power operated relief valves (PORVs) and the steam generator atmospheric dump valves (ADVs) for deptessutization and cooldown in ordet to limit offsite doses to within 10 CFR 100 guideline values. Similarly, your cooldown evaluation in FSAR Section 5.4.7 performed to show compliance with BTP RSB 5-1 telied on the availability and opetability of the PORVs and ADVs to provide the necessary depressurization and cooldown functions. Your proposed technical specifications however, appear to be
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inconsistent with your FSAR assumptions in that they allow the PORVs to be taken out of service for an indefinite period of time and, on the other hand, they do not contain an operability requirement for the steam generatot ADVS.
Please demonstrate how you comply with the requirements of 10 CFR 50.36 regarding how your technical specifications for the PORVs were derived from the FSAR safety analyses. Specifically, we believe it is necessary to show that the steam generator tube rupture criteria and the RSB 5-1 criteria can be met assuming inopetable PORVs and ADVs consistent with your proposed technical specifications. Otherwise, you should demonstrate that yout technical specification is consistent with the FSAR analyses.
RESPONSE
Concerning the Pressutizer PORVs, revised action statements for Technical Specification 3.7.7 have been made and incorporated into the " Final Draft" of the Technical Specifications.
Commonwealth Edison recognizes the need to address this issue which will be presented to the Westinghouse Owners Group for resolution on a genetic basis.
At the conclusion of this review, Commonwealth Edison will incorporate the m ults/tecommendations of the owners Group as applicable to Byton Station.
In the event that the Owners Group elects not to address this issue genetically, commonwealth Edison will have a review performed and incorporate tto findings of this review as appropriate.
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QUESTION 4:
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Plant Systems. Main Steast Isolation valves 3.7.1.5 (page 3/4 7-9)
The Technical Specifications do not requite manual isolation capability for the Main Steam Isolation valves in mode 4 (below a RCS temperature of 350*F).
Justify that in the event of a steam generator tube rupture in mode 4 that the offsite dose consequences calculated in the PSAR would not be exceeded.
RESPONSE
Commonwealth Edison recognizes the need to address this issue which will be presented to the tiestinghouse Owners Group for resolution on a genetic basis.
At the conclusion of this review, Commonwealth Edison will incorporate the results/tecommendations of the Owners Group as applicable to Byron Station.
In the event that the Owners Group elects not to address this issue genetically, Commonwealth Edison will have a review performed and incorporate the findings of this review as appropriate.
(0563M)
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QUESTION 5b:
Reactor Coolant System, Pressurizet 3.4.3 (page 3/4 4-9)
The Technical Specifications limit the pressurizer level to less than 92% for operation it modes 1, 2 and 3 and impose no limits for operation below mode 3.
Justify that the recommendations of Branch Technical Positions RSB 5-1 (cold shutdown) and RSB 5-2 (TOP) can be met within the above limits in view of the following considerations.
b.
A pressurizet vapor space corresponding to an indicated water level of 25%
is required to permit boration to cold shutdown without letdown.
(0212.154 P.7)
RESPONSE Sb):
In the scenario described in 0212.154 the" plant is assumed to be operating at power when the teactor is tripped and brought to zero load hot standby conditions. At zero percent power, pressutizer level is automatically maintained by the pressucizer level control system at 25% in accordance with the pressurizer level program.
If the pressurizet level deviates from the program level by i 5%, then an annunciator alarms making the operator aware of the pressurizer level deviation so cortective action can be taken. This can include establishing normal or excess letdown, closing the charging flow path valve or stopping the charging pumps.
If the operator were manually controlling pressurizer level, he would strive to duplicate the pressurizer level program (i.e. maintain 25% level at zero percent power).
In'accordance with Byron Station's Plant Shutdown and Cooldown proceduce BGP 100-5, the RCS will be borated to cold shutdown, zenon-free boron 7'
concentration before the cooldown h initiated which is started in Mode 3.
The cooldown can be initiated while boration is in progress if adequate Shutdown Margin is available from xenon for the duration of the boration operation. Upon teaching Hot Standby, the Shutdown Margin will be verified and calculated once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while in Modes 3, 4 or 5 pet Technical Specifications 3.1.1.1 and 3.1.1.2.
Also in Hot Standby the boton concentration required for Cold Shutdown will be calculated and an evaluation of the plant conditions and the availability of equipment and systems that can be used in the shutdown will be made.
As mentioned above, at no load condition the pressurizer level will be maintained at 25% so that pressurizer level can be increased to 95% of span to provide sufficient boron to ccupensate for xenon decay at Hot Standby. To allow some margin for the pressucizer level not being exactly 25% the pressurizer level could be taken to 100% of span without a concern for taking the pressurizer solid. There is a steam volume in the pressurizer above the upper level tap when pressurizer level indication is exactly 100%.
In addition, as the cooldown is started, the pressurizet level will shrink. This also allows additional margin if the level is not exactly 25% of span. In accotdance with Byron Station procedures for a loss of normal letdown the operator will be taking actions to restore normal letdown oc to establish excess letdown as soon as possible.
(0563M) i
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QUESTION 10:
Table 3.3-1, Reactor Trip Instrumentation (page 3/4 3-2)
For tod withdrawal accident at subccitical conditions, staff is under the impression that reactor trip is initiated by t,he power range neutron flux ttip. Howevet, the powet tange neutton flux ttip needs only to be opetable in modes 1 and.2 according to the Technical Specifications. Please explain this apparent discrepancy. If your explanation takes credit for either the intermediate range or source range trips, then the setpoint methodology will have to be amended to reflect this.
RESPONSE
en==nnwealth Edison recoghizes the need to address this issue which will be presented to the Westinghouse Owners Group for resolution on a genetic basis.
At the conclusion of this review, commonwealth Edison will incorporate the results/tecommendations of the owners Group as applicable to Byron Station.
In the event that the owners Group elects not to address this issuo genetically, Commonwealth Edison will have a review performed and incorporate the findings of this review as appropriate.
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r o-QUESTION 11:
Reactor Coolant System Hot Shutdown 3.4.1.3 (page 3/4 4-3)
Technical Specification 3.4.1.3 permits operation in taode 4 with one RHR loop in operation. Justify that the consequences of an inadvettent control tod withdrawal event with one RHR loop in operation in mode 4 would be bounded by the FSAR analysis which assumes two teactor coolant pumps in operation in mode 2.
In yout evaluation considet the effect of non uniform flow distribution through the core on minimum DNBR.
RESPONSE
Commonwealth Edison recognizes the need to address this issue which will be presented to the Westinghouse Owners Group for resolution on a genetic basis.
At the conclusion of this review, commonwealth Edison will incorporate the results/tecommendations of the owners Group as applicable to Byton Station.
In the event that the owners Group elects not to address this issue genetically, commonwealth Edison will have a review pectormed and incorporate the findings of this review as appropriate, t
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Additional Topic: Containment Pressure High-1 in Mode 4 (Table 3.3.3, item IC, page 3/4 3-15)
Concerning the issue of adding Mode 4 to the Applicable Modes column for Containment Pressure-High-1. Commonwealth Edison recognizes the need to address this issue which will be presented to the Westinghouse Owners Group for resolution on a genetic basis. At the conclusion of this review, Commonwealth Edison will incorporate the results/ recommendations of the Owners Group as applicable to Byton Station.
In the event that the owners Group elects not to address this issue genetically, Commonwealth Edison will have a review performed and incorporate the findings of this review as appropriate.
(0563M)
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