ML20039E193
| ML20039E193 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 07/21/1977 |
| From: | Taylor J BABCOCK & WILCOX CO. |
| To: | Boyd R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201060632 | |
| Download: ML20039E193 (36) | |
Text
Mub/l n, n77 7Ev
'Balicock&WilCOX power ceneraan croup P.O. Box 1260, Lynchburg, Va. 24505 Telephone: (804)384 5111 DOCKET STN 50-561 July 21, 1977 Office of Nuclear Reactor Regulation Attention: Mr. Roger S. Boyd, Director Division of Project Management U. S. Nucicar Regulatory Commission Washington, D.C. 20555 Subj ect: B-SAR-205 - Outstanding Issues
Reference:
(1) Report to the Advisory Committee on Reactor Safeguards by the Office of Nuclear Reactor Regulation, U. S.
Nuclear Regulatory Comission, in the matter of Babcock and Wilcox Company Reference Safety Analysis Report B-SAR-205, Docket No. STN 50-561, July 8, 1977 (Draft)
(2)
J. H. Taylor to R. S. Boyd, "B-SAR-205 - Outstanding Issues", July 15, 1977
Dear Mr,
Boyd:
In accordance with the commitment made in Reference (2), B&W is herewith submitting revised B&W positions concerning items 6, 7. 8 and 32 in Section 1.6 of Reference (!).
The attachment previously transmitted with Reference (2) and the attachments to this letter provide a B&W resolution of all 13 open issues listed in Section 1.6 of Reference (1). We hope that our revised or c3arified positions as addressed in these attachments will resolve all of these open iss ces prior to the ACF.S Committee meetings.
The Babcock & Wilcox Company commits to formally including modified material as described in the attachmeat in a future amendment to the B-SAR-205 which is scheduled for submittal prior to August 31, 1977.
fe truly yours, a
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l James H. Taylor Manager, Licensing
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Attachment
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DOCKET STN 50-561 July 21, 1977 Paga 1 of 6 s
ATTACIDfENI (6) Overpressure Protection Staff Position
- t "We will require Babcock & Wilcox to supply additional information-regarding the B-SAR-205 capability to protect the reactor pressure vessel from excessive pressures during low temperature operation such as startup or shutdown. This issue is discussed further in Section 5.2.2 of this report."
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B&W Position B&W will revise B-SAR-205 to demonstrate the capability to protect the reactor pressure vessel from excessive pressures during low temperature operation such as startup or shutdown. The following discussion outlines the program to ca'rry out this demonstration in compliance with the criteria identified by the Staff in Section 5.2.2, page 5-6, items (1) through (5) of the Report to the ACRS 4
dated July 8, 1977.
(1) B&W will revise B-SAR-205 to specify that during shutdown plant operations with one decay heat removal train in operation, all t
four decay heat removal suction valves (two in each train) will remain open. This requirement be included as an interface require-ment in B-SAR-205 and in the final technical specifications. This I
resolves the Staff concern that a single failure (spurious closure) of one of the valve 7 during an overpressure transient could defeat i
the overpressure pt otection provided by the decay heat removal system suction relief valves.
(2) B&W will revise B-SAR-205 to require that during plant heatup operations, the decay heat removal suction isolation valves will remain open until the reactor coolant system temperature' reaches a value such that the Appendix G limit would not be exceeded in the event of an overpressure trainsent, or 305 F, whichever is less.
(3) B&W will amend B-SAR-205 to provide additional information to identify and verify that the appropriate worst case overpressure events have been selected and analyzed.
Items (1), (2) and (3) above will demonstrate that the reactor pressure vessel will be protected from excessive pressures by the decay heat removal suction relief valves during heatup, cooldown, and shutdown operation when the reactor coolant system temperature is 3050F or less.
These relief valves have a setpoint of 455 psig. This information will be included in an amendment to B-SAR-205 to be submitted to the Staff on or about 8/31/77.
- Report to the Advisory Committee on Reactor Safeguards by the Office of
- Nuclear Reactor Regulation, U. S. Nuc1 car Regulatory Commission, in the matter of Babcock and Wilcox Company Reference Safety Analysis Report B-SAR-205, Dockcc No. STN 50-561, July 8, 1977 (Draft)
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DOCKET STN 50-561 July 21, 1977 Page 2 of 6 B&W will also provide, prior to the FSAR, information to demonstrate that the reactor pressure vessel is protected from excessive pressures at temperatures greater than 305 F, when the decay heat removal system isolation valves are closed and overpressure protection is provided by the pressurizer safety valves. ' The pressurizer safety valves have a setpoint of 2500 psig. We will demonstrate that the Appendix G pressure limit is greater than the maximum reactor coolant system pressure at reactor coolant system temperatures above 3050F. This demonstration will be made possible by taking credit of the following factors:
1.
The very low residual element content (copper and phosphorus) of the reactor vessel beltline region materials.
2.
The use of more realistic values for the design neutron fluence. The expected neutron fluence for the end of service condition are lower then the current design values, furthermore, the neutron fluence will be monitored by the reactor vessel surveillance program.
3.
Use of more realistic unirradiated and irradiated material properties such as RTNDI will be employed in the Appendix G analysis. The RTNDT for the end-of-service condition will be predicted using realistic radiatien-induced shif ts in all-ductibility transition temperature.
4.
The Reactor Surveillance Program for applicants referencing B-SAR-205 will employ the compact fracture type specimen, which will be used to directly measure the important material properties such as static and dynamic fracture toughness.
In order to provide backup in the unlikely event that it should not be possible to demonstrate that the Appendix G pressure limit is greater than the pressure permitted by the pressurizer safety valves at temperatures greater than 3050F, B&W will make provisions in the B-SAR-205 design and balance of plant interf ace criteria so as not to preclude the future installation of additional pilot operated relief valve capacity. This additional capacity, should it be required, would be designed in accordance with the Staff's criteria specified in Section 5.2.2, page 5-6, item (1) through (5) of the Report to the ACRS dated July 8, 1977. We note that, even using current methods for calculation of the Appendix G limit, the limit is above 2500 psig at temperatures niet-e 4ees than 3050F for approximately the first 20 years of effective full power operation, and installation of additional relief capacity would not in any event be necessary until that time.
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DOCKET STN 50-561
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July 21, 1977 Pzga 3 cf'6 (7) Provisions for Shutdown q
' t Staff Position "We will require Babcock & Wilcox to provide additional interface specifications for a referencing applicant to assure that such applicant can provide the balance-of-plant capability (1) to achieve and maintain cold shutdown conditions using the decay heat removal system, assuming the use of only safety grade equipment and a singic active failure, and (2) to reach and maintain hot shutdown conditions from normal operating conditions, using only safety grade equipment and assuming the loss of offsite power. This issue is discussed further in Section 5.4.3 of this report."
B&W Position (1) An interface requirement will be included in Section 10.5.9 to require that a sufficient quantity of auxiliary feedwater be provided to reach and maintain hot shutdown for.a sufficient period of time to allow restoration of main feedwater or to initiate cooldown, with only safety grade equipment and assuming loss of off-site power.
(2) In order to reduce the probability of a nalfunction of the decay heat removal heat exchanger bypass valves that could cause an increase in cooldown rate, B&W will change the valve actuator for the bypass valves from a pneumatic to a electric actor operator. These actuators will be designed with' a fail-locked (in position) feature, so that a failure of the power supply to the valve will not result in an unwanted closure of the valye.
(3) As shown in Table 15C-1 of B-SAR-205, the auxiliary supporting systems or functions for the decay heat removal system (low pressure injection system) are the Class 1E power system, the HVAC system, and the component cooling water system. Interface requirements for decay heat removal system power are included in Section 9.3.6.6.1 and require Class IE power which is safety grade and single failure proof. Section 9.3.6.6.19 specifies HVAC (environmental ) require-ments, and these provisions require that this function be safety grade and singic failure proof.
Section 9.3.6.6.8 requires safety grade, single failure proof component cooling water. We believe that these requirements are in compliance with the Staff position.
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DOCKET STN 50-561 July 21 1977 I
P2ge 4 ef 6 8
(8) _(DifRS ISOLATION):
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Staff Position "We require additional information from Babcock & Wilcox to demonstrate that decay heat removal system isolation from the reactor coolant system can be accomplished in the event of a postulated pipe failure in the decay heat removal system outside containment, during shutdown cooling, assuming a single active component failure, in accordance with Standard Review Plan 3.6.1".
B&W Position B&W will demonstrate that decay heat removal system isolation can be accomplished in the event of a postulated pipe failure in the decay heat removal system outside containment, during shutdown cooling, assuming a single active component failure, in accordance with Standard Review Plan 3.6.1.
For the purposes of this analysis, we will assume as a single failure, the postulated failure of an electrical bus.
The design changes necessary to accomplish this will be documented in an Arendment to B-SAR-205 scheduled for submittal on or about
-Jmlr w^ ^. 19 77.
/tagmf 31,1971 llowever, B&W does not consider failure of the electrical bus to be a credible active failure. This component can be designed such that it has no moving parts whose single failure would cause the loss of the bus, and the loss of structural integrity is precluded by definition of a single active component failure given in Appendix A to APCSB 3-1.
The postulation of a passive failure, such as the existence of a foreign object creating a 3-phase short, would be an event independent of' the postulated piping failure and not a required, assumption per the SRP and BTP noted above.
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DOCKET STN 50-561 July.21, 1977 Page 5 of 6 4
i (12) Anticipated Transients i'
Staff Position
" Babcock & Wilcox must demonstrate that, for moderate frecacney events, non-safety grade equipment action is not necessary to meet applicable limits on system pressure and fuel damage. This matter is discussed in Section 15.4 of this report."
B&W Position The use of non-safety grade equipment for anticipated transinets is shown on Table 15.1-4 of B-SAR-205 to be limited to use of the turbine bypass system. The anticipated transients (moderate frequency) so identified in the table are rod group withdrawal at startup and at power, control rod misoperation, chemical and volume control j
system malfunction (Boron Dilution), turbine trip, loss of normal feedwater, excessive heat removal, and inadvertent operation of ths i
ECCS.
For the above listed transients, except turbine trip, a reactor trip
[
is initiated prior to tripping the turbine. The significance of this is that the heat demand during the transient and Lamediately following the reactor trip is independent of the use of the turbine bypass system.
In fact, the heat demand for approximately 2 seconds following turbine trip is the same whether or not the bypass system is assumed to function.
InaMIElon,theheatdema%withoutturbinebypassthantheheatdemand with turbine bypass. Analyses performed with the heat demand simulating no turbine bypass system action showed negligible differences to the analysis presented in B-SAR-205, Chapter 15.
This statement will be summarized in a footnote to Table 15.1-4 which reads " Turbine bypass
' used as part of its normal role, but is not a required function.
Adequate secondary steam pressure relief capacity is available without bypass action through the atmospheric ducp and/or safety valves with negligible effect on transient response (except for turbine trip analysis where failure of the bypass is discussed)."
The turbine trip transient is the only one in the category of a turbine trip prior to reactor trip. The turbine trip is addressed in Reference (2), item 13, with and without turbine bypass.
Section 15.1.7 of B-SAR-205 also discusses turbine trip with and without operation of turbine bypass. The arbitrary use of an initial power Icvel of 112% combined with a pc ;itive coderator coefficient produces DNB results that bound all cases of turbine trip from 102% power, regard-less of the assumptions used for turbine bypass operation.
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DOCKET STN 50-561 July 21. 1977 Pag 2 6 of 6 l
The only anticipated transients listed in Table 15.1-4 that assume turbine trip without turbine bypass action (dis ~ cussed above) are the loss of four pumps, break in primary system penetration lines, and control room uninhabitability. The equipment assumed to function in the event of control room uninhabitability is dependent on the initiating event. A spectrum of postulated events is listed in BSAR Section 15.1.23, and for these events, the systems assumed to operate are separately identified in Table 15.1-4.
The loss of four pumps and the break in primary system penetration lines both cause a reactor trip prior to turbine trip. Therefore, the effect of turbine bypsss is negligible as it was for the bypass action discussed above.
Furthermore, as these two transients are under-cooling in nature, assuming a turbine trip is conservative.
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