ML20040C694
| ML20040C694 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/25/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-03, TASK-15-3, TASK-RR LSO5-82-01-061, LSO5-82-1-61, NUDOCS 8201290159 | |
| Download: ML20040C694 (8) | |
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January 25, 1982 Docket No.50-029)
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Mr. J&mes A. Kay IN JM MI982*
Senior Engineer - Licensing D nrenermx:::ss 'I Yankee Atomic Electric Company N Ef* H 1671 Worcester Road ci s
Framingham, Massachusetts 01701
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Dear Mr. Kay:
SUBJECT:
YANKEE R0WE - SEP TOPIC XV-3, LOSS OF EXTERNAL LOAD. TURBINE TRIP, LOSS OF CONDENSER VACUUM AND STEAM PRESSURE REGULATOR FAILURE By letter dated June 30, 1981, you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes the review of this topic for Yankee Rowe.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility de-sign is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
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Mr. Jar s A. Kay YANKEE R0WE Docket flo. 50-29 Cc Mr. James E. Tribble, President
. Yankee Atomic Electric Coapany 25 Research Drive Westborough, Massachusetts 01581 Greenfield Community College 1 College Drive Greenfield, Massachusetts 01301 Chai rman Board of Selectmen Town of Rowe Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o U.S. NRC Post Office Box 28 Monroe Bridge, Massachusetts 01350 i
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YANKEE R0WE, SEP TOPIC XV-3 EVALUATION LOSS OF EXTERNAL LOAD, TURBINE TRIP, LOSS OF CONDENSER VACUUM, AND STEAM PRESSURE REGULATOR FAILURE I.
Introduction These events result in an unplanned decrease in heat removal by the secondary system. They could lead to excessive pressure in the system if suitable protec-tion is not provided.
The consequences of a loss of external load event are bounded by a turbine trip.
A loss of external load initiates a fast closure of the turbine control valves.
However, a turbine trip initiates a fast closure of the turbine stop valves.
Since-the turbine stop valves close faster than the turbine control valves, the result-ing pressure transient is more severe for a turbine trip.
The Yankee Atomic Electric Company (YAEC) submitted several analyses of the turbine trip event. These are listed in References 1 through 3.
YAEC also sub-mitted an assessment of all the events under SEP topic XV-3 on June 30, 1981 (Reference 4).
II.
Review Criteria Section 50.34 of 10 CFR Part 50 requires that each applicant for a construc-tion permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from' the operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facil i ty.
, Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occur-rences.
GDC 26 " Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated opera-tional occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
III. Related Safety Topics Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
. I V.
Review Guidelines The review is conducted in accordance with SRP 15.2.1,15.2.2,15.2.3, and 15.2.5.
The evaluation includes review of the analysis for the event and iden--
tification of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required. The extent to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
V.
Evaluation At the Yankee Rowe plant, the following components and systems aid in prevent-ing excessive pressure in the primary system:
1.
Steam dump system (steam bypass valve) 2.
Charging volume and control system 4.
Pressurizer spray 5.
Power operated relief valve (PORV) on pressurizer 6.
Main coolant loop, pressurizer, and steam generator safety valves.
The steam generator and pressurizer safety valves are sized to protect the main coolant system and steam generators against overpressure for all load lossesiwithout operation of the steam dump system, pressurizer spray, PORV, automatic rod control, or direct reactor trip following a turbine trip.
In order to demonstrate that the primary coolant system is adequately protected from overpressurization during a complete loss of load transient, the analyses did not take credit for the operation of either the steam dump system or the PORV.
Furthermore, it was assumed that the reactor was not tripped at the time the turbine
. was, but that it was tripped later by a high water level signal from the pressurizer.
l This is the most limiting transient for the XV-3 events _ in this plant.
By exclud-ing the use of the steam dump system, which a loss of condenser vacuum would do, the effects of a loss of condenser vacuum are included in this analysis. This plant does not have a steam pressure regulator so this cannot fail. With these assumptions, the pressura in both the primary and secondary systems starts to rise immediately after the turbine trip. The reactor is not tripped until about twenty seconds after the turbine trip. The pressurizer safety valves open to keep the primary coolant system pressure below 2545 psia, and the steam generator safety valves open to limit the secondary pressure to 1040 psia. The departure from nucleate boiling (DNB) ratio stays above the limit and the' fuel temperatures de-crease during the transient.
The analyses performed for References 1 and 2 were similar. The initial conditions, which were used for these analyses, included:
(a) A 3% allowance for calorimetric error in core power and an additional 4% for instrumentation error.
(b) A 4 degree allowance for error in the measurement of the average core temperature.
(c) A 75 psi allowance for deadband and measurement error in the main coolant pressure.
l-These analyses were performed with the GEMINI-II computer program, which gave the average core and system response, and the COBRA-III-C program, which gave l
the conditions in the hot channel of the core. The NRC in Reference 5 found i
these analyses acceptable.
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. In 1978, the NRC requested YAEC to determine the sensitivity of the conse-quences of a loss of load event to a range of moderator temperature and Doppler reactivity coefficients. The results of this sensitivity study were submitted to the NRC via Reference 3 and approved by the NRC in Reference 6.
VI. Conclusions.
As part of the SEP review for Yankee-Rowe, all of the events under topic XV-3, which could result in an unplanned decrease in heat removal by the secondary system, have been evaluated. We have concluded that the consequences of all of these events in the Yankee-Rowe plant are bounded by those of a turbine trip without steam dump and without an immediate reactor trip. We have also concluded that the con-sequences of this limiting event are in conformance with the criteria of SRP section 15.2.1 - 15.2.5 and therefore acceptable.
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REFERENCES 1.
Yankee Nuclear Power Station Final Safety Analysis Report; Yankee Atomic Eiectric Company, Westboro, Massachusetts; January 3,1974; Volume III, Section 15.2.5 2.
YAEC Letter to Directorate of Licensing, USAEC on Core XI Refueling; March 29,1974.
3.
YAEC Letter to NRC on Additional Information on Core XIV Refueling; November 21, 1978.
4.
YAEC letter to NRC on Systematic Evaluation Program Topic Assessments; June 30,1981.
5.
USAEC; Safety Evaluation by the Directorate of Licensing Supporting Amend-ment No. 9 to Facility License No. DPR-3, Yankee-Rowe Atomic Power Plant; Washington, D. C.; July 30, 1974.
6.
U.S.NRC; Safety Evaluation By The Office of Nuclear Reactor Regulation Supporting Amendment No. 54 to Facility Operating License No. DPR-3, Yankee Nuclear Power Station (Yankee-Rowe); Washington, D. C.; December 6,1978; page 6.
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