ML20040F845
| ML20040F845 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/08/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Vandewalle D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-15-05, TASK-15-5, TASK-RR NUDOCS 8202100370 | |
| Download: ML20040F845 (7) | |
Text
.
February 08, 1982 S) g Docket No. 50-155 s'/
LS05-82 043
/ '
.a
' Q: -
e??-Qs.A
suRu4:e>...EMcKeDDU.d.k..G l' D.4........WB.yp.tc.U.......... BERM............0.0.ru4 f.1.el.c...Gl,Kfnu......
o4re >.. 2/.f../82........2/!/./82.........2/.M/. 8 2........... 2/.f/82.......h?.......... 2/..[/.82......
NRC FORM 318 00-80) NRCM Ouo OFFICIAL RECORD COPY usom mi-meso
I j
Mr. David J.
VandeWalle CC Mr. Paul A. Perry, Secretary U. S. Environmental Protection Consumers Power Company Agency 212 West Michigan Avenue Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN: Regional Radiation Representative Judd L. Bacon, Esquire 230 South Dearborn Street Consumers Power Company Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Peter B. Bloch, Chairman Atomic Safety and Licensing Board Joseph Gallo, Esquire U. S. Nuclear Regulatory Commission Isham, Lincoln & Beale Washington, D. C.
20555 1120 Connecticut Avenue Room 325 Dr. Oscar H. Paris Washington, D. C.
20036 Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Peter W. Steketee, Esquire Washington, D. C.
20555 505 Peoples Building Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Atomic Safety and Licensing Board Alan S. Rosenthal, Esq., Chairman U. S. Nuclear Regulatory Commission Atomic Safety & Licensing Appeal Board Washington, D. C.
20555 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Big Rock Point Nuclear Power Plant ATTN: Mr. C. J. Hartman Mr. John O'Neill, II Plant Superintendent Route 2, Box 44 Charlevoix, Michigan 49720 Maple City, Michigan 49664 Chri sta-Maria l
Charlevoix Public Library Route 2. Box 108C 107 Clinton Street Charlevoix, Michigan 49720 Charlevoix, Michigan William J. Scanlon, Esquire Chairman 2034 Pauline Boulevard County Board of Supervisors Ann Arbor, Michigan 4B103 Charlevoix County Charlevoix, Michigan 49720 Resident Inspector Big Rock Point Plant Office of the Governor (2) c/o U.S. NRC Room 1 - Capitol Building RR #3, Box 600 Lansing, Michigan 48913 Charlevoix, Michigan 49720 Herbert Semmel Mr. Jim E. Mills Counsel for Christa Maria, et al.
Route 2, Box'10BC Urban Law Institute Charlevoix, Michigan 49720 Antioch School of Law 263316th Street, NW Washington, D. C.
20460
I
~-
Mr. David J. VandeWalle cc Dr. John H. Buck Atomic Safety and Licensing Appeal Board a
U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Ms. JoAnn Bier 204 Clinton Street Charlevoix, Michigan 49720 Thomas S. Moore Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
- James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III Office of Inspection and Enforcement 799 Roosevelt Road Glen Ellyn, Illinois 60137 s
O 3
~
BIG ROCK POINT PLANT SEP YOPIC XV-5 EVALUATION LOSS OF NORMAL FEEDWATER FLOW I.
Introduction Loss of normal feedwater flow to the reactor could occur as a result of feed-water regulating system failure, condensate system failure, or feedwater pump failure. On loss of normal feedwater flow, steam drum level will decrease and a low level alarm will annunciate in the control room at 4 inches below normal water level.
Reactor scram will occur when water level reaches 8 inches below normal. The reactor depressurization system will be initiated at a steam drum level of 17 inches below normal water level. Manual operation by the operator will be required to identify the cause of the event and start the standby feedwater pump or the reactor will scram.
In the event of a total loss of feedwater, core cooling can be accomplished with the control rod drive (CRD) cooling water pumps and the main condenser.
One CRD pump is sufficient for decay heat removal.
If the CRD pumps and/or the main condenser is not available, core cooling can be accomplished by the emergency condenser.
Should all of the above components fail or are not available, the reactor de-pressurization system and the core spray system may be used for cooling the reactor core, following safety relief valve operation.
II.
Review Criteria Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction perr.it or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with i
the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety 1
l during normal operations and transient conditions anticipated during the life of the facility.
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-colled 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 uperational occurrence.
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 occurrences.
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 ar.ticipated operational 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 4
o
. IV.
Review Guidelines _
The review is conducted in accordance with SRP 15.2.6.
The evaluation includes review of the analysis for the event and identifica-tion 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.
V.
Evaluation The loss of normal feedwater was evaluated in Reference 1, using the RETRAN-01 version of RETRAN, a one-dimensional transient thermal hydraulic analysis com-puter program.
The results of the analysis indicate that the termination of feedwater flow results in an increase in enthalpy in the system downcomer.
When this higher enthalpy fluid reaches the reactor core, it will cause a rise in core void fraction and a corresponding reduction in core power. The core power drops rapidly to about 45% of initial value and subsequently re-covers to approximately 60% at the time reactor scram occurs on low steam drum l evel.
The results also indicate that the maximum primary pressure peaks at 1350 psia which is well below the maximum acceptable reactor coolant system pressure of 1870 psia and the MCPR for this event is 1.65.
VI. Conclusion As part of the SEP review of Big Rock Plant, the analysis for loss of feed-water event has been evaluated against the criteria of SRP Section 15.2.6.
We conclude, based on our evaluation, that the results of the analysis for the loss of main feedwater flow accident are acceptable.
t
i f
REFERENCES 1.
Letter from R. C. Vincent to D. M. Crutchfield dated July 15, 1981.
Enclosure entitled " Plant Transient Analysis of the Big Rock Point Nuclear Reactor".
1 4
9 4
i t.
I b
f i
1 4
l i
i e
"'D"'h-~**--*..
. _,..~
,e
,.e y
..m
...mm..
--...m.,- - - -,,,,.
r.
- m,.w._-m
-rg w r.-y.c-
..7m,,...~.,
-. =
e
.