ML19282A821
| ML19282A821 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/06/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Cavanaugh W ARKANSAS POWER & LIGHT CO. |
| References | |
| NUDOCS 7903070135 | |
| Download: ML19282A821 (5) | |
Text
.
e#
4, UNITED STATES j
yg 2*.
NUCLEAR REGULATORY COMMISslON j
,,I 1() j WASHINGTON, D. C. 20555
%,^ p.f
~
j' February 6, 1979
' w,0
....+
Docket No. 50-313 Mr. William Cavanaugh, III Executive Director, G.eneration and Construction Department Arkansas Power & Light Company P. O. Box 551 Little ilock, Arkansas 72203
Dear Mr. Cavanaugh:
We have reviewed your submittal dated November 8,1978, concerning the requested changes to the Arkansas Nuclear One
' Unit 1 Technical Specifications for Cycle 4 operation. Consistant with our telephone conversations with your staff on January 9 and January 18, 1979, enclosed, herewith,.are the questions whicn were discussed during these conversations.
We request that you provide the information to us within 15 days or receipt of this letter.
Sincerely, r). /
-~,
~
./,'u k h'. /.jls liobert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page 790307 6 /3 f
Arkansas Pow;er & Light Company CC:
Phillip K. Lyon, Esquire House, Holns & Jewell 1550 Tower Bu~il ding Little Rock, Arkansas 72201 Mr. Daniel H. Williams Manager, Licensing Arkansas Power & Lig' t Company Post Of fice Box 551 Little Rock, Arkansas 72203 fir. John W. Anderson, Jr.
Plant Superintendent Arkansas f4uclear One Post Office Box 608 Russellville, Arkansas 72801 fir. Thoras F. 'desterman U. S. fluclear Regulatory Commission P. O. Box 2090 Russelville, Arkansas 72801
~
Mr. Robert B. Borsum Babcock & Wilcox t'uclear Power Generation Division Suite 420, 7735 Old Gecrgetown Road Bethesda, Maryland 20014 Arkansas Polytechnic College Russellville, Arkansas 72801
Enclosure Recuest for Additional Infomation Concerning Cycle 4 Operation for Arkansas Nuclear ~0ne - Unit 1 Docket No. 50-313
- 1) The current power di,stribution reliability factor, RF, shown in BAW 10119 is based on comparisons of measured and predicted power distributions of cores utilizing conventional three batch, out-in, fuel management schemes. An in-out-in fuel management scheme has been proposed for cycle 4 Hence the current RF is not, a priori.
applicable to cycle 4 To support use of the current RF,confirma-tory analyses should be proposed. Specifically, a statistical test, such as 'but not necessarily the F test or Bartlett test, and at-ceptance criteria should be proposed which will test the hypothesis that ANO-1 cycle 4 comparisons of measured and predicted power dis-tributions are members of the family of comparisons which fom the data base for the current reliability factor. Such comparisons and statistical testing should be made at at-least month *1y intervals (monthly surveillance is currently required) and a running tally maintained throughout the cycle.
Results of these tests need not be reported if acceptance criteria are met.
2)
Please provide a description of your planned quality assurance program to insure that the proposed reprogramming of control rods to altered bank designations will be successfully perfomed. Reference to specific approved procedures will suffice.
3)
Provide a complete analysis of the moderator dilution accident.
- 4) Provide the computational basis for the revision of the flux / flow setpoint.
Response to items (3) and (4) should include a discussion of methods, models, and numerical results of comparable detail to that which appears in your FSAR.
Extensive use of references to previous sub-mittals, the FSAR, and topical reports as applicable, should be made.
- 5) The ordinate of Figures 8-7, 8-8, and 8-9 should be labelled " Power,
% of Allowable." Please revise them.
- 6) Extensive use of lumped burnable poisons to hold down excess reactivity and tailor power distributions, as has been proposed, is a potentially more difficult problem in a reload core than a fresh cure. This potential problem has been addressed as question (1). An alternate approach is to carefully monitor reactivity anomolies. Please provide a detailed description of your reactivity anomoly check, renomaliza-tion procedures, if any, and review criteria.
It is understood that you employ a review criteria of 1/2%Ap, as distinct from a Technical Specification limit of 1%Ap.
Please confim.
-2 7)
Please provide the predicted maximum batch and maximum assembly
, burnup at end of cycles 4, 5 and 6.
- 8) Table 7-2 of your submittal shows val' ; of the maximum ejected and dropped rod worths for cycle 4 art within the bounds of the reference analysis. Please confirm that associated peaking factors are also bounded and provide the predicted values.
9)
Similarly, please show calculated values of the prompt neutron lifetime,1*,.and delayed. neutron fraction, 8, predicted for cycle 4, and the corresponding values used in the reference safety analysis.
10)
Values df the predicted moderator temperature coefficient are shown in Table 7-2 of your submittal and shown to bound values used in the safety analyses.
It is understood that the safety analyses do not employ values of these coefficients directly, but rather curve fits of reactivity vs density and temperature over a, range of densities and temperatures spanning nominal and upset conditions.
Please confirm that predicted values for cycle 4, over the full range of postulated states, are bounded by the values of reactivity vs density and reactivity vs temperature used in the safety analyses, and provide the cycle 4 predicted values and the values assumed in the safety analyses.
The following four items are related to startup testing.
Response by reference to previous submitted information will suffice.
11)
Please commit to provide a physics startup test report similar to the report for cycle 3.
12)
Section 9.4 of your submittal describes actions to be taken if acceptance criteria are not met.. The Control Rod Group Reactivity Worth Test description is of insufficient detail. Are the planned actions for this test the same as those for cycle 3, as stated in your March 20, 1978 submittal ? If not, please state how they differ and why.
13)
Section 9.4 describes actions to be taken if Acceptance Criteria are not met.
Please state who will perform the " evaluation." Will the on-site safety committee review the evaluation prior to power escalation? What auditable records are to be maintained?
~ 14) Section 9.2.4 describes the ejected control rod reactivity worth test.
It does not address " swap" of symmetric rods with the measured rod as you' did during the cycle 3 startup. Do you intend to perform rod swap tests? If not, please explain why and describe alternate tests which will be performed to validate core symmetry prior to exceeding 5% of rated power.
O O
O e
O W O
w.A s.
a
, g._ '
- s '
e b***
-es,..