ML19296C557
| ML19296C557 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/19/1980 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8002260641 | |
| Download: ML19296C557 (10) | |
Text
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^)sMuD SACRAMENTO MUNICIPAL UTILITY distr CT C 6201 s street, Box 15830. sacramento, California 95813; (916) 452-3211 February 19, 1980 Director of Nuclear Reactor Regulation Attention:
Mr. Robert W. Reid, Chief Operating Reactors, Branch 4 U.S. Nuclear Regulatory Corrmission Washington, D.C. 20555 Docket 50-312 Rancho Seco Nuclear Generating Station, Unit 1 Additional Infomation Cycle 4 Reload
Dear Mr. Reid:
In your letter of February 11, 1980, you recuested additional information on Rancho Seco Unit 1, Cycle 4 Reload.
The answers to your questions are provided in the attachment.
Respectfully submitted, f *. WTW John J. Mattimoe Assistant General Manager and Chief Engineer Attachment
{
3002260
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RANCHO SECO UNIT 1 CYCLE 4 RELOAD kiPORT BAW-1560 1.
Question:
The current power dist ribution rel iabil i ty 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, without analyses, applicable to cycle 4.
To support the use of the current RF, confirmatory analyses should be proposed.
Specifically, a statistical test and acceptance criteria should be proposed which will test the hypothesis that Rancho Seco cycle 4 comparisons of measured and predicted power distributions are members of the family of comparisons which form the data base for the current, reliability factor. Such comparisons and statistical testing shoul'd be made on at least a monthly interval and a running tally maintained throughout the cycle.
Resul ts of these tests need not be reported if acceptance criteria a re me t.
Response: The topical report BAW-10119P, " Power Peaking Nuclear Relia-bility Factors" has been approved by the NRC as an acceptable reference to describe the methods used by B&W to determine the reliability associated with B&W nuclear calculations.
a.
The following is a quote from the SER on BAW-10119P.
"The resul ts of BAW-10119P which include the calculational nuclear relia-bility factor (CNRF) functional dependencies, are acceptable for use in B&W licensing applications for all cycles."
Thus, B&W has approval for use of the functional dependencies shown in BAW-10119P. However, B&W has chosen not to take credit for the functional dependencies but to continue to use the conservative values of 1.075
, for peak hot pellet and 1.05 for radial hot pin in Licensing Rancho Seco Cycle 4.
It is suggested that the NRC consider the functional dependencies of proprietary Figures 1-1 and 1-2 f rom BAW-10119P with a radial hot pin peak of 1.5 and a peak hot pellet of 2.0 as a reminder j ust how conserva-tive the 1.05 and 1.075 values are.
b.
As additional information on why Rancho Seco falls within the "all cycles" of the SER, the following is provided.
The data for the topical included the usage of LBP in the first cycle analysis of TMl-1 and SMUD. The data also included the verification of Oconee 1 Cycle 2 where one-thi rd of the f resh fuel was loaded into the inte rior of the core. Therefore, the data base and the techniques used to quantify the CNRF values leads to the conclusion that the resul ts reported in BAW-10119P are valid for the in-out-in shuf fle scheme for Cycle 4 at Rancho Seco.
c.
Finally, comparisons of measured and predicted power distribution for the first-of-a-kind reload implementation of the in-out-in fuel management schene in an operating B&W reactor have confirmed that the conclusions presented in BAW-10119P apply equally to this in-out-in scheme.
2.
Question:
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 performed.
Response
Rancho Seco Technical Speci fication 4.7.2 is speci fic to verification of proper control rod assignments.
BAW-1560 provides the assignments applicable to cycle 4.
Rancho Seco surveillance procedure SP 208.03, CRD Program Veri fication, is a prerequisiste to the 80C-4 startup. To date, every cycle at Rancho Seco has involved at least one such reprogramming.
. 3 Questica:
De fi ne in more specific terms the APSR pull near EOC and how the stability and control of the core in this mode have been analyzed.
Res ponse : The same techniques used for Cycle 3 were employed for the Cycle 4 analysis of the near End of Cycle APSR removal.
(See the answers to questions 4 and 5 in the reference).
Re fe rence: J. J. Ma t t i moe to Robe r t W. Re i d, "P ro po s e d A me n dme n t No.
60','
November 15, 1978 4.
Question:
Please provide the numerical values for the calculated stability index wi thout APSR's if they differ from Cycle 3 and the reasons for the variance.
Response: The stability index ca'.ulated for Cycle 4 is -0.0404 hr-I.
This was calculated in the same manner as for Cycle 3 and is described in the above reference. The dif ference in stability index between Cycles 3 and 4 is due to differences in xenon worth, doppler coefficient, and the axial burnup distribution between the two cycles.
5 Question:
Please explain why the values for the Average fuel temperature at nominal LHR in Table 4-2 for batches 4 and 5 differ from those reported in the Cycle 3 reload report.
Response
The Cycle 3 calculations assumed densi fication f rom the Lower Tolerance Limit to 96.5% TD, whereas the Cycle 4 calculations assemed densification from nominal densi ty to 96.5% TD.
6.
Question: The nominal linear heat rate, KW/f t at 2772 MWt in Table 4-2 is dif ferent f rom what was listed on page 3-1.
Please revise.
Re s ponse : The value for nominal Linear Heat Rate given on page 3-1 is a core average value and is the average of the individual batch LHR values given in Table 4-2.
No revision is required.
. 7 Question:
Extensive use of lumped burnable poison (LBP) to hold down excess reactivity and tailor power distributions, as has been proposed, is a potentially more dif ficult problem to analyze in a reload core than a fi rst core. This potential problem has been addressed as question (i).
An alternate approach is to carefully monitor reactivity anomolles.
Please provide a detailed description of your reactivity anomoly check, renormal ization procedures, i f any, and review cri teria.
Res ponse :
Reactivity anomoly surveillance is requi red by Rancho Seco Technical Specification 4.9 Compliance with this requirement is satisfied in detail by Surveillance Procedure SP 209.01.
While renormalization is allowed by this specification, i t has never been deemed necessary and/or desirable to do so in the previous three cycles of ope rat ion. The anomoly surveillance procedure requires evaluation and corrective action should the anomoly exceed 80% of the Technical Specification.
8.
Question:
Please provide the predicted maximum batch and maximum assembly burnup at end of cycles 4, 5 and 6.
Response
The predicted maximum batch and maximum assembly burnups (MWD /MTU) are tabulated below, based on 288 EFPD in Cycle 3 Cycle Batch Average Maximum Assembly 4
32300 35100 5
30100 36000 6
33500 37500 9
Question:
Figure 2.1 Identi fy the speci fic credits taken in modifying the power imbalance tent in Cycle 4 f rom that used in Cycle 3 Provide a comparison of actual RPS imbalance limits for Cycles 3 and 4 which would include variations in burnup, controi rod and APSR posi tions and xenon concentration.
. Res ponse: The speci fic credi ts taken in modi fying the power imbalance tent in Cycle 4 from that used in Cycle 3 are due to the dif ferences in the three dimensional power distributions.
If the limiting criteria is Centerline Fuel Mel t (as it generally is for the negative RPS imbalance limit) in both cycles, then the limiting total peak would be the same.
However, due to the di f ferences in three dimensional isotopic distributions and control rod and APSR positions, the actual power shape (and conse-quently imbalance) which generates the limiting peak will change from cycle to cycle. As stated in the response to question 10 in the reference, the imbalance limits are derived f rom a matrix of N300 FLAME cases.
These include variations in burnup, xenon concentration, and control rod and APSR posi tions.
From this matrix of cases, the imbalance corresponding to the limiting power peak is conservatively interpolated. The following table compares the actual RPS limits for Cycles 3 ar,14.
Comparison of Actual RPS Imbalance Limits for Cycles 3 & 4 Cycle 3 Cycle 4
% Power Nea. lmb.
Pos. l mb.
Ne g. l mb.
Pos.Imb.
112
-32.5
+35.8
-37.0
+35.8 100
-45.0
+60.0
-43.0
+61.0 80
-56.8
+83.2
-47.6
+80.0 Note that the changes are relatively minor and in general are due to the change to the LBP shuffle schene.
10.
Question:
Section 9.2.1 critical boron concentration states that the acceptance criteria for this test will be +100 ppm.
Please state the review criteria.
General Comment on the Establishment of " Review Criteria".
For the following reasons, the District does not establish " Review Criteria" for operations at Rancho Seco:
. a.
The test program at Rancho Seco is administered in compliance with the District's quality Assurance Program as specified in 10 CFR 50.34.
This program requires Acceptance Criteria be established on test results important to the design as verification the assumptions in the Safety Analysis are satisfactory. " Review Criteria" is apart of these criteria or commitments.
b.
B&W engineers evaluate all pertinent test data, both on site and at thei r home office. The District's nuclear engineers perform their review and evaluation of test data before approving any escalation of power.
c.
The District's fuel supply / fuel management contract provides for a B&W test engineer to be on-site during the reactor startup.
d.
We believe that this level of review by B&W exceeds that being done by other reactor vendors, and should therefore be considered a more-than-adequate substitute for review criteria.
Res ponse : Whatever the measured versus predicted boron concentration di fference may be, i t will be reviewed.
11.
Question:
Section 9.2.2 Temperature Reactivity Coefficient states an acceptance cri teria of +0.4 x.10-b Ak/k'F.
Please state the review criteria for this test.
Re s ponse :
a.
See " General Comment" above b.
Evaluation to determine the compliance with stated acceptance criteria constitutes a review.
12.
Ques t ion:
Please state what further rod worth tests will be performed if the sum of the measured values for groups 5, 6 and 7 is more than 10%
less than the predicted value for this sum.
. Response: The action to be taken in the event the total measured worth of Groups 5-7 differs from the predicted value by more than 10% is to perform an evaluation consisting of one or more of the following items as appropriate to the situation:
a.
Review of measurement datc and d9ta analysis.
b.
Veri fication that the available shutdown margin based on the measured data satisfies the minimum shutdown margin requirement.
c.
Review of the resul ta of other zero power physics test, d.
Review of calculations ised to obtain the predicted value.
e.
Evaluation of the impact of the discrepancy on safety of operation and on Technical Specifications limits, i f any.
f.
Determination as to whether retest of one or more of the regulating groups would be required.
g.
Determination as to whether measurement of one or more of the safety groups would be required based on considerations of the extent and nature of the discrepancy and of item e above.
If it is determined that measurement of one or more of the safety groups would be required to resolve the discrepancy, such measurements will be pe r fo rme d.
13 Question:
Your description of the ejected control rod reactivity worth test does not state tha t four symmetric control rods will be measured.
As stated in BAW-1477 "0conee 1 Cycle 4 quadrant Flow Tilt" page 12, this test "has proven to be an indicator of core symmetry." Please indicate that the measurement of ejected rod worth of four symmetric loca t ions is part of your test program and state the review criteria.
. Response:
B&W no longer recommends that the symmetric ejected rod worth be part of the startup test program. The following reasons are given:
a.
The incidents and severity of quadrant power til t problems have decreased substantially on BEW reactors since elimination of the cross core shuffle, b.
The symmetric ejected rod test provides only a preliminary indication of quadrant power til t which wculd then be
~ i rmed by powe r d i s t ri-bution data at 40% power. The poaer dis 2 tion test at 40% power is a proper test f rom which to determi
.he core power til t charac-teristics, c.
For the above reasons, the symmetric ejected rod test is not con-sidered necessary, but is optional at the discretion of the utility.
The zero power physics test program is being prepared to allow symmetric ejected rod worths to be measured should condi tions permi t.
When such meas u remen ts a re made, the resul ts will be included in the physics startup test report, see response to your question #16.
14.
Que s t ion: The acceptance criteria stated in Section 9.3.1 Core Power Distribution Verification at N40, 75 and 100% FP with Nominal Control and Position is acceptable.
Please state a review cri teria for these tests.
i t is normally stated as a percentage which the RMS of the detector readings will net exceed.
Response
a.
See " General Cor. ment" Question 10.
b.
Evaluation to determine the compliance with stated acceptance cri teria ccnsti tutes a review.
_9 15 Question:
Please state both acceptance anil review criteria for the critical boron concentration comparion (met:sured vs. predicted) at steady-state full power.
Res ponse :
Critical boron concentration is evaluated at the steady-state full power condition per requirements for reactivity anomoly surveillance.
See response to question 7 for specific critaria.
16.
Question:
Indicate your commitment to submit a physics startup test report wi thin 45 days of completion of the tests.
Respo3se: The results of the startup physics program for Cycle 4 will )e compiled and submitted wi th one of the renthly reports filed within 97 days af ter completion of the physics test program. Startup test data is available for review at the station by NRC personnel prior to the formal submittal.