ML19256E140
| ML19256E140 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/25/1979 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7910290281 | |
| Download: ML19256E140 (9) | |
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~ i 400 Chestnut Street Tower II October 25, 1979 Director of Nuclear Reactor Regulation Attention:
Mr. Thomas A. Ippolito, Chief Branch No. 3 Division of Operating Re:ctors U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Ippolito:
In the Matter of the
)
Docket No. 50-296 Tennessee Valley Authority
)
This is in response to your request for additional information concerning the reload analyses for Browns Ferry unit 3 cycle 3 operation. Enclosed is the information requested by your September 19, 1979, letter to H. G. Parris.
If you have any further questions, please get in touch with us.
Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Regulation and Safety Enclosure N4 7 9102 9 0 Sl S I O
TVA RESPONSE TO NRC'S
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- 8 ht REQUEST FOR ADDIT ~ I'.AL INFORMATION i3 g3 N
FOR BROWNS FERRY UNIT NO. 3 RELOAD FOR CYCLE 3 Question 1: Your letter dated August 6, 1979 (*JVA BFNP TS 127) transmitted analysis for operatirn of Browns Ferry Unit No. 3 (BF-3) in the third fuel cycle. During the current outage, we under-stand you are modifyira the end-of-cycle recirculation pump trip (RPT) system on Unl: No. 3 to be the same as the RPT systems on Units Nos. 1 and 2.
The transient analyses in your August 6, 1979 submittal, which include credit for the RPT system in establihsing operating limits, has been performed with the REDY code. During the past year, there have been a number of discussions and correspondence between the TVA and NRC staffs on whether the REDY Code vs. ODYN Code is the better predictor of plant behavior as transient severity is reduced (by the RPT system). Our position - with which we believe you concur - is that the ODYN Code, which uses a more physically correct model of the plant is probably a better predictor of changes in critical power ratios.
During the previous Unit No. 3 reload submittal, as well as the subsequent reload submittals on Units Nos.1 and 2, we had requested an ODYN analysis of the limiting pressurization transients to establish operating limit minimum critical power ratios (OLHCPRs). This subject has been discussed at length in our safety evaluations supporting the most receat reload amendments on Units Nos. 1, 2, and 3.
We would still prefer and are requesting, for the current Unit No. 3 reload a reanalysis of the load rejection without bypass (LRWOBP) and the feedwater controller failure transients with the proposed licensing basis ODYN Code, as applied in the letter from E. D. Fuller, General Electric Company, to D. F. Ross, NRC, dated June 26, 1979, " Impact of CDYN Transient Model on P.5 ant Operating Limits". Your previous position has been that you would perform any reload analysis with either REDY or ODYN - but not both - because of the time and cost required for duplicate analysis. You also raised the question of the acceptability of ODYN analyses until such time as the ODYN Code is approved by NRC. On the most recent reload amendments for Units Nos. 1, 2, and 3 we have resolved this issue by adding a margin to the OLMCPRs calculated by the REDY Code to account for possible lack of conservatism at the end of the fuel cycle when transient effects are most severe. You are requested to provide an ODYN analysis of the limiting pressurization transients as discussed above.
If.
you do not propose to provide these ODYN analyses, explain the basis for your position and propose appropriate margins to the OLMCPR's with justification therefore.
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. Response 1: Until NRC's review of ODYN is complete, TVA's. transient analyses will be bcsed on 7.he accepted REDY code.
TVA acknowledges that so.e uncertainty may exist in the ability of REDY to accurately predict pressurization rate in all cases. However, overall, we believe that REDY provides a conservative cal-culation for the current licansing basis transients. Never-theless, in order to account for any possible nonconservatisms of REDY, TVA ommits to implement a ACPR penalty of.03 for the pressurization events for E0C-2000 mwd /t.
The unit 3 cycle 3 operating limit MCPR's should be revised as follows:
8x8 fuel; BOC 3 through EOC 3: 1.28 8x8R fuel; BOC 3 through EOC 3-2000 mwd /T: 1.22 EOC 3-2000 mwd /t through EOC 3:
1.25 P8x8R fuel; BOC 3 through E0C 3-2000 MRd/t:
1.23 E0C 3-2000 mwd /t through EOC 3:
1.26 puestion 2:
T'..e proposed Technical Specifications for Cycle 3 include a change (pg. 75) which would allow BF-3 to operata at up to 85% power with neither RPT operable for an indefinite period of time.
The present Technical Specifications require that the unit be brought to below 30% power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if both RPT systems are inoperable. Provide justifi-cation for this change via a plant and cycle specific analysis of the most severe pressurization transients occurring from 85% power without taking credit for the RPT feature.
Show that the safety limit MCPR will not be violated by any fuel type assuming the respective proposed operating limit MCPRs.
If the proposed specification is intended to be applied to all future cy..es of BF-3 the analysis should bound expected future core characteristics, otherwise cycle specific analyses may be required.
Response 2: Perfor=ance studies of a BWR at reduced power and flow shows that the ACPR for pressurization events (load rejection without bypass, etc.) can be reduced to 60 percent of the value based on the 100 percent power / flow stated when the core power has been reduced to approximately 85 percent by flow control.
These studies were perfor=ed for several transient events including the Load Rejection Without Bypass using a standard computer program and' licensing conservatism. The study did not include Recirculation Pump Trip (RPT but other studies relatJag improvements with RPT have shown that RPT reduces ACPR to approximately 60 percent of that without RPT.
The study was performed at end-of-cycle conditions (i.e.,
all rods out) at other operating conditions (lesser exposures, etc.), where events such as Rod Withdrawal Error which are not affected by the operation of the RPT system would be limiting.
Thus, reducing core power to 85 percent without RPT by flow control would result in t trade-off not requiring MCPR operating limit changes.
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. Additional conservatism will also appear in a -form of higher operating Minimum Critical Power Ratio when the core power is reduced. This additional conservatism is worth approximately 7 percent at the reduced core power and flow described previously.
Thus, should the unlikely pressurization event (Load Rejection Without Bypass) actually occur with RPT inoperative, the safety limit MCPR would not be violated as long as operating power was limited to 85 percent.
Questicn 3: Describe or reference the physics startup test program which will be used for the restart of BF-3 for cycle 3 operation.
Response 3: The startup test program that will be conducted at the beginning of unit 3 cycle 3 operation is the startup program that was presented by BWR utilities to the Reactor Safety Branch of the Division of Operating Reactors in a March 29, 1979, meeting.
This test program has been submitted by Nebraska Public Power District for Cooper Nuclear Station cycle 5 and was approved by the NRC.
These tests are in addition to the technical specification requirements for startup and are attached for your information.
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, 1.
CORE LOADI!;C VERIFICATIO!!
I.
PURPOSE The purpose of this test is to visually verify that the core is loaded as intended.
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DESCRIPTIO!!_
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An undervater television camera or suitabic viewing device will be
. employed to verify both proper orientation and location of each fuel assembly in the reactor core.
At least one independent person cust also either participate in performing the verification or review a vidcotape of the verification prior to startup.
III.
CRITERIA AND ACTIO!!S_
The as-loaded core must conform with the referenced core upon which the licensing analysis was perfor=ed.
Any discrepancies discovered in the loading vill be pro =ptly corrected and the affected areas re-verified to be properly loaded prior to startup.
Conf omance to the ref erence loading vill be de=onstrated by a permanent core serial nu=ber =ap, and document.ed by the signatures of the verifiers.
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2.
CONTROL Ron nPERABII.ITY AND SUBSCRITICAI.ITY Cl:ECi' I.
PURPOSE This test is performed to ensure that no gross local reactivity irregularitics exist and that all operable control rods are f'nctioning properly.
II.
DESCRIPTION The control red cobility test will be performed after the four bundles surrounding the given control rod are leaded. The suberiticality check will be perforced after the core loading has been completed.
'The control red cobility check =ay be performed concurrent with the subcriticality check after core loading has been completed.
Performance of this test will provide assurance that criticality will not occur due to the withdrawal of a single rod. Each control rod in the core will be withdrawn and inserted one at a time to ensure its mobility with drive pressure. Also, the nuclear instrumentation will be monitored during the novement of each control rod to verify sub-criticality.
III.
CRITERIA AND ACTIONS For those control rods that will not cove under drive pressure, appropriate repairs or adjust =e.ts will be made or the red will be declared inoperable as described in the techc cal c re.ifications.
e If criticality were to be achieved by the withdrawal or' a singic control rod, the control od would be inserted and all further rod movements would cease and an investigation would be conducted to determine the cause.
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3.
TIP SICNAL in CERTAINTY TEST ik[5]ik UNu NlL-p.
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PURPOSE i
The purpose of this test is to determine the Traversing In-Core Probe (TIP) system total uncertainty using a statistical analysis.
II. DESCRIPTION Total TIP signal uncertainty consists of geo=ctric and random noise components.
Data to perfor= the analysis is obtained at in t e r=cd ia t e power levels and/or power levels greater than 75*: with the reactor operating at steady state in an octant sym=ctric rod pat tern (if possible).
This data will be additionally used to perform a gross TIP syr=ctry check, which is a comparison of integrated readings fro = sy==ct:ically, located TIP's.
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l III. _ CRITERIA AND ACTIONS The total TIP signal uncertainty (random noise plus geo=ctric A.
.uncertaintics) obtained by averaging these uncertaintics for all
. data sets should be less than 9%.
A cinimum of two or up to six data sets may be used to meet the above criterion.
If the 9% criterion is l
not met and the calculations have been rechecked, the calibration of i
TIP system (e.g. axial alignment) sha'll be checked.
It =ay be necessary to omit data pairs frem the analysis if exact octant sy==etry is not attainable in fuel loading or control rod patterns.
In such cases, offline code predictions of exposure or control rod induced asy==ctry may prove useful in explaining the uncertainty.
B.
The gross check of TIP signal sy= metry should yield a maximu=
deviation between sym=ctrically located pairs of less than 25':.
If the criterion cannot be met, the cause of the asy =etry must be investigated and an explanation attc=pted as per Criterion A.
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.. Question 4: The staff stated in Section 6.2.2 of its safety evaluation of the Generic Reload Fuel Application (which you,have referenced in your reload application) that " Additional data should be submitted by CE to the staff for review, to justify the conservatism of the GEXL correlation for the second and subse-quent cycles of operation of the retrofit 8x8 bundles, when local peaking factors may increase sufficiently to cause non-conservative CPR calculations." Your reload submittal has not addressed this issue. Accordingly, we request you provide either directly or through reference, adequate information which speaks to this concern. Your response should include:
The extent to which individual heater rods are instrumented in steady-state critical power tests for the retrofit fuel design.
For each test bundle provide measured and predicted results in tabular form for the various test conditions.
Provide trend plots (measured :ritical power / predicted critical power vs h1N, G, P, critical power, test bundle)
Maximum R-Factor for each test bundle (new and old R-Factor definitions)
I Thermocouple locations (rod-by-rod axially)
I Spacer-grid locations Provide power and heat flux for all plots of transient CPR cases.
Response 4: This question was asked during the review of the Cooper Nuclear Station Unit 1 Reload 4.
General Electric provided a generic response to this question in Reference 4-1.
This referenced letter supplied additional information similar to that given in the approved GETAB Licensing Topical Reports NED0-10958-A and NEDE-10958-P-A, and also demonstrated that the additive constants used in the GEXL correlation for the 8x8R fuel design was conservatively derived using the methods approved in these two documents.
The NRC Safety Evaluation Report for the Cooper Reload 4 (Reference 4-2) concluded that ".
when viewed over its range of applicability, the 8x8R GEXL correlation (uith new additive constants) has somewhat better precision in predicting 8x8R critical bundle powers than the 7x7 and 8x8 GEXL formulations are for predicting 7x7 and 8x3 critical bundle powers, respectively. Furthermore, from these results it may also be concluded that the 3.6 percent standard deviation and best estimate assumption of the GEXL correlation (which were actually used in the GETAB statistical analysis to derive the 1.07 safety limit MCPR) bound the
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1 statistical characteristics associated with the subject 8x8R GEXL correlation." Therefore, the conservatis: of GETA3 for the 8x8R fuel design based on the previously approved licensing topical report has been recognized by the NRC staff. _ Reference 4-1 Letter, Ronald Engel to Dar sll G. Eisenhut and Robert L. Tedesco, " Additional Information, 8x8R Fuel GETA3 R-Factors," March 30, 1979. Reference 4-2 " Safety Eva;.uation by the Of fice of Nuclear Reactor Regulation s"/ port Amende:nt No. 55 to Facility Licensing No. DPR-46, .ebraska Public Power District, Cooper Nuclear Station, Docket No. 50-298," April 27, 1979. ! 5 312 ...}}