ML20216C781
| ML20216C781 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/07/1998 |
| From: | Raghavan L NRC (Affiliation Not Assigned) |
| To: | Zeringue O TENNESSEE VALLEY AUTHORITY |
| References | |
| TAC-M99711, TAC-M99712, NUDOCS 9805190407 | |
| Download: ML20216C781 (5) | |
Text
May 7, 1998 Mr. O. J. Zeringue Chief Nuclear Officer And Executive Vice President Tennessee Valley Authority i
6A Lookout Place
{
1101 Market Street Chattanooga, Tannessee 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3: REQUEST FOR ADDITIONAL INFORMATION RELATING TO TECHNICAL SPECIFICATION l
CHANGE NO. TS-384 - POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)
I
Dear Mr. Zeringue:
By letter dated October 1,1997, the Tennessee Valley Authority (TVA) proposed revisions to the Browns Ferry Nuclear Plant Units 2 and 3 Technical Specifications to permit operation of l
the units at the uprated power level of 3458 Mwt. The staff has determined that additional information is required to complete the review of the proposed action. A description of the information required is provided in the Enclosure.
j Please provide the required information no later than June 5,1998.
Sincerely, l
L.kaghavan, Senior Project Manager Project Directorate 11-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-259,50-260, and 50-296
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Enclosure:
Request for Additional Information I
cc w/ encl: See next page DISTRIBUTION:
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%,..... p May 7, 1998 Mr. O. J. Zeringue Chief Nuclear Officer And Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3: REQUEST FOR ADDITIONAL INFORMATION RELATING TO TECHNICAL SPECIFICATION CHANGE NO. TS-384 - POWER UPRATti OPERATION (TAC NOS. M99711 AND M99712)
Dear Mr. Zeringue:
By letter dated Octo'.;i 1,' 1997, the Tennessee Valley Authority ('IVA) proposed revisions to the Browns Ferry Nuclear Plant Units 2 and 3 Technical Specifications to permit operation of the units at the uprated power level of 3458 Mwt. The staff has determined that additional information is required to complete the review of the proposed action. A description of the information required is provided in the Enclosure.
Please provide the required information no later than June 5,1998.
Sincerely, t
H L. Raghavan, Senior Project Manager Project Directorate ll-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-259,50-260, and 50-296
Enclosure:
Request for Additional information cc w/ encl: See next page l
l 1
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i REQUEST FOR ADDITIONAL INFORMATIQB BROWNS FERRY NUCLEAR PLANT. UNITS 2 AND 3 DOCKET NUMBERS 50-260 AND 50-296 1.
Tennessee Valley Authority's (TVA's) submittal for the power uprate amendment did not address several aspects of a design basis loss of coolant accident (LOCA). The U. S.
I Nuclear Regulatory Commission (NRC) staff is of the opinion that the following need to be considered in a design basis LOCA analysis.
(a)
TVA assessed main steam isolation valve (MSIV) leakage with regard to the control room. The same release affects persons at or beyond the site boundary.
Therefore, please address exclusion area boundary (EAB) and low population zone (LPZ) doses due to MSIV leakage.
(b)
NUREG-0737 ltem lil.D.3.4 specifically states that the design basis LOCA source term should be consistent with Appendix A and B of Standard review Plan (SRP) Chapter 15.6.5. The SRP, Appendix B addresses emergency core cooling system (ECCS) leakage. Please provide an assessment of EAB, LPZ, and Control Room doses due to leakage from ECCS outside of primary containment. Otherwise, please provide a technicaljustification why TVA believes these credible release paths need not be addressed.
Please notz mat although the above issues may not be currently addressed in the Browns Ferry Nuclear Plant (BFN) Updated Final safety Analysis Report (UFSAR) accident analyses, the staff believes that they must be addressed to enable the staff to make a finding that public protection is not adversely affected by the proposed power upgrade.
1 2.
While TVA is performing the requested ECCS leakage dose analysis outside the primary containment as requested in item 1 above, to facilitate our continued review, please provie a design basis value for the ECCS leakage (gpm) outside of the primary containment. The staff will multiply this value by 2 as provided for in the SRP.
i 3.
The control room infiltration at BFN is stated to be 3717 cfm. The filtered control room intake fans are rated at 3000 cfm. Does any of the 3000 cfm filtered intake displace any of the 3717 cfm unfiltered infiltration, i.e., is the net intake 717 cfm or 6717 cfm?
I l
4.
The UFSAR describes the control room pressurization fans starting on a primary j
containment isolation signal or on high radiation signal. Please describe which signalis l
l applicable for each of the analyzed accidents; LOCA, main steam line break (MSLB),
rod drop accident (RDA), and fuel handling accident (FHA). If the isolation is based on a high radiation signal, please describe the basis of the setpoint and the expected time delay between the onset of the event and isolation of the control room. There have been recent 50.72 reports in several facilities that describe conditions under which the l
expected monitor alarm setpoints would not be reached for certain analyzed accidents potentially resulting in doses in excess of the General Design Criterion (GDC) 19 limits.
If these data are not readily available, the NRC staff will assume manual operator action after 30 minutes.
l ENCLOSURE i
e
. 5.
The power uprate was specifically requested for Units 2 and 3. TVA calculation ND-Q0031-920075 Rev. 7 (control room doses) explicitly discounts Unit 1 (sheet 7, 8 of 37). Accordingly, the staff's review will be limited to Units 2 and 3. However, the staff will consider the postulated doses to Unit 1 control room operators resulting from Units 2 and 3 design basis accidents. The control habitability issues as they apply to Unit 1 should be resolved prior to its operation. Please inform us of your schedule for revising the TVA calculation ND-00031-920075 Rev. 7 to address Unit 1 issues and providing the missing Unit 1 data for staff review.
6.
TVA analysis ND-Q0031-920075 Rev. 7 (control room doses) indicates on sheet 13 of 38 that the turbine building volume of 2,100,000 cubic feet was used as a dilution volume for MSIV leakage. Data items d and e on page 14.6-29 of the UFSAR imply that dilution credit was taken. The staff does not consider this assumption acceptable in a design basis calculation. The condenser leak could occur at a single location.
Convective forces will cause the release to ascend to the nearest roof vent. These analysis descriptions contradict the information in the referenced General Electric (GE) report NEDC-32091, dated August 1992, which states on page B-14 that "for MSIV calculations, no credit has been taken for the turbine building.. " While the turbine building volume is entered, the purge rate is set to 1E6 %/ day, which establishes a near-instsntaneous release to the environment. The computer input sheets in the GE letter to TVA, dated August 28,1992, indicate that credit was effective!y not taken.
Please update ND-Q0031-920075 Rev. 7 and the UFSAR to be consistent with the supporting calculation, such that the BFN design basis does not credit turbine building dilution for MSIV leakage.
7.
ND-00031-920075 Rev. 7 sheet 12 of 38, states that TVA considers that all piping from the MSIVs to the condenser remain intact even though it is not seismically qualified.
The calculation notes that: "This appears to be in accordance with the Boiling Water Reactor Owners Group (BWROG) position as well as others in industry." No further justification is provided. However, the GE NEDC-31858P report addresses the need to justify integrity of the main steam lines following a seismic event. The NRC has approved licensee applications of the GE methodology as a basis of increasing MSIV leakage. Each of these requests addressed these integrity concerns. Please provide a technical justification for 1 ur proposed approach or re-analyze the doses without using the GE BWROG methodology.
8.
TVA's submittal of October 1,1997 and letter dated April 1,1998 described the methodology for analyzing the radiological impacts due to the proposed power uprate.
This methodology involved a scaling factor based on the change in power level that was then applied to the results from prior analyses. In reviewing your submittal, the NRC staff has identified apparent discrepancies between the TVA's proposal and the existing UFSAR analyses. For example:
a.
Table 3 on Page 8 of the environmental assassment (EA) for the proposed uprate provides results for the MSLB analysis and contains the text: " lodine concentration in coolant = 26 pCi/g dose equivalent 1-131." Section 14.6.5.2.1.b of BFN-15 tabulates the iodine concentrations assumed in the current analyses.
The existing analysis assumptions are not based on 26 pCi/g.
a o
9 b.
Table 3 reports that the pre-uprate analysis result for the exclusion area boundary was 0.66 rem whole body and 32.1 rem thyroid. Contrary to this, l
Section 14.6.5.3 of the UFSAR reports 0.0012 rem whole body and 0.65 rem thyroid. Table 14.9-1 of the UFSAR provides EAB results of 0.017 rem whole body and 2.9 rem thyroid.
Table 5 on Page 9 of the EA for the proposed uprate provides pre-update EAB c.
doses for the RDA analysis as 0.055 rem whole body and 1.62 rem thyroid.
Table 14.6-2 of the UFSAR provides limiting values of 0.0056 rem whole body and 2.4E-4 rem thyroid. Table 14.9-1 of the UFSAR provides EAB results of 0.012 rem whole body and 6.1 rem thyroid.
d.
Table 4 on Page 8 of the EA for the proposed uprate provides results for the FHA analysis and contains the text: " Fuel Handling Accident (single fuel bundle and handling equipment dropped)." The assumptions in Section 14.6.4.1 do not appear to address the weight of the handling equipment.
Section 14.6 of the UFSAR does not report numerical values for the analysis e.
results for the LOCA and FHA. Table 14.9-1 does provide numeric results. The data in the corresponding tables in the environmental assessment are not consistent with the Table 14.9-1 results.
f.
Section 14.6.2.8.4 and TVA letter dated April 4,1994, address release paths for the RDA that do not appear to be reflected in the power uprate submittal. The April 4,1994 letter postulated a thyroid dose of 18.1 rem. This is not cons,nent with the power uprate submittal which indicates the present (pre-uprate) thyroid i
l dose to be 1.62 rem.
Therefore, please:
i explain why the pre-uprate data in the submittal is not consistent with the current design basis as described in the UFSAR.
ii provide a certification that the post-uprate data in the submittal is based on the current design basis as stated in the UFSAR, modified only by those factors explicitly identified in the submittal.
I iii provide a tabulation, in a format similar to that of Table 15-4 of Regulatory Guide 1.70, the analysis assumptions and parameters ;*1at will constitute the design basis of Browns Ferry Units 2 and 3, following the uprate, with regard to the Chapter 14 design basis radiological accident consequences to persons offsite and in the control room.
9.
GE Nuclear Energy, " Generic Guidelines for General Electric Boiling Water Reator Power Uprate," Licensing Topical Report, NEDC-31897P-A, May 1992, Section 5.11.9,
" Power Uprate Testing" indicates that performance testing will be conducted for systems and components which have revised performance requirements. Please identify the systems and components that will be tested 1
10.
Please describe how preoperational testing will be conducted in accordance with the requirements of 10 CFR 50, Appendix B, Criterion XI, " Test Control."