ML20115A116
| ML20115A116 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Waterford |
| Issue date: | 03/14/1984 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082170562 | List:
|
| References | |
| FOIA-84-143 NUDOCS 8404050128 | |
| Download: ML20115A116 (10) | |
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W hti E M L. RUBENSTEIN C. BERLINGER ttDl0RANx2 Fud.
T. M. hovas, Assistant 01 rector H. BA W IAN for Licensing, LL
[. HANG Fhut:
L. S. hetenstein, Abststant Director L. PHILLIPS for Core and Plant Syntes4, LSI R. LOBEL Sd!JCCT:
ATERFORp 3 SLR SUPPLEl:Lt.T lib. 6 Plesit tempe; ater ord-3 uucket itsmber.
bO-3"2 Ltcanstn, Stage:
UL wtsponsible tirancn.
LS-3 Pru,ect haneSer.
J. H. uIison (A-27702)
.,SI r',cvttw oranct:
Loru Performance brancn hus i=w St4tus:
Ceiplete ciclohco is inp.t for the. heterfoN Unit 3 Supp1 ment tw 6 to the Safety tvaluation Ntiert.
Tnts input incluoes Section 4.2.2 Design Lvaluation,
.wisritc-end-LuGA taading. desed on the applicants recent sets 41ttel (Drurtund
, quat 12. a'43), ime conclude tnat the seimic-ane-LOCA loading issue has o==ri rwsulved and snus ts.c licesiw cundttion of w1mic-and-LOLA lueuing is ru:.otus. for beterforo unit 3.
For Si.ction 4.4, Therual-$oraulic Design, the applicant sies provided slightly
..o rt conservative rou tas DiCR penalttes than previously reportou in Supplesient
.~. ! n.nich is acceptable. Also, the applicant has provideo an acceptable eisolvts tu.;ustify the design value of 2.Gk for bypass flow, ar rsv1=w of acd1tional inforination sulaattted in respone to Iteen II.F.2, "4..tru.or.tation for Detection of inacewate Core Cooling" of liUREG-0737
- ciarificatter of Till Action Plant Rewirement" tuas found to be acceptable.
Su,+l is-en t o had identif feu e deficiency in the wterford Unit 3 LPC sottur= te6ere a pre-atterutned penalty factor for botn-falleo CEAC condition auss not oppitua for the local pouer density calculation.
The dettotency tus l
Contacts:
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- 1. Huang, CPb:DSI
- 5. L. L,u, CPb.051 A-L A22 A-27064 1-29476
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MAR 14 Re4 been corrsste.c 4:s Ltwreton. ttw wtertura Lnit s s.vt/ Lust is occstt4 Lit as ri.purted in tac letter of Decta.2ber 6, bo. sent to T. ti. t.ovs k t ur a p pi ss is n t Original signed by ((
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- k. S. Rubenstci, L. 5.; 49 Rutenstein. Assistant tirector for Core end Plant Systu..s. 051 4.nclosure;.
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ENCLOSURE SUPPLEMENT NO. 6 TO THE WATERFORD-3 SAFETY EVALUATION REPORT 4.2.2 Design Evaluation Seismic-and-LOCA LOADING An important aspect of the behavior of the reactor core during a loss-of j
coolant accident (LOCA) is the response of the fuel assablies to asynnetric blowdown loads and the SSE. The applicant has subnitted (Drumond, August 12, i
1983) a plant-specific analysis described in the attached report CEN-159(C)-P Rev.1-P, " Final Assessment of Waterford-3 Fuel Structural Integrity Under Faulted Conditions" dated July 15, 1983. The CEN-159(C)-P Rev. 1-P'results are based on the models and acceptance criteria described in the approved report CENPD-178. Revision 1.
Table 1 of CEN-159(C)-P Rev.1-P shows that the peak canbined loads are well below the allowable prescribed limits in almost all cases except for the case of peripheral assemblies under a one-sided load. The' peak combined load on these perthperal assablies under one-sided loading condition is about 300 lbs higher than the allomble limit. The applicant performed an ECCS analysis to determine the coolability of these assablies by assaing grid deformation according to the SRP 4.2 Appendix A recommendations. The result shows that coolability is maintained mainly because these assablies are located at a low power density area.
Therefore, based on the fact that (1) the peak cabined loads on the grids are below the prescribed limits for most cases, and (2) the coolability is main-tained for those fuel assemblies with combined load exceeding the prescribed limit, we conclude that the selmulc-and-LOCA loading on fuel assablies satisfies the intent of Sir 4.2 Appendix A and this license condition can
'therefore be removed for Waterford thtt 3.
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, 4.4 Thermal-Hydraulic Design 4.4.1 Thermal-Mydraulic Design Criteria and Bases l
By a letter from K. W. Cook, Louisiana Power and Light to G. W. Knighton, NRC, dated Phrch 1,1984, Louisiana Power and Light (LP&L) provided a revised table of values for rod bow penalty which is slightly more conservative (0.5%) than the previous values given in Supplanent No. 5 to the Waterford SER (Section 4.4.1).
These values which follow the guidelines of the CE Topical Report CENPD-225, an approved report, are as given below:
Burnup DNBR Penalty (GWD/MTU)
(1) 0-10.0 0.5 10.0-20.0 1.0 20.0-30.0 2.0 30.0-40.0 3.5 40.0-50.0 5.5 The applicant has stated that the thermal margin reductions for rod bow will be put in the DNBR limit basis of the Core Protection Calculator (CPC) system.
Tney will be verified to be included in the DNBR limit calculations in the Core Operating Limits Supervisory System (COLSS) and the CPC systen at least once per 31 days. Thereforv, the appropriate provisions will be incorporated into the Technical Specifications. The applicant should also insert into the basis of the Technical Specifications any generic or plant specific margin
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that may be used to offset the reduction in DieR due to rod bowing, and reference the source and staff approval of each generic margin. With these rewirements satisified by the applicant the staff concludes that they have l
adewately accounodated the reductions listed above.
The core flow design basis rewires a minimim flow which will pass through the fuel region and be effective for fuel rod cooling as a percent of the Primary coolant flow rate er 148.0x10'lb/hr. The remainder of the flow.
. called bypass flow, will he ineffective for cooling because it will take the following bypass paths:
(1) outlet nozzle / core support barrel (CSB) gap (2) core shrouo/ CSS annulus l
(3) alignment keys (4) guide tubes l
The design and best estimate bypass flow rates for Waterford-3 were reduced in Arwndment 30 of the FSAR fras earlier values as shown below:
l Previous Now Design bypass flow 3.5%
2.6%
Calculated best estimate bypass flow 2.7%
2.1%
We mwested that LP&L provide a description and justification of changes resulting in the reduced bypass flow. This mes provided in a letter from K. W. Cook (LP&L) to G. W. Knighton. NRC dated February 22, 1984.
The infonnation indicated that two design changes were made to reduce the bypass flow through the guide tubes. The bypass flows in the other leakage paths remain the same. The best estimate bypass flow rate in the guide tubes ses reduced by:
(1) reducing the overall flow area of inlet flowholes and (2) by adding sleeves in the upper ends of the guide tubes. These changes reduce the bypass flow by increasi g the hydraulic resistance.
This is summarized below:
Best Estimate Bypass Flow Rate treass flow sath Previous Now h tlet nearla/ CSS gap 0.61 0.61 Com shroud /C58 annulus 0.62 0.62 Alignment keys 0.09 0.09 Guide tubes 1.38 0.78 2.70%
2.10%
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The applicant presented a description of the guide tube design chanjes and a breakdown of the bypass flow through the guide tubes including the flow networks used to calculate the present best estimate guide tube leakage rate of 2.11.
An additional 0.51 increnent over the best estimate value of 2.10s i
accounts for the effects of core crudding, tolerances and other unknown factors and results in a design value of 2.65.
The staff concludes from the calculations presented that reduction from 3.51 to 2.6% for design bypass flow is acceptable.
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SSER INPUT FOR WATERFORD UNIT 3' RESPONSE TO ITEM !!.F.2 0F NUREG-0737 INSTRLMENTATION FOR DETECTION OF INADEQUATE CORE COOLING Discussion SER Suppleent 5 of Waterford Steam Electric Station. Unit No. 3 stated the staff findings that two open items remain to be resolved prior to issuing an operating license as follows:
(1) The response to Iten (2) of II.F.2. Attachent 1 (on a primary operator display) is incomplete.
It should be clarified; and (2) The response to Itas (4) of II.F.2 documentation required. is not complete.
l It should include each subsyste of the final ICCI systen.
In response to the staff findings in the SSER No. 5. the Waterford Unit 3 applicant (Louisiana Power & Light (LP&L)) has provided additional infonnation in both the October 31, 1983 and the FSAR Amendment No. 34 dated January 1984.
LP&L has clarified its use of the Qualified Safety Parameter Display Systen (QSPDS) for primary and backup ICC display in the Waterford-3 control room.
The QSPDS perfonas safety grade signal processing and display of the ICC parameters, and is located on the main control panel for reactor protection,
in order to facilitate operator use. The QSPDS accepts sensor inputs, processes the signals, and transits the output to its own alphanumeric display and to the plant computer through which the :line printer is accessible.
'I Ci non-Class 1E inputs and interface with the plant computer are isolated i
fran the Class IE QSPDS egsipment, l
A spatially oriented CET temperature map is available on demand from each train of the QSPDS (primary and backup) providing a uniform representative picture of core exit temperature obtained by utilizing 28 CETs (7 per quadrant) dedicated j
only to that train. A strip chart recorder is provided to allow trending of representative CET temperature for the primary display (QSPDS train A).
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' Direct readout and hard copy capability is provided for all thennocouple taperatures (direct readout for the 28 CETs associated with each train of tw QSPDS can be obtained fran the display associated with that train; hard copy capability is via the line printer). Selective readings of core exit taperature, continuous on deand, is available fran both the primary and backup displays. Based on our review, we have found that the applicant's clarification is acceptable.
LP&L has also provided the infonnation in response to Iten (4) of the docu-mentation retired by NUREG-0737 Itan II.F.2, including an evaluation on the confonnance of the ICC instrument system to !!.F.2 Attacionent I and Appendix B of NUREG-0737. We have reviewed and found it in compliance with the rew irm ents.
Conclusfor.
i Based on our review, we have concluded that the applicant's response to the open itms stated above is acceptable for an operating license.
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ENCLOSURE 2 SALP INPUT FR(M THE C0pE PERFORMANCE BRANCH FOR WATERt.'O UNIT 3 A.
Licensing Activities 1.
Managment Involvement in Assuring Qual 16.-
No basis i
Rating 2.
Appr to Resolution of Technical Issues from a Safety
- Standp, We
- d three telecons with the applicant for Section 4.2.2.
Th applicant has shown an adequate approach in resolving i
the license condition on seismic-and-LOCA loading.
Rating:
Category 2 i
3.
Responsive to NRC Initiatives Acceptable resolution of II.F.2 Quick and acceptable response for bypass flow resolution Rating: Category 2.5 i
4.
Staffing (including knagment).
No basis Rating:
5.
Reporting and Analysis of Reportable Events.
l No basis i
Rating:
6.
Training and Qualification Effectiveness.
No basis Rating:
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2-7.
Overall Rating for Licensing Activity Functional Area.
Adequa te Rating:
2.0 h
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