ML20093D672
| ML20093D672 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/12/1984 |
| From: | Woolever E DUQUESNE LIGHT CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-4-102, NUDOCS 8407170134 | |
| Download: ML20093D672 (6) | |
Text
.-
e>
'Af 4 2) 787-41 Nuclear Construction Division Telecopy Robinson Plaza, Building 2, SLite 210 Pittsburgh, PA 15205 July 12, 1984 United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Mr. George W. Knighton, Chief Licensing Branch 3 Of fice of Nuclear Reactor Regulation
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Response to Outstanding Issues Gentlemen:
This letter forwards responses to the issues listed below.
The following items are attached: : Response to Outstanding Issue 44 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report. : Response to Outstanding Issue 45 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report. : Response to Outstanding Issue 47 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report. : Response to Outstanding Issue 73 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report.
DUQUESNE LIGHT COMPANY SUB CRIBED AND SWORN TO BEFORE ME THIS
//
DAY OF M
, 1984.
N l' do E.
. Woolever W
bu 'verm Y
c Vice President
,g KAT/wjs Attachments cc:
Mr. H. R. Denton, Director NRR (w/ attachments)
Mr. D. Eisenhut, Director Division of Licensing (w/ attachments)
Ms. M. Ley, Project-Manager (w/ attachments)
Mr. M. Lititra,-Project Manager (w/ attachments) 0g Mr. G. Walton, NRC Resident Inspector (w/ attachments) f0 I
k 8407170134 840712 PDR ADOCK 05000412 E
United States Nuclear Regulatory Commission Mr. George W. Knighton, Chief Page 2 COMMONWEALTH OF PENNSYLVANIA )
)
SS:
COUNTY OF ALLEGHENY
)
On this
//
day of
/f7
, before me, Notary - Public in' and for saId hommonwealth and County, personally a
appeared E. J. Woolever, who being duly sworn, deposed and said that (1) he is Vice President of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of said Company, and (3) the statements set forth in the Submittal are true and correct to the best of his knowledge.
- du st/
Notary Public ANITA ELAINE f EiTER. NO. 'SY PUBLIO -
HOBINSON TOWNSHIP. ALLEGHE;iY COUNTY MY COMMISSION EXPIRES 00TOBER 20,1986 k
r
ATTACHMENT 1 Response to Outstanding Issue 44 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 4.4.8:
Conclusion (excerpt)
Address the concerns regarding the effeet of rod bow on DNBR as described in Section 4.4.4.1 of the SER.
Response
The phencmenon of fuel rod bowing, as described in WCAP-8691, " Fuel Rod Bow Evaluation," must be accounted for in the DNBR safety analysis of Condition I-and Condition II e*/ents for each plant application. Appli-cable generic credits for margin resulting from retained conservatism in the evaluation of DNBR and/or margin obtained from measured plant oper-F[H r core flow) -- which are less limit-ating parameters (such as ing than those required by the plant safety analysis -- can be used to offset the ef fect of rod bow.
The safety analysis for Beaver Valley Unit 2 maintained sufficient mar-gin (9.1 percent)* to accommodate full and low flow DNBR penalties iden-tified in References A and B
(< 3 percent for the worst case which occurs at a burnup of 33,000 MWD /MTU).
The fuel rod diameter, pitch, and bowing variation (including inpile e f fect s) was considered in the preparation of the THINC input values such as axial flow area, equivalent hydraulic diameter, and lateral i
cross-flow area for the hot ch annel.
This ef fect (pitch reduction) was used as part of the margin to of fset rod bow penalties.
The maximum rod bow penalties accounted for in the design safety analy-sis are based on an assembly average burnup of 33,000 MWD /MTU. At burn-ugs greater than 33,000 MWD /MTU, credit is taken - for the effeet of F
burndown, due to the decrease in fissionable isotopes and the bu!1 dup of fission product inventory, and no additional rod bow penalty is required.
Design Limit DNBR of 1.30 vs. 1.28 Grid Spacing (K,) of 0.046 vs. 0.059 Thermal Diffusion Coef ficient of 0.038 vs. 0.051 DNB Multiplier of 0.865 vs. 0.88 L
Pitch Reduction Reference A: " Partial Response to Request No. I for Additional Information on WCAP-8691, Revision 1,"
- letter, E.
P.
Rahe, Jr. (Westing-house) to J.
R.
Miller (NRC), NS-EPR-2515, dated October 9, 1981 Reference B: " Remaining Response to Request No. I for Additional Information on WCAP-8691, Revision 1,"
- letter, E.
P. Rahe, Jr. (Westing-house) to R.'J.
Miller (NRC), NS-EPR-2572, dated March 16,.
1982-i
ATTACHMENT 2 Response to Outstanding Issue 45 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 4.4.3.2:
Crud Deposition Operating experience on two pressurized water reactors (not of Westing-significant reduction in the core flow house design) indicate that a rate can occur over a relatively short period of time as a result of crud deposition on the fuel rods.
In establishing the Technical Speci-fications for Beaver Valley Unit 2, we will require provisions to assure that the minimum design flow rates are achieved.
We also require that the applicant provide a description of the flow measurement capability for Beaver Valley Unit 2 as well as a description of the procedures to measure flow.
Response
Operating experience to date has indicated that a flow resistance-allow-ance for possible crud deposition is not required.
There has been no detectable long-term flow reduction reported at any Westinghouse plant.
Inspection of the inside surfaces of steam generator tubes removed from significant surface operating plants has confirmed that there is no deposition that would affect system flow.
The small piping friction contribution to the. total system resistance and the lack of significant deposition on piping near steam generator nozzles support the conclusion that an allowance for piping deposition is not necessary. The ef fect of crud enters into the calculation of core pressure drop through the fuel rod frictional component by use of a surf ace roughness factor.
Present analyses utilize a surface roughness value which is a factor of three.
greater than the best estimate obtained from crud sampling from several operating Westinghouse reactors.
The operator has at his disposal several methods of detecting signifi-cant RCS flow reduction; these are:
a.
Flow meter on each RCS loop, b.
If operating in an automatic control rod mode (T held constant) a e
reduction in reactor power would be present for significant reduc-tions in RCS flow, c.
If operating 'in a manual control rod mode (power held constant) an increase in 21 T across the core would be present for significant reductions in flow, d.
Local changes in flow could be indicated by incore flux maps (assum-ing significant changes in local power), and e.
Core exit thermocouple readings.
Technical Specifications are being prepared for Bes?er Valley Unit 2.
These are being draf ted to require the operator to verify flow, perform c alorimetric power checks, and generate incore flux maps as specified by the Standard Technical Specifications (NUREG-0452, Rev. 4).
7
ATTACHMENT 3 Response to Outstanding Issue 47 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 4.4.7:
ICC Instrumentation (excerpt)
We have reviewed the applicant's submittal of the instrumentation for indication of inadequate core cooling (Section 4.4.6.4).and found it i
insufficient; therefore, the staff will require the applicant to provide the itemized documentation of a complete ICCI system on a schedule which will permit completion of our review prior to fuel load.
Response
A description of the ICC instrumentation,. a core cooling monitor (TsaturationMeter) and a Reactor Vessel Level. Instrumentation Sys-tem (RVLIS) has been provided in FSAR Section 7.7.2.
These meet the requirement of NUREG-0737 Item II.F.2 to provide ins trumentation for the direction of inadequate core cooling.
For more detailed information on the system, see the summary report titled " Westinghouse Reactor Vessel Level Instrumentation System fo r Monitoring Inadequate Core Cooling" (Westinghouse 1980).
4 4
d I
h e
t 9
t e
+
ATTACHMENT 4 Response to Outstanding Issue 73 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 7.6.2.2:
RCS Loop Isolation Valve Interlocks FSAR Section 7.6.6 describes the RCS loop isolation ' valve interlocks.
The description is incomplete and additional information is required to clarify that the design is in conformance with IEEE-279.
Additionally, the staf f is concerned that, during operation with N-1 loops, the cri-teria for testing and single failure may not be met due to reduced pro-tection logic. This is an open item.
Response
a desc ript ion of how the Reactor Section 7.2.2.2 of the FSAR provides Trip System provides automatic core prot ect ion during non-standard operation with a loop isolated by the Reactor Coolant System Loop Isola-tion Valve interlocks.
Isolation of a loop is under strict administrative control.
One of the actions required to continue to meet the single failure criterion during this operation is to place the si TOP and the 4 TOT channels (associated with the loop not in service) in their tripped condition.
This instru-mentation would continue to be in partial trip except during surveil-lance. The surveillance testing requirements will be described in Tech-nical Specifications similar to those being developed for Beaver Valley Unit One.
.._.