B17512, Provides Response to RAI Re Review of Mnps,Unit 2 Steam Line Break Analysis.No Regulatory Commitments Contained within Ltr

Provides Response to RAI Re Review of Mnps,Unit 2 Steam Line Break Analysis.No Regulatory Commitments Contained within Ltr

ML20155G534
Person / Time
Site:	Millstone
Issue date:	10/30/1998
From:	Bowling M NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B17512, TAC-MA3410, NUDOCS 9811090090
Download: ML20155G534 (5)

Text

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Nortlie:st 4 * ""7 " * " ' *** " " "

Nuclear Energy win. tone Nudear Power Station Northeast Nudear Energy company P.O. Box 128 i ," ,, Waterford, Cr 06385-0128 (860) 447 1791 Faz (860) 444 4277

% Northeast Utilities System OCT 3 01998 Docket No. 50-336 B17512

. U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Response to The Request For Additional information Relating to The Review of Millstone Unit No. 2 Steam Line Break Analysis (TAC NO. MA3410)

This letter provides Northeast Nuclear Energy Company's (NNECO) response to the request for additional information relating to the review of Millstone Unit No. 2 Steam Line Break Analysis, in a letter dated August 121998,W NNECO proposed to amend Operating License DPR45 : by incorporating changes into the Millstone _ Unit No. 2 Technical Specifications. The proposed changes update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b. The plant-specific analysis used by NNECO for the Steam Lino

-Break (SLB) utilizes the r_evised Siemens Power Corporation Methodology, which is currently being reviewed by the NRC staff, in a [[letter::B16951, Application for Amend to License DPR-65,changing TS 3.3.2.1, Instrumentation - ESFAS Intrumentation, 3.4.9.3, RCS - Overpressure Protection Sys & ECCS - ECCS Subsystems - Tavg 300 F|letter dated October 22,1998]],* the NRC requested -additional information regarding the Technical Specification Amendment request which contained two questions relating to the Millstone Unit No. 2 SLB analysis. Attachment 1 provides NNECO's response to these two questions.

There.are no regulatory commitments contained within this letter.

W M. L Bowling to the Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Changes to Technical Specifications, Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.18b," dated 1 i l August 12,1998.-

• D. G. Mcdonald to M. L Bowling, " Request For Additional information Regarding

' Technical Specification Amendment Request - Millstone Nuclear Power Station, Unit i No. 2," dated October 22,1998. l m , I 9811090090 99103 8 p .0 ~

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ADOCK 05000: 36 l PDR P PL R,

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U.S. Nuclear Rrgulatory Commission B17512/Page 2 Should you have any questions regarding this submittal, please contact Mr. Ravi G.

Joshi at (860) 440-2080.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY

/ 7 Martin L. Bowling, Jr. )

Recovery Officer - Technical Services Attachment cc: H. J. Miller, Region i Aaministrator D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 E. V. Imbro, Director, Millstone ICAVP inspections S. Dembek, NRC Project Manager, Millstone Unit No.1

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Docket No. 50-336 B17512 4 ..

Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Response to The Request For Additional Information Relating to The Review of Millstone Unit No. 2 Steam Line Break Analysis October 1998

s , U.S. Nucl:tr R:gulatory Commission B17512/Attrchm:nt 1/Prga 1  !

Request for Additional Information for Review of the

, Millstone Unit No. 2 Steam Line Break Analysis Question 1:

The values assumed in the steam line break (SLB) analysis are listed in Table 3.1 (EMF-98-036) for thermal-hydraulic parameters and in Table 3.3 for different trip setpoints and delay time for actions of various safety systems. The values t.,e different from the current FSAR values listed in FSAR Tables 14.1.5-3 and 14.1.5-4. The break size for the limiting case is also different from the FSAR value.

Please identify all the changes to the input parameters that are important to the results i of the SLB analysis and provide the technical bases to demonstrate the acceptance of l each change.

Response

Section 14.1.5 of the Millstone Unit 2 FSAR has been completely revised to reflect the new steam line break analysis. The revisions to FSAR Section 14.1.5 have been determined to conctitute an Unreviewed Safety Question with respect to 10CFR50.59 and has been submitted to the Nuclear Regulatory Commission (NRC) for review in conjunction with a Technical Specification change for the Control Room Ventilation System. The revised FSAR section 14.1.5 is included in Attachment 5 of NNECO's l [[letter::B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA|letter dated September 28,1998]]) 0 to the NRC. The revised FSAR section 14.1.5 identifies the changes to the input parameters of the SLB analysis and provide the technical bases to demonstrate the acceptance of each change.

Question 2:

The results of the analysis indicate that 0.5 percent of the fuel in the core will fail during an SLB event while the FSAR SLB calculations show no fuel failure.

Please provide technical bases to show the acceptance of the proposed SLB analysis with fuel failure. In addition, provide an example calculation to show that the fuel failure will be limited to 0.5 percent of the fuel in the core due to fuel centerline melt.

Response

The revised FSAR Section 14.1.5, which is included in Attachment 5 of NNECO's [[letter::B17413, Application for Amend to License DPR-65,revising TS Sections 3.3.2.1,3.4.6.2,3.4.8,3.6.2.1,3.6.5.1,3.7.6.1 & 3.9.15 as Result of Revised MSLB Analyses & Revised Determinations of Radiological Consequences of MSLB & LOCA|letter dated September 28,1998]] to the NRC, provides the basis for acceptance of the proposed steam line break analysis with fuel failure. A subsection has been added addressing the offsite dose and control room dose consequences.

") M. L. Bowling to The Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications, Control Room Ventilation System," dated September 28,1998.

• , U.S. Nucirr Regul tory Commission B17512/Att:chment 1/P gs 2 The following example is provided by Siemens Power Corporation. This example l provides a sample calculation for determination of the extent of fuel failure.

i

Subject:

Example Linear Heat Generation rate (LHGR) Fuel Failure Calculation The LHGR centerline melt limit represents the core-wide maximum allowable LHGR on 2

a UO rod to preclude centerline' melt on either a UO2 or gadolinia rod. The fuel I centerline melt LHGR limit for Mills. tone Unit 2 is 21.0 kW/ft. The peak LHGR is '

calculated as follows:

LHGRg = LHGR,, x F, x F, Where: ,

1 LHGRm = Maximum post-scram core average LHGR (based on core power) ,

from ANF-RELAP l

F, =

Nuclear heat flux hot channel factor from a " snap-shot" steady- I state XTGPWR run performed at the time of maximum post- l scram reactor power l F. = Engineering uncertainty factor (F. is typically set to 1.03)

The limiting LHGR for the Main Steam Line Break (MSLB) analysis, Hot Full Power (HFP) with offsite power available, is:

LHGRg = 0.869 x 27.116 x 1.03 = 24.27 If the LHGRg value is below the centerline melt limit, no fuel failures are predicted to occur. If the calculated LHGR % is above the centerline melt limit, a fuel failure  ;

assessment is performed. This calculation indicates that there are fuel failures. To I determine the number of assemblies predicted to fail, the F, associated with the 21 kW/ft must be calculated:

LHOR w p' ,,

LHGR,, x F,

=

F* "= (0.869 1.03) 23.46 The steady-state XTGPWR run also calculates the maximum F, for each quarter assembly in the core. The F, map provided by XTGPWR is reviewed to determine the number of quarter assemblies that exceed the F," limit. If the peak F, in an assembly is

- determined to exceed the F," limit, the entire quarter assembly is conservatively predicted to fail. The HFP MSLB with offsite power available case conservatively predicts that 4 quarter assemblies, or 1 full assembly, may fail. The number of failed assemblies, reported as a percentage of the total core, is 0.46% (1 failed assembly /

217 total assemblies).

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