ML20217E123
| ML20217E123 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/25/1997 |
| From: | Wetzel B NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9710060310 | |
| Download: ML20217E123 (49) | |
Text
.
C' o p Rat:0 '
/
3 4 UNITED STATES
{, } NUCLEAR REGULATORY COMMISSION go . g.2
$ WASHINGTON. D.C. 30006-4001
.o September 25, 1997 LICENSEE: Northern States Power Conpany (NSP)
FACILITY: Prairie Island Nuclear Cenerating Plant
SUBJECT:
REVISED MAIN STEAM LINE BREAK METHODOLOGY AmeetingwasheldatNRCHeadquartersonAugust 26, 1997. with Northern States Power (NSP) to discuss Prairie Island s revised main steam line break (MSLB) containment 3ressure response in antici)ation for a license amendment request to reduce tie boron concentration in t1e boric acid storage tanks.
Enclosure 1 provides a list of attendees. Enclosure 2 provides the licensee's handout. Enclosure 3 provides background information of the May 26, 1994 MSLB meeting summary with NSP.
In June 26. 1997. NSP submitted a report revising the MSLB methodology for staff review. The report includes credit taken for the closure of the broken loop non return check valve and liquid entrainment modeled for the broken steam generator (SG). The report corrects some weaknesses identified during a plant Final Safety Analysis Report (FSAR) review 3erformed in 1994.
Discussion of problems identified in the FSAR and tie use of DYN00E and CONTEMPT codes to calculate mass and energy releases profiles and containment peak pressure during an MSLB event are included in Enclosure 3.
NSP requests that' output of the DYN00E code that calculates energy and mass release profiles be approved for use, utilizing the liquid entrainment model, as input to the CONTEMPT code for containment response analysis. The revised MSLB methodology will be used for future core reload evaluations, equipment qualification, and plant modifications as well as other related main steam line events. NSP concluded that the revised MSLB analysis addressed known weaknesses identified in the FSAR and is an im)rovement over the current MSLB analysis. Also, the use of the new MSLB metlodology demonstrates that the j plant licensing basis is still conservative. /j There were some questions from the staff regarding entrainment modeling and the effects of 3hysical impacts of liquid entr61nment during an MSLB event.
NSP indicated tlat the staff's approved reference industry report. WCAP-8822.
" Mass and Energy Releases Following a Steam Line Break." was used to address the entrainment phenomena.
9710060310 970925
- " ^ cxo5oo=2 yp, wu m g 7_73 gW
September 25, 1997 The review schedule was also briefly discussed during the meeting. The staff.
is currently performing its review of the topical report, which will be used to support Prairie' Island's future amendments. The staff expects to have the .
Prairie Island plant review completed by January 1998.
ORIGINAL SIGNED BY.
Beth A. Wetzel Senior Project Manager Project Directorate 111-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-282-50-306 i
cc w/ encl: See next page
Enclosures:
- 1) List of attendees
- 2) Licensee's Handouts
- 3) Background information HARD COPY (w/atts)
Docket File (50-282/50-306)
PUBLIC PD# 3-1 Reading OGC ACRS.
D.- Diec B.-Wetzel E MAIL (w/o atts)
S. Collins /F, Miraglia (SJC1/FJM)
R. Zimmerman (RPZ)
E. Adensam (EGA1)
C. Jamerson (CAJ1).
T. Martin (SLM3)
T. Collins (TEC)
F. Orr (FR0)
W. Long-(WOL)
C. -Ber linger- (CHB)
D. Terao (DXT)
E. Weiss (EWW)-
G .Tracy (GMT)-
J. McCormick. RIII (JWM)
S. Ray. RIII-(SPR) .
DOCUMENT.NAME: G:\WPDOCS\ PRAIRIE \PIMTG897. SUM To receive a copy of this document, indicate in the box C= Copy w/o attachment / enclosure E-Copy with attachment / enclosure N - No copy 0FFICE PM:PD31 nE 'PM:PD31 LA:PD31 E; D:PD31 ; BC:SRX,B/\ OCSB k NAME DDiec(d[ -BWetzel60% CJamersor([/JHannFP F0rrY,Y WLong I DATE 9/W/97 9/pj/97 9/l7/97/) 9/2 6 /97 9/26/97 9/SI/97 0FFICIAL REK)R) COPY
1 C+ 4 Northem States Power Company Prairie Island Nuclear Generating Plant cc:
J. E. Silberg, Esquire Site Licensing Shaw, Pittman, Potts and Trowbridge Prairie Island Nuclear Generating 2300 N Street, N. W. Plant Washington DC 20037 Northem States Power Company 1717 Wakonade Drive East Plant Manager Welch, Minnesota SS089 Prairie Island Nuclear Generating Plant Tribal Council Northem States Power Company Prairie Island Indian Community 1717 Wakonade Drive East ATTN: Environmental Department Welch, Minnesota 55089 5636 Sturgeon Lake Road Welch, Minnesota 55089 Adonis A. Nebiett Assistant Attomey General Mr. Roger O. Anderson, Director Office of the Attomey General Nuclear Energy Engineering 455 Minnesota Street Northem States Power Company Suite 900 414 Nicollet Mall St. Paul, Minnesota 55101-2127 Minneapolis, Minnesota 55401 U.S. Nuclear Regulatory Commission Resident inspector's Office 1719 Wakonade Drive East We'ch, Minnesota 55089-9642 Regional Administrator, Region ill U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Mr. Jeff Cole, Auditor / Treasurer Goodhue County Courthouse Box 408 i Red Wing, Minnesota 55066-0408 Kris Sanda, Commissioner Department of Public Service 121 Seventh Place East Suite 200 St. Paul, Minnesota 55101-2145 November 1996
< s MEETING ATTENDEES ON AUGUST 26, 1997 Hame Oraanization D. Diec NRC B. Wetzel NRC F. Orr NRC
~D. Terao NRC D. Vincent NSP K. Higar NSP C. Bonneau NSP
- 0. N31 son NSP D. Kern NetCorp ENCLOSURE 1
Northern States Power -
Steam Line Break Methodology NSPNAD-97C O2, June 1997 Presentation on 8/26/97 i
m
^i i
i Outline t
1 i
I a
! e Introduction
'l l:
o Licensing Background j -
e Format of Report l
, o Codes e Methodology e Reload Evaluation Approach l
^
introduction m ._ _ . .
e Describes NSP's MSLB methodology e Evaluation of future core reloads and plant modifications e Similar but separate methodology for containment and core response e Addresses weaknesses in current licensing basis analyses e Tied to approved Topicals
Licensing Background
' ~~ '
**f fMM4;MM&%Mdt%' GXMfC' C
eFSAR e IE Bulletin 80-C4 e NSP's RSE Topicals e Weaknesses identified in 1994
FSAR e Single analysis for both core and containment response e Modeled to maximize return to power
.._.__: l l
e Additional water sources to SGs e May 1980 Submittal e Single analysis for both core and containment response e Modeled to maximize return to power o SER issued in October 1982
f l NSP's Topicals
]!
1 e NSPNAD-8101 and NSPNAD-8102
! e SER issued February 1983 e Mass and energy release from DYNODE not qualified as input to CONTEMPT e FSAR and May 80 submittal remain l licensing basis for containment response
FSAR Weaknesses identified in -
1994
' MAwebwu awwa,gy.gggggj =ry
=vr--
e Low Feedwater temperature l e Engineering analysis concluded FSAR remains conservative and bounding
] e Met with NRC in May 1994 L
u o Documented in JCO 4
4 t
.l Format of Report e Sections 1,2 & 3 provide background and descriptions of codes e Section 4 defines the MSLB Containment Methodology e Section 5 defines the MSLB Core Methodology l e Section 6 defines the SSLB Core 4 Methodology i
Format of Report cont.
l
- o Section 7 provides examples of results
, using new methodologies e Section 8 describes reload physics parameter comparisons e Section 9 lists the acceptance criteria for Steam Line Break i e Appendix A compares new containment methodology to ANSI /ANS-56.4
e a N
o
- PH CO CD D
U
, CD l
- e-4 1 O
l l
1 n l M
m i O
- PH h
o N
9 H
Codes l e Dynode l main and Aux. Feedwater flow split modeled entrainment modeled e VIPRE used per Topical NSPNAD-8102 e Contempt i Uchida correlation, per NUREG-C588 Appendix B
Initial Conditions e Analyzed a spectrum of l
power levels break locations & sizes entrainment assumptions
~
o Used consistent input values
b .
Initial Conditions cont.
o Physics parameters and SDM calculated c
per Topicals e SI boron concentration e Initial SG liquid inventory i
e Initial containment pressure and temperature l e PZR level & pressure
Single Failures
.4;,aul .,.~ , . , _ . . . _._ _ ___ __ .
e With and without offsite power .
l main FW Reg valve non-Return Check valve 1
one train of safeguards i
I r
I i
r i
i 1
o FW & AFW oFW modeled valves opening and closing o AFW flow -
iow based on system hydraulic characteristics run-out protection conservatively modeled f 1
Plant Systems
- - _ _ _ _ _ _ _ =
l l
e SG tube height 1 e RCS, SG, and containment heat structures .
o Reverse SG heat transfer e Containment spray and FCU conservatively modeled e Non-Return check valves modeled 1
i
. 1 1
Entrair ~:ent
== \
e Entrainment modeled for broken SG e WCAP-8822, applicable for Model 51 SG e Used recommended uncertainties
Asymmetrical Temperature
~
Distribution e DYNODE assumes perfect mixing e Containment response; conservatism due to tube height assumption bounds non-conservatism due to temperature asymmetries e Core response; adjust the K vs. T curve such that the core inlet is effectively at the broken loop's temperature
0 Misc.
e Decay heat is modeled e Boron Transport penalty e SG Tube plugging conservatively modeled e Appropriate time delays are used for Engineered Safeguard components e Conservatively modeled Void & Doppler flattening e Evaluated operator initiated RCP trips
O A O
Reload Evaluations e Calculate cycle specific physics parameters per Topicals e Compare them to values used in last MSLB analysis e If inequalities are not met will reanalyze MSLB
em Acceptance Criteria
+
e Listed in Section #9 e Slightly different than in NSPNAD-81C2 o More consistent with Standard Review Plan e Shows how analyses in section 7 meet the criteria
Summary
= = _ _ _
e Addresses know weaknesses e Models physical phenomena not previously modeled e More rigorous analysis e Is an improvement over current MSLB analyses
J
. * *"84 i ' *
. 5 -
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. McMoot Docket Hos. 50-282 and 50-306 ,
LICENSEE: Northern States Power Company '
FAtlLITY: Prairie Island Nuclear Generating Plant
SUBJECT:
MAIN STEAM LINE BREAK METHODOLOGY A meeti*ng was held at NRC Headquarters on May 19, 1994, with Northern States !
Power (NSP) to discuss Prairie Island's (P! Main Steam Line Break analysis.
Enclosure 1 provides a list of attendees. nelosure 2 provides t.
handout. 11censee's During a recent licensee review of the PI Main Steam Line Break (MSLB) analysis, some problems were identified. The review determined that inputs in the analysis were incorrect. These incorrect assumptions included:
- 1) using 110% of the Technical Specification Reactor Coolant System flow instead of 115%;
- 2) using an incorrect safety injection line volume; i
1 3) using 50 Stu/lbm as a value for feedwater enthalpy instead of 40$ Btu /lbm.
When these inputs were corrected and the analysis was rerun with all other conservative assumptions retained, the result exceeded the containment pressure limit of 46 psig by approximately 1 psig. Additional cases of the analysis were performed removing some conservatism.
The reduced conservatism included credit for the closure of the broken loop non-return check valve and inodeling Followingofa liquid Steam entrainment using WCAP-8822,
- Mass and Energy Releases Line Bre k." for model D steam generators. P1 has model 51 steam generators; however, the data for the model D steam generators is more conservative.
of Enclosure 2. The results of the additional cases are summarized in Table 5 The licensee is using a new methodology for containment response to a MSLB.
- jThe Dynode code is used to calculate mass and energy release profiles which I are input to the CONTEMPT code for containment response analyses. The -
licensee has been previously found qualified for use of CONTEMPT; however., '
the .
staff's review of the licensee's qualifications to use Dynode (in 1983) is limited to core response analyses. The licensee indicated that it plans to submit for approval a report describing the use of Dynode for containment mass and energy release analysis.
There were questions from the staff regarding equipment qualification (EQ) and long-term containment high temperature response resulting from the higher
. calculated mass and energy releases. NSP s,tated that the loss,of coolant accident event is limiting for EQ; therefore, there werd no new EQ concerns.
ENCLOSURE 3
May 26, 1994 The licensee performed an internal Justification for Continued Operation (JCO) and concluded that there was r.o reduction in the margin of safety and there were no unreviewed safety questions.
The staff informed the licensee that the changes discussed at the May 19, 1994, presentation were acceptable. The JC0 should be documented 45 an existing plant record and should be available for future audits.
If you have any questions regarding this summary, please contact me at (301) 504-3024.
l}/
(
wk wW :
Marsha Gamberoni, Project Manager Project Directorate 111-1 Division of Reactor Projects - !!!/IV
Enclosures:
Office of Nuclear Reactor Regulation l 1. List of Attendees t
- 2. Licensee Handout cc w/ enclosures:
See next page to
- w e
N
.. =e w. ,- A:.---M.<ww. w wgM M e4>*M -'- ' '"
O L
NSP - NRC Meeting Prairic Island F Main Steam Line Break Analysis Probicm Resolution and JCO
.: i il Washington, D.C. -
1; May 19,1994 o
a, i
Agenda
+ Background
+ Problem Identification 1
- + New Analysis i
+ Results and Conclusions i
?
ENCLOSUFI 3 !
i NSP -NRC Meeting Prairie Island
- Main Steam Line Break Analysis Problem Resolution and JCO i
Washington, D.C.
l l
May 19,1994
+ Objective
. Describe Analysis Done for Main Steam Line Break
. Obtain Justification for Continued Operation PI Unit 2 Cycle 16 - operating PI Unit I Cycle 17 - refueling, startup in mid-June
.., .......,,....u . . ., . .. . . _ . . . . . . . . . . . . . , .
List of Acronyms i
l Aux FW Auxilliary Feedwater BAST Boric Acid Storage Tank DE Double-Ended (break)
EOC End Of Cycle FSAR Final Safety Analysis Report ;
! HFP Hot Full Power >
HZP Hot Zero Power MSIV Main Steam Isolation Valve
! MSLB Main Steam Line Break M&E Mass and Energy release j N-1 All control rods inserted, most reactive stuck out NSP Northern States Power Co.
PI Prairie Island RCS Reactor Coolant System RSE Reload Safety Evaluation SDM Shut Down Margin SG Steam Generator i
SI Safety Injection TS. Technical Specification -
r, . . . - , , , ,. , ,,,-.,--r,-wn- ,, - -- ,- - , - - - - - , , - , , e --. ,-w - y,---- ,
w-
- ,i'
Background
! Prairie Islarid description:
. Westinghouse 2-Loop PWR
. Rate'd power is 1650 MWth West Vantage + 4.95 w/o U-235
- . Gadolinia at 6 and 8 w/o .
. West Model 51 steam generators l Prior to 1980, FSAR was licensi's basis for Main Steam Line Break (MSLB).
1
. Vendors responsible for Reload Safety Evaluations (RSE) l .
In 1980, IE 80-04 issued to address water sources:
-l
. Increased Reactor Coolant System (RCS) flow
! - Aux Feedwater flow at runout
May 1980 Assumotion FSAR Submittal New Analysis Break Type DE @ SG Exit FSAR FSAR Physics Parameters HZP, EOC FSAR FSAR SDM 2% FSAR FSAR Control Rods , N-1 FSAR FSAR Offsite Power Available FSAR FSAR FW Isolation 10 sec FSAR FSAR FW Enthalpy 50 Btu /lbm FSAR 405 Btu /lbm Aux FW Flow ,
200 gpm Runout Runout -
SG Level ,
33 % FSAR FSAR RCS Flow 100% ofTS Min 110% ofTS Min 115% ofTS Min Single Failure Safeguards Train FSAR FSAR MSIV Closure 10 sec FSAR FSAR Check Valve Closure Not credited FSAR 5.5 sec SI Time. 6 sec FSAR FSAR Entrainment 85% M&E FSAR WCAP-8822 I
m__ ________ _ _ -
NSP response (May 1980 Submittal) accepted by NRC.
~
May 1980 Submittal became part oflicensing basis.
In 1982, NSP RSE methods topical submitted.
l
. Same MSLB assumptions as FSAR except Aux FW at runout
. Utilizes the DYNODE and CONTEMPT codes DYNODE models primary and secondary systems.
. Simulates pipe breaks
. Computes mass and energy release (M&E) i g
i
CONTEMPT models the containment structure.
. Inputs DYNODE mass and energy release (M&E)
. Computes pressure and temperature 4 NRC did not approve DYNODE mass and energy (M&E) input into CONTEMPT.
. Review was not completed.
. NSP would therefore verify the previous analysis bounding. .
Initially, NSP did not evaluate MSLB containment response:
. Response relatively insensitive to fuel type L__
i Starting in 1987, compared M&E to FSAR.
. Approved to generate M&E
. Gave assurance that FSAR still bounding o
O O
i
Problem Identification ,
During 1993 several minor problems identified.
. Led to detailed review of MSLB methods.
This review determined:
. Higher RCS flow limiting. (115% Tech Spec min) .
- Previous analysis used 110% Tech Spec min
- Actual flow is ~111% ofTech Spec min
. SI line volume not conservative for Unit 1 1
. Higher feedwater enthalpy limiting (found in 1994)
- FSAR and May 1980 used 50 Btu /lbm. .
- 405 Btu /lbm exists at full power -
~
When corrected, May 1980 Submittal M&E exceeded.
.i
~
Led to reknalysis of MSLB Containment Response, including:
. .PI 1 Cycle 16 (just completed)- .
. PI 2 Cycle 16 (operatic;)
. PI 1 Cycle 17 (next reload)
) .
New analysis used DYNODE / CONTEMPT link.
. Not approved in the topical.
i New Analysis _ .
i Problems Cc rected
+
. . CS Flow at i15% oft.S. Min R
. SI Line Volumes Modeled for Each Unit
' . FW Enthalpy at 405 Btu /lbm (*ze N %)
Result.- Table 5 ,_im r 40 psicy PI 1 Cycles 16 & 17 PI2 Cycle 16
. Case A 47.4 psig 47.1 psig
@ 160-200 sec @l60-200 sec This was Expected.
All Conservative Assumptions were Retained.
" " -_es-:-- _ _ - mmm-__1____ _ m__ _
i 3
(New Analysis)
I I
+- Remove Some Conservatism t t
i j Use Experience From Boric Acid Storage Tank (BAST) Project t i . i
~
~
.. Credit the Closure of the Broken Loop Non-Return Check Valve j
. Model Liquid Entrainment .
4 i
Mini,mize Deviations from Licensing Basis
.. t 1 !
Use Methods that are Not Approved for NSP Goal : Demonstrate Licenang Basis is Still Conservative ;
i .
~
f:
~
I . !
~.
(New Analysis)
+ Non-Return Check Valves Not Credited in FSAR and May 1980 Submittal MSIVs Provide Isolation
. Safety Grade (QA Class I)
. Meet ASME Section XI testing criteria - test valve travel . .
. Similar to MSIVs - no air cyclinder actuator
. Conservative Closure Time of 5.5 seconds
1 i
. (New Analysis)
+ Entrainment Specify Break Exit Quality vs Time in Dynode Code .
WCAP-8822 Overview:
. Tranfl0 Code - Break Exit Quality
. Model D SGs - U Tube, Integral Preheater
. Sensitivity Study found:- .
applicable to Model 51 SGs Model Ds have less entrainment
. 1.4 sqft DE Break (PI: 4.6 sqft DE Break)
. WCAP-8822 has been Reviewed byNRC 4
___m.- - . _.___..___ _ . _ - .
~
d Results
! + Case B: .
l l . Errors Corrected
- . Credit Taken for Non-Return Check Valves l
Result - Table 5 PI 1 Cycles 16 & 17 PI 2 Cycle 16 .
i i Case B 46.2 psig 46.2 psig
~
@ 170-200 sec @l70-200 sec 4
. . Containment Pressure Slightly Over Limit
. Worth of Check Valve = 1 psi
~
(Results)
+ . Case C-
.'- Errors Corrected
. Credit Taken for Non-Return Check Valves
. Liquid Entrainment Modeled .
Result - Table 5
< PI I Cycles 16 & 17 PI 2 Cycle 16 .
Case C .28.7 psig 28.4 psig
@ 12 sec @ 12 sec
. Margin of 17 psi to Limit
. Worth of Entrainment Method = 16 psi
(Results)
+ Case D:
. Errors Corrected
- Credit Taken for Non-Return Check Valves
- Sensitivity Study on Entrainment Result - Table 5
~
PI I Cycles 16 & 17 PI 2 Cycle 16 -
Case D 46.1 psig . 45.9 psig
@ 80 sec @ 80 sec
. Entrainment Reduction of 37% Required to Reach Limit
. WCAP-8822 Average Uncertainty is 17%
. Provides Confidence in Making Conclusions e
l Table 5 Results ofNew Analysis .
Containment Design Pressure Limit = 46 psig PI I Cycles 16 & 17 PI 2 Cycle 16 Case A allerrors resolved 47.4 psig 47.1 psig
@ 160-200 sec @l60-200 sec Case B check valves 46.2 psig 46.2 psig .
@ 170-200 sec @l70-200 see Case C cheak alves, entrainment 28.7 psig 28.4 psig
@ 12 sec @ 12 sec Case D check valves, 63% entrainment 46.1 psig 45.9 psig
@ 80 sec @ 80 sec
~
Conclusion
+ Current Licensing Basis is FSAR and May 1980 Submittal
+ Problems Corrected
. Used Unapproved Methods in a Conservative Manner l + -No Reduction in the Margin of Safety L
+ No Unreviewed Safety Questions
+ Justification for Continued Operations:
Unit 2 Cycle 16 - operating l
Unit 1 Cycle 17 - refueling, startup in mid-June i
5
- . Future Actions i
i + Long Term Solution:
. New MSLB Methods BAST Project work in Progress
. Finish in Fall 1994
. PI Unit 2 Cycle 17 - startup June 1995 .
l I
. -