ML19211B954

Notifies of Recent Evidence Suggesting Small LOCAs for PWRs May Be More Limiting W/Respect to Vessel Integrity at Low Temp than Normally Assumed Steam Line Break or Large Break LOCA Events.Div Will Pursue on Generic Basis

ML19211B954
Person / Time
Issue date:	03/19/1979
From:	Novak T Office of Nuclear Reactor Regulation
To:	Noonan V Office of Nuclear Reactor Regulation
References
REF-GTECI-A-11, REF-GTECI-RV, TASK-A-11, TASK-OR NUDOCS 8001070255
Download: ML19211B954 (24)

Text

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UNITED STATES b g p t NUCLEAR REGULATORY COMMISSION g WASHINGTON, D. C. 20656

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MEMORANDUM FOR: V. Noonan, Chief, Engineering Branch, D0R FROM: T. M. Novak, Chief, Reactor Systems Branch, DSS %W0

SUBJECT:

INFORMATION MEM0 - VESSEL INTEGRITY FOR SMALL LOCAs

== Introduction:== 1 Evidence has arisen recently which suggests that small LOCAs for PWRs may be more limiting with respect to vessel integrity at low temperatures than the normally assumed steam line break or large break LCCA events. { Problem: An analysis of reactor vessel repressurization fracture mechanics performed by Westinghouse for Alabama Power's Sequoyah Units 1 and 2 revealed that the limiting events were small LOCAs, 4 and 6 square inches in area. Two dimensional flaw analyses of the pressure vessel indicated that following S these small LOCAs, the faulted stress limits would be exceeded at 27 and t 28 calendar years assuming a load factor of 0.8 (see Table 1). Previously, the staff had considered large break LOCA and steam line break events to be the most limiting events which challenge vessel integrity. For vessels with high copper concentration or marginal welds, a small LOCA calculation may prove to be more limiting than the previously analyzed events which are nonnally required by the staff. In the vessel integrity analyses perfonned for Sequoyah by Westinghouse, material fracture properties were based on a copper content of 0.15 weight percegt, a phosphorus content of 0.011 weight percent and an initial RT NDT of 73 F obtained from vessel material certification for the pressure vessel. The fluence used in these analyses was supplied by Westinghouse for a four loop plant similar to Sequoyah. The applicant's acceptance criteria was that flaws less than 0.1156 a/t (1.0 inch) would be arrested within 75 percent of the vessel wall thickness. No credit was given for operator action prior to 10 minutes after the first alarm. DSS Actions: DSS intends to pursue this matter on a generic basis through Generic Task t. A-11, a study on the resistance of reactor vessel materials to brittle fracture. g r

Contact:

Glenn Kelly, NRR ? i 49-27591 n M = c>. 1 ~y 3759

37 p'

3 mU 8001070 -

V. Noonan IAAR 191979 We request that D0R supply us with a description of any action or actions they intend to pursue to resolve this issue for operating plants. This information will be included in a board notification being prepared by DSS related to yessel integrity. }&:r,- Thomas M. Novak, Chief Reactor Systems Branch Division of Systems Safety

Enclosure:

As stated cc: R. Mattson G. Kelly V. Stello R. Tedesco T. Novak S. Israel G. Mazetis t i e 1765 118 99M

SEQUDYAH NUCLEAR PLANT . FIN AL SAFETY Added by Amendment 58, December 22, 1978 ' ANALYSIS REPORT \\ SMALL LOCA AND LSB VESSEL INTEGRITY ANALYSES 2 DIMENSIONAL FRACTURE " ^ ' ^"^ PLANT LIFE (YEARS) T ble CASE' BS - LG BS - LG WD - CF CU = 0.15% CV = 0.13% CU =. 3B (. 33)% P = 0.011% P. = 0.015% P = 0.021% l RTHDTI,= 40*F RTHDTI = 73 F. RTHDTI : 40*F i 2 INCH 40 SMALL 40 40 LOCA I 3 INCH SMALL 40 40 40 l LOCA 4 INCH 40 SHALL 40 27 LOCA i l 6 INCH 28 j 40 SMALL 40 LOCA ( i ? LSB l ( WITH REACTOR 40 40 COOLANT PURPS 40 RUNNING i LSB WITH REACTOR 40 40 COOLANT PUHPS 40 g n TRIPPED l-BS - BASE HATERIAL LC - LONGITUDINAL FLAW 1765 119 i WD - WELO HATERIAL CF - CIRCUHFERENTIAL FLAW g I }- ANOCid7-

SNP-58 1 w l 6.54 During long term cooling following a steamline break, feedwater line break, or small LOCA, the operator must control primary system pressure to preclude overpressurizing the pressure vessel after it has been cooled off. ~ Describe the instructions given the operator to perform a. long term cooling. b. Indicate and justify the time frame for performing the required action. c. List the instrumentation and components needed to perform this action and confirm that these components meet safety grade standards. d. Discuss the safety concerns during this pericd and the design margins available. This should include pocential adverse hydraulic conditions leading to inadequate cooling-or mech-anical damage, e. Provide temperature, pressure, and RCS inventory graphs that would show the important features during this period. 49 t The above discussion should account for the following: a. loss of offsite power b. operator error or single failure c. small LOCA's may occur in the cold leg or in the hot leg / pressurizer, d. small LOCA's may result in nitrogen blanketing of the steam generators. e. long term cooling for a small LOCA may depend on alternating forced convection and vaporization depending on the break location and size.

Response

The response to this question as submitted on the D.C. Cook Unit 2 docket is an appropriate approach to the generic issues which hav 58 been raised. See D.C. Cook FSAR amendment 78. l 1765 120 Q6.54-1 Decenber 22, 1978

SNP-58 To address the issue of reactor vessel repressurization a fracture mechanics study on the integrity of the Sequoyah Units 1 and 2 reac-The tor vessel beltline under faulted conditions was performed. faulted conditions evaluated were the large steamline break (LSB) and the small loss-of-coolant accident (LOCA). These analyses sup-plement previous studies done for normal,' upset, emergency, and faulted conditions as described in FSAR section 5.2. The LSB transients used for this analysis are generic transients for a UHI four loop plant that have been modified to approximate the impact of pressurizer thick metal heat of the Sequoyah Units. Two LSB transients were evaluated: a case which assumes a loss of off-site power that causes the main reactor coolant pumps to stop (pumps tripped case) and a case where offsite power remains available that allows the main reactor coolant pumps to continue operation (pumps running case). The RCS response for the LSB is shown in Figures Q6.54-1 through Q6.54-5. The transients used in the analysis of small LOCAs were taken from work for the British performed by West-inghouse in 1974. The RCS responses for the 2, 3, 4, and 6 inch diameter small LOCAs are given in Figures Q6.54-6 through Q6.54-13. These RCS responses were used to determine the temperature, thermal t' stress, and pressure stress profiles through the vessel wall in the beltline region as a function of time. these profiles were then used in performing the fracture mechanics analyses. In these analyses the following material properties were used for a longitudinal flaw in the base material: Copper Content = 0.15 weight percent Phosphorus Content = 0.011 weight percent Initial RT'T = 730F For a circumferential flaw, the following material properties of the core region circumferential weldment were used: Copper Content = 0.33 weight percent Phosphorus Content = 0.021 veight percent Initial RT'T = -400F These properties were obtained from the vessel fabrication material test certification for the Sequoyah Units. The fluence used in these analyses was that calculated for a generic four loop vessel similar to the Sequoyah Units and satisfactorily approximates the fluence levels of these units. The irradiation damage of the material is correlated by trend These curves were developed by Westinghouse to relata the curves. magnitude of the shif t of RT'T to the amount of neutron fluence and are a function of copper content. The final RT'T values are then 1765 121 December 22, 1978 Q6.54-2

SNP-58 used to calculate the plane strain fracture toughness (K7@) and the re ference fracture toughness (K7@) as a function of the fractional depth through the vessel vall. A two dimensional combined flaw analysis that is an approximation of a three dimensional flaw is used. The results of this fracture mechanics analysis are presented in Table Q6.54-1. These results are presented in terms of the maxi-mum number of calendar years (.8 load factor is assumed) the plant will conform to the following criteria: Minimum critical flaw is greater than 0.1156 a/t (1.0 inch) or flaw arrest is within 75 percent of the vessel wall thickness. From Table Q6.54-1 it can be seen that for the two dimensional flaw method the vessel integrity can be shown for only about 30 years of plant operating life for two cases. All other cases indicated ves-sel integrity is assured for at least 40 years. For the 4 and 6 inch small LOCAs, the maximum number of calendar years the plant will conform to the vessel integrity criteria is 27 and 28 years respectively. TVA is reviewing plans to perform a 10 CFR 50 Appen-dix C analysis of the Sequoyah vessels prior to the one quarter service life surveillance. This more realistic analysis is fully expected to verify reactor vessel integrity for the full 40-year plant life. Based on the anticipated outcome of the upcoming Appen-dix C analysis and the analyses already performed showing vessel integiry for nearly 30 years, vessel integrity is assured with ade-quate margin for the first 10 years of its service life. To provide guidance to the operator to be alert to the potential for vessel repressurization af ter an accident and also to be able to respond quickly, the plant operating procedures provide explicit instructions. The operator is instructed to be continuously aware of primary system pressure and temperature comparing them to 10 CFR 50 Appendix C pressure-temperature curves for Sequoyah which are provided in the Technical Instructions. The procedures also iden-tify the qualified instruments necessary for this monitoring action. 1765 122 Q6.54-3 December 22, 1978

l ,SEQUDYAH NUCLE AR PL ANT 22, 1978 Added by Amendment 58, December ANALYSIS REPORT SMALL LOCA AND LSB VESSEL INTEGRITY ANALTSES 2 DIMENSION AL FRACTURE MECHANICS ANALYSIS PLANTLIFE(YEARS) TABLE 06.541 CASE BS - LG BS - LG WD - CF 'CU = 0.15% CU = 0.13 % CU =. 38 (. 33)% P = 0.011% P = 0.015% P = 0.021% RTNDTI = 40*F RTHDTI = 73*F RTHDTl " 40"F l t 2 INCH 40 40 40 SMALL LOCA 3 INCH 40 40 40 C',' SMALL LOCA 4 INCH 40 27 40 SMALL LOCA 6 INCH 40 28 40 SMALL LOCA LSB WITH REACTOR 40 40 COOLANT PUMPS 40 RUNNING LSB 40 WITH REACTOR 40 40 COOLANT PUMPS TRIPPED LC - LONGITUDINAL FLAW BS - BASE HATERIAL CF - CIRCUMFERENTIAL FLAW WD - WELD HATERIAL 1765 123

Added by Amendment 58, December 22, 1978 TEMPERATURE VERSUS TlHE 600 400 L. - g u 5Ii 3 200 w 1,765,124 8 8 1 0 3000 1000 2000 TlHE(SECONDS) SEQUOYAH NUCLEAR PLANT FIN AL S AFETY ANALYSIS REPORT LARGE STEAMLINE BREAK WITH REACTOR COOLANT PUMPS RUNNING FIGURE 06.541

Added by Amendment 58, Decec:ber 22, 1978 m PRESSURE VERSUS TIME 2400 1600 7 r u i 5? 0 2 800 1765 125 I I i e O 1000 2000 3000 TlHE(SECONDS) SEQUOYAH NUCLEAR PLANT FIN AL SAFETY ANALYSIS REPORT LARGE STEAMLINE BREAK WITH REACTOR COOLANT PUMPS RUNNING FIGURE 06.54 2

Added by Amendment 58, December 22, 1978 TEMPERATURE YERSUS TIME 600 1400 - C E E5 E5 200 1765 126 I t I 1000 2000 3000 TlHE (SECONDS) SEQUOYAH NUCLEAR PLANT FIN AL SAFETY ANALYSIS REPORT LARGE STEAMLINE BRE AK WITH REACTOR COOLANT PUMPS TRIPPED FIGURE 06.54 3

O Added by Amendment 58, December 22, 1978 m PRESSURE VERSUS TlHE 2%0 e GE ua S E 800 8 1765 127 I I I I I O 1000 2000 3000 TlHE(SECONDS SEQUOYAH NUCLEAR PLANT FIN AL SAF ETY ANALYSIS REPORT LARGE STEAMLINE BREAK WITH REACTOR COOLANT PUMPS TRIPPED FIG URE 06.544

Added by Amendment 58, December 22, 1978 FLOW VERSUS TlHE I.0 NOMINAL REACTOR COOLANT FLOW t 94420 GPH/ LOOP j j 0.8 - i 2 g 0~ 6 o g p y 0.4 Od

0. 2 -

1 I i 0 200 400 1765 128 TlHE(SECONDS) SEQUDYAH NUCLEAR PLANT FINAL SAF ETY ANALYSIS REPORT LARGE STEAMLINE BREAK WITH REACTOR COOLANT PUMPS TRIPPED FIGURE 06.54 5

Added by Amendment 58, December 22, 1978 TEMPERATURE VERSUS TlHE 600 400 - C ~ La 505 a. 5 200 1765 129 I I I I I O 2000 4000 6000 1 TlHE (SECONDS) SEQUOYAH NUCLEAR PLANT FIN AL SAF ETY AN ALYSIS REPORT SMALL LOCA 2 INCH DI AMETER BREAK FIGURE 06.544

Added by Amendment 58, December 22, 1978 PRESSURE VERSUS TlHE 2400 1600 p. S u 3 C E 800 1765 130 L i i i 0 2000 4000 60 % TIME (SECONDS) SEQUDYAH NUCLEAR PLANT FIN AL SAFETY ANALYSIS REPORT SMALL LOCA 2 INCH OIAMETER BREAK l. FIGURE 06.54 7 ?

Added by Amendment 58, December 22, 1978 ~ ~ TEMPERATURE VERSUS TIME 600 400 CL E i" E E5H 200 t I i i t 1744 131 0 2000 4000 6000 TIME (SECONDS) SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT SMALL LOCA 3 INCH DIAMETER BREAK FIGURE 06.544

Added by Amendment 58, December 22, 1978 PRESSURE VERSUS TlHE 2400 1600

c. '7, GA u

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n. 800 I I I g i O 2000 4000 60 % TlHE(SECONDS) 1765 132 SEQUOYAH NUCLEAR PLANT FINAL SAF ETY ANALYSIS REPORT SMALL LOCA 3 INCH DIAMETER BREAK FIGURE 06.54 9

Added by Amendment 58, December 22, 1978 TEMPERATURE VERSUS TIME 600 140 0 C ,' L E R a.5r 200 I I I I O 1000 2000 3000 1765 133 TIME (SECONDS) SEQUOYAH NUCLEAR PLANT FINAL SAF ETY ANALYSIS REPORT SMALL LOCA 4 INCH DIAMETER BREAK FIGURE 06.54 10

Added by Amendment 58 December 22, 1978 PRESSURE VERSUS TlHE 2H)0 ~ 1600

v. ~

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E wg 800 Su n. f I I I I I 3000 2000 1000 } O TIME (SECONDS) 1765 134 SEQUOYAH NUCLEAR PLANT FINAL SAFETY AN ALYSIS REPORT SMALL LOCA 4 INCH DI AMETER BREAK FIGURE 06.54 11

Added by Amendment 58, December 22, 1978 TEMPERATURE VERSUS TIME 600 I 400 ~ D, ', b w x 4 oc E5 r 200 I I I I I 1765 135 600 O 200 400 TlHE(SECONDS) SEQUOYAH NUCLEAR PLANT FIN AL SAFETY ANALYSIS REPORT SMALL LOCA 6 INCH DIAMETER BREAK FIGURE 06.5412

Added by Amendment 58, December 22, 1978 PRESSURE VERSUS TIME i 2%0 1600 l 7a 5 !3 = 800 1765 136 l I I e 0 200 400 600 TlHE(SECONDS) SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT SMALL LOCA 6 INCH DIAMETER BREAK FIGURE 06.54 13

561.' REGULATORY INFORMATION DISTRIBUTION SYSTEM DOC DATE: 791n16 DOCKET NBR:050-029 YANKEE ROWE ACCESSION NBR:791026016n RECIPIENT: MILLER, W.O. COPIES RECEIVED: ORIGINATOR:VANDENBURGH, D. LTR 1 _ ENCL 1 COMPANY: YANKEE' ATOMIC ELEC SIZE:5

SUBJECT:

Objects :to continuous NRC demands that util meet addl criteria re overpressurization events, while NRC review process is uniustifiably lengthy & results only in reauests for addl info. Believes $4000 fee request is unauthorized. NOTARIZED'- DISTRIBUTION CODES A001 DISTRIRUTION TITLE GrNERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICEN3E. FOR ACTION NAME ENCL 7

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"I BR CHIEF a/7 ENCL W/ ENCL REG FILE NRC Pnp w/ ENCL I & E W/P ENCL OELD tTR nNLY HANAUER W/ ENCL CORE PERFnR"ANCE BR w/ ENCL AD CDD SYS & DRDJ w/ ENCL 6 GIN M y w/ ENCL MtacInD SAFETY BR W/ ENCL PLANT SYSTE*S RR

• / ENCL EER w/ ENCL EFFLUENT TREAT SYS N/ ENCL J MCGnttGH w/ ENCL LPOR W/ ENCL TERA w/ ENCL NSIC w/ ENCL ACRS w/tb ENCL REBA DIGGS W/ ENCL TOTAL NUMPER OF CORIES RECUIREDI LTR u0 ENCL 39 1765 137

~ NOTES: g

... ~ e Telephone 617 366-9011 rw 700 390 0739 YANKEE ATOMIC ELECTRIC COMPANY ,Y E xts 20 Turnpike Road Westborough, Massochusetts 01581 m wYR 78-8s October 16, 1978 RECEIVED E.Y WiS United States Nuclear Regulatory Commission Date.M9 $'N... Washington, D. C. 20555 , Time.. b..~.... I D-Attention: William O. Miller, Chief Ro m....... Licensing Fee Management Branch Office of Administration l Ci ;0 -

. c..u n %n

t..........

References:

(a) License No. DPR-3 (Docket No. 50-29) (b) YAEC Letter to USNRC dated June 5,1978 (WYR 78-46) (c) USNRC Letter, Daniel J. Donoghue, to Licensee, dated February 28, 1978 (d) USNRC Letter to YAEC dated October 2, 1978, certified mail to D. E. Vandenburgh

Subject:

License Amendment Fee for Proposed Change No.161 (Reference b) It has always been Yankee's practice to operate its plants in a safe and responsible manner. Long before the NRC expressed any outward concerns regarding overpressurization events, Yankee, through detailed operating procedures, would not permit any operator action which could develop into a potential Low Temperature Over-Pressurization (LTOP) event. Our operating record is a testimony of our commitment to this philosophy, as Yankee Rowe has never experienced a LTOP transient in its seventeen years of operation. During the past two years the staff has escalated once common utility /NRC concerns regarding potential overpressurization events to the point where the only acceptable solution is a full commitment to design and install costly systems, in order to meet evolving cri, 7 g g 3JU zo teria. The staff has taken the position that regardless of current 1 I U J l plant designs and procedural commitments, utilities must redesign to mitigate a LTOP event, rather than prevent its occurrence. Alternate plant specific designs are ignored by the staff as they continue develop-ing additional criteria. Yankee takes exception to this practice, and has offered for review alternative designs which meet the intent of your staff's criteria. Following an NRC/ utility meeting the staff's growing concerns over LTO DUPLICATE DOCUMENT procedures were again reviewed and up controls were issued to further preel Entire document previously possibly result in a LTOP; and we sub entered into system under: requirement (then current practice) t qQfa g/ tor in the cc,ntrol room, whenever the ANO / U/W WI No. of pages: 7h0M o!Go

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DRAFT MAycock 6/18/79 TABLE OF CONTENTS - A-ll NUREG I. Introduction and Summary A. Description of Problem B. Discussion of its Applicability to operating plants and newer plants C. Sun: mary of information and results provided in the report. Overview 01 approach to thE problem (proVides logic of presentation of remainder ofNUREG). II. Materials Properties A. Summary of data gathered in operating plants and description of data ccmputerization. B. Summary of condition of operating vessels (in terms of NDTT or some-thing). C. Discussion of irradiation effects studies. D. Conclusions regarding materials properties to be used in calculations of vessel toughness. III. Consideration of Postulated Accidents and Transients A. Discussion of past treatment of accidents ar:d transients B. Disctssion of HSST Program and the relevance of its results C. Discussion of Event Sequences that have potential for challeng-ing vessel integrity D. Description of those event sequences that should be considered in developing loads for use in calculations of vessel toughness 1765 146

a IV. Acceptance Criteria A. Acceptance criteria for normal operation (safety margins) B. Acceptance criteria for postulated accidents and transients (includes a discussion of probabilities of the event sequences developedinIII). V. Calculations of Toughness for Low Toughness Vessels A. Description of Calculational Technique developed by Paris, et al B. Discussion of experimental verification of the validity of the technique C. Conclusions regarding its use in licensing actions D. Recommendations for Rule Making VI. Discussion of Possible Methods of Improving Vessel Toughness or Limiting Toughness Degradation A. Discussion of the Feasibility of Annealing (chances of success, environmental aspects, etc.) B. Discussion of the possibility of Core Mods - European (Gennan?) Experience VII. Conclusions and Recomendations Appendices A. Summary of Operating Data B. Proposed Branch Technical Position re: Consideration of postulated accidents and transients C. Others as needed 1765 147

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