ML14188B880

Notification of 820224 Meeting W/Westinghouse Owners Group & Utils in Bethesda,Md Re Pressurized Thermal Shock Issue & plant-specific 150-day Responses

ML14188B880
Person / Time
Site:	Turkey Point, Robinson, San Onofre, 05000000
Issue date:	02/18/1982
From:	Vissing G Office of Nuclear Reactor Regulation
To:	Stolz J Office of Nuclear Reactor Regulation
References
TASK-2.K.2.13, TASK-TM NUDOCS 8203040422
Download: ML14188B880 (9)

Text

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REG4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 February 18, 1982 Dockets Nos.

50-261, 50-251 and 50-206 MEMORANDUM FOR:

John Stolz, Chief, Operating Reactors Branch #4, DL

)

FROM:

Guy Vissing, Project Manager, Operating Reactors Branch #4, DL

SUBJECT:

FORTHCOMING MEETING WITH WESTINGHOUSE OWNERS GROUP CAROLINA POWER & LIGHT COMPANY, FLORIDA POWER COMPANY AND SOUTHERN CALIFORNIA EDISON, CONCERNING PRESSURIZED THERMAL SHOCK ISSUE Time &"ODate:

'i8):v5am-5:00pm ed'nesday, February 24, 1982 iLocation -oliday Inn, Bethesda, MD

-3rd floor, Maryland Room Pu pose :

T dqiscuss the Westinghouse Generic Report followed by discussions on the plant specific "150 day" responses for Robinson 2, Turkey Point 4 and San Onofre 1.

See attached agenda.

Requested

Participants:

NRC: H. Denton, S. Hanauer, T. Novak, W. Johnston R. Woods, R. Klecker, W. Hazelton, C. Johnson, R. Johnson, J. Clifford, L. Lois, N. Randall, E. Throm, C. Morris, C. Serpan, A. Oxfurth, M. Vagins, J. Strosnider.

Licensee: WOG-Daniel Speyer.etlal.

Southern California Edison Carolina Power Company Florida Power Company Other: Westinghouse representatives Guy Vissing, Project Manager Operating Reactors Branch #4 Division of Licensing cc:

See next page 8203040422 820218 PDR ADOCK 05000206

MEETING NOTICE DISTRIBUTION OPERATING REACTORS, DIVISION OF LICENSING Dock~et File Regional Administrator U. S. N. R. C., Region L PDR TERA NSIC ORB#4 Rdg JStolz Project Manager TNovak JHeltemes, AEOD 0ELD IE-3 OSD-3 SShowe, IE (PWR) or CThayer, IE (BWR)

Licensing Assistant 9

Receptionist, Bethesda ACRS-10 Program Support Branch PTKuo (seismic reviews only)

HGaut, State Programs RMa tts on ORAB, Rm.

542 Meeting Notice File BKGrimes, DEP SSchwartz, DEP FPagano, EPLB SRamos, EPDB MJambour NRC

Participants:

AGENDA FOR MEETING WITH WESTINGHOUSE OWNERS GROUP AND THE WESTINGHOUSE SSS OWNERS WHO RECEIVED THE AUGUST 21, 1981 LETTER.

CONCERNING PRESSURIZED THERMAL SHOCK FEBRUARY 24, 1982 I. Session Concerning Westinghouse Generic Report WCAP-10019

1. Introduction

2. Summary of Basic Staff Concerns:

Information is needed regarding sensitivity of remaining EFPY to principal parameters including:

a) Operator delayed action b) Decrease credit for warm prestressing and uncertainities in material perperties.

c) Decrease thermal mixing credits See attachment fdr detailed concerns

3. Owners Group Responses (Owners Group opportunity to question staff for clarifications and to discuss concerns).

4. Owners group Caucus - if necessary.

5. Identification of what more information will be provided and a schedule for submittals by WOG.

6. Concluding Remarks.

II. Session Concerning 150 Day Responses

1. Introduction

2. Staff concerns (see attachment).

3. Licensee's Responses (opportunity for licensees to question staff for clarifications and discuss concerns.

a) Robinson 2 b) Turkey Point 4 c) San Onofre 1

4. Caucus if necessary

5. Identification of what more information will be provided and schedule for submittals by licensees.

6. Concluding Remarks.

CONCERNS RELATED TO OPERATOR ACTIONS A. Applicable to Generic Report WCAP-10019

1. In WCAP-10019, an analysis of an isolatable LOCA was conducted to study the effect of operator response time on vessel integrity to demonstrate that sufficient time exists to allow operator isolation of the PORV before the vessel is challenged. The presentation in WCAP-10019 shows that if the operator isolates the PORV in 30 minutes the vessel would not be challenged, but does not show that 30 minutes is sufficient time for the operator to take'action. Provide justification that 30 minutes is sufficient time for correct operator action. Provide an evaluation of the sensitivity 30 minute ooerator time.

2. In the WCAP-10019 Large Steam Line Break analysis, it is assumed that the operator terminates AFW and injection flow at 10 minutes.

Control of AFW does not seem to significantly affect the cooldown.

Injection flow termination, however, appears to occur at a time critical to prevent significant repressurization. Provide an evaluation of the sensitivity of the time assumed for operator action (i.e., if the operator acts at 15 minutes, or 20 minutes, or 30 minutes, what are the resulting pressure/temperature transients?).

B. Applicable to Licensees 150 day Responses

3. In the San Onofre 1, the H. B. Robinson 2, and the Turkey Point 3 and 4 evaluations, the actions described do not provide the operator with clear direction for dealing with the conflicting.

concerns that need to be evaluated when considering the operation of HPI and charging flow as it relates to vessel integrity and maintaining core cooling. Each licensee should provide an evaluation of the need and effectiveness of procedure modification to clearly identify the concerns in the emergency operating procedures themselves, in addition to upgrading operator training.

4. Carolina Power and Light stated that a formal training program will be completed by March 31, 1982, and Southern California Edison indicated that training would be performed.in February 1982.

The programs described by the two licensees are adequate to address the NRC's concerns for short term action, except that the procedureal guidance provided in the training program may require emergency operating procedure modification to clearly define the conflicting concerns.

5. Florida Power and Light does not consider any procedural modifications or upgraded training programs necessary. The licensee should be advised that operator knowledge of current industry information regarding Pressurized Thermal Shock, and clearly outlined options in emergency operating procedures, could prove very valuable if an unforeseen condition challenges the vessel integrity.

CONCERNS RELATED TO FRACTURE ANALYSIS Applicable to WCAP 100-19 A. Justification for taking credit for warm prestressing for small break LOCA and other transients. 8ow can it be assured that pressure will not fluctuate up (if the system is repressurized) or down to such an extent that warm prestressing is negated.

Both excursions up and large excursions down may negate warm prestressing. Operating experience has shown us that such excursions do in fact occur during transients.

B. Justification for Assumptions:

1. No clad contribution to K thermal

.I

2. Semielliptical crack, initially C. Describe steps in the analysis of small steam break -

show curves of:

1.

Metal temperature through the wall at critical times

2. K, KIc, and KId as functions ofd/t at critical times, and

3.

"Football curves" for fluence values of interest.

Applicable to WCAP 10019 and 150 day responses

1. Initial RT

- Generic values for typical metal wire and flux types could gene y values be obtained?

CONCERNS RELATED TO SYSTEM ANALYSIS A. Applicable to Generic Report WCAP-10019

1. Concerning the Thermal-Hydraulic Complete Program Provide a description of the models used to:

1.1 Evaluate the ECC mixing in the cold leg and downcomer.

1.2 Evaluate the non-symmetrical temperature distribution in the downcomer (resulting from blowdown of a steam generator).

1.3 Evaluate the primary-to-secondary heat transfer (and reverse heat transfer).

1.4 Evaluate how voiding in the primary system and the subsequent collapse of primary system voids is treated.

1.5 Evaluate the repressurization of the primary system.

1.6 Evaluate any other thermal-hydraulic phenomena important to the PTS problem.

1.7 Provide a description of the verification of the thermal-hydraulic computer program to applicable experimental data for repressurization and overcooling transients.

1.8 What additional verification is required to demonstrate that the thermal-hydraulic computer program adequately models the phendmena important to PTS evaluations.

2. Concerning the mixing models.

2.1 Provide the references for the mixing computer programs.

2.2 Provide a description of any changes made in order to treat water as the working fluid.

2.3 Provide a description of the input data used to perform the mixing.

analysis. Of particular importance are any coefficients supplied by the user to treat convective terms.

2.4 Provide a description of the limitation of the mixing model (with respect to relative velocities, fluid states, and flow regime, e.g.

annular, slug, stratified).

2.5 Provide verification of the mixing program with suitable experimental data covering a wide range of injection and coolant flow rates, as might be expected for potential PTS scenarios.

2.6 Provide a description of the method used to determine when "non-mixing" needs to be considered to a PTS overcooling event.

B. Applicable to Generic Report and Licensee's "150 day" Responses 1.

Concerning Input Data and Assumptions 1.0 Provide a description of the models or data used for:

(a) Heat sources (or sinks),

(b) Decay heat, (c) ECC and feedwater temperatures (enthalpies) and flow rates, (d) Primary and secondary relief capacities, (e) Empirical correlation coefficients used for PTS evaluations, (f) Operator Actions, (g) Initial conditions 1.2 Provide a list of all transients or accidents by class (for example:

excessive feedwater, operating transients which result from multiple failures including control system failures and/or operator error, steam line break and small break LOCA) which could lead to inside vessel fluid temperatures of 300 F or lower. Provide any Failure Modes and Effects Analyses (FMEAs) of control systems currently available or reference any such analyses already submitted.

Estimate the frequency of occurrence of these events and provide the basis for the estimates. Discuss the assumptions made regarding reactor operator actions.

For a given initiating event, potential multiple and consequences failures need to be considered to identify those transients which could lead to a PTS problem.

1.3 Identify all potential PTS events which have occurred at your facility. Include a designation of the operator actions and identify potential additional failures (including operator) which could have resulted in a more severe event.

CONCERNS RELATED TO IRRADIATION INFORMATION A. Applicable to Generic Report WCAP-10019

1. WCAP-10019 Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants. The material covered in this review is (a) Fluence Methodology pp 22-42 and Fluence calculations pp 126-127.

1.1

p. 26, What is the fundamental cross section set used to generate the 21-group cross sections and the zone dependent spectra? How is the P1 scattering expansion justified, particularly for Fe?

1.2 p. 29, It is stated that "...radial power distributions applicable to long term operation are derived from a statistical analysis of calculated distributions..."

Bias factors are used for the correction of observed to calculated differenced. (a) How are these biases applied and how large are they? and (b) With the trend to low leakage loadings where will the data base be found to calculate bias factors for the outer limits?

1.3 p. 30, To what extent is the common axial distribution applicable to 4, 3 and 2-loop plants?

1.4 p. 30 and p. 38, An uncertainty of +20% is given. What are the components and how are they combineTto yield this overall value?

1.5 p. 38, Data bases for support of long term core power distribution and measurements in the reactor cavity are mentioned.

(a) What are these data bases?

(b) Does the data bases include low leakage loading? and (c) (also p. 41) from what W-plAnts are the reactor cavity measurements?

1.6 pp 38-41 Saturated activity of Fe-54, Ni-58, Np-237 and U-238 from ten 2-loop and eight 4-loop plant surveillance capsule are used in tables 11.2-2 and 11.2.3. Are these the only available data?

1.7 p. 41, Is the uncertainty level of +20% applicable to the single plant predictive uncertainty?

B. Applicable to Licensee's 150 Day Response

2. The licensee submittals for the W-plants are essentially identical with no plant specific details. The following questions are applicable to H. B. Robinson (HBR), Turkey Point-3 (TP), and San Onofre (SO).

2.1 The power distributions used for fluence predictions are said to be "statistically based." What is the statistical basis for

HPB, TP, SO and how was the plant specific information and data accounted for?

2.2 The rod by rod power distributions do not reflect plant specific information. Are these distributions bounding so as to be acceptable for a conservative estimate of the fluence?

2.3 Is the geometic information on the core plant specific with as built demensions?

2.4 For future low leakage operation will there be a sufficient statistical base data for reliable fluence prediction?