ML20216B611

Forwards Response to NRC 980427 RAI to Assist Review of 980325 TS Change Under Exigent Circumstances.Topical Rept Cenc 1139, Analytical Rept for Pilgrim Reactor Vessel, Encl

ML20216B611
Person / Time
Site:	Pilgrim
Issue date:	05/05/1998
From:	Desmond N BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20216B615	List: ML20216B611 ML20216B631
References
BECO-2.98.063, NUDOCS 9805180272
Download: ML20216B611 (7)

Text

' '

10 CFR 50.90 FR 50.91 0 1 Boston Edisore Pilgrim Nuclear Power station Rocky Hill Road Plymouth, Massachusetts 02360-5599 May 5,1998 Nancy L. Desmond Regulatory Relations Group Manager BECo Ltr. 2.98.063 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington DC 20555 Docket No. 50-293 License No. DPR-35 Response to NRC Reauest for AdditionalInformation Reaardina PNPS Proposed Technical Specification 3.6.A.1 By letter dated March 25,1998, (BECo Ltr. 2.98.023), Boston Edison Company (BECo) proposed a change to PNPS Technical Specification 3.6.A.1 to eliminate a 145 F differential temperature limit between the reactor vessel flange and adjacent vessel shell.

BECo letter dated April 8,1998, (BECo Ltr. 2.98.043), requested Nh1 review of the above submittal under exigent circumstances.

The NRC submitted a request for additional information (RAl) in NRC letter dated April 27,1998, to assist the review under exigent circumstances.

In response to the NRC RAI, BECo is providing the following enclosures to this letter:

Enclosure 1 Response to NRC Questions on Reactor Vessei DT T. S. Change Enclosure 2 Applicable Pages of Combustion Engineering Report No. CENC 1139,

" Analytical Report for Pilgrim Reactor Vessel", dated March 9,1971.

Please refer any questions regarding this submittal to C.S. Brennion, Sr. Regulatory Affairs Engineer, at (508)830-8674. j (

RQ.

'en N i hsw n ** Y rv N.L. Desmond Attachments: Enclosure 1 f Enclosure 2 '/

CsB/deg /'

.3 l

298063 SL ,'

7 9805180272 980505 PDR ADOCK 05000293 P PDR i

l -

Regional Administrator, Region 1 Peter LaPorte, Director l U.S. Nuclear Regulatory Commission Massachusetts Emergency Management i Agency Office of Emergency Preparedness i

475 Allendale Road 400 Worcester Road King of Prussia, PA 19406 P.O. Box 1496 Framingham, Ma. 01701-0317 Senior Resident inspector Pilgrim Nuclear Power Station '

l

Mr. Alan B. Wang l Project Manager Project Directorate 1-3 Office of Nuclear Reactor Regulation Mail Stop

OWFN 1482 1 White Flint North 11555 Rockville Pike

! Rockville, MD 20852 l

l Mr. Robert M. Hallisey, Director Radiation Control Program Center for Communicable Diseases Mass. Dept. of Public Health 305 South Street

! Jamaica Plain, MA 02130 l

j

~ '

ENCLOSURE 1 1

RESPONSE TO NRC QUESTIONS ON REACTOR VESSEL DT TS CHANGE l

NOTE: The details of the thermal / structural analysis of the Pilgrim reactor vessel ;

(Combustion Engineering report CENC-1139) have been attached to this  ;

response. This report should provide additional details regarding the i analysis of the reactor vessel closure region and will assist in providing answers to the four NRC questions.

1. Question: DT Calculations - Page 4 of Attachment indicates that the flange to adjacent shell DT is -83 F (corresponding to a stress level of 38 ksi) for cooldown per CENC 1139 and -166 F (corresponding to a stress level of 80 ksi) per M-778. Confirm that they represent results due to the sole effect of the cooldown transient as was suggested in Attachment A.

Provide the cooldown rates for both calculations. Provide also the stress due to the flange bolt load and the stress due to the vessel pressure during cooldown. Assess the impact of the DT's due to these additional stresses.

Answer: The stress levels are not due solely to thermal effects. For example the 38 ksi maximum range of stress intensity at the outside surface of body 4 at cut V, the intersection of the flange and adjacent shell, is derived from the combined effects of forces and moments caused by applied pressure (as appropriate), flange reactions due to bearing (identified as load "B"), bolt load ("F") consisting of the variable portion of bolt load due to relaxation from pressure and thermal effects (load "V or Fc"), and the initial tensile load (preload "Fp"), plus O-ring gasket seating load ("Po"), including the pressure associated with seal leakage ("po").

Eccentricities in these load applications force the flange and shell to displace and rotate radially. The load descriptions are addressed on sheets 19,35,51, and 60 of section S-101 of the CE stress report (CENC-1139-attached). The loads are tabulated on sheets 61 through 63 of this report. A discontinuity moment ("Mv") is developed which takes into account all the above mentioned loads plus discontinuity loads associated with differential thermal expansion obtained from the unit thermal loads applied to the " Seal-Shell" structural model. The moment contributions are listed in the table sheet 65 of S-101. This moment produces a bending I stress which makes up the 38 ksi stress reported in both CENC-1139 and M-778. The various load combinations are applied as required by the '

ASME code (Section Ill) . The load combinations considered for the i maximum range of stress intensity are the highest positive stress '

combined with the lowest negative stress for design operation, design bolt  !

tension, design pressure test, steady state operation, inner seal leakage, vessel overpressure, startup, shutdown flooding, end of shutdown, rapid heating, rapid cooling, hydro test bolt tension, hydro bolt preload, and hydro pressure test. The load combinations are described mathematically in Section S-101 of the attached CE stress report, CENC-1139, and are reproduced in the BECo analysis (M-778).

Page I or4 I

[

l The stresses due to the flange bolt load and stresses due to the vessel pressure are also addressed in this section of the CE stress report. The i impact of the DT's due to these additional stresses are summarized on pages 79 through 82 of CENC-1139 Section S-101. The BECo analysis followed the same approach as the CE stress report but with the exception that the DT's are higher. The effects of differential thermal growth between the flange and adjacent shell were extracted from the closure region interaction analysis of CENC-1139. The difference between the CE analysis and the BECo analysis is that the differential thermal growth portion or contribution to secondary stress was extrapolated out to the maximum 3Sm limit (80 ksi) allowed. The increase in secondary stress between 38 and 80 ksi is due solely to differential expansion effects between the flange and adjacent shell considering an increase in DT of 83 F. Since the method of CENC-1139 and BECo analysis M-778 is identical, only differential thermal growth between the flange and adjacent shell, (i.e., the restraint of free differential thermal expansion) is required to be superimposed on the 3Sm limit. The "heatup/cooldown conditions which result in the maximum differential thermal growth for the BECo M-778 analysis are not necessarily limited to any specific normal heatup or cooldown ramp rate. However,100 F/hr. is considered the upper bound limit.

2. Question: Heatup Rate Greater Than 100 F/hr. - Page 3 of Attachment A indicates that heatup rates of greater than 100 F /hr. are assumed to be step temperature changes.... Mathematically, what does this mean?

Physically, what makes the rate of 100 F /hr. the separation point for different analytical treatments for cases with rates larger and smaller than 100 F /hr.

Answer: Thermal loads were derived from the coolant temperatures defined by specification (i.e., GE Spec. 22A1110). The operating conditions are evaluated as follows:

Quoting from Section S-101 sheet 20 of the CE stress report :

e) Heatup rates of greater than 100 F/hr. are assumed to be step temperature changes and are combined into a composite condition called " rapid heating." The structure is assumed to be at the minimum fluid temperature of any of the included conditions, and the surface in contact with the coolant is taken as the maximum temperature of the included conditions.

f) ' Rapid cooldown' is used to define conditions with cooldown rates greater than 100 F/hr. These temperature changes are treated as step changes with the structure assumed at the maximum temperature of the included conditions, and the surface in contact with the coolant at the minimum temperature of the included conditions."

The above combined " rapid heatup/cooldown" conditions were analyzed by Combustion Engineering (Ref. CENC-1139 ) and, based on the limited range of the fluid step temperature change, were found not to govern for Page 2 of 4

the maximum range of primary plus secondary stress intensity. Step l changes in fluid temperature generally result in locked in thermal (skin) l stress which are considered in the fatigue evaluation whereas heat /cooldown ramps result in stresses from restraint of free differential thermal expansion between bodies. Holding the metal temperature

constant and applying an instantaneous step change in fluid temperature results in stresses which are less than those calculated during the normal i heatup or cooldown conditions because these ramps and vessel response ,

experience larger temperature excursions over a longer period of time. The mathematical expression is described in the CE stress report (CENC-i 1139) Appendix B. The 100 F/hr ramp rate is the upper bound limit for normal heatup and cooldown specified in the Tech. Specs. This ramp rate is an administrative limit established by the NSSS and by regulation. The reactor vessel, however, is not physically limited to this ramp rate. The vessel can withstand higher ramp rates without exceeding structural or fracture toughness limits.

3. Question: Linear and Mean Body Temperatures - Page 12 of Attachment D indicates that the linear and mean body temperatures are needed in the thermal analyses. Refer to the figure on Page 17 and explain the size of these bodies or components for which you have performed the calculations of the linear and body temperatures. Have you derived the linear and mean temperatures separately for the heatup and cooldown transient?

Answer: The size of the bodies can be found on the attached drawings and excerpts provided in the attached CE Stress report. Liner and mean body temperatures were calculated separately for heatup and cooldown. l The mean and linear body temperatures are weighted averages based on the temperature profiles provided in the CE stress report for "End-of-Heatup" and "End-of Cooldown". The weighted distributions reported in BECo report M-778 were calculated independently from the temperature ,

profiles provided in the CE stress report and are in good agreement with I the body temperatures listed in the CE stress report (CENC-1139). I

4. Question: Structural Modeling - Page 12 of Attachment D shows the modeling of the structure. Is the model axisymmetric? What are the l boundary conditions on both ends of the structure?

i Answer: The model is axisymmetric about the vertical centerline (x l direction) of the reactor vessel. The "r" or radial direction is transverse to the axial centerline and "q" represents the hoop or tangential direction.

The " structural" boundary conditions are as follows.

l l

Page 3 of 4

a) Too boundary condition: The upper head (Body 1) is a "long" sphere.

The top of the sphere (crown) is free to move in the axial (x) direction l relative to the reactor vessel vertical centerline but restrained from l displacement and rotation in the radial (r) direction. For specific loading l conditions such as pressure, a blowoff load (when appropriate) is

applied to the sphere which represents the upward force due to internal pressure and is necessary for equilibrium. l l

l b) Bottom boundary condition: The lower shell (Body 4) is a "long" cylinder. The bottom edge of the cylinder is free to move in the radial i (r) direction and restrained in the axial (x) direction. The bottom edge is i also restrained from rotation in the three orthogonal directions. The t

boundary condition is required for restraint against rigid body motion.

l l NOTE: The upper (spherical) and lower (cylindrical) boundary conditions do not influence the body movements or redundant edge loads at the intersections of the elements which make up the closure region because there is no interaction between the near and far edges since l

the distances exceed one attenuation length. (Ref. " Theory of Plates and Shells" by S. Timoshenko) l l

1 Page 4 of 4

ENCLOSURE 2 APPLICABLE PAGES OF COMBUSTION ENGINEERING REPORT NO.1139, ,

" ANALYTICAL REPORT FOR PILGRIM REACTOR VESSEL", DATED 3/09/71 1 Number of pages provided 177.

e.g., Abstract 1.000 Introduction l 2.000 Design Criteria 3.000 Vessel Geometry 4.000 Summary of Detailed Analysis 5.000 References l

Appendix A Detailed Structural Analysis (Closure Region) l Appendix B Thermal Analysis l (Closure Region)

(Vessel Shell) l l

t l

l l

l I

1 of 1

.