ML18003A838
| ML18003A838 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/22/1982 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | Utley E CAROLINA POWER & LIGHT CO. |
| References | |
| NUDOCS 8301070356 | |
| Download: ML18003A838 (27) | |
Text
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Docket Nos.:
50-400 and 50-401 gEG 2 2 l982 hr.
E.
E. Utley Executive Vice President Carolina Power 8 Light Company Post Office Box 1551
- Paleigh, North Carolina 27602
Dear IIr. Utley:
DISTRIBUTION Document Control 50-400/
'LPDR NSIC LB¹3 Reading JLee NPKadambi
- Attorney, OELD ACRS (16)
ELJordon JNTaylor RBosnak
Subject:
Request for Additional Information - Shearon Harris
Reference:
Letter from G.
M. Knighton, NRC to E.
E. Utley, Carolina Power and Light Company "Confirmatory Piping Analysis and Design Documentation Review for SHNPP by NRC IIEB" Dated December 10, 1982 As part of the staff's review of the Shearon Harris FSAR, the Viechanical Engineering Branch has identified i tems of needed additional information.
The enclosure contains a listing of the items together with the applicable section numbers.
This enclosure is substantially the same as one which has already been informally communicated to you.
From past experience of reviews in this area, it has been found advantageous to fulfillthe information needs using a meeting which brings tooether appropriate people from the NSSS vendor, the AE and your staff.
The discussion would focus on theiissues raised by the items in the enclosure.
Me request that such meeting be held starting on February 1,
1983 at the Ebasco offices in New York.
In addition to the resolution of questions in the enclosure, the meeting would have a parallel objective of accomplishing the Design Documentation Review described in the referenced letter.
Please inform us by January 14, 1983 of your concurrence for this meeting.
In
- addition, you are requested to prepare an agenda for the meeting and transmit to us any information which might aid in the efficient accomplishment of the objectives.
Please make available the ag'enda and any new information at least one week before the meeting.
830i07035h 82l222 PDR ADOCK 05000400 A
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OFFICEI SURNAME $
DATE $
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You may contact Dr. N. P.
Kadambi of my staff if you have any questions.
Siincerely,
Enclosure:
As stated George ll. Knighton, Chief Licensing Branch Ho.
3 Division of Licensing cc:
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Mr. E.
E. Utley Executive Vice President Power Supply and Engineering and Construction Carolina Power
& Light Company Post Office Box 1551
- Raleigh, North Carolina 27602 cc:
George F. Trowbridge, Esq.
Shaw, Pittman, Potts Trowbridge 1800 M Street, NW Washington, DC 20036 Richard E. Jones, Esq.
Associate General Counsel Carolina Power
& Light Company 411 Fayetteville Street Mall
- Raleigh, North Cardlina 27602 M. David Gordon, Esq.
Attorney Associate General State of North Carolina" P. 0.
Box 629 Raleigh, North Carolina 27602 Thomas S. Erwin, Esq.
115 W. Morgan Street Raleigh, North Carolina 27602 Mr. George Maxwell Resident Inspector/ Harris NPS c/o U.S. Nuclear Regulatory Commission Route 1, Box 315B New Hill, North Carolina 27562 Charles D. Barham, Jr.,
Esq.
Vice President
& Senior Counsel Carolina Power
& Light Company Post Office Box 1551
- Raleigh, North Carolina 27602 hr.
John Runkl'e,'xecutive'oordinator onservation Council of North Carolina 307 Granvi lie Road Chapel Hill, North Carolina 27514
<r. Wells Eddleman 18-A Iredell Street Durham, North Carolina 27705 "ir. George Jackson Secretary Environmental Law f roject School of'aw, 064-A University of North Carolina Chapel Hill, North Carolina 27514 Dr. Phyllis Lotchin 108 Bridle Run Chapel Hill, North Carolina 27514 t
Mr. Travis Payne, Esq.
723 W. Johnson Street P. 0.
Box 12643
- Raleigh, North Carolina 27605
~ Mr, Daniel F.
- Read, President CHANGE
,P. 0.
Box 524 Chapel Hill, North Carolina 27514 Ms. Patricia T.
Newman, Co-Coordinator Mr. Slater E.
Newman, Co-Coordinator Citizens Against Nuclear Power 2309 Weymouth Ct.
- Raleigh, North Carolina 27612 Richard D. Wilson, M.D.
725 Hunter Street Apex, North Carolina 27502 Regional Adminstrator - Region II U. S. Nuclear Regulatory Commission 101 Marietta Street Suite 3100 Atlanta, Georgia 30303
ENCLOSURE REQUEST FOR ADDITIONAL INFORNTION RESULTING FROM REVIEW OF SHEARON HARRIS FSAR DOCKET NOS 50-400 5 50-401
210. 0 MECHANICAL ENGINEERING j
g 210.01 3.2.1.1, Table 3.2.1-1 What code was used in the design of reactor vessel internals?
Why is there no quality group required for the reactor vessel internals?
~ 'y g 210.02 3.2.1.1, Table 3.2.1-1, Pages 3.2.1-28, 29 Several waste processing system components that are identified safety class 3 are not seismic Category I.
Explain this apparent inconsistency.
g 210.03 3.2.1.2, Page 3.2.1-2 Identity safety class 2 systems or components that are part of the reactor coolant pressure boundary.
g 210.04 3.6.1.2.3, Page 3.6.1-2 Provide details of the portions of the safety injection system that you have excluded for break and, through-wall leakage cracks by reason of not being normally pressurized.
g 210.05 3.6.1.2.4, Page 3.6.1-10 The criteria you have used for the effects of jet impingement forces is intended for postulating the effects of unrestrained whipping pipe.
Provide justification for applying this criteria to the effects of jet impingement.
0 210. 06
- 3. 6. 1. 3 Specify the assumed damage by an unrestrained whipping pipe to an impacted pipe of equal size with thinner wall thickness.
3.6. 2 Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 210-1
frt 44 h
g 210.07 3.6.2.1.1.2, Page 3.6.2-1 Branch Technical Position MEB 3.1 requires that pipe rupture in Class 1 piping in areas other than containment penetration areas be postulated at:
(a) terminal ends.
(b) intermediate locations where the maximum stress range as calculated by Eq.
(10) and either (12) or (13) exceeds 2.4 Sm.
(c) intermediate locations where the cumulative usage factor exceeds
- 0. 1.
Revise your ASME Section III Class 1 piping break postulation criteria to conform to this position.
g 210.08 3.6.2.1.1.2, Page 3.6.2-2 Clarify your position with respect to branch connections being considered terminal ends.
What is meant by "two overlapped models?"
g 210.09 3.6.2.1.1.3, Page 3.6.2-2 It is the staff's position that breaks should be postulated at all terminal ends in ASME Class 2 and 3 piping, excluding piping in containment penetration
- areas, regardless of whether or not they are adjacent to the protective structure.
Change your FSAR to conform to this criteria.
g 210.10 3.6.2.1.2, Page 3.6.2-2 The break exclusion region for the main steam line should only extend to the inboard or outboard isolation valves.
Modify your break criteria to include main steam piping between the outboard isolation valve and the f'irst pipe rupture restraint.
g 210.11 3.6.2.1.4, Page 3.6.2-5 Justify not evaluating pipe whip and jet impingement loads for main steam and feed water lines in the steam tunnel.
2l'0-2
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g 210.12 3.6.2.2 Insufficient detail for a complete review of your dynamic analysis of jet thrust exists.
Provide information regarding your time-dependent function representation of the jet thrust force, your assumptions concerning rise time, and the time variation of the jet thrust forcing function's relation to pressure, enthalpy and volume of the fluid in any existing upstream reservoir.
g 210.13 3.6.2.2.3, Page 3.6.2-9 In order for the staff to complete its review of FSAR Section 3.6.2, more detail of the methods used to perform piping dynamic analysis is required.
Specifically, the following information is required:
(1)
The loading condition assumed prior to rupture.
(2)
Methods employed to account for the effects of:
a.
Mass inertia and stiffness b.
Impact and rebound c.
Elastic and inelastic deformation of piping d.
Support boundary conditions (3)
A representative mathematical model of the piping system or piping and restraint system.
(4)
The analytical method of solution selected.
(5)
Solutions for the most severe responses among the piping breaks analyzed.
(6)
Solutions with demonstrable accuracy or justifiable conservatism.
The extent of mathematical modeling and analysis should be governed by the method of analysis selected.
21,0-3
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g 210.14 3.6.2.3 Verify that all possible targets of unrestrained whipping pipes and jet impingement have been considered.
g 210.15 3.6.2.3.4.2, Page 3.6.2-16 It is the staff's position that jet expansion is not acceptable when used to evaluate jet impingement forces due to saturated water or sub cooled water blowdown.
Justify your jet expansion model for saturated water blowdown or change youp FSAR to conform to the staff's position.
g 210.16 3.6.2.5, Page 3.6.2-18 Justify the use of limited area circumferential or longitundinal
- breaks, provide a list showing where limited break areas'ave been postulated.
g 210.17 3.6.2.5, Page 3.6.2-18 Provide details and examples of the analysis performed with respect to piping restraints.
q 210.18 3.6.2.5, Page 3.6.2-18 Provide details of dynamic testing performed to determine the energy dissipating capacity of crushable material used in pipe restraints.
Verify that the allowable capacity is limited to 80K of the energy dissipating capacity determined by dynamic testing.
g 210.19
- 3. 6. 2. 5. 1, Table 3. 6. 2-2 Provide primary-plus-secondary stress intensity ranges in the main reactor coolant loop fatigue analysis and also the cumulative usage factors for our review.
g 210.20 3.6.2.5.2 Provide for our review a summary of the data developed to select postulated break locations for balance of plant piping.
Include 210-'4
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calculated stress intensities, cumulative usage factors, and the calculated primary-plus-secondary stress range.
3.7.3 Seismic Subsystem Analysis g 210. 21
- 3. 7. 3. 1. 1, Page
- 3. 7. 3-1 How have you determined "a sufficient number of degrees of freedom to closely simulate the dynamic behavior of the subsystems?"
g 210.22 3.7.3.1.1, Page 3.7.3-1 Define the term "significant modes" as you have applied it to seismic subsystem analysis.
g 210.23 3.7.3.1.1, Page 3.7.3-1 Mhat special devices have been used to eliminate the effects of relative displacements?
Mhere have they been employed?
g 210.24 3.7.3.5.1, Page 3.7.3-6 If the equivalent static load method is not used on piping systems where has it been used?
g 210.25 3.7.3.8.1, Page 3.7.3.3-8 Further discussion of your approach to determining modal acceleration is required.
It is not apparent from the material presented that the alternate response spectra are conservative.
g 210.26 3.7.3.8.1.1, Page 3.7.3-9 Provide an example of your computer method analysis using the response spectra method of Section 3.7.3. 1. 1.
Include a case in which the peak with the lowest period was used.
Also provide an example of your frequency based static method.
Justify the use of 70K of the period of the peak response as a cutoff criteria.
210-5
g 210.27 3.7.3.9.1, Page 3.7.3-12 Stress in component supports due to differential seismic motion are not treated as 'secondary by ASME Subsection NF.
Provide a basis for its acceptability.
3.9 Mechanical Systems and Components g 210.28 3.9.1.1, Page 3.9.1-1 Provide for our review the ASME Code service limits you have specified for transient loading conditions or load combinations with respect to Code Class 1 and CS components.
g 210. 29
- 3. 9. 1, 2. 2, Page
- 3. 9. 1-11 Specify cases where you have combined loads by algebraic addition.
g 210.30 3.9.1.2.2, Page 3.9.1-11 NUREG/0800 requires that computer programs in analyses of seismic Category I Code and non-Code items have the following information provided to demonstrate their applicability and validity:
a.
The author,
- source, dated version and facility.
b.
A description and the extent and limitation of its application.
c.
Solutions to a series of test problems which shall be demonstrated to be substantially similar to solutions obtained from any one of sources 1 through 4, and source 5:
1.
Hand calculations.
2.
3.
4.
Analytical results published in the literature.
Acceptable experimental tests.
By an MEB acceptable similar program.
The benchmark problems prescribed in Report NUREG/CR-1677, "Piping Benchmark problems."
210-6
Demonstrate compliance with these requirements and prov'ide summary comparisons for the computer programs used in seismic Category I analyses.
g 210.31 3.9.1.4.7, Page 3.9.1-15 Clarify your statement, "Ifplastic component analysis is used with elastic system analysis or with plastic system analysis, the deforma-tions and displacements of the individual system members will be shown to be no larger than those which can be properly calculated by the analytical methods used for the system analysis."
g 210,32 3.9.2.1, Page 3.9.2-1 Your discussion of piping vibration and thermal expansion tests is too general.
Provide specific acceptance criteria for piping vibration.
Reactor coolant system transients must also include turbine stop valve closure and pressurizer pressure relief valve operation.
0 210.33 3.9.2.3 Your discussion of seismic system analysis lacks sufficient information for the staff to complete their review.
Information must be provided concerning the following:
(1)
Consideration given to maximum relative displacements between supports.
(2)
Procedures used to separate fundamental frequencies of components and equipment from the forcing frequencies of the support structure.
(3)
Procedures for consideration of the three components of earthquake motion.
(4)
Methods to consider differential piping support movements.
210-7
I h
(5)
Methods for seismic analysis of equipment and components supported at different elevations within a building or between different buildings with distinct inputs.
(6)
Justification for the use of constant vertical static factors, if any.
(7)
Procedures used to consider torsional effects due to eccentric masses.
(8)
Methods used to analyze Category I buried piping, if any.
(9)
Methodology to account for the seismic motion of non-Category I piping systems in the design of Category I piping.
0 210.34 3.9.3 Provide a discussion giving a more detailed justification to Regulatory Guide 1.48; Regulatory Positions C.6, C.7, C.8, and C. 10.
0 210. 35
- 3. 9. 3. 1 Your discussion of loading combinations, system operating transients, and stress limits lacks sufficient information for the staff to complete its review.
Assurance must be provided that all operating gJ)oc lc(
categories>include plant events and service loading combinations required by Appendix A to Standard Review Plan 3.9.3.
Provide appropriate service limits for Code Class 1, 2, and 3 and Class CS core support structures.
Provide for our review your piping components functional capability program.
I'nclude ASME Code Section III allowable stress limits.
g 210. 36
- 3. 9. 3. 1~~I~
Provide>allowable stresses for bolts used in ASME Code components.
210-8
Q 210.37 3.9.3.1, Table 3.9.3-8 Valve discs are considered part of the pressure boundary and as such should have allowable stress limits.
Provide these limits for our review.
Q 210.38 3.9.3.1, Table 3.9.3-10, 3.9.3-11 The actual stress limits used should be clarified rather than a
reference to the appropriate Code paragraph.
Q 210.39 3.9.3.1.2.2, Appendix 3.9A-2 Define closely spaced modes as you have used them in Response Spectra Analysis.
Q 210.40 3.9.3.3 Identify any areas where the piping and support system for pressure relief devices uses hydraulic snubbers.
Provide the snubbers performance characteristics, if any, for our review.
Q 210.41
- 3. 9. 3. 3. 1, Page
- 3. 9. 3-7 Confirm that pressure-relieving devices are spaced according to Regulatory Guide 1.67.
Q 210.42 3.9.3.4.2, Table 3.9.3-1 Provide information showing where you have considered thermal stresses.
Q 210.43 Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, and due to the potential safety significance of failed snubbers in safety related systems and com-ponents, it is requested that maintenance records for snubbers be documented as follows:
210-9
J T,
Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9.
This examination should be made after snubber installation but not more than six months prior to initial system pre-operational
- testing, and should as a minimum verify the following:
(1)
There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.
(2)
The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifications.
(3)
Snubbers are not seized, frozen or jammed.
(4)
Adequate swing clearance is provided to allow snubber movement.
(5) If applicable, fluid is to the recommended level and is not leaking from the snubber system.
(6)
Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
If the period between the initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1, 4, and 5 shall be performed.
Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
210-.10
Pre-0 erational Testin During pre-operational
- testing, snubber thermal movements for systems whose operating temperature exceeds 250' should be verified as follows:
(a)
During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
(b)
For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubb'er will accommodate the projected thermal movement.
(c)
Verify the snubber swing clearance at specified heatup and cooldown intervals.
Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test programs.
The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion.
This test program should be specified in Chapter 14 of the FSAR.
There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure.
There are also some systems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and 210-11
II
the low pressure systems.
The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IW-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action i.e.,
shutdown or system isolation when the final approved leakage limits are not met.
Also surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.
Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and for systems rated at less than 50%%uo of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.
The testing interval should average to be approximately one year.
Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling
- outage, maintenance and etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.
Sig-nificant increases over this limiting valve would be an indication of valve degradation from one test to another.
210-12
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The Class 1 to Class 2 boundary will be considered the isolation point'hich must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested.
Mhen three or more valves provide isolation, only two of the valves need to be leak tested.
Provide a list of all presssure isolation valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isola-tion valves.
Also discuss in detail how your leak testing program will conform to the above staff position.
0 210.45 Provide your complete program for the inservice testing of pumps and valves including any requests for relief from ASNE Section NI requirements.
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