ML17054A703

Forwards Request for Addl Info Re Mark I Containment long-term Program for Plant Unique Analysis Rept & Loads & Structural Evaluations.Meeting Requested within 4 Wks of Ltr Date

ML17054A703
Person / Time
Site:	Nine Mile Point
Issue date:	05/03/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Rhode G NIAGARA MOHAWK POWER CORP.
References
NUDOCS 8405210456
Download: ML17054A703 (28)

Text

May 3, 1984 Docket No. 50-220 Mr. G.

K. Rhode Senior Vice President Niagara Mohawk Power Corporation 300 Erie Boulevard West

Syracuse, New York 13202

Dear Mr. Rhode:

DISTRIBUTION NRC PDR Local PDR ORB5'2 Reading DEisenhut OELD SNorris RHermann BSiegel ELJordan JNGrace ACRS (10)

Gray File SPCarfagno, FRC

SUBJECT:

MARK I CONTAINMENT LONG TERM PROGRAM - PLANT UNIQUE ANALYSIS REPORT, LOADS AND STRUCTURAL EVALUATIONS Re:

Nine Mile Point Nuclear Station, Unit No.

1 The NRC staff and its consultants, Brookhaven National Laboratory (BNL) and Franklin Research Center (FRC), are reviewing the loads and structural aspects of your plant unique analysis report.

As a result of our review to date we have prepared the enclosed Request for Additional Information.

To expedite this review it is requested that within four weeks of the date of this letter a meeting between the NRC and our consultants, and you and your contractor be held to discuss your response to these issues.

Since it is our intent to resolve these issues at this meeting, it is imperative that you have a representative at this meeting that has the authority to make the decisions necessary to accomplish this goal.

It is suggested that this meeting be held at your contractor's office;

however, we are amenable to having it wherever it is most convenient.

Please notify your project manager within seven days of receipt of this letter with a proposed meeting date.

If you cannot meet the four week

schedule, please propose an alternative one.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0091 which expires October 31, 1985.

it Sincerely,

Enclosure:

As stated cc w/enclosure:

See next page Original signed by/

Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing DL:0$Bg2 SNoFPfs ajs 05/8,/84 DL:

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B82 DVassallo 05/~/84 8405210456 840503 PDR ADOCK 05000220 P

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Mr.

G.

K. Rhode Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station, Unit No.

1 cc Troy B. Conner, Jr.,

Esq.

Conner 8 Wetterhahn Suite 1050 1747 Pennsylvania

Avenue, N.

W.

Washington, D. C.

20006 Mr. Robert p. Jones, Supervisor Town of Seri ba R.

D. g4

Oswego, New York 13126 Niagara Mohawk power Corporation ATTN:

Mr. Thomas Perkins Plant Superintendent Nine Mile Point Nuclear Station Post Office Box 32

Lycoming, New York 13093 Thomas A. Murley Regional Administrator Region I Office U. S.

Nuclear Regulatory Commission 631 Park Avenue King of prussia, pennsylvania 19406 Mr. Jay Dunkleberger Division of Policy Analysis and planning New York State Energy Office Agency Building 2, Empire State plaza

Albany, New York 12223 U. S. Environmental protection Agency Region II Office Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Resident Inspector U. S. Nuclear Regulatory Commission post Office Box 126

Lycoming, New York 13093 John W. Keib, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West

Syracuse, New York 13202

REQUEST FOR INFORMATION NINE MILE POINT

ITEN I:

PUAR Section 3.2 The vent-header main vent junction in Nine Mile Point Unit I is appre-ciably different than in other Mark I plants.

The pool swell phenomena must be affected by this substantial difference in geometry.

ITEM 2:

Give details of the vent header deflector configuration in the vicinity of the main vent.

Give details of the procedures for applying the gSTF pool swell results to the bays containing the main vents.

PUAR Section 3.2 The LDR and NUREG-0661 specify a minimum of four tests as a data base for obtaining net torus loads.

Does Nine Mile Point Unit I operate under zero ap conditions?

If this is so, explain how the single gSTF zero ap test is used to derive the pool swell loads such as torus vertical load, impact, and drag.

PUAR Section 3.3.1 The calculated shell membrane stress reported on p.

35 of the PUAR is very close to the allowable stress.

Specify the fractional contribu-tion to this stress from the DBA.CO Loading.

Is the allowable stress more restrictive than is necessary for the specific load combination (case 20)?

Have any bounding assumptions beyond those required by the Acceptance Criteria been used in the calculated load?

PUAR Sections 3.2.4, Appendix 1

Provide the following additional information regarding the in-plant SRV tests conducted at Nine Mile Point and the SRV design loads extrapo-latedd from the tests:

1.0 Description of the tested quencher Device 1.1 Drawings showing details of the quencher geometry - plan, eleva-tion, arm length, arm diameter, hole arrangement,

spacing, size, etc.

1.2 Location of quencher device relative to suppression pool bound-aries and suppression pool surface.

1.3 Any difference between the tested quencher configuration and the Monticello version (as described in GE NEDE-24542-P) should identified.

2.0 A description of the loads observed during testing-

2. 1 Peak overpressure (POP) and underpressure (PUP) recorded on the torus shell during each relevant SRV actuation.

PAGE TWO 2.2 A measure of the frequency content of each pressure signature.

3.0 A description of the test conditions-3.1 Geometry of the tested SRVDL (diameter, length, free volume, and routing below pool surface).

3.2 Geometry of any SRVDLs in the plant that differ significantly from the tested SRVDL.

3.3 SRV steam flow rate (MS), pool temperature (TPL), pipe temperature (TP), water leg length (LW) and pressure differential (~P), if any,.for each test.

3.4 Minimum aP permitted by NRC Technical Specification and corres-ponding LW for all SRVDLs.

4.0 A description of the design conditions for each load case used for de-sign-4.1 Geometry of all SRVDLs involved and their azimuthal location in the torus.

4.2 TP, TPL, MS, DP and LW for al'1 SRVDLs involved.

5.0 A description of the design loads for each load case-

5. 1 Normal ized pressure signature.

5.2 Single valve POP/PUP values.

5.3 Spatial attenuatio'n of the POP/PUP values (if this differs from the LDR methodology, sufficient additional torus shell pressure data must be supplied to justify such deviation).

5. 4 Frequency range considered.

5.5 Procedure used to combine loads for multiple values.

ITEM 5:

Section 4.3.5, Appendix I Use of SRV test data extrapoiated to design cases for submerged struc-ture drag loads represents an exception to the Acceptance Criteria.

Identify the al lowed SRV drag multipliers to meet code allowables for all structures for which the multiplier is belo~ 3.

Identify the structures for the two Nine Mile Point data points on Figure Al-7 of the PUAR.

Explain how data fr om other plants with totally different quencher gecmetry and location is relevant to Nine Mile Point SRV drag load cal cul ation.

PAGE THREE ITEN 6:

Sections 3.2.1, 4.3.1, 5.3 Because of the significant differences in geometry of Nine Mile Point Unit 1 from many other Mark I plants, the staff needs to examine how certain generic loads were applied.

Provide G.E.

Report NED0-24574, Rev.

1 "Mark I Containment Program-Plant Unique Load Definition-Nine Mile Point 1 Nuclear Generating Plant"-July 1981.

ITEM 7:

Section 2.2. I Provide the exact location of the twentysix sensors used to monitor pool water temperature.

Include the elevation of the sensors with respect to the pool water level and the SRV quencher device.

APPENDIX B ADDITIONAL INFORMATION REQUIRED Franklin Research Center A Division of The i.rankiin Institute

%he Benjamin Frank)in Parkway, Phi)a,. Pa. )9)03 (2) 5) 448 l000

C I

,,f','l I 'I ~

II I I III

~ I TER-C5506-3 31 Item 1:

~ ~

I RE UEST FOR INFORMATION Provide piping report TR-5320-2.

,The'following items must be covered adequately in order to satisfy",the criteria<

I o

The analysis of applicable portions,'of ECCS and other piping systems required to maintain core,cooling after a LOCA, vacuum breaker piping, and piping that provides drywell-to~etwell pressure differential.

o The classification of piping sy'stems as essential or nonessential and by code class.

I o

Analytical models representing;piping and supports from first rigid anchor (or where the effects,of torus 'motion is insignificant).

( f o

The analysis of torus piping penetrations.

o The use of time history or response spectrum analysis for dynamic effect of torus motion at piping attachment points.

o The code classification of piping supports and welds.

Item 2:

Provide a summary of the analysis with regard to vacuum breaker valves; indicate whether they are considered Class 2 components as required by the criteria [1].

Item 3:

Indicate whether the fatigue usage factors for the SRV piping and the torus-attached piping are sufficiently small that a plant-unique fatigue analysis is not warranted for piping.

The NRC is expected to review the conclusions of a generic presentation

[4] and determine whether it is sufficient for each plant-unique analysis to establish that the expected usage factors for piping are small enough to obviate a plant-uniqhe fatigue analysis of the piping.

Item 4:

Indicate whether all active equipment associated with piping, such as pumps and valves, has been evaluated for operability and discuss the operability criteria.

Item 5:

Indicate whether net tensile forces are produced in the torus support columns due to the upward phase of loading.

If so, provide a nonlinear time history analysis using a spring mass model of the torus support as required by the criteria fl].

Item 6:

Indicate how loads resulting from different dynamic phenomena were combined.

Item 7:

Soecify the code class of the vent ring header supports.

00 Franklin Research Center A Dhisen 4 Tht Franklin Insatuce

TER-C5506-3 31 Item 8:

Specify code sections and/or formulas used to derive all allowable stresses and loads presented in the PUA report

[5] in the following sections:

3.3.1 through 3.3.5 4.4.1 through 4.4.6 4.4. 8 5.4.1 and 5.4.2 6.4.1 through 6.4.3 7.1.3 7.3 3

7.4 pages 35 through 40 pages 60 through 62 page 63 pages 74 and 75 pages 83 and 84 pages 87 and 88 pages 91 and 92 page 92 Item 9:

Provide and justify the bounding controlling load cases presented following sections:

3.3.1 through 3.3.5 4.4.1 through 4.4.6 4.4.8 5.4.1 and 5.4.2 6.4.1 through 6.4.3 7'.3 7.3."3 7.4 technique used to determine the in the PUA report [5] in the pages 34 through 39 pages 59 through 62 page 63 pages 74 and 75 page 83 page 87 page 91 page 92 Item 10:

With respect to Section 3.1 of the PUA report [5], justify the, use of the 1/20 segment model instead of the 3604 beam model to analyze the torus shell for stresses due to asymmetric loads (horizontal earthquake, SRV, chugging).

Item 11:

Provide and justify the reasons for not considering a 180 beam model of the vent system, as required by Section 6.5c of Reference 1, in order to determine the effects of seismic and other asymmetric loads.

Item 12:

The PUA report

[5] indicates that the following stresses are very close to the respective allowables:

o torus shell primary membrane stress o

saddle-to-shell weld stress o

ring girder weld to torus stress (near outer column and saddle regions) o monorail column and base plate weld stress.

Indicate conservatisms in the analysis to show that these calculated values would not be exceeded if a different analytical approach were to be used.

ljIlllFranklin Research Center A Divison or %be Franklin insoonc

TERM5506-331 Item 13:

With respect to Section 3.1 of the PUA report [5], provide the technical basis for obtaining the static degrees of freedom for the torus model.

Also, provide a brief description of the boundary conditions.

Item 14:

With respect to Section 3.2.2 of the PUA report [5], explain why the absolute values of the four highest harmonics were considered for shell stresses due to condensation oscillation, while only the three highest absolute values were considered in evaluating support loads.

Also, clarify whether 31 or 32 Hz (as specified in Appendix 2 [5]) was used as the cutoff frequency.

Item 15:

With respect to Section 3.2.3.2 of the PUA report [5], explain how the pre~hug load bounds the post-chug for column and saddle loads while post-chug stress exceeds pre~hug stress by 538 and why the analysis for postmhug uses the pre-chug stress value.

It is recommended that the explanation be detailed enough to thoroughly clarify this issue.

Item 16:

With respect to Section 3.2.4 of the PUA report [5], provide the data and plots showing the correlation between calculated and measured shell stress in the dry structure analysis.

Indicate locations where the correlation was made and whether the comparison was obtained based on time history traces and/or frequency distributions.

Item 17:

Regarding the fatigue evaluation of the torus presented in Section 3.2.7 [5], provide precaution measures (if any) to be taken in case the operator fails to act after 15 minutes.

Item 18:

With respect to Section 4.2 of the PUA report [5], provide a brief description of the stress calculation method for each of the following:

o vent header support columns o

vent pipe/vent header intersection o

vent pipe/drywell intersection o

vent header mitre joint o

main vent pipe.

Item 19:

Provide a description of the hand analysis used to evaluate the effects of pool swell water impact on the vent header deflector.

Also, provide and justify the dynamic load factors used with the impact forces (Section

4. 2 [5] ).

l)ll Franklin Research Center ADiva<on ot 1be Fee&in Inseute

TER-C5506-331 Item 20'ith respect to Section 4.'3.1.1 of the PUA report [5], indicate how relative timing between the pool swell water impact loadings on the vent system was maintained to preserve an accurate representation of the longitudinal and circumferential wave sweep.

Item 21:

With respect to the vent header beam model shown in Figure 4-4 in the PUA report [5], indicate how the stiffnesses representing the vent header/vent pipe intersection were selected.

Also, provide the technical basis and justification for the selection of these stiffnesses.

Item 22:

Clarify whether the seismic and thermal response of the drywell was considered in evaluating the main vent/drywell intersection (Section 4.4.6 [5]) ~

Item 23s Provide and justify the stress intensification factor used in the fatigue evaluation of the vent system in Section 4.4.9 of the PUA report [5].

Item,24:

Provide and justify all dynamic load factors used in the analysis of the ring girder (Section 5.3.3 [5]) for loads due to the attachment of the following structures:

o quencher support beam I

o vent header support columns o

monorail supports o

spray header.

Item 25:

With respect to Section 6.3.3 of the PUA report [5], explain and'rovide the set of harmonic analyses.

The Licensee stated, "Results for individual load conditions were determined by scaling individual frequency results of the computer analysis by the appropriate pressure amplitude."

Please elaborate on this statement.

Item 26:

With respect to the computer analysis of the quencher and supports given in Section 6.3.3 [5], explain how the mass of structure was adjusted to account for the "added mass" effect of the surrounding water.

Item 27:

Indicate why only four rnaximurn frequency contributions were considered in the condensation oscillation analysis of the quencher and supports (Section 6.3.4 [5]), whereas five were considered in the chugging analysis (Section

6. 3. 3 [5] ).

(1ll FranMin Research Center h~ el The Fcanklin tnsaeute

TERW5506-331 Item 28:

With respect to Section 7.1.1 of the PUA report [5], indicate and justify all factors used to account for dynamic effects in the internal spray header analysis.

Item 29:

In the description of the dynamic characteristics of the bellows in Section 7.2.1 [5], the following passage appears:

"We also expect that the convolutions will produce complex modes and stress patterns that will not couple efficiently with specific input frequencies, i.e., high dynamic response is not expected.

Further, the "pogo" and "rolling" modes of the convolutions are non-linear, highly cross-coupled modes that would not be predicted by ordinary structural codes."

Provide a detailed explanation to clarify this passage.

Item 30:

With respect to the verification of the computer model used to evaluate torus shell stresses and support system loads due to SRV actuation (Appendix A, p. Al-3 [5]), it was noted that the correlations of predicted column loads and measured column loads was generally off by about 508.

Provide some possible reason for this discrepancy.

l)ll FranMin Research Center A Oirisen of The Franklin fneecute

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TER-C5506-3 31 REFERENCES FOR APPENDIX B 1.

NEDO-24583-1 "Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide" General Electric Co.,

San Jose, CA October 1979 2 ~

NUREG-0 661 "Safety Evaluation Report, Mark I Containment Long-Term Program Resolution of Generic Technical Activity A-7" Office of Nuclear Reactor Regulation July 1980 3.

NEDO-21888 Revision 2

"Mark I Containment Program Load Definition Report" General Electric Company, San Jose, CA November 1981 4.

P.

M. Kasik "Mark I Piping Fatigue" Presentation at the NRC meeting,

Bethesda, MD September 10, 1982 5.

Nine Mile Point Unit 1 Plant-Unique Analysis Report of the Torus Suppression Chamber TR-5320-1, Revision 0

Niagara Mohawk Power Company Teledyne Engineering Services October 1983 00 Frankiin Research Center A Divinon of Tha Franklin inaonna

0 1