ML20106A574

Technical Evaluation of Nine Mile Point Unit 1 Plant-Unique Analysis Rept

ML20106A574
Person / Time
Site:	Nine Mile Point
Issue date:	09/30/1984
From:	Bienkowski G, Economos C, Lehner J BROOKHAVEN NATIONAL LABORATORY
To:	NRC
Shared Package
ML17054B366	List: ML20106A552 ML20106A574
References
CON-FIN-A-3713 BNL-04243, BNL-04243-08, BNL-4243, BNL-4243-8, NUDOCS 8410030397
Download: ML20106A574 (30)

Text

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Technical Evaluation of the Nine Mile Point Unit 1 Nuclear Generating Station Plant Unique Analysis Report George Bienkowski John R. Lehner Constantino Economos i

Reactor Safety Licensing Assistance Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 September 1984 i

FIN A-3713 l

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ABSTRACT This Technical Evaluation Report (TER) presents the results of the post-im-plementation audit of the Plant Unique Analysis Report (PUAR) for the Nine Mile Point-Unit 1 Nuclear Generating Station. The contents of the PUAR were compared against the hydrodynamic load Acceptance Criteria (AC) contained in NUREG-0661.

The TER summarizes the audit findings (Table 1), and discusses the nature and -

status of the exceptions to the AC identified during the audit (Table 2).

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l ACKNOWLEDGEMENTS l The cognizant NRC Technical Monitor for this program as Dr. Farouk Eltawila of the Containment Systems Branch (DSI) and the NRC Project Manager ws Mr. Jack N. Donohew of the Technical Assistance Program Management Group of the Division of Licensing. Mr. Byron Siegel of the Operating Reactors Branch Number 2 (DL) acted as Head Project Manager.

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[ List of Acronyms i.

AC ' Acceptance Criteria BNL Brookhaven National Laboratory BWR Boiling Water Reactor i

CO Condensation Oscillation DBA Design Basis Accident DL Division of Licensing I- DSI Division of Systems ' Integration

, FSTF Full Scale Test Facility

LDR Load Definition Report LOCA Loss-of-Coolant Accident LTP Long Term Program NRC Nuclear Regulatory Commission l

j PUAR Plant-Unique Analysis Report RFI Request For Information SMA Structural Mechanics Associates i

SRSS Square Root Sum of the Squares SRV Safety Relief Valve STP Short Term Program TER Technical Evaluation. Report 4 TES Teledyne Engineering Services T/Q T-Quencher l

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-Table of Contents Page No.

Abstract- 1 I

Acknowledgements 11 List of Acronyms iii

1. Introduction 1

-2. Post-Implementation Aud',t Summary 3

3. Exceptions to Generic Acceptance Criteria 12 3.1 Harmonic Phasing for C0 Response 13 3.2 Harmonic Phising for Post-Chug Response 14

, 3.3 C0/ Chugging Ring Girder Drag Loads 15

3.4 In-Plant SRV Data for Submerged Structure Drag 15 3.5 SRV Torus Loads 17 l 4. Conclusions 17

5. References 19 i

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1. INTR 00VCTION-The supprassion pool hydrodynamic loads associated with a postulated loss- 1 3 '

of-coolant accident (LOCA) were first identified during large-scale testing cf an advanced design pressure-suppression containment (Mark III). These addition-  !

_ al loads, dich had not explicitly been included in the original Mark I contain- l i  !

I ment design, result from the dynamic effects of drywell air and steam being rap-  !

idly forced into the suppression pool (torus). Because these hydrodynamic loads had not been considered in the original design of the Mark I containment, a de-tailed reevaluation of the Mark I containment system was required.

t A historical development of the bases for the original Mark I design, as 1

well as a summary of the two-part overall program (i.e., Short Term and Long -

Term Programs) used to resolve these-issues can be found in Section 1 of Refer .

ence 1. Reference 2 describes the staff's evaluation of the Short Term Program

(STP) used to verify that licensed Mark I facilities could continue to operate

safely while the Long Term Program (LTP) ms being conducted.

The objectives of the LTP were to establish design-basis (conservative) loads that are appropriate for the anticipated life of each Mark I BWR facility (40 years), and to restore the originally intended design-safety margins for -

each Mark I containment system. The principal thrust of the LTP has been the development of generic methods for the definition of suppression pool hydrody-namic loadings and the associated structural assessment techniques for the Mark

] I configuration. The generic aspects of the Mark I Owners Group LTP were com-pleted with the submittal of the " Mark I Containment Program Load Definition Re-l port" (Ref. 3) and the " Mark I Containment Program' Structural Acceptance Guide" (Ref. 4), as well as supporting reports on the LTP experimental and analytical tasks. The Mark I containment LTP Safety Evaluation Report (NUREG-0661) l l .

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presented the NRC staff's review of the generic suppression pool hydrodynamic load definition and structural assessment techniques proposed in the reports cited above. It as conclude.d that the load definition procedures utilized by the Mark I Owners Group, as modified by NRC requirements, provide conservative estimates of these loading conditions acA that the structural acceptance crite-ria are consistent with the requirements of the applicable codes and standards.

, The generic analysis techniques are intended to be used to perform a plant-unique analysis (PUA) for each Mark I facility to verify compliance with the acceptance criteria (AC) of Appendix A to NUREG-0661. The objective of this study was to perform a post-implementation audit of the plant-unique analysis for the Nine Mile Pofnt Nuclear Generating Station (Reference 5) against the hydrodynamic load criteria in NUREG-0661.

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2. POST-IMPLEMENTATION AUDIT SUNMARY

! The purpose of the post-implementation audit was to evaluate the hydrody-inic loading methodologies which were used as the basis for modifying the pres- l 1 sure suppression system of the Nine Mile Point Nuclear Generating Station. The Nine Mile Point PUAR methodologies (Reference 5) were compared with those of the LDR (Reference 3) as' approved in the AC of NUREG-0661 (Reference 1). The audit

' procedure consisted of a moderately detailed review of the plant unique analysis report (PUAR) to verify both its completeness and its compliance with the 8

acceptance criteria. A list of requests for further information was submitted

-(Reference 6), and answers were obtained at a meeting with the licensee

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(Reference 7).

Table 1 summarizes the audit results. It lists the various load categories

specified in the AC, and indicates plant-unique information through the refer-

ences, in the right-hand column, to the notes which follow ii, the text.

This audit did not include the Torus Attached Piping Report.

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CRITERIA G$ $ fr i 8p si se 0 hh 30 MET'

[{ O a J LOADS Z4 4 #<

CONTAINMENT PRESSURE a TEMPERATURE -2.1 /

VENT SYSTEM THRUST LOADS 2.2 /

, POOL SWELL

! TORUS NET VERTICAL LOADS 2.3 /

-] TORUS SHELL PRESSURE HISTORIES 2.4 /

j VENT SYSTEM IMPACT AND DRAG 2.6 / /

IMPACT AND DRAG ON OTHER STRUCTURES 2.7 /

FROTH IMPINGEMENT 2.8 /

) POOL FALLBACK 2.9 /

l LOCA JET 2.14.1 /

, LOCA BUBBLE DRAG 2.14.2 /

l VENT HEADER DEFLECTOR LOADS 2.10 /

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TABLE 1. LOAD CHECKLIST FOR POST-IMPLEMENTATION AUDIT

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l . CRITERI A s=

m9 1 m- sz 4 0 m os 94 24 w MET- O no a- J 4 Z4 #

• LOADS CONDENSATION OSCILLATION .

i TORUS SHELL LOADS 2.11.1 / 1 LOADS ON SUBMERGED STRUCTURES 2.I4.5 / 3

. . VENT SYSTEM LOADS 2.11.3 /. If I DOWNCOMER DYNAMIC LOADS 2.11.2 / g j CHUGGING i TORUS SHELL LOADS ,

2.I 2.1 / (,

LOADS ON SUBMERGED STRUCTURES 2.14.6 y 7 VENT SYSTEM LOADS 2.12.3 / g LATERAL LOADS ON DOWNCOMERS 2.12.2 /

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I TABLE 1. (CONTINUED) i

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CRITERIA WI m

m8 3 OF- 9 4 m 50 z< m w

MT a 8- z 30 a Jd LOADS

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T-QUENCHER LOADS DISCHARGE LINE CLEARING 2.13.2 /

TORUS SHELL' PRESSURES 2.13.3 / Y g JET LOADS ON SUBMERGED STRUCTURES 2.I4.3 / /0

' AIR BUBBLE DRAG 2.14.4 y fg THRUST LOADS ON T/Q ARMS 2.I3.5 V S/RVDL ENVIRONMENTAL TEMPERATURES 2.13.6 /

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IABLE.l. (CONTINUED)

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CRITERIA BZ $ Wx m9 os m Do m g4 2 4 w MT a z  ;

20 a J DESCRIPTION Z4 4 4 4 1

SUPRESSION POOL TEMPERATURE LIMIT 2.13.8 / // ,

. SUPRESSION POOL TEMPERATURE 1 p MONITORING SYSTEM 2.13.9 /

DIFFERENTIAL PRESSURE CONTROL SYSTEM FOR THOSE PLANTS USING A DRYWELL-TO-WETWELL PRESSURE DIFFERENCE AS A POOL SWELL 2.16

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MITIGATOR l l

I SRV LOAD ASSESSMENT BY IN-PLANT TEST 2.13.9 /

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Table 1 Notes i

.1. Nine Mile Point main vents enter the vent header through large spherical shells which'are not directly included in the LDR methodology for pool

! swell . The' applicant's modelling of the spherical portion in terms of

, cylindrical-sections is considered conservative. The staff expressed l- concern about potential enhanced pool swell and impact on the vent and l bellows due to large obstruction which was not included in the data base

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, used for the LDR' methodology. The applicant showed that potential l

i non-conservatisms associated with the obstruction are compensated by conservatisms in the LDR pool swell specification. In addition, the applicant showed substantial margins in the critical vent stresses and e provided assurances that the bellows can withstand a worst case pool swell

, impact. See Reference 7 for details.

l 2. The AC requires absolute summation of the C0 load harmonics (from 1 to 50 Hz) for the analysis of structures affected by CO loads. Nine Mile Point used a random phasing methodology instead, where the absolute sum of the i

four highest component responses is added algebraically to the SRSS of the l remaining component responses to get a total shell response. Loads on support and anchor systems were determined by adding the absolute value of

the three highest harmonic contributors to the SRSS of the others.

Combination of individual harmonic stresses into total element stress was done by considering frequency contributions at 31 Hz and below. This methodology was found acceptable. See Section 3.1 for additional details.

3. For condensation oscillation loads on submerged structures, the AC requires that loads be computed on the basis of both the average of all sources and maximum nearest source as derived from FSTF data. FSI effects must be included. In Nine Mile Point, for those structures where C0 loads were not easily bounded by pool swell loads, C0 loads including local FSI effects were based on random phasing methodology as in Note 2 above. Final loads were determined by adding the four maximum frequency contributors to the SRSS sum of the others up to 31 Hz. See Section 3.1 for additional details.

4. Instead of using a sinusoidal load superimposed on a static load for a C0 vent system load,'both loads were applied in a static panner to calculate pressures for Nine Mlle Point. The low frequency of the applied pressure s

was cited as justification. This analysis was found acceptable by the 1

staff.

5. The licensee states that an evaluation was performed which showed that the combined effects (i.e. hori.zontal and vertical components) of the C0 i downcomer load was comfortably bounded by the chugging lateral loads, therefore, the licensee used chugging lateral load results for all load '

cases in place of C0 downcomer loads. This analysis was found acceptable by the staff.

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6. The AC requires that total response to post-chug loads is obtained by sum-ming steady state response from each frequency from 1 to 50 Hz. The Nine Mile Point post-chug response was bounded by using pre-chug stresses in the torus shell. .This was justified on the basis of comparisons done on another TES plant and the low level of loads involved. This methodology was found acceptable. See Section 3.2 for further discussion.

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7. For chugging loads on submerged structures, the approach used for Nine Mile Point differs from that approved in the AC. As for CO, source strength for post-chug loads is based on a phasing methodology. However, for post-chug l loads five maximum frequency contributors are added to the SRSS sum of the

others. This method was found acceptable. See Section 3.2 for further -

i discussion.

l 8. For internal vent system loads due to chugging, the licensee states that an

evaluation was performed dich showed that internal vent system pressures were substantially less than internal vent pressures resulting from pool swell. The licensee used the pool swell pressure values in all combined load cases involving chugging pressures. This analysis mis found accept-able. -

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9. Torus shell pressures due to T-quencher loads were based on data collected during in-plant SRV tests. While this is in accordance with the AC, the way in dich the design loads were developed represents an exception to the AC requirement of conservative interpretation of the test data. The meth-odology was found acceptable in part because of existing margins in the structure capacity. See Section 7,.5 for additional information regarding this issue.

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10. SRV drag loads on submerged structures in Nine Mile Point were not computed according to AC approved methods. Instead, drag loads were based on data j collected during in-plant SRV tests. Test data was scaled to correct for appropriate SRV conditions and then applied to the structural model to determine stress. This methodology represents an exception to the AC and was discussed with the licensee. It was found acceptable largely because

of existing margins in the loads. See Section 3.4 for additional information regarding this issue.

11. The local suppression pool temperature limit was defined in NUREG-0661 as i

200*F for the generic Mark I T-quencher as described in Appendix A. Section 2.13.8. Subsequently, NUREG-0783 provided procedures whereby the limit -

could be increased if certain restrictions could be met. Conformance with.

the above criteria was indicated in the PUAR. However, the applicant utilized a local pool temperature model whose overall methodology provides i

a conservative may of computing pool temperature transients for purposes of demonstrating compliance with the provisions of NUREG-0783.

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3. EXCEPTIONS TO GENERIC ACCEPTANCE CRITERIA Nine Mile Point is one of several plants analyzed by Teledyne Engineering Services based on an essentially common hydrodynamic loading methodology

(Vermont Yankee, Millstone, J. A. Fitzpatrick and Pilgrim are other plants in

, this group). The methodology differs from the generic acceptance criteria of NUREG-0661 in five major areas which are listed in Table 2.

In what follows, each of these areas is discussed in detail, and the bases for the resolutions of the differences indicated.

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Table 2: Issues Identified During Audit as Exceptions to ,

the Generic Acceptance Criteria Issue No. Description Status Resolved Open i 1. Phasing of load harmonics used to analyze X structures affected by C0 loads

! 2. Phasing of load harmonics used to analyze X l structures affected by post-chug loads.

3. C0/ chug drag loads on the ring girder X

! 4. Submerged drag loads due to SRV water jet X and air clearing

! 5. SRV torus shell loads

• X 1

*While use of in-plant SRV test data to develop 1 SRV torus shell pressure. loads is permitted by I

the AC, the use of the data by Nine Mile Point rep-

, resents an exception to the AC requirement of con-servative interpretation of the in-plant test data.

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3.1 Harmonic Phasing for C0 Response The C0 torus shell load is an oscillating load caused by periodic pressure I oscillations superimposed upon the prevailing local static pressure. The LDR defines the load in terms of a rigid well pressure amplitude versus frequency a

spectra from 0 to 50 Hz dich is to be used in conjunction with a flexible all coupled fluid structure model. In addition, three alternative sets of spectral I

amplitudes are provided in the range from 4 to 16 Hz and the alternative which maximized the response is to be used. The resulting reponses from applying the amplitude at each frequency given in the total spectra to be analyzed are to be

sunened. The above procedure was found acceptable in the AC because the high

! degree of conservatism associated with the direct summation of the Fourier components of the spectrum was more than sufficient to compensate for any j uncertainties associated with the FSTF data from dich the load specification j was developed. Direct application of the above methodology to the Nine Mile Point torus proved to be too conservative and so an alternate approach based on a study performed in Reference 8 was used. The alternative approach obtains the 4

total response for CD by taking the absolute sum of the four highest harmonic l component responses and adding algebraically the SRSS of the remaining component .

l responses for shell stresses. Loads on the support and anchor systems are j determined by adding the three highest harmonics to the SRSS of the others. For

] C0 drag loads on submerged structures the four maximum harmonic contributors l added to the SRSS sum of the others are used for source strength. In all these

cases only harmonics of 31 Hz or below are considered, sile the AC requires i

harmonics to 50 Hz.

The Nine Mile Point procedure is one of several variations for implementing l phasing in the CO load definition discussed in Reference 8 and subsequent SMA

! Reports (References 9, 10) dich account for data obtained after Reference 8 f

tes published. Reference 11 reviews the various design rules and their justification as given in References 8, 9 and 10 and discusses ey they are acceptable alternatives to the LDR procedure. The method used for Nine Mile Point shell stresses and torus loads was one which was found to be marginally i

acceptable in Reference 11 provided stresses are not within a few percent of

! allowables. Since critical stresses in the Nine Mile Point shell and its l l support system are substantially below allowables for the controlling load j --

combinations dich include CO, the alternative approach for obtaining shell and support system C0 response has been found acceptable. Using phssed C0 sources i j for submerged structure drag loads has been found acceptable since one can i

expect the C0 pressurit ' signals to be considerably more desynchronized for this

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loading than for the shell pressure loads.  !

i 3.2 HarmonidPhasingforPost-ChugResponse ,

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! Post-chugging is defined as a spectral load across a wide band of i

l frequencies, similar to CO, but lower in amplitude. The AC requires that total response to post-chug loads is obtained by summing steady state response from  !

j each frequency from 1 to 50 Hz. The response of the torus shell and associated i support system was obtained for one TES plant by combining the 4 maximum -

harmonic responses with the SRSS of the others for frequencies below 32 Hz. The i i licensee states in the PUAR that post-chug stresses were small and loads due to l post-chug were always bounded by pre-chug values. Therefore, the licensee used 4

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! pre-chug stress values for all analysis involving post-chugging. In order to account for the 32-50 Hz harmonics in the post-chug spectrum, the PUAR further '

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states that these pre-chug stresses may be increased by 535 and still meet allowables. Based on these statements by the licensee and the fact that chugging is generally acknowledged to be an asynchronous load, the use of I

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pre-chug stresses for all load combinations involving post-chugging to evaluate shell and support system stresses has been found acceptable.

For submerged structure drag due to post-chug sources, a phased methodology, using the five maximum harmonic contributors plus the SRSS sum of the others, has been employed for Nine Mile Point. Since post-chug loads for I submerged structure drag loads can be expected to be even more desynchronized than for shell loads and since absolute sunning of the five maximum harmonics is a fairly conservative phasing approach, this method has also been found acceptable, j 3.3 C0/ Chugging Ring Girder Drag Loads The theoretical hydrodynamic mass coefficient used for the Ring Girder C0 ,

and chugging drag analysis of Nine Mile Point is not the limiting one required t

i by the AC, i.e., a circumscribed cylinder of diameter equal to d L,,x in the maximum transverse dimension. Instead, a circumscribed cylinder of diameter L ,

is used, justified by the relatively low ratio of fluid motion to ctructural dimension. The staff finds this,modelling conservative because of the i overprediction of the " effective buoyancy" term for an I-beam like structure, l

and the great overprediction of the flange force. The reduction of the

• l interference factor for web forces is not totally justified by the argument in i

! Appendix 3. The even greater reduction that could be used for flange forces, l coupled to the remaining conservatism in the acceleration volume, provides j adequate conservatism in the PUAR calculation of ring girder loads.

! 3.4 In-Plant SRV Data for Submerged Structure Drag i The AC and LDR require T-quencher bubble-induced drag loads on submerged

, structures to be calculated on the basis of an analytical model dose major l assuptions are summarized in Section 5.2.5.1 of the LDR (Ref. 3). For Nine Mile i

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Point a completely different approach was used: During in-plant SRV tests in Nine Mile Point and three other plants, strains were measured on two or three submerged structures in each plant. From these data (a total of 10 points) an equivalent static load was computed for each structure. This was done by calculating the static pressure load.which would produce the same bending stresses as those measured, Wien applied uniformly to the structure. From these calculations a curve was developed showing static pressure values versus horizontal distance from the quencher. The curve is supposed to represent the equivalent static drag pressures, including quencher jet loads. To account for

] other SRV load cases besides those tested, the curve is scaled by the ratio of

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the calculated shell ' pressures for the various cases to the test case.

] The staff had several concerns with this methodology, particularly since it did not account for bubble frequency content or structure reponse I characteristics in a direct manner.

l However, the stresses computed for Nine Mile Point with this methodology were well below the allowable limits. The ratio by which the SRV-induced drag could be increased before the allowable load was reached ranged from 2.25 in the case of the downcomers to 41 in the case of outer columns, and was greater than 5.0 for all submerged structures other than the downcomers (7). For the

downcomers,1eledyne has presented results calculated by a separate agent for i

7 another plant owned by a different utility which used a test calibrated version of the LDR methodology for its SRV drag loads. For a similar downcomer at a similar distance from the quencher, the loads calculated for this plant were compa'rable to those obtained by the Teledyne methodology, giving the staff l confidence that a factor of 2.25 load margin as quite adequate for the downcomer in Nine Mile Point.

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i Based on the large margins. therefore, and on the favorable comparison of loads on a similar downcomer computed using a method accepted by the NRC, the Nine Mile Point submerged structure drag loads were found to be satisfactorily computed.  ;

l 3.5 SRV Torus Loads )

1 The design value used by Nine Mile Point for SRV shell pressure loads was derived by extrapolation of the peak pressure (3.5 psid) observed during a se-t ries of four SRV actuations performed in the Nine Mile Point plant. If the overall structural capability of the containment is not considered, this is an i inadequate procedure in that it does not represent a " conservative interpreta-tion of the in-plant test data" as required by the acceptance criteria (2.13.9.2.3).

Whenever the available data base is limited, and particularly when the data exhibit large variability, a bounding approach is not appropriate. Conventional engineering practice dictates the application of statistical methods for data interpretation to provide sufficient confidence that the loads used for design will not be exceeded.

For the Nine Mile Point tests, the average of the four positive pressures that were recorded was 2.7 psid with a standard deviation of almost 25%. The corresponding (95-95) nonexceedance value of positive pressure is 6.0 psid, a value 70% higher than the peak observed value. Extrapolation of this pressure to design conditions implies a design pressure of about 8 psid. This is twice the value used by the applicant (4.0 psid), but is well below the value that can be accommodated by the structure according to the information supplied to us in Reference 7. On this basis we find the proposed design acceptable.

4. CONCLUSIONS A post-implementation pool dynamic load audit of the Nine Mile Point PVAR has been completed to verify compliance with the generic acceptance criteria of

NUREG-0661. Five major differences were identified between the PVAR and the generic acceptance criteria. Based on additional information supplied by the applicant as detailed in the previous section all of these issues were resolved. The review of the Nine Mile Point PUAR Torus suppression chamber has been completed with no issues or concerns outstanding.

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5. REFERENCES

, References cited in this report are available as follows:

! Those items marked with one asterisk (*) are available in the NRC Public Document Room for inspection; they may be copied for a fee. )

l Material marked .with two asterisks (**) is not publicly available because l it contains. proprietary information; however, a nonproprietary version is available in the NRC Public Document Room for inspect. ion and may be copied for a fee.

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l Those reference items marked with three asterids (***) are available for purchase from the NRC/GPO Sales Prograin, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, and/or the National Technical Information Service, Springfield, Virginia 22161.

T All other material referenced is in the open literature and is available through public technical libraries.

(1) " Safety Evaluation' Report, Mark I Long Term Program, Resolution of Generic Technical Activity A-7", NUREG-0661, July 1980.***

i (2) " Mark I Containment Short-Term Program Safety Evaluation Report". '

NUREG-0408, December 1977.***

'(3) General Electric Company, " Mark I Containment Program Load Definition Report" General Electric Topical Report NEDO-21888, Revision 2, November 1981.*

(4) Mark I Owners Group, " Mark I Containment Program Structural Acceptance Criteria Plant-Unique Analysis Applicaticns Guide, Task Number 3.1.3",

General Electric Topical. Report NEDO-24583, Revision 1 July 1979.*

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" Plant-Unique Analysis Report of the Torus Suppression Chamber for Nine

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Mile Point Unit 1 Nuclear Generating Station",- Technical Report TR-5320-1, Teledyne Engineering Services, October 5,1983.*

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'(6) Letter from J. R. Lehner, BNL, to F. Eltawila, IstC, dated April 16, 1984.

Subject:

Request for Information Regarding Nine Mile Point PUAR.*

-(7) " Mark I Torus Program: Review of- Planti-Unique Analysis Report for Nine Mile Point Unit 1" Teledyne. Response to BNL Questions, August 23,.1984.* l (8) " Mark I Containment Program Eval $ ton of Harmonic Phasing for Mark I Torus Shell Condensation Oscillation Lca.u", NEDE-24840, prepared by Structural Mechanics Associates for. General Electric Company,' October 1980.*

I (9) Kennedy, R. P., " Response Factors Appropriate for Use with C0 Harmonic Response Combination Design Rules". SMA 12101.04-R002D, prepared by Structural Mechanics Associates for General Electric Company, March-1982.*

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t (10) Kennedy, R. P., "A Statistical Basis for Load Factors Appropriate for Use with C0 Harmonic Response Combination Design Rules". SMA 12101.04-R003D, 1 y prepared by Structural Mechanics Associates for General Electric Company, March 1982.* l l

(11) Bienkowski, G., " Review of the Validity of Random Phasing Rules as Applied to C0 Torus Loads" Internal BNL Memo, August 1983.

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sr ENCLOSURE 2

. SALP .

prepared by the Containment Systems Branch NINE 111LE POINT 1 Evaluation Narrative Description -

Criteria Category issue.

I* #8 M I"'"I "" 2 Managtynt took positive steps to assume timely resolution of the

z. npproacn to Resolution Sound Technical Understanding of the issue. Worked closely with the staff and cf Technical Issues 2 its consultant toward resolution of the issue. .--

3. Responsiveness Met with the staff and its cons utant shortly after receiving the RAl.

-4 Enforcement History , .

M/A .

~ 5. Reportable Events N/A

' 6. Staffing N/A

7. Training ,

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p ree uq kg UNITED STATES NUCLEAR REGULATORY COMMISSION

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U :j WASHINGTON, D. C. 20555

...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO MARK I CONTAINMENT LONG-TERM PROGRAM STRUCTURAL REVIEW NIAGARA M0 HAWK POWER CORPORATION DOCKET N0. 50-220

1.0 INTRODUCTION

The capability of the boiling water reactor (BWR) Mark I containment structures and piping systems to withstand the effect of hydrodynamic loads resulting from a loss of coolant accident (LOCA) and/or a safety relief valve (SRV) discharge was not considered in the original design of the structures. The resolution of this issue was divided into a short-term .

program and a long-term program.

Based on the results of the short-term program, which verified that each Mark I containment would maintain its integrity and functional capability when subjected to the loads induced by a design-basis LOCA, the NRC 5taff granted an exemption relating to the structural safety requirements of 10 CFR 50.55(a). The study was_ reported in NUREG-0408, " Mark I Containment Short Term Program".

The objective of the long-term program was to maintain a margin of safety when the Mark I containment structures and piping system are subjected to additional hydrodynamic loads. The detailed guidance of the long-term pro-gram are contained in the NRC Safety Evaluation Report, NUREG-0661, " Mark I Containment Long-Term Program" and its supplement which describe the ,

generic hydrodynamic load definition and structural acceptance criteria consistent with the requirements of the applicable codes and standards.

To fulfill the objective of the long-term program, Niagara Mohawk Power Corporation (NMPC) has completed all modifications on the Nine Mile Point, Unit I containment and torus attached piping. The adequacy of these modi-fications was documented in reports prepared by Teledyne Engineering Service (Teledyne) titled, TR-5320-1, " Mark I Containment Program, Plant-Unique Analysis Report of the Torus Suppression Chamber for Nine Mile Point Nuclear Station Unit 1" and TR-5320-2, " Mark I Containment Program, Plant Unique Analysis Report of the Torus Attached Piping for Nine Mile Point Nuclear Station Unit 1".

The Franklin Research Center (FRC) was contracted to review the structural adequacy issue for compliance with the staff's acceptance criteria.

2.0 EVALUATION i The Mark I long-term program of the Nine Mile Point, Unit I was described in j the plant-unique analysis report (PVAR) prepared by Teledyne. This report l

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describes modifications performed on containment structures and torus attached piping at the Nine Mile Point Unit 1. Areas covered by the report ir.clude the torus shel.1, external support system, vent header system, internal structures, torus attached pipings, SRV lines and vent pipe penetrations. The materials.

design and fabrication requirements of the modifications were-in accordance ,

with the American Society of Mechanical Engineers (ASME) Boiler and Pressure '

Vessel (B&PV) Code, Division 1.Section III with Addenda through Summer 1977 and Code Case N-197, " Service Limits for Containment Vessels".

Modifications'were performed in accordance with the requirements of Section XI of the same code. To determine the appropriate code allowable service  :

limits for the_specified loading combinations, the report followed guidelines of NUREG-0661 and the GE report, NED0-24583-1, " Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide."

The portion of the report applicable to loadings and loading combinations was audited by BNL, and results of that audit are discussed in a separate Safety Evaluation. - -

l' Using the properly determined loadings and loading combinations, Teledyne employed the computer program, STARDYNE, as a major tool to perform the l analyses. STARDYNE is a program which has been used widely in the industry

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1 for similar purposes and was approved by NRC. Results of the analyses were sumarized to show that modifications are adequate under various loading '

combinations.

The adequacy of the modified containment structures.and torus attached piping was audited by the FRC. FRC developed audit procedures for all Mark I long-term program users, which is described in detail'in the FRC TER-C5506-308,

" Audit Procedures for Mark I Containment Long-Term Program - Structural Analysis." The review performed by FRC followed this document closely.

Results and conclusions of this effort were reported in FRC TER-C5506-331, .

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" Audit for Mark I Containment Long-Term Program - Structural Analysis for Operating Reactors - Niagara Mohawk Power Corporation - Nine Mile Point

Nuclear Station, Unit 1." The audit verified analyses by examining mathe-

matical models and loading combinations'used, and summarized the results to see whether the modifications met the required criteria. A check list was compiled to ensure the completeness of the auditing. The staff has reviewed the FRC report and concurs with its conclusions that the modifications meet-i the Mark.I Containment Long-Term Program objective. .An augmented. fatigue.

evaluation method for ASME Code Class 2/3 piping was developed by MPR for GE in MPR Report-751, titled, " Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus attached on SRV Piping ~ System", dated November.

1982. -This report was reviewed by the staff and the conclusion that all

torus piping systems have a fatigue usage of less than 0.5 during the plant

[- . life is acceptable for the Nine Mile Point, Unit 1. l 1

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3.0' CONCLUSIONS

- The modifications performed at the Nine Mile Point, Unit 1 followed the guide-lines of NUREG-0661 and its supplement and met the respective requirements of Sections III and XI of the ASME Boiler and Pressure Vessel Code and are,

. therefore, acceptable. NMPC analyses have been. verified by the FRC audit and approved by the staff under the LOCA and SRV discharge loads.

Principal Contributor: H. Shaw Date: January 22, 1985 Attached: TER prepared by Franklin Research Center, dated September 26, 1984 f

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