ML20083J323

Technical Evaluation of Oyster Creek Plant Unique Analysis Repts

ML20083J323
Person / Time
Site:	Oyster Creek
Issue date:	09/30/1983
From:	Bienkowski G, Lehner J, Lin C BROOKHAVEN NATIONAL LABORATORY
To:	NRC
Shared Package
ML20083J300	List: ML20083J323
References
CON-FIN-A-3713, RTR-NUREG-0661, RTR-NUREG-661 04243, 4243, NUDOCS 8401160284
Download: ML20083J323 (26)

Text

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Technical Evaluation of the Oyster Creek Plant Unique Analysis Reports John R'. Lehner George Bienkowski C. C. Lin Constantino Economos Reactor' Safety Licensing Assistance Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 September 1983 FIN A-3713 BNL'No'. 04243 8401160284 840113 PDR ADDCK 05000219 P _ _

PDR

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ABSTRACT This Technical Evaluation Report (TER) presents the results of the post-implementation audit of the Plant Unique Analysis Report (PUAR) for the Oyster Creek Nuclear Generating Station. The contents of the PUAR were compared against the hydrodynamic load Acceptance Criteria (AC) contained in NUREG-0661.

The;TER contains a summary of the audit findings, as well as_a more detailed discussion of special issues or exceptions to the AC identified during the au-dit. Two tables are provided. The first is a checklist of PUAR loads versus AC specifications. The second highlights each special issue or AC exception along with an indication of the type and status of each issue.

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ACKNOWLEDGEMENTS The cognizant NRC Technical MonitoF for this program was Dr. Farouk .

Eltawila, of the Containment Systems Branch (DSI) and the NRC Project Manager was Ms. Beverly Barnhart of-the Technical Assistance Program Maaagement 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 AC Acceptance Criteria ADS Automatic Depressurization System ,

BNL Brookhaven National Laboratory BWR . Boiling Water Reactor C0- Condensation Oscillation ,

DL Division of Licensing DSI ., Division of System Implementation FSI Fluid Structure Interaction FSTF Full Scale Test Facility GE General Electric Company

. GPU General Public Utilities LDR Load Definition Report LOCA Loss-of-Cooland Accident LTP Long Term Program NRC- Nuclear Regulatory Commission PUAR Plant-Unique Analysis Report QSTF Quarter Scale Test Facility RFI Request For Information SBA Small Break Accident SMA Structural Mechanics Associates SRV Safety Relief Valve

'SRVDL Safety Relief Valve Discharge Line STP Short Term Program TER . Technical Evaluation Report T/Q T-Quencher

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Table of Contents

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. Page No.

Abstract i Acknowledgements ii List of Acronyms. iii l' . Introduction  !

2. Post-Implementation Audit Summary -

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3. Summary of the NRC Request For Information 11 Regarding the Oyster Creek PUAR

4. Conclusions 39

-.5. References 20

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1. INTRODUCTION The suppression pool hydrodynamic loads associated with a postulated loss- '

of-coolant accident (LOCA) were first identified during large-scale testing of an advanced design pressure-suppression containment (Mark III). These addi-tional loads, which had not explicitly been included in the original Mark I containment design, result from the dyaamic effects of drywell air and steam being rapidly forced into the suppression pool (torus). Because these hydro-dynamic' loads had not been considered in the original design of the Mark I containment, a detailed reevaluation of the Mark I containment system was re-qui red.

A historical development of the bases for the original Mark I design as 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 Ref-erence 1. Reference 2 describes the staff's evaluation of the Short Term Pro-gram (STP) used to verify that licensed Mark I facilities could continue.to operate safely while the Long Term Program (LTP) was 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-

l. namic loadings and the associated structural assessment techniques for the Mark l

l 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 Report" (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) 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 was concluded that the load definition procedures utilized by the Mark I Owners Group, as modified by NRC requirements, provide conservative estimates of these loading conditions and that the structural ac-ceptance criteria 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 ac-ceptance criteria (AC) of Appendix A to NUREG-0661. The objective of this study ,

was to perfonn a post-implementation audit of the Oyster Creek plant-unique an-alysis (References 5 & 6) against the hydrodynamic load criteria in NUREG-0661.

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2. POST-IMPLEMENTATION AUDIT

SUMMARY

The purpose of this post-implementation audit is to evaluate the hydrody-namic loading methodologies used to modify the suppression chamber, vent system, internal _ structures and the torus attached piping of the Oyster Creek Nuclear

-Generating Station. The methodologies of the Oyster Creek PUAR (References 5 &

6) are compared to those presented in the LDR (Reference 3) which were approved in the AC of NUREG-0661 (Reference 1). The audi.t procedure consists of a mode-rately detailed review of the plant-unique analysis report to verify both its completeness and its compliance with the acceptance criteria. A checklist of the various load categories specified in the AC, as shown in Table 1, is used to facilitate this task. Beside.i providing an overview of the audit, Table 1 sup-plies plant-unique information through the notes in the right-hand margin which are explained at the end of the table.

The next section of this TER, Section 3, identifies the exceptions to the i AC, as well as those special areas, detailed during the Oyster Creek PUAR audit, where additional information was needed.

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. CRITERIA

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W9 m k O. m O l- 'pq g 4 w .,

MET T z ao o. J4 f LOADS CONTAINMENT PRESSURE a TEMPERATURE 2.1 V VENT SYSTEM THRUST LOADS 2.2 v /

POOL SWELL TORUS NET VERTICAL LOADS 2.3 / 1 l

TORUS SHELL PRESSURE HISTORIES 2.4 V 2--

j VENT SYSTEM IMPACT AND DRAG 2.6 /

l IMPACT AND DRAG ON OTHER STRUCTURES 2.7 /

j FROTH IMPINGEMENT 2.8 y

! POOL FALLBACK ,2.9 #

j LOCA JET 2.14.1 / ,

! LOCA BUBBLE DRAG 2.14.2 # +

VENT HEADER DEFLECTOR LOADS 2.10 V i

TABLE 1. LOAD CHECKLIST FOR POST-IMPLEMENTATION AUDIT i

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CRITERIA WI 52 $

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Ot- p q z< w MT z ao a J 4 Z4 #

LOADS i

l CONDENSATION OSCILLATION l TORUS SHELL LOADS 2.11.1 / 3

&' LOADS ON SUBMERGED STRUCTURES 2.14.5 /.

VENT SYSTEM LOADS 2.11.3 /

j DOWNCOMER DYNAMIC LOADS 2.11.2 /

1 CHUGGING

! . TORUS SHELL LOADS 2.12.1 / 4 LOADS ON SUBMERGED STRUCTURES 2.14.6 /

VENT SYSTEM LOADS 2.12.3 /

LATERAL LOADS ON DOWNCOMERS 2.12.2 /

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'I CRITERIA j WI gz m9 m Du m OH 94 z< w MET z h 30 a J 4-l LOADS T-QUENCHER LOADS 5'

, DISCHARGE LINE CLEARING 2.13.2 / (,

? TORUS SHELL PRESSURES 2.13.3 / *

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! JET LOADS ON SUBMERGED STRUCTURES 2.14.3 / 8 AIR BUBBLE DRAG 2.14.4 / $

j THRUST LOADS ON T/Q ARMS 2.I3.5 / fo l S/RVDL ENVIRONMENTAL TEMPERATURES 2.13.6 /

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TABLE 1. (CONTINUED) i 4

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. q CRITERIA -

BZ $ - Wc  :

e9 os g4 m Do 2 4 m

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MET z o ao Z4 a

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DESCRIPTION SUPRESSION POOL TEMPERATURE 2.13.8 I

LIMIT

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, SUPRESSION POOL TEMPERATURE 2.13.9 Fy 2 MONITORING SYSTEM 7 fg l

DIFFERENTIAL PRESSURE CONTROL SYSTEM FOR THOSE PLANTS USING A l 3 DRYWELL-TO-WETWELL PRESSURE DIFFERENCE AS A POOL SWELL 2.16 /

MITIGATOR SRV LOAD ASSESSMENT BY 4 IN-PLANT TEST 2.13.9 / /g e

TABLE 1. (CONTINUED)

Table 1 Notes

1. Since the vent line is reduced in diameter at the vent line-drywell inter-section of the Oyster Creek Plant, a thrust load can occur at this reduc-tion. This thrust load was not addressed generically in the LDR but was adequately accounted for in the Oyster Creek analysis.

2. The AC requires the torus net vertical loads and the shell pressure histor-ies to be b'ased on four QSTF tests. Since the decision was made to oper-ate Oyster Creek with 0 op between drywell and wetwell, the Oyster Creek vertical loads were based on only a single test. This was the test with the

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greatest su'bmergence conducted at 0 ao. Oyster Creek loads were increesed to account for the larger statistical variance associated with the smaller number tests. The amount of increase was found acceptable. See Section 3.1-for additional discussion.

3. The' AC requires absolute summation of the C0 load harmonics for the analysis of structures affected by C0 loads. Oyster Creek used a random phasing methodology instead where individual harmonic responses are added assuming

. random phase' angles. . Shell stresses and strains are multiplied by 1.3, other responses by 1.15. This methodology was found acceptable. See Sec-tion 3,2 for additional aetails.

4. The post-chug load was defined in the AC as the absolute summation of 50 separate harmonic loads from 1 to 50 Hz. Response above 30 Hz was very small for Oyster Creek structures so the final analysis procedure used for Oyster Creek only absolute summed the responses up to 30 Hz for calcula-tional convenience. The summation up to 30 Hz only was found acceptable for this load.

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5. The Qyster Creek SRV system differs from that of most other Mark I plants in two principal ways: Several SRV lines converge to a common header, and plant-unique Y-quenchers are used instead of GE T-quenchers. Therefore, all SRV suppression pool loads in the Qyster Creek PUAR are based on,or are modified by plant unique in-plant SRV tests of the Y-quenchers.

6. SRV discharge line clearing loads were calculated in the Qyster Creek PUAR according to AC methodology but modified to account for Qyster Creek's spe-cial Y-quencher and for several lines entering a common header.

7. SRV torus shell pressure loads were calculated in the Qyster Creek PUAR ac-cording to the guidelines given in the AC for calculations based on in-plant SRV tests. See Section 3.3 for further discussion.

8. Jet loads on submerged structures were calculated using the analytical method of the LDR and approved in the AC but with the specific Qyster Creek Y-quencher geometry.

'9. SRV air bubble drag loads were calculated using the LDR analytical model but adapted- to the Qyster Creek Y-quencher by development of an empirical factor to bound all test data similar to the one developed for the GE T-quencher. Frequency of bubble oscillation was taken from the SRV shell load analysis.

10. An analytical model was developed to model the sparger arms on the Qyster Creek Y-quencher and calculate water thrust loads on them. To ensure its adequacy, the Qyster Creek model was used to calculate water thr jst loads on a GE T-quencher and these were compared to the thrust loads calculated by the generic analytical model from the LDR. Loads calculated by the Qyster Creek model were at least 20% greater than those calculated by the LDR model.

- 11. The Qyster Creek pool temperature monitoring system which meets AC approval

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wil1 be incorporated during the next scheduled shutdown of the plar,t (Cycle 11).

12. The Qyster Creek pool temperature ahalysis was found-acceptable. See Section 3.3 for more-discussion on the Pool Temperature Limit analysis for Qyster

-Creek.

13. The Qyster Creek in-plant SRV tests were conducted using the guidelines given in the AC for the use of such tests in developing a load methodology.

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Besides the information presented in the PUAR, a sepafite report dealing with the conduct and measured results of the Y-anencher discharge tests (Reference 7) was also reviewed (Reference 8).

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3.

SUMMARY

OF THE NRC REQUEST FOR INFORMATION REGARDING THE OYSTER CREEK PUAR.

During the post-implementation audit of the Oyster Creek PUAR, various i.s-sues were identified as either exceptions to the AC or as areas where additional information was required. To resolve these issues, a request for information (RFI) (Reference 9) was sent to the licensee in order to obtain further. details to supplement the information contained in the PUAR. Most of the requested de-tails were presented by the licensee at a meeting in Washington, D. C. on July 14, 1983. This meeting was attended by GPU Nuclear, MPR Associates, as well as NRC and its BNL consultants. More information on a few items was furnished by the licensee at a later date. The material presented at the meeting .has .been be formally documented .

An overview of the RFI sent to Oyster Creek is presented in Table 2 along with~ an indication of the type and status of each item. As the table shows, two exceptions to the AC have been identified in the Oyster Creek PUAR. Both of them have been resolved. All items relative to the Mark I LTP have been resolved. For completeness, following Table 2 a brief description of any ex-ceptions to the AC and their justification is provided.

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ISSUES IDENTIFIED DURING l TABLE 2. '

1 POST-IMPLEMENTATION AUDIT -

TYPE OF' ISSUE STATUS"0FISSUL EXCEPTION REQUESTS FOR TO ADDITIONAL

ITEM DESCRIPTION NUREG-0661 AC INFORMATION, RESOLVED OPEN 1 SPECIFICATION OF'PROCE- X ' X-DURES BY WHICH OPERATOR WILL IDENTIFY SBA AND INSURE MANUAL OPERATION

, OF ADS.

2 THRUST LOADS AT VENT LINE- X X

DRYWELL INTERSECTION WHERE VENT LINE DI AMETER IS RE-DUCED. ,

3 TORUS NET VERTICAL LOADS X X BASED ON SINGLE QSIF IEST -

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14 NUMBER OF LOAD CASES AND X X

[ EXCEEDANCE PROBABILITY USED TO EVALUATE MULTIPLE DOW.NCOMER LOADING.

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t TABLE 2 (CONTINUED). ,

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TYPE OF ISSUE STATUS OF-ISSUE- .

EXCEPTION REQUESTS FOR-TO ADDITIONAL ITEM DESCRIPTION NUREG-0661 AC INFORMATION. RESOLVED OPEN 5 RANDOM PHASING OF LOAD X X l

HARMONICS TO ANALYZE STRUCTURES AFFECTED BY C0 LOADS.

i 6 TORUS PRESSURES LOAD -

X X DISTRIBUTION DURING L POOL SWELL.

Y 7 METHOD USED TO INCLUDE X X FSI EFFECTS FOR SUBMERGED STRUCTURES.

8 DETAILS OF POST-CHUG SuB- X X MERGED STRUCTURE LOAD CAL-CULATION.

9 APPLICATION OF VENT SYSTEM X X LOADS DuRING C0 AND CHUG-GING. ,

10 MODEL USED FOR Post-CHUG X X LOAD CALCULATION.

TABLE 2-(CONTINUEb) ..

. TYPE OF ISSUE STATUS'0F LSSHE .

EXCEPTION REQUESTS FOR TO ADDITIONAL-NUREG-0661 AC ITEM DESCRIPTION INFORMATION RESOLVED OPEN 11 COMPUTER MODELS USED-TO X X l CALCULATE SWELL AND VENT

, SYSTEM LOADS.

l 12 FURTHER INFORMATION ON IN- X X PLANT SRV TESTS AND METH-ODOLOGY USED TO DEVELOP

! $ SRV LOADS, AS WELL AS IN- .

FORMATION ON THE POOL TEM-PERATURE MONITORING SYSTEM AND POOL TEMPERATURE LIMITS.

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3.1 Single OSTF Test at 0 Ap. (Item 3 of Table 2).

The LDR (3) and the AC (1) specify a minimum of f,ur OSTF tests as a data base for obtaining net torus vertical loads. Most of the QSTF tests conducted for Dyster-Creek and repeated with identical conditions were carried out with a pressure differential (op) between the drywell and wetwell. The Oyster Creek PUER states that Oyster Creek will operate with zero op between drywell and wetwell and therefore this condition was the one selected for calculating net

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torus vertical loads. The PUAR also states that " loads were increased to account for the larger statistical variance associated with the smaller ~ number of tests at the 0 Ap conditions". The RFI asked for the number of tests the loads were based en and the amount the loads were increased, as well as the statistical basis for the increase. Tha licensee replied that the pool swell loads were based on a single QSTF test at 0 op, tne one with the greatest submergence to maximize the loads. The licensee further stated that GE has provided generic factors for single test statistics based on doubling the

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uncertainty for the torus shell load due to having one test instead of four in the data base. Oyster Creek used this doubling of the uncertainty to obtain new margins for their shell loads. The total margin increased from 1.215' to '1.28 for the upload and from 1.096 to 1.192 for the download in Oyster Creek. No increased margins were used for other QSTF single test derived loads.

B'NL and its consultants, including those involved in approving the original uncertainty margins for four QSTF tests, reviewed the licensee's arguments and found them acceptable. The conclusion was that doubling the uncertainty was a conservative way to -account for reducing the data base from four tests to one.

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3.2 Harmonic Phasing for C0 Re.sponse. (Item 5 of Tab 1e 2).

The C0 torus shell load is an oscillating load caused by periodic pressure oscillat' ions superimposed upon the prevailing local static pressure. The LDR

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defines the load in terms of a rigid wall pressure amplitude versUs frequency spectra from 0 to 50 Hz which is to be used in conjunction with a flexible wall coupled fluid structure model. In addition, threr alternate sets of spectral

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amplitudes are provided in the range from 4 to 16 Hz and the alternate which maximizes the response is to be used. The resulting responses from applying the amplitude at each frequency given in the total spectrum to be analyzed are to be summed. The above procedure was found acceptable in the AC because the high de-gree of conservatism associated with the direct. summation of the Fourier compo-nents of the spectrum was more than sufficient to compensate for any uncertain-ties associated with the FSTF data from which the load specification-was devel-oped. Direct application of the above ' methodology to the Oyster Creek torus proved to be too conservative and so an alternate approach based on a study per-formed in Reference 10 was used. This alternate approach considers a random phasing of the 50 harmonics rather than an absolute summation. Individual har-monic responses are added assuming random phase angles. Results are multipi f ed by 1.3 for shell stress and strain values and by 1.15 for other responses. This procedure is one of several variations for implementing phasing in the C0 load

. definition discus' sed in Reference 10 and subsequent SMA Ryorts (References 11,

12) which account for data obtained after Reference 10 was published. Reference 13 reviews the various design rules and their justification as given in Referen-ces 10,11 and 12 and discusses why they are acceptable alternatives to the LDR procedure. The method used by Oyster Creek is one which was found acceptable in Reference 13. It should also be noted that while the design rules of References 10,11 and 12 were developed from FSTF data, the Oyster Creek plant was selected in these studies as the Mark I example plant on which the design rules were ap-plied. Therefore, the effect of applying the C0 phasing methodology has been extensively documented for Oyster Creek.

m 3.3 Plant Unique SRV System and Pool Temperature Limits. (Item 12 c / Table 2). '

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The Oyster Creek SRV system differs from that of most other Mark I plants . _

in two principal ways: Several SRV lines converge to a common header, and plant '

unique Y-quenchers are used instead of GE T-quenchers. Information contained in the PUAR and Reference 7 indicated that the Oyster Creek SRV methodology was de-  ;-

veloped from in-plant tests in accordance with the guidelines of section 2.13.9 -

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of the AC. The RFI (Reference 9) requested additional d ' etails on test instru- s:

mentation and initial conditions, as well as specific numerical examples of mod- - -

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el calibration and extrapolation to design case amplitudes and frequencies. At _

the July 16, 1983 meeting in Washington, D.C. most of these details were pro-  :

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vided by the licensee. Additional information was received in a written com- 1.

munication from MPR Associates (Reference 14) and a conference call involving l All of this information confirmed NRC, BNL, MPR Associates and the licensee.

that the SRV load methodology used for Oyster Creek did indeed conform to AC -

guidelines and was developed in a conservative manner. Based on the licensee's 1 statements in, References 5 and 14, as well as verbal communications during the above-mentioned conference call,'the Oyster Creek methodology for obtaining SRV 5' loads on the torus shell and associated support systems has been found accept-  ; )a able.

3 Item 12 of the RFI also pointed out that nu discussion had been provided in 1' the Oyster Creek PVAR demonstrating that certain pool temperature limits will _

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not be exceeded during certain SRV discharge transients (Reference 15). The t

licensee subsequently supplied infonnation addressing this issue at a meeting s held on Septenter 28, 1983. The information included a description of the -

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methods used to derive suppression pool temperature response to the selected

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transients. The methods involved the use of conventional plant transient I--

calculational procedures to develop bulk pool temperature histories

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9 (Reference 16) coupled with the Monticello pool temperature test data base (Ref-erence 17) to derive local temperature. A plant-unique feature which was high-lighted at this meeting was the absence of a submerged RHR return in the Oyster

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Creek plant. This precludes reduction of the local-to-bulk pool temperature

- difference due to suppression pool circulation. Thus, only the data base which derives from the Monticello test results obtained without RHR operation was employed by the applicant. Although we are not in total agreement with the way in which the applicant employs the Monticello results, we conclude, based on our evaluation of the total computational procedure that local pool temperature will not exceed the limits dictate'd by the AC during the most severe SRV transients of interest. A' detailed description of these findings will be provided in a report which will be issued by the NRC staff in the second quarter of 1984.

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The applicant has also supplied us with a Design Report (Reference 18) in' which a detailed description of the applicant's Suppression Pool (SPTMS) Tem-perature' Monitoring System is presented. We have reviewed this material and conclude that the proposed SPTMS is designed in accordance with the AC require-ments and will provide a reasonable measure of the pool bulk temperature.

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4. CONCLUSIONS A post-implementation pool dynamic load audit of the Oyster Creek PUAR was conducted to verify compliance of the plant unique analysis with the acceptance criteria contained in NUREG-0661. As a result of the audit, several items wre identified which required additional infomation for resolution. A request for information was sent to the licensee in February,1983. At a meeting with the

. licensee in July,1983, most of the outstanding items were satisfactorily re-solved and the general conformance of the PVAR with the requirements of the ac-cept.ance criteria was confirmed. All issues relative to the Mark I long Term Implementation Program are closed as of the present date.

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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.

Material marked with two asterisks (**) is not publicly available be'cause it contains proprietary information; however, a nonproprietary version is avail-able in the NRC Public Document Rocm for inspection and may be copied for a fee.

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

All other material referenced is in the open literature and is available through public technical libr. aries.

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

(2) " Mark I Containment Short-Term Program Safety Evaluation Report", NUREG-

. 0408, December 1977.***

(3) General Electric Company, " Mark I Containment Program Load Definition Re-port", General Electric Topical Report NED0-21888, Revision 2, November 1981.* .

-(4) Mark I Owners Group, " Mark I Containment Program Structural Acceptance

, Criteria Plant-Unique Analysis Applications Cuide, Task Number 3.1.3",

General Electric Topical Report NE00-24583, Revision 1, July 1979.*

(5) "Qyster Creek Nuclear Generating Station Mark I Containment Long Term Program Plant Unique Analysis Report - Suppression Chamber and Vent Sys-tem", MPR-733, prepared by MPR Associates, Inc. for General Public Utilities Nuclear, August 1982.*

(6) "Qyster Creek Nuclear Generating Station Mark I Containment Long-Term Program Plant-Unique Analysis Report - Torus Attached Piping", MPR-734, prepared by MPR Associates, Inc. for General Public Utilities Nuclear, August 1982.*

(7) "Qyster Creek Nuclear Generating Station Test Report - Effect of Modified Discharge Device on Response of Suppression Chamber to Relief Valve Actua-tion", MPR-550, prepared by MPR Associates, Inc. for Jersey Central Power &

Light Co. , May 1978.**

(8) C. C. Lin, " Technical Evaluation Report on Qyster Creek In-Plant SRV Test Results", Draft. To be published August 1983.**

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'o (9) Letter from J. R. Lehner, BNL, to F. Eltawila, NRC dated February 25, 1983,

Subject:

Request for Information Regarding Qyster Creek PUAR.* ,

(10) " Mark I Containment Program Evaluation of Harmonic Phasing for Mark I Torus Shell Condensation Oscillation Loads", NEDE-24840, prepared by Structural Mechanics Associates for General Electric Company, October 1980.** ,

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

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

(13) To be published Third Quarter 1983.

(14) " Responses to a Request For Information From the Brookhaven National Laboratory Concerning the Qyster Creek Nuclear Generating Station Mark I Containment Long-Term Program Plant-Unique Analysis Reports", prepared for GPU Nuclear by MPR Associates, Inc., August 1983.*

(15) Letter dated December 9,1977 from George Lear to I. R. Frinfrock, en-titled "Qyster Creek Nuclear Generating Station Unit 1 - Suppression Pool Temperature Transients".

(16) Smith, P. S., Lanese, L. C. , " Report on the Effects of Electromatic Relief Valve D,ischarge on Torus Water Temperature", GPU Service TDR, Rev.1, March 1981. .

(17) Patterson, B. J. , " Mark I Containment Program - Monticello T-Quencher Thermal Mixing Test Final Report", GE Report No. NEDE-24542-P, April 1979.

(18) Hwang, J. G. , et al . , " Design Report - Suppression Pool Temperature Moni-toring System Sensor Selection and Placement - Qyster Creek Nuclear Generating Station", NUTECH Report GPN-02-101, Rev. O, January 19, 1983.

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