Similar Documents at Maine Yankee, 05000000 |
---|
Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20211M4841999-08-31031 August 1999 Replacement Pages 2-48,2-49 & 2-50 to Rev 14 of Defueled Sar ML20211D7111999-08-0909 August 1999 Rev 17 to Maine Yankee Defueled Safety Analysis Rept (Dsar) ML20155G9591998-11-0303 November 1998 Rev 1 to Post-Shutdown Decommissioning Activities Rept ML20236X1751998-08-0303 August 1998 Rev 16 to Defueled Sar ML20247D3451998-05-0606 May 1998 Rev 15 to Defueled SAR, Replacing List of Effective Pages ML20216D7681998-02-25025 February 1998 Rept to Duke Engineering & Services,Inc,On Allegations of Willfulness Related to Us NRC 971219 Demand for Info ML20202E0541998-01-30030 January 1998 Rev 14 to Myaps Defueled Safety Analysis Rept ML20198R6371997-11-0606 November 1997 Yankee Mutual Assistance Agreement ML20155G9511997-10-31031 October 1997 Rev 1 to M01-1258-002, Decommissioning Cost Analysis for Myaps ML20217R1051997-08-27027 August 1997 Myaps Post Shutdown Decommissioning Activities Rept ML20151M1351997-07-21021 July 1997 Rev 0 to Technical Evaluation 172-97, Cable Separation Safety Assessment Rept ML20149M3221996-12-10010 December 1996 Response to Independent Safety Assessment of Myap ML20135D3571996-11-30030 November 1996 Rev 1 IPEEEs for Maine Yankee Atomic Power Station ML20127P2561993-01-31031 January 1993 Rev 0 to Licensing Rept for Maine Yankee Atomic Power Company High Density Spent Fuel Pool Reracking Project ML20141M1201991-12-0202 December 1991 Criticality Analysis of Maine Yankee Spent Fuel Storage Racks to Allow Up to 3.95 W/O U-235 Fuel ML17347B4621989-12-31031 December 1989 App a to USI A-46 & Generic Ltr 87-02. ML20246D6871989-08-14014 August 1989 Rev 1 to Criticality Analysis of Byron & Braidwood Station High Density Fuel Racks ML19327B4011989-07-31031 July 1989 Safety Evaluation for Byron/Braidwood Stations Units 1 & 2 Transition to Westinghouse 17 X 17 Vantage 5 Fuel. ML18008A0311989-07-31031 July 1989 NTH-TR-01 Decrease in Heat Removal by Secondary Sys. ML20247H0711989-06-30030 June 1989 Description & Verification Summary of Computer Program, Gappipe ML20246D6711989-06-30030 June 1989 Criticality Analysis of Byron/Braidwood Fresh Fuel Racks ML20247H0791989-06-22022 June 1989 App to Description & Verification Summary of Computer Program,Gappipe ML20247N0621989-05-31031 May 1989 Production Training Dept,Braidwood,Malfunctions & Initial Conditions ML20247K3011989-05-12012 May 1989 Leak-Before-Break Evaluation for Carbon Steel Piping ML20247L1841989-05-12012 May 1989 Leak-Before-Break Evaluation for Stainless Steel Piping, Byron & Braidwood Nuclear Power Stations Units 1 & 2 ML18094A3551989-04-30030 April 1989 Assessment of Impacts of Salem & Hope Creek Generating Stations on Kemps Ridley (Lepidochelys Kempi) & Loggerhead (Carretta Caretta) Sea Turtles. ML20247F1321989-03-23023 March 1989 Post-Tensioning Sys Evaluation,Callaway Unit 1 Containment & Wolf Creek Unit 1 Containment ML20244B7851989-02-28028 February 1989 115-kV Capacitor Bank Design Rept ML20005G4211989-02-28028 February 1989 Reactor Vessel Heatup & Cooldown Limit Curves for Normal Operation. ML17251A4811989-02-28028 February 1989 Ultrasonic Indication Sizing Technique Development. Related Info Encl ML20206C7451988-11-30030 November 1988 ATWS Mitigation Sys Specific Design for Byron/Braidwood Stations, Rev 5 ML20205T7501988-11-0404 November 1988 Detection & Skin Dose Evaluation for Characteristic X-Ray in Activation Product Contamination ML20206K3021988-10-31031 October 1988 Rev 1 to Impact of Reg Guide 1.99,Rev 2 on Peach Bottom Atomic Power Station Unit 3 ML20206K3101988-10-31031 October 1988 Rev 1 to Impact of Reg Guide 1.99,Rev 2 on Limerick Generating Station Unit 1 ML20154N6081988-09-30030 September 1988 Rev 1 to Identification of Unisolable Piping & Determination of Insp Locations ML20154K2091988-09-0909 September 1988 Rev 0 to Response to NRC Bulletin 88-005,Nonconforming Matls Supplied by Piping Supplies,Inc at Folsom,Nj & West Jersey Mfg Co.... Proprietary Procedure 1404.1, Leeb Hardness Testing (Equotip).... Encl.Procedure Withheld ML20154K3971988-08-23023 August 1988 Environ Qualification Enforcement Conference ML20245B4181988-08-17017 August 1988 Investigation Rept,Design & Operation of Sampling Sys for Analysis of High Purity Water ML17347A7981988-06-16016 June 1988 Radiological Data Prepared for Resolution of USI A-46. ML17347A7971988-06-16016 June 1988 Seismic Hazard Data Prepared for Resolution of USI A-46. ML20150F2941988-05-31031 May 1988 Rev 4 to, ATWS Mitigation Sys Specific Design for Byron/ Braidwood Stations ML20196L6281988-05-20020 May 1988 Rev 2 to ATWS Mitigation Sys Actuation Circuitry (Amsac) ML20196L6421988-05-0606 May 1988 ATWS Mitigation Sys Actuation Circuitry Response to Unit Transients ML20151H9581988-04-30030 April 1988 CASMO-3G Validation ML20151B8061988-03-31031 March 1988 Thermal Shield Status Rept ML20150E4111988-02-29029 February 1988 Criticality Analysis of Maine Yankee Spent Fuel Pool & New Fuel Vault ML20151T4681988-01-31031 January 1988 Experimental & Finite Element Evaluation of Spent Fuel Rack Damping & Stiffness ML20148G5431988-01-15015 January 1988 Nonproprietary Mods to Critical Flow Model in RELAP5YA ML20155K1391987-12-30030 December 1987 Final Rept MSIV 3-Way Dual Solenoid Valve Failures ML20237F0161987-12-23023 December 1987 Rev 1 to MYC-430, Maine Yankee Auxiliary Power Sys Voltage Study 1999-08-09
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20211M4841999-08-31031 August 1999 Replacement Pages 2-48,2-49 & 2-50 to Rev 14 of Defueled Sar ML20211D7111999-08-0909 August 1999 Rev 17 to Maine Yankee Defueled Safety Analysis Rept (Dsar) ML20196K4821999-07-0606 July 1999 Safety Evaluation Concluding That Because of Permanently Shutdown & Defueled Status of Myaps Facility,Confirmatory Orders No Longer Necessary for Safe Operation or Maint of Plant ML20206H1611999-05-0505 May 1999 Safety Evaluation Supporting Amend 164 to License DPR-36 ML20206G5731999-05-0303 May 1999 Safety Evaluation Supporting Amend 163 to License DPR-36 ML20205D5261999-03-26026 March 1999 SER Accepting Util Rev 1 to CFH Training & Retraining Program for Maine Yankee.Rev 1 to CFH Training & Retraining Program Consistent with Current Licensing Practice for Facilities Undergoing Decommissioning ML20204C4631999-03-16016 March 1999 Safety Evaluation Supporting Amend 162 to License DPR-36 ML20206D7491998-12-31031 December 1998 Co Annual Financial Rept for 1998. with ML20155G9591998-11-0303 November 1998 Rev 1 to Post-Shutdown Decommissioning Activities Rept ML20155D8651998-10-28028 October 1998 Public Version of, Maine Yankee Emergency Preparedness Exercise ML20197C8231998-09-0303 September 1998 Safety Evaluation Supporting Request for Exemption from Certain Requirements of 10CFR50.54(q),10CFR50.47(b) & (C) & App E to 10CFR50 Re Emergency Planning ML18066A2771998-08-13013 August 1998 Part 21 Rept Re Deficiency in CE Current Screening Methodology for Determining Limiting Fuel Assembly for Detailed PWR thermal-hydraulic Sa.Evaluations Were Performed for Affected Plants to Determine Effect of Deficiency ML20236X1751998-08-0303 August 1998 Rev 16 to Defueled Sar ML20236U8451998-07-24024 July 1998 Safety Evaluation Accepting Rev 14 to Maine Yankee Atomic Power Co Operational QA Program ML20248D3011998-05-26026 May 1998 Rev 14,page 16 of 17,Section II of QA Program ML20247J0211998-05-11011 May 1998 Revised Page 16 of 17 of Section II of QA Program,Rev 14 ML20247D3451998-05-0606 May 1998 Rev 15 to Defueled SAR, Replacing List of Effective Pages ML20217J9811998-04-28028 April 1998 Part 21 Rept Re Incorrect Description of Drift Specification for Model 1154,gauge Pressure Transmitters,Range Code 0 in Manual Man 4514,Dec 1992.Cause Indeterminate.Will Issue & Include Errata Sheets in All Future Shipments to Users ML20217H5241998-03-30030 March 1998 Safety Evaluation Supporting Amend 161 to License DPR-36 ML20217D9691998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Maine Yankee.W/ ML20216D7681998-02-25025 February 1998 Rept to Duke Engineering & Services,Inc,On Allegations of Willfulness Related to Us NRC 971219 Demand for Info ML20202F3771998-01-31031 January 1998 Annual Rept of Facility Changes & Relief & Safety Valve Failures & Challenges ML20203A3011998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Maine Yankee ML20202E0541998-01-30030 January 1998 Rev 14 to Myaps Defueled Safety Analysis Rept ML20199K3211998-01-27027 January 1998 Rev 13 to Maine Yankee Atomic Power Co,Qa Program ML20199K3471998-01-22022 January 1998 Rev 14 to Maine Yankee Atomic Power Co QA Program ML20199C2281997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Maine Yankee ML20217R2301997-12-31031 December 1997 Myap Annual Financial Rept for 1997 ML20203F8551997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Maine Yankee ML20202D3181997-11-26026 November 1997 Safety Evaluation Supporting Amend 160 to License DPR-36 ML20198R6371997-11-0606 November 1997 Yankee Mutual Assistance Agreement ML20155G9511997-10-31031 October 1997 Rev 1 to M01-1258-002, Decommissioning Cost Analysis for Myaps ML20198P9431997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Maine Yankee ML20217K5191997-10-24024 October 1997 Part 21 Rept Re Five Valves That May Have Defect Related to Possible Crack within Forging Wall at Die Flash Line.Caused by Less than Optimal Forging Temperatures.Newer Temperature Monitoring Devices at Forging Area Heating Ovens Procured ML20211N0571997-10-0707 October 1997 Revised Pages to Jul/Aug 1994 SG Insp Summary Rept ML20198K0201997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Maine Yankee ML20199H1861997-09-25025 September 1997 Rev 12 to Maine Yankee Atomic Power Co,Qa Program ML20217A9251997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Maine Yankee ML20217R1051997-08-27027 August 1997 Myaps Post Shutdown Decommissioning Activities Rept ML20216F1511997-08-0808 August 1997 Safety Evaluation Supporting Amend 159 to License DPR-36 ML20210M4451997-07-31031 July 1997 Monthly Operating Rept for July 1997 for Myaps ML20151M1351997-07-21021 July 1997 Rev 0 to Technical Evaluation 172-97, Cable Separation Safety Assessment Rept ML20141H2691997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Maine Yankee ML20141B9951997-05-31031 May 1997 Monthly Operating Rept for May 1997 for Maine Yankee.W/ ML20141G6511997-05-19019 May 1997 Safety Evaluation Supporting Amend 158 to License DPR-36 ML20138G3541997-05-0202 May 1997 Safety Evaluation Supporting Amend 157 to License DPR-36 ML20141G2871997-04-30030 April 1997 Monthly Operating Rept for Apr 1997 for Maine Yankee ML20137Z1971997-04-14014 April 1997 Forwards to Commission Results of Staff Evaluation of Performance of Licensees W/Ownership Structure Similar to Plant ML20138B1491997-03-31031 March 1997 Monthly Operating Rept for Mar 1997 for Maine Yankee ML20137N7541997-03-31031 March 1997 Rev 11 to Operational Quality Assurance Program 1999-08-09
[Table view] |
Text
.- .. ,, - ,_
W h. ,
/ , *[
%I
,posine vNa D*W /,-al- 7/ & >'
3, d b' y7 JUN 2 8W7W 7 aw -u 77
- i. crr Wu Ed%"'l 9, un ::a p tu s REVIEW OF THE SE SMIC ANALYSIS AND DESIGN
- FOR THE i MAINE YANKEE ATOMIC POWER STATION (DocketNo.50-309)
{.
.m c
, W 1
June 18, 1971 l
) -
?
l JOHN A. BLUME & ASSOCIATES, ENGINEER 3
( San Francisco, California 1
~
~d G .l
[ 8508130233 850703 9 PDR FOIA i
{
HERRMAN85-301 PDR q
_ Y
_a : f 7 _ .. ;... x ;;.:&
k 11 h REVIEW OF THE SEISMIC ANALYSIS AND DESIGN FOR THE MAINE YANKEE ATOMIC POWER STATION (Docket No. 50-309)
This report summarizes our review of the engineering factors pertinent to the seismic and structural adequacy of the Maine Yankee Atomic Power Station. The plant is located on the west shore of the Back River approximately 3 9 miles south of.the center of Wiscasset, Maine. The design and construction of the plant was performed by Stone and Webster i Engineering Corporation. The Nuclear Stea i Supply System (NSSS) was :
supplied by Combustion Engineering, Inc. Westinghouse Electric Corporation ,
has supplied and erected the turbine generator. The pressurized water
~
type reactor plant has a capacity of cenerating 2,440 MWt (830 MWe).
Application for an operating IIcense has been made to the Atomic Energy 1
Commission (Docket No. 50-309) by the Maine Yankee Atomic Power Company.
I The Final Safety Analysis Report (FSAR) has been submitted in support of f the applicatinn to show that the plant has been designed and constructed 1
In a manner which will provide for safe and reliable operation and safe f shutdown in case of a loss-of-coolant accident. Our review is based upon the iriformation presented in the FSAR and is directed specifically
- j towards an evaluation of the seismic design of Class I structures, I
systems, and components. The list of reference documents upon which l ,
' this report is based is given at the end of this report. 11 DESCRIPTION OF FACILITY '
TheMaineYankeeAtomicPowerPlantshteislocatedinastablegeologic region where no major geologic changes other than those produced by glactation have occurred. The overburden at the site consists of medium t soft to medium stiff silty clay varying from 15 to 20 ft in thickness, g ,,
Underlying this layer is steeply dipping schistose bedrock of the Cape
- p Elizabeth formation interlayered with granite and coarse crystalline ,-
pegmatite.
It is stated that the major structures are directly founded I on hard crystalline rock.
i E .
s I
I l
The containment structure is a reinforced concrete cylindrical structure topped with a hemispherical dome. The reinforced concrete foundation slab is 10 ft in thickness and is bearing on the bedrock. The cylinder is of 135'0" inside radius with 4'6" wall thickness. The hemispherical dome is of the same inside radius with 2'6" wall thickness. The spring line of the dome is at 102'0" above top of the foundation stab. The containment structure is completely lined on the inside with a steel liner plate which is 1/4", 3/8" and 1/2" thick at top of the foundation slab, vertical walls, and dome respectively.
The primary auxiliary building and fuel building are shear wall type reinforced concrete structures located north of the containment structure.
e STRUCTURAL DESIGN CRITERI A AND LOADS All structures, equipment, systems, and piping are classified according to function or consequence of failure as either Class I or it, as defined in Section 5.4 of the Safety Analysis Report. Class I structures, systems, and equipment'are those whose failure could cause uncontrolled release of radioactivity or are those essential for immediate and long-term operation following a loss-of-coolant accident. They are designed to withstand the
. appropriate seismic loads simultaneously with other appilcable loads without loss of function. Class 11 structures, systems, and equipment are those .
whose failure would not result in a release of radioactivity and would not prevent reactor shutdown but may Interrupt power generatlon.
h The design loads for the Maine Yankee Atomic Power Station are basically l: divided in two categories. The first category includes dead loads, ice
and snow loads, normal live loads, operating loads, normal temperature
.j loads, hydrostatic loads, etc. The second category includes seismic loads due to Design Earthquake (DE) and Hypothetical Earthquake (HE), tornado ,
i loads and tornado-blown misslie loads, turbine generator missile loads, accidental pressure and temperature loads, accidental pipe rupture loads a
.i 8 'b..
v~ ,
gl etc. The structure design loads were increased by load factors based I
, on probability and conservatism of the predicted design Icads. It is stated
.c y,- [j that these Increased designed loads were used for the design of Class I structures by the ultimate strength method. Capacity reduction factors
~
,s y j.,]. ;[ were applied to the yield stresses allowed by the applicable codes. ,
... . .r p', ,
t f ADEQUACY OF THE SEISMIC ANALYSIS AND DESIGN We have reviewed the Final Safety Analysis Report, Volumes I and 2, and
, d Amendments numbered 14 and 17.to 25. Our review also Ir.cluded discussions with DRS and DRL staff during meetings on January 21, 1971 and May 19, 1971; n ,; r' L
data gained during a site visit on May 18, 1971; and discussions with DRS J : ' and DRL staff and the applicant during a conference call on June 18, 1971.
4, _..e. <
F'? We have the following coments regarding the seismic analysis and design.
- . 2.
c.' l. The maximum horizontal ground acceleration (Section 2.5.1) used by !,
"z the applicant in the seismic design of the Class i Items was 0.059 i
- ..i N,j; e, se w .'
for the Design Earthquake (DE), and 0.10g for the Hypothetical Earth-
% M.' quake (HE) . The maximum vertical ground acceleration was assumed to l.
- ..y:, be equal to two-thirds of the maximum horizontal ground acceleration. p
't a t IN G '
Horizontal and vertical ground accelerations were assumed to act simul-
'm .4 .
taneously.
?
so
>n #
s' The horizontal acceleration response spectra cruves (Section 2.5.~4) used in the seismic design of the Class I items are shown in Figure d
2.5-6 and 2.5-7 for the DE and HE respectively. These spectra are i as originally defined in TID 7024.
] !
These criteria were accepted prior to the issuance of the construction permit.
,1
- 2. The applicant has performed dynamic analysis of all Class I structures. t Although certain approximations were made in the analytical techniques, we understand that the resulting seismic stresses are quite low, and 1
therefore the seismic design of these structures should be adequate.
i 3
e.
$ = .I l. ,
0 $ -*, O *. p g.4) " % * *e . A# p9i,W . -
~ ~ ' ~ ~ ~ ~ ~ ~ ' '
1.___ __ _ ~_________. _ . _ . . ~~_____ ____ _____ _ _____ .
I I
\ .
- 3. . w 4 ' . * . m .. c e : e._ % e.m o ; h g e ;. _e . , . . . . . . . /
v vy .' di
~$& u -
- ,+ .M.4>pustAt!Tn %enx4n..a v:-g w. u.s.-< .c g._3.f . . e g ; . .. . v. t .. ,jIne . We understand that modification will be made if required by the follow-up analysis.
- 4. For Class I piping systems other than the reactor coolant system, and for Class I equipment, the applicant has stated that he has initiated a program to verify the adequacy of this piping and equipment and/or l Identify those systems which require modification.
l
> a. The applicant has proposed to perform static analysis of Class I piping and equipment systems, by using a static horizontal accel-eration equal to 1.3 gX' peak ground acceleration.
@ t
- c. The applicant will revise, expand and update Tables 4.6.1 and 4.6.2 showing comparison of actual stresses and allowable stresses in the NSSS. These revised tables will be submitted for review.
- d. The applicant will use appropriate damping ratio in the analysis of
.all Class I piping.
This program, when completed, will provide assurance of the adequacy of the seismic design of Class I ;:! ping and equipment.
- 5. The applicant has stated that Class li structures are either designed to adequately withstand Class I loads without failure, or in case of i a collapse of a Class il structure, the Class il structure will col- l 1
lapse away from the adjacent Class I structure. Accordingly, the ap- i plicant has stated that the Turbine Building and the adjoining Class l Control Room Building have been designed for Class i design criteria.
.j -4
.: f . s h : ( $ Q M; 3 " 3s 4 :- 4
. *6 ' l.
5$.01. *? $:l.Y.%k-
bi'. - 83 -
. t I CONCLUSION _
On the basis of the information presented by the applicant in the Final f
- Dj_ Safety Analysis Report and Amendments 14 and 17 to 25, and provided that adequate analysis and evaluations and implementation in construction are
[
performed for the reactor coolant system and all Class I piping systems, discussed in Comments equipment, instrumentation and control panels, as i
T~
3 and 4, it is our opinion that the approach to the seismic analysis and I
.f design for the Maine Yankee Plant as outlined in the FSAR, and Amendments l
j will have resulted in a design that is adequate to resist the earthquake j l conditions postulated for the site,
~
.5 JOHN A. BLUME r, ASSOCIATES, ENGINEERS
[-
(} jm 3 vv~-
. r Natvar M. Chauhan l
s
.g y.f .
- hfrrft Roland L. Sharpe 0
-(
.a.
O.-
o
- \
"', a 7, .l
.f* m l ,
t 8
k d I
. i II l
I I-s
, ) ':
i 3, l .
? . Q. ~
~~ ~
.3 e ., . , ,-
,t s " ; . _
~
l' ,? *, , / % 4.0 .* ** ,,
(
~
i ,.
- YC.$ .
je ,
). .
- ' N.
. bf;.
REFERENCES
. 1u .se ...
. .' s
, 1. " Maine Yankee Atomic Power Station, Final Safety Analysis Report."
,- )
Volumes I and 11,* Including Amendment no. I4, and 17 through 25 by Maine Yankee Atomic Power Company, Wiscasset, Maine.
S>.
n ... -
-["['E 3
1
, '4 i
. i i
v u i
.I -
if 1
!! d l,
3 1
l t: :i I
7 l
l