ML20148T464
| ML20148T464 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 12/04/1978 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20148T460 | List: |
| References | |
| NUDOCS 7812050203 | |
| Download: ML20148T464 (146) | |
Text
{{#Wiki_filter:t ( ACNGS- PSAR i' HOUSTON LIGHTING & PORER CCMPANY ALLENS CREEK NUCLEAR GENERATING STATION - UNIT NO.1 PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT NO. 48 INSTRUCTION SHEET This amendment contains additional information which is submitted in response to outstanding safety review issues identified by NRC Ictter dated November 7,.1978 as well as updated information. Each revised page bears the notation Am. No. 48,12/4/78 at the bottom of the page. Vertical bars with the number 48 representing Amendment No. 48 have been used in the margin of the revised pages to indicate the location of the revision on the page. The revised ' pages beve the question number (eg Q010.5) next to the appropriate information which responds to the question. The following page removals and insertions should be made to incorporate Amendment No. 48 into the PSAR. CHAPTER 2 Remove Insert _(Existing Pages) (Amendment No. 48 Pages)
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() 3* 14* 3* 14* 2.1-2a 2.1- 2a 2.2.A-1 2.2.A-1 2'.5-46b 2. 5-46b CHAPTER 3 1* 1* 1a* 1a* 5* 5* 6* 6* 8* 8* 12* 12* 15* 15* 17* 17*
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ACNGS-PSAR Remove Insert (Existing Pages) (Amendment No. 48 Pages)
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5* 5* 13.1-3 13.1-3 13.1 13.1-9 13.1-10 13.1-10 13.1-11 13.1-11 13.1-11a - 13.1-12 13.1-12 13.1-12a 13.1-12a 13.1-12b Figure 13.1-7 Figure 13.1-7 Figure 13.1-8 Figure 13.1-8 13.1A-2 13.1A-2 13.1A-9 13.1A-9 13.1A-11 13.1A- 11 13.1A- 12 13.1A-12 CHAPTER 17 1* 1* pg 2* 2* 3* 3* () 17.1-2 17.1-2 17.1-4 17.1-4 17.1-5 17.1-5 17.1-6 17.1-6 17.1-70b 17.1-70b Figure 17.1A-1 Figure 17 lA-1 Figure 17.1A-2 Figt re 17.1A-2 Figure 17.1A-3 Figure 17.1A-3 APPENDIX C 1* 1* 3* 3* C1.124-1 C1.124-1 C1.130- 1 C1.130-1 , 1 APPENDIX I 1110.17-2 1110.17-2 1110.17-3 O 111 Am No. 48, 12/4/78
ACNGS-PSAR Remove Insert (Existing Pages) (Amendment No. 48 Pages) APPENDIX K ( K130.18 K130.18 , APPENDIX M M110.3 and 4 M110.3 and 4 M110.6(3)-1 M110.6(3)-1 M130. 20- 1 M130.20-1 M211.2 M211.2 M211.3 M211.3 M211.22 M211.22 M211.26 (1)-1 M211.26(1)-1 M211.26(1)-2 M211.26(1)-2 M361.4-1 M361. 4- 1 APPENDIX N
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EFFECTIVE PAGES LISTING (Cont'd) O CHAPTER 2 ! j SITE CHARACTERISTICS Page No. Amendment No. 2.5-46 20 2.5-46a 20 2.5-46b 48 2.5-47 - 2.5-48 20 2.5-48a -20 2.5-48b 20 2.5-48c 20 2.5-48d 20 2.5-48e 20 2.5-48f 20 2.5-48g 20 2.5-48h 20 2.5-49 36 2.5-50 -- 2.5-51 38 2.5-51a 36 2.5-52 38 2.5-52a 36 2.5-53 4 O '2.5-54 2.5-54a 2.5-55 41 41 42 2.5-55a 42 2.5-56 42 2.5-57 36 2.5-58 36 2.5-59 20 2.5-60 20 2.5-61 20 2.5-62 20 2.5-62a 20 2.5-63 20 2.5-63A 20 2.5-64 20 2.5-65 20 2.5-66 42 2.5-66a 20 2.5-67 20 2.5-67a 20 2.5-68 38 2.5-68a 3 2.5-68b 4 O V 14 Am. No. 48, 12/4/78
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r ACNGS-PSAR property by directional drilling or other means, so long as the Company's ,1 5 use of this property is not interfered with, disturbed or damaged. By reason of having full and legal surface title to the land, Houston Lighting & Power Company has the authority to regulate any and all activi-ties within the exclusion area, including the exclusion of personnel and property. The implementation of this authority is the respo,nsibility of the plant supervisory staff, who will be aided by (1) a barbed-wire fence , around the land portion of the exclusion area, and (2) floating buoys i around the lake portion. Both the fence and the buoys will be clearly 42 designated as the limits of the exclusion area. In addition, the buoys (U) will be connected by nylon rope and illuminated at night. There are to be no persons living within the exclusion e ea. There is to be no one working within the exclusion area except employees of HL6P and its authorized agents, Arrangements will be made for the immediate evacua-tion (of all non essential personnel) in case of a major accident. 2.1.2.2 Boundaries for Establishing Ef fluent Release Limits ' The coincident restricted / exclusion area is irregular in shape and encom-passes a nominal 2,064 acres, 1,137 acres of which is on land, -l 35 (U) O a 2.1-2a (U)-Update Am. No. 48, 12/4/78
ACNGS-P- l im ( ) l APPENDIX 2. 2-A s v/ i l PIPELINE BREAK EVALUATION 1.0 BREAK OF 6" LPG LINE FILLED WITH PROPANE The consequences of a complete severance of the 6 inch LPG line, assumed to be pipina propane have been evaluated on the basis of the following assump-tions: a) Double ended rupture of the line occurs instantaneously and at the closest point to plant Catecory I structures (7,000 feet). b) The released petroleum liquid-gas mixture escapes from the break at the critical velocity for two phase flow, and at the design pressure of the line, 1,000 psin (a conservative assumption since the oper-atinc pressure is only 750 psi). c) The temperature of the atmosphere is assumed to be 72 F. Higher temperatures would lead to higher vaporization of escaping propane, 01 but the flow rate would be less due to the higher quality at the 9 exit plane. 2.6 d) Five percentile meteorolony is assumed, which is equivalent to a Pasquill F inversion with wind speed of 0.8 mps in the direction of the plant structures. U 1.1 CALCULATION OF FLOW RATE OUT OF THE BREAK P The propane in the line will, upon the instant of the break, decompress isenthalpically to a saturation pressure of 125 psia immediately because of the very large speed of sound in the liquid. A decompression wave will travel very rapidly away from the break leaving the fluid behind at the saturation pressure. Since propane would issue from the break at 72 F, approximately 1/3 of it would quickly vaporize, coolina the remainder to its boiling point of about - 44 F. Hence the process of decompression is described by the throttline process shown in Figure 1.1 . From that figure the exit plane cuality, x, of the fluid can be estimated from: v=v +xv 48 I E (U) 3 3 3 where v=2.4 f t /lb, v =.0275 ft /lb, v =6.6 f t /lb, v =v -v f e fg g f Hence x = 0.36 # To estimate the flow rate out of the break, Fauske's equation 1/ for critical tuo-phase mass velocity is used: G = (-c/(kydvg /dp+k dx/dp+k d# dp)) Crit 2 3 f ( L , (U)-Update 2.2.A-1 Am. No. 48, 12/4/78 i
l ACNCS-PSAR
/~'g t The net result of this exploration in the site area has been negative. In
\s_s light of _ the exhaustive efforts and negative results of oil exploration in this site area, the likelihood of future hydrocarbon production here is 3 minimal . Q2.63 A review of oil lease maps in the site vicinity indicate very few leases purchased. In addition, no leases were in effect on any of the 11,152 acres of land purchased by HL6P for the ACNGS site. This is further evi- l dence that significant future hydrocarbon production should not be expected i in the vicinity. l All surface rights have been acquired within the revised exclusion area boundary (Fig. 2.1-2) . Also, HL6P controls all mineral interests within 48 the revised exclusion area boundary. (U)' The mineral interests that are subject to HL6P's agreement with former 8 landowners give the Company broad enough powers to enforce any substantive Q1-criteria or restrictions placed on mineral extraction. Brie fly these cri- 13.67 teria are encompassed with the scope of the words that the property must not be "interf ered with, disturbed , or damaged ." O (
'N (U)-Update 2.5-46b Am. No. 48, 12/4/78
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ACNGS-PSAR~ EFFECTIVE PAGES LISTING O DESIGN OF STRUCTURES2 COMPONENTS. EQUIPMENT AND SYSTEMS CHAPTER 3 , Page. Amendment 3.3-1 35 3.3-2 35 3.3-3 48 3.3-4 35 3.3 35- _3.3-6 35 3.3-7. 35 3.3-8 35 3.4-1 45 3.4-la 35 1 3.4-2 35 l 3.4-3 35 l 3.4-4 37 3.4-5 35 3.5-1 35 3.5-2 48 3.5-2a 35 3.5-3 35 3.5-4 35 ( 3.5-5 35 l 3.5-6 37
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ACEGS-PSAR EFFECTIVE PAGES LISTING CHAPTER 3 DESIGN OF STRUCTURES, COMr' EQUIPMENT AND SYSTEMS Page taendment 3.9-17 3 i 3.9-18 3 I 3.9-19 - I 3.9-20 - 1 3.9-21 5 3.9-22 8 3.9-22a 12 3.9-22b 35 l 3.9-22c 35 3.9-22d 35 3.9-22e 35 3.9-22f 35 3.9-22g 47 3.9-22h 48 3.9-23 - r~ f
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ACNGS-PSAR EFFECTIVE PAGES LISTING CHAPTER 3 MGN OF STRUCTURES. COMPONENTS. EQUIPMENT AND SYSTEMS PAGE AMENDMENT NO. 3.11-8 35 , 1 3.11-9 '35-3.11-10 35 l 3.11-11 35 3.11-11a 35 3.11-12 - 3.11-13 -- 3.11-14 45 3.11-15 45 3.11-16 45 3.11-17 45 3.11-18 48_ 3.11-19 35 3.11-'20 35 3.11-21 - 3.11-22 35 3.11-23 - 3.11-24 - 3.11-25 - 3.11-26 - 3.11-27 17 : 15 Am. No. 48,12/4/78
ACNGS-PSAR' EFFECTIVE FIGURE LIST
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ACNCS-PSAR TABLE OF CONTENTS CHAPTER 3 (CONT'D)
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Section Title- Fane 3.7.4 ' SEISMIC INSTRUMENTATION PROGRAM ' 3.7-28a 3.7.4.1. Comparison with AEC Regulatory Guide 1.12 3.7-28a. 3.7.4.2 Location and Description of Instrumentation 3.7-28d 3.7.4.3 Control Room Operator Notification 3.7-26d 3.7.4.4 Comparison of 'Nhasured and Predicted Responses 3.7-28d 3.7.5 SEISMIC DESIGN CONTROL' MEASURES 3.7-28d 3.7.5.1 Seismic Input Data of Purchase Speci-fications of Seismic Category I Components and Equionent 3.7-28d 3.7.5.2 Program for Auditina Vendor Seismic Analyses and Tests of Seismic Category I Components and Equipment 3.7-29
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3.7 REFERENCES
3.7-30 3.7.A SEISMIC DESIGN CONSIDERATIONS - 3.7.A-1 3.7.A REFERENCES 3.7.A-7 48(U) ) l 1 i Il ks ' (U)-Update xix Am. No. 48,12/4/78
ACNGS LIST OF TABLES (CONT'D) Table Title Page es train Da ta 3.6-37
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') 3.6-1
- 3. 6- 2 Comparison of PDA and NSC Codes 3.6-38
- 3. 6-3 Type Restraint Testing Program Results 3.6-39 3.6-4 Piping Design Parameters 3.6-40 3.6-5 Table of High Energy (1) Systems Outside Containment 3.6-42
- 3. 6- 6 ' Table of Moderate Energy (1) Systems Outside containment 3.6-43 3.6-7 Stress Level Criteria for No Breaks in Class I Piping Between Isolations Values 3.6-46 3.7-1 Damping Factors Used in Seismic Analysis 3.7-31
- 3. 7- 2 Summary of Results of the Fuel Vibration Damping Measurements 3.7-32 3.7-3 Comparison of Analytical and Test Results 3.7-32a 44(C) 3.7-4 Maximum Stress Comparison 3.7-32b O
f4 v 3.7-5 Structural characteristics for Category I Buildings 3.7-32c
- 3. 7- 6 Earthquake cumulative Distribution 3.7-32d 3.7.A-1 Identification of the Various Analyses Perfomed 3.7.A-8 3.7.A-2 Input Parameters for Elastic Half-Space Approach 3.7.A-9 48(U) 3.7.A-3 Comparison of Maximum Horizontal Accelerations Flush - a vs Flush - b 3.7.A-10 3.7.A-4 Comparison of Maximum Horizontal Accelerations Flush - c vs Spring - c 3. 7. A- 11 3.7.A-5 Comparison of Maximum Horizontal Accelerations Flush - a vs Spring - b 3.7.A-li 3.7 A-6 Comparison of Maximum Horizontal Accelerations Flush - a vs Spring - a 3.7.A-13 3.7.A-7A Comparison of Shears and Moments Reactor Shield Building Spring - a vs Flush - a 3.7.A-14 3.7.A-7B Comparison of Shears and Moments RPV and RPV Pedestal
[] Spring - a vs Flush - a 3.7.A-15 k,) xxxiii (C)-Consistency (U)-Update Am. No. 48, 12/4/78
ACNGS-PSAR LIST OF TABLES (CONT'D) Table Title Page /T 3.7.A-7C Comparison of Shears and Moments Containment (s_,/ Spring - a vs Flush - a 3.7.A-16 48 3.7.A-7D Comparison of Shears and Moments Drywell (U) Spring - a vs Flush - a 3.7.A-17 3.7.A-8 Deconvoluted Zero Period Acceleration 3.7.A-18 3.8-1 Stress Limits for containment Vessel 3.8-79 46 3.8-2 Buckling Criteria for Containment vessel 3.8-81 (C) 3.8-3 Load Combinations and Load Factors 3.8-82 3.8-4 Load Combinations for Cuard Pipes - Deleted 3.8-83 3.9-1 ASME Code Class 2 and 3 Components 3.9-11 3.9-2 Design Loading Combinations for ASME Code Class 2 and 3 Components 3.9-16 3.9-3 Allowable Stresses for ASME Class 2 and 3 Components 3.9-17 3.9-4 Safety Related Components Not Covered by ASME Code 3.9-19 g
) 3. 9- 5 Applicable Codes and Standards for Heating, \s_/ Ventilating and Air Conditioning Systems and Components 3.9-22a ,m
(
) (C)-Consistency xxxiiia (U)-Update Am. No. 48, 12/4/78
ACNGS- PSAR LIST OF FIGURES (CONT'D) O v) ( Figure 3.7-22 Title Seismic Protection Analysis Sample Problem No. 2 Sheet 2- l 3.7-23 Seismic Protection Analysis Sample Problem No. 3 3.7-24 -Analytical Model 3.7-25 Convergence of the Wilson-0 Method for a Single Degree of Freedom Undamped System 3.7-26 , Time History Response - Cantilever Beam 3.7-27 Loading - Free End of Beam 3.7-28 Distributed Load 3.7-29 Shear at Base - Constant. Load case 3.7-30 Moment at Base - Constant Load Case 3.7-31 Shear at Base - Multi Input Case 3.7-32 Moment at Base - Multi Input case 3.7.A-1 North-South Croos-Section For Analyses 3.7.A-2 Representative Nodal Points Reactor Building 3.7.A-3 Spectra Comparison Flush-a VS Flush-b 3.7.A-4 Spectra Comparison Flush-a VS Flush-b 3.7.A-5 Spectra Comparison Flush-a VS Flush-b 48(U) , 3.7.A-6 Spectra Comparison Flush-a VS Flush-b 3.7.A-7 Spectra Comparison Flush-a VS Flush-b 3.7.A-8 Spectra Comparison Flush-a VS Flush-b - 3.7.A-9 Spectra Comparison Flush-C vs Spring-C
- 3. 7. A- 10 Spectra Comparison Flush-C vs Spring-C 3.7.A-11 Spectra Comparison Flush-C vs Spring-C 3.7.A-12 Spectra Comparison Flush-C vs Spring-C 3.7.A-13 Spectra Comparison Flush-C vs Spring-C ,
xxxvii (U)-Update t Am. No. 48,12/4/78
ACNGS-PSAR LIST OF FIGURES (CONT' D) /% l \ Figure Title
- 3. 7 '. A Spectra Comparison Flush-C vs Spring-C 3.7.A-15 Spectra Comparison Flush-A vs Spring-B
- 3. 7. A- 16 Spectra Comparison Flush-A vs Spring-B 3.7.A-17 Spectra Comparison Flush-A vs Spring-B ,
3.7.A-18 Spectra Comparison Flush-A vs Spring-B 3.7.A-19 Spectra Comparison Flush-A vs Spring-B 3.7.A-20 Spectra Comparison Flush-A vs Spring-B 3.7.A-21 Spectra Comparison Flush-A VS Spring- A 3.7.A-22 Spectra Comparison Flush- A VS Spring-A . 3.7.A-23 Spectra Comparison Flush-A VS Spring-A 3.7.A-24 Spectra Comparison Flush-A VS Spring-A . 48(U) 3.7.A-25 Spectra Comparison Flush-A VS Spring-A h 3.7.A-26 Spectra Comparison Flush-A VS Spring-A 3.7.A Axisymmetric and Plane Strain Finite Element Models of Soil Structure System 3.7.A-28 Material Properties Used in Axisymmetric and Plane Strain Analyses 3.7.A-29 Finite Element Mesh for Axisymmetric and Plane ', Strain Analyses i 3.7.A-30 Comparison of Horizontal and Vertical Response Spectra From Axisymmetric and Plane Strain Analyses 3.7.A-31 Structural Response (#= 2%)
- 3. 7. A- 32 Response Obtained in the Simplified 3-D Analysis Compared to Free-Field Response (p = 2*/.)
3.7.A-33 Response Obtained in tie Axisymmetric Analysis Compared , to Free-Field Response ( = 2%) i 3.7.A-34 Comparison of Horizontal and Vertical Response Spectra for i Axisymmetric Superstructure Cut at Ground Surface , r I ( xxxviis (U)-Update Am. No. 48 12/4/78 r- e - . - --m-,- , , . ,-e + ,-w- -e, ea --,-w ,~e+-,--* .- 1 r+ e ~ g ,. ,&c e,--9.- -, ,,,,,,-e---+,e-em,s.my -
ACNGS- PSAR LIST OF FIGURES (CONI'D)
\ Figure Title
- 3. 8- 1 ~ Containment Vessel Sheet 1 3.8-2 Containment vessel Sheet 2
~
3.8-3 Reactor Building Containment Internal Structures-Base Details 3.8-4 Reactor Building Piping Penetrations
- 3. 8- 5 Reactor Building Steel Plate RPV Pedestal 3.8-6 Reactor Building Reactor Shield Wall ;
3.8-7 Typical Equipment Foundation
- 3. 8- 8 Typical Reinforcement Detail
- 3. 8- 9 Drywell Base Detail 3.8-11 Reactor Building Dome - M&R
~
3.8-12 Containment Vacuum Breaker A/fK vs. Maximum Containment Negative Pressure , b 3.8-13 Small Line Break Inside the Containment 3.8-14 Containment Response Af ter Drywell and Containment Vacuum Breaker Initiation 3.8-15 Drywell Negative Pressure vs. Time for t Steam Condensation Following Small Primary System 3.8-16 Relative .rfects of Heat Sinks and Spray Depressurization i 3.8-17 Inadvertent Spray Activation. l 3.8-18 Preoperational'High Pressure Drywell Leak Rate Test
- 3. 9- 1 RPV and Internals Vertical Dynamic Model 3.9-2 The Amplification Factor p as a Function of the Frequency Ratio w for Various Amounts of Viscous Damping (U)-Update xxxviii Am. No. 48,12/4/78
I
'ACNGS-PSAR lp. where C is the internal pressure coef ficient. Detail test values of for Pg certain buildings are listed in Reference 3.3-3.
C pg In the. design of walls and roof the pressure coefficient includes the , summation of the external and the' internal pressures. Considering Eq. (3) ! and Eq. (4), the total dynamic pressure is: l 5 l P " * (5) 91_3,1 t- L i P t
=q (C pe
+ Cpt.) (GF}
The total pressure (P ) for the building, in the direction of the wind, 10 given by: 5
.P =C
- D F where C is the average drag or shape coefficient for the building and q is the ynamic pressure at the given height. C ' includes the effects of i positive pressure on the windward wall and negahive pressure on the leeward wall. G7 is the Gust Pressure Factor described in' Section 3.3.1.3. 5 lQ1-3.1 CD and pressure distribution around the Cylindrical Reactor Containment Building is determined by using References 3.3-3 and 3.3-5. Pressure dis-tribution on the dome is determined using Reference 3.3-5.
Table 3.3-1 lists the applied force magnitude and distribution calculated for each plant structure. 3.3.2 TORNADO LOADINGS j 3.3.2.1 Applicable Design Parameters ; Parameters applicable to the design tornado for Category I structure design ; are as follows: 35(D) I; a) External wind forces resulting from a tornado funnel with a horizon- f tal peripheral tangential velocicy of 290 mph and a horizontal ' translational velocity of 70 mph. The maximum resultant wind speed is therefore 360 mph. 48 '[ b) A decrease in atmospheric pressure of 3 psi at a rate of 2 psi pe r (C) se c o nd . c) Tornado generated impact loads resulting from the postulated tornado driven missiles indicat'ed in Section 3.5. 35 , 3.3.2.2 Determination of Forces on Structures Tornado wind speed is converted into equivalent dynamic pressure loadings and the computations for' wind pressures, their distribution on surface area of buildings, shape factors and drag coefficients are based on the O
, procedures outlined in Reference 3.3-3. Because of the unique characteris-tics of. tornados, gust pressure factor and velocity variation with height (D)-Des ign (C)-Consistency 3.3-3 Am. No. 48, 12/4/78
ACNGS-PSAR ,m designed to contain or deflect the missiles from the safety related feature; ( ) but, as a minimum requirement, penetration of the missile through the U barrier will reduce the missile energy to levels which cannot cause failure of the safety feature function. The above design criteria precludes missile failure of.the Containment. Capability to achieve safe - shutdown will be maintained., This is consistent with the philosophy that ' missiles generated in coincidence with a loss of coolant accident or main . steam line accident shall not cause loss of func-tion of Engineered Safe'ty Features or lot s of Containment integrity. The Reactor Building, Fuel Handling Building, Reactor Auxiliary Building, Control Center Building and Diesel Generator Building will be designed to withstand postulated tornado missiles. Wherever possible , advantage will be taken of walls and structures arising from functional requirements, other than missile considerations, by judi-cious arrangement of equipment. Table 3.5-1 lists the plant structures, shields and barriers that will be designed to withstand missile ef fects and the protection afforded. The types of missiles for wnich each structure is designed are indicated. General Arrangement Figures 1.2-4 through 1.2-9, 1.2-17 and 1.2-18, and 1.2-29 through 1.2-35 indicate the safety related equipment which is pro-tected by the structures, shields and barriers indicated in Table 3.5-1.
. Missile protection at doors and access points will be handled on a case- 21 by-case basis, and will depend on the consequences of a missile penetrating a particular door or access point. No door will be designed to withstand a spectrum of missiles in excess of that which the structure itself is de signed to wi thst and . In some cases , special missile barriers will be provided to protect the doors.
In addition to missiles, where necessary as determined by a case by case analysis, the doors and access points will be designed to withstand the 360 48 mph wind and/or 3 psi internal pressure drop at the rate of 2 psi per second (C) as implemented in Section 3.3.2.2. In our evaluation of the consequences of l 35 missile penetration at doors, we will not consider a simultaneous tornado (C), i missile and independent nuclear incident. However, safe shutdown will be 39(U) ! required considering a simultaneous tornado missile plus single additional l active failure. 3.5.2 MISSILE SELECTION In analyzing the plant missile protection required , several sources of mis-siles were evaluated to determine their potential for damage to Category I structures and components. These include: a) Internal missiles generated as a result of failure of pressurized or rotating equipment of plant systems
.O . b) Turbine missiles t
v) (C)-Consistency (U)-Update 3.5-2 Am. No. 48, 12/4/78
ACNCS-PSAR could raise the piping operational stresses to unacceptable l [~ levels. However, the inboard valve piping will be restrained ( to prevent valve motion under pipe rupture loading. The loca-tion of the anchor and guard pipe concept fully isolates the isolation valve from pipe rupture ef fect ; i.e. the inboard l valve is unaffected by breaks outside containment , and the outboard valve is unaf fected by breaks inside containment . The integrity of the containment penetration will be afforded by designing this area to carry the ultimate loading capacity 9 in bending of the associated process pipe without failure of Ql-1 the penetration. 48 Q110.2
)
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f 1 l 1 f 1
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O i ( 3.6-5b(1)' Am. No. 48, 12/4/78
I ACNGS-PSAR- lQ 110. 2 48
- 3) Piping welds in systems supplied by CE will beJ subject to the 17
-/, x\ same. inservice inspections and: the same inspection interval as Q2-2.3 described in Section 3.6.2.2.4 e)3).
- l. V l-
- 4) Welded pipe, support attachments to these portions of piping should be avoided to eliminate stress concentrations.
- 5) The number of piping circumferential and longitudinal welds and branch connections should be minimized.
The extent of piping run should be reduced to the minimum
~
6) length practicable.
- 7) The design of reliable high quality welded flued heads'is an 9.
acceptable approach. We feel that a properly designed , - shop q1_1,1 welded, heat treated, and volumetrically inspected welded , flued head is as safe as an integral forged design, and this ' option will be considered in the. penetration design. We also recog-nize that poorly designed field welded construction pipe-anchors are not acceptable.
- 8) Geometric discontinuities, such as at pipe-to-valve section transitions, at branch connections, and at changed in pipe wall thickness ehould be designed to reduce the discontineities stresses to the minimum practicable.
,- 9) GE will evaluate the operability of ~ the Main Steam Isolation f
\ Valves (MSIV's) to insure that isolation can be assured under loadings from the SSE and postulated pipe break events. The specific load combinations which are to be considered are still being evaluated and will be defined in the future.
To. insure operability of the MSIV, the criteria that must be satisfied is 41 defined below: M i 2.0 Sm (torsion) ___2Zp MB+MC f 2.0 Sm (bending) Zp FA+ A . m (axiaD Ap where: F A = Axial loads due to pipe reaction. M = Torsion load about the valve centerline. B
= Bending load about the horizontal axis.
MC = Bending load about the vertical axis. Zp = Section modulus of the attached pipe. P = Pipe axial load due to internal pressure. A p-t 3.6-5c Am. No. 48, 12/4/78 i
I
'ACNGS-PSAR f"'g APPENDIX'3.7.A i SEISMIC DESIGN CONSIDERATIONS \.s)
1.0 INTRODUCTION
This appendix presents an in-depth discussion of the soil structure interaction analysis methodology employed to design Allens Creek NGS - Unit No. I for~ earthquakes. It demonstrates that the analytical methods presented herein result in a conservative treatment of the seismic design of those structures, systems and components important to safety.
' Table 3.7. A-1 provides a - summary listing of the various analyses performed.
2.0 INPUT MOTION 2.1 INTENSITY As indicated in PSAR Section 2.5.2.9, the highest horizontal site acceler- 48 ation is'obtained by conservatively assuming the nearby largest event N130.6 (Bonham 1882 VII) could migrate to the closest approach of an associated or similar structure. This would place an 1882 Bonham-type event at the southern limit of the Wichita /Ouachita tectonic province, 205 miles from the site. . Attenuation of this event using iso-seismal data would result in an earthquake of Modified Mercalli Intensity IV at the site. Using Figure 1 of Reference 1, the horizontal SSE ground acceleration for the Allens Creek site would be, considering average foundation considerations, '/'~N less than 0.018g. Thus, the SSE horizontal ground acceleration of 0.lg 5 selected for the site is considerably higher than the maximum horizontal ground acceleration that could be, even remotely, expected at the site. 2.2 DESIGN RESPONSE SPECTRA Design response spectra were obtained in accordance with guidelines provided in Regulatory Guide 1.60. The Regulatory Guide 1.60 spectral r shape is considered conservative over certain frequency ranges when applied to the deep alluvium deposit Allens Creek site. As such, the use of the Regulatory Guide 1.60 response spectra rather than site-specific ; response spectra provides additional conservatism for the ACNGS seismic soil-structure interaction analyses. 2.3 CONTROL MOTION ELEVATION The control mo tion, i .e. , time histories consistent with Regulatory Guide 1.60 response spectra, were defined at the ground surf ace (Elevation + 142.0, MSL). This is considered appropriate in view of the fact that the records used to develop the Regulatory Guide 1.60 response spectra were obtained at or near the surf ace. Definition of the control motion at the bottom of the Reactor Building mat level, would result in SSE response levels equivalent to'a'0.2 - 0.3g SSE for both the Control Building and Diesel Generator Building. Th us , the acceleration levels for buildings founded at or near the ground sur- t ['_s f ace become unreali stically high.
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3.7.A-1 Am. No. 48, 12/4/78
ACNGS-PSAR / With respect to the Reactor Building, a comparison of accelerations ( ' obtained at various points (refer to Figure 3.7. A-2) indicates a relative-ly close agreement between results obtained with the control motion de-fined at the bottom of the Reactor Building mat (FLUSH-b) vs. at the ground surface (FLUSH-a). Maximum accelerations at various points from the above two cases are presented in Table 3.7. A-3 and comparisons of floor response spectra are presented in Figures 3.7. A-3 through -8. The design of structures is based on the maximum accelerations obtained at the various floor levels. These maximum accelerations are the peaks of the time histories obtained at the various floors and are identical to the high frequency range accelerations shown in the corresponding spectra. The di f ferences in maximum accelerations presented in Table 3.7. A-3 are in the range of 25% or less. Systems and subsystems located on a certain floor are designed for the 48 corresponding floor response spectra. The floor response spectra com- N130.6 parisons presented in Figures 3.7. A-3 through 8 indicate a more accentuated dif ference in floor spectral accelerations in the frequency range of 3 to approximately 7 cycles per second which is caused by di fferences in the control motion input in the 3 to 6 cycles per second frequency range. There is no equipment in the Ebasco scope-of-supply that has a frequency between 3-6 cps and it is not anticipated that other equipment will be in thi s frequency range. i Nevertheless, should essential systems or components having natural fre-g quencies less than 8 cps be later identified, the ACNGS analysis method-ology will be appropriately modi fied to reflect, in the 0 to 8 cycles per second frequency range, the horizontal floor response spectral values obt ained when the control motion is defined at the bottom of the Reactor Building mat and a + 15% peak broadening is performed. This modification of horizontal floor response spectra in the 0-8 cps frequency range shall be performed for the Reactor Building, Reactor Auxiliary Building, and Fuel Handling Building. 3.0 EMBEDMENT EFFECTS AND STRUCTURE - STRUCTURE INTERACTION One of the reasons for selecting the FLUSH type finite element analyses to establish the soil-structure interaction effects is that the analysis comes closest to representing in a rational way all the important aspects of the problem. While a FLUSH type finite element analysis allows for the adequate re-presentation of structure-structuta interaction and embedment effects, an elastic half space type approach does not. As an example, a FLUSH type analysis of the Reactor Building, where there is no embedment and no adjacent structures, was performed (FLUSH-c) and result s were compared with responses obtained from an elastic half space solution (Spring-c). A comparison of maximum horizontal accelerations at various points is provided in Table 3.7. A-4, and compari sons of response p spectra are presented in Figures 3.7.A-9 through -14. obtained from the two analyses are in excellent agreeme nt in terms of both The responses maximum accelerations and response spectra. 3.7 A-2 Am. No. 48, 12/4/78
ACNGS-PSAR However, when the embedment and adjacent structures.are considered, more ( pronounced di fferences, caused by embedment and structure-structure effects become present. A comparison between FLUSH type- finite element analyses i (FLUSH-a) versus a half space approach using as input moticn the time history. Obtained at the bottom.of the Reactor Building mat for the three shear moduli considered (Spring-b) indicate more accentuated di fferences both in terms of maximum accelerations (refer to Table 3.7.A-5) and response spectra at the various points (refer to Figures 3.7. A-15 through -20). These pronounced di f ferences are caused by not consider-ing, anong others, the embedment and structure-structure effects. Lastly, a comparison between FLUSH finite element analyses (FLUSH-a), and an elastic half-space soluiion (Spring-a) is presented in Table 3.7. A-6 and Figures 3.7. A-21 through -26 in terms of maximum accelerations and floor response spectra, respectively. The comparison indicates maximum acceleration di fferences in the range of 25-30% or less. A comparison of the design shears and moments resulting from the FLUSH-a/ Spring-a analyses i s presented in Table 3.7. A-7. The differences in the design moments and shears obtained at various points in the Reactor Building from the two 48 analyses are in the range of 25-30% or less and parallel closely the N130.6 di f ferences in the maximum accelerations. 4.0 SOIL LAYERING An appropriate seismic soil-structure interaction analytical procedure should be able to take into account, among others, the variations of soil characteristics with depth. While a FLUSH type finite element analysi s is capable of adequately representing the soil layering, an elastic half space approach does not. For the Allens Creek site, the soil layering, as presented in PSAR Figure 2.5.4-7A, makes the use of a FLUSH type finite element analysis preferable. 5.0 TREATMENT OF THE THREE-DIMENSIONAL SOIL-STRUCTURE SYSTEM WITH A TWO-DIMENSIONAL MODEL A nuclear power plant represents a complex three-dimensional system which is very difficult to represent analytically. While an elastic half space approach considers three-dimensional ef fects, the lack of adequate repre-sentation of the structure-structure, embedment, and soil layering effects makes the approach undesirable for the Allens Creek site. The finite element FLUSH type analysis selected for the ACNGS (FLUSH-a) is a two-dimensional idealization using plane-strain elements. While a three-dimensional finite element analysis is theoretically possible, its practical implementation is difficult and the benefits derived are minimal. Theoretically, it should be possible to analyze. with finite elements any arbitrary three-dimensional geometry. However, as discussed in Reference 2, restrictions of cost and available computer capacity and the lack of good 3-D stress-strain relationships for foundation raterials make a full three-dimensional analysis impractical. Fortunately, indi-a cations are that while 3-D effects might be significant for the above - U ground parts of structures they are relatively unimportant for the 3.7 A-3 Am. No. 48, 12/4/78
.ACNGS-PSAR O parts below ground level. This was; illustrated by Berger et-al., 1975 G (Reference" 3) who, compared the results of 2-D plane-strain analyses of the structures shown in Figures 3.7.A-27, 28 and~29 with those'of the corresponding 3-D axisymmetric structure (see Figure 3.7. A-27) for a '
seismic environment consisting ' of vertically propagating she'ar. waves. A summary of. the result s from which this conclusion was made is 'sh'own in i Figure 3.7.A-30. The results clearly indicate that at and below the-ground level 2-D and 3-D ~ analysis give similar spectra. One might even consider the 2-D spectra slightly conservative in this region. As shown
- by Hwang'et. al. ,1975 (Reference 4) .even better agreement can be obtained '
by the introduction of viscous boundaries to~ simulate 3-D effects in a plane-strain analysis. Hwang's results .for the structure used by Berger et. al. are shown in Figure 3.7.A-31. In this figure the spectra marked 2D and AXI are identical to Berger's results shown in Figure 3.7.A-30 and the curves marked'S3D correspond to the 3-D simulation which was- 48 ! actually.. used for' the FLUSH analyses of the Allens Creek Nuclear N130.6 Generating Stat' ion. Hwang~et. al. also showed that the simpli fied 3-D analysis leads to far-field results which agree with the 3-D axisymmetric response. Refer to Figures 3.7.A-32 and -33. . For points abo' ve the ground surf ace Berger's results (see Figure 3.7. A-31) , indicate that plane-strain analysis may not be conservative and this observation has also been made by 'other investigators. - However, Berger et. al., 1975, also showed that this effect is mainly due to the dynamic characteristics of the- above-ground part of the structure and that ex-b cellent approximations to che 3-D response of the upper part of a structure can be obtained by performing a 3-D analysis of the upper part of the - structure alone subjected to a rigid base motion obtained from a 2-D FLUSH analysis (see Figure 3.7. A-34) which corresponds to the soil- , structure system shown in Figure 3.7.A-28. This method was in fact used for the analysis of 'the individual structures of the Allens Creek Nuclear Generating Station and there is therefore reasonable confidence that major. 3-D effects have been accounted for in the analysis.
}
6.0 SEISMIC INPUT PROPAGATION l The FLUSH type finite element analysis used for ACNGS assumes that the ! ground surface motion results from upward propagation of shear waves -l through the soil deposit. This assumption gives results shich are in i reasonable accord with available, recorded data. l 7.0 VARIATIONS IN PROPERTIES As seen in PSAR Figures 2.5.4-7A, 7B, 7D, 7L, and 7K, the soil properties vary not only with depth but also as a function of the strain rates
' involved. While an elastic half space approach cannot adequately re-present the above variations, the FLUSH type finite element analyses chosen for the Allens Creek site do consider the variations of various i parameters not only as a function of depth but also, in the case of the j shear modulus and damping, as a function of the strain rate involved.
In addi tion, for each and every one of the cross-sections considered, three
\ analyses will be performed with shear moduli curves corresponding to 1
3.7.A-4 Am. No. 48, 12/4/78
ACNCS-PSAR (pU j C AVE, where OAVEisrgpresentedbythgcurvesasindicatedinFigures 2.5.4-7B and 7D, 1.5 AVE, and 0.667 AVE. The envelope of the three floor response spectra obtained at a mass point shall be used for seismic qualification and design purposes of systems and sgbsystems with the re-striction that the peak frequency shifts from the AVE case should be a mi nimum of + 10% . Also, for the design of structures, the highest of the three maximum accelerations obtained at a given mass point for each direction shall be used. In vi ew of the above, it is felt that the FLUSH type finite element analysis chosen for the ACNGS site is the most adequate. Furthermore, the implementation procedures used will significantly increase the con-servatism level of the ACNGS seismic analyses. 8.0 DECONVOLUTION ANALYSES For the ACNGS seismic soil-structure interaction analyses, the control earthquake motion is defined at the finished grade, elevation 142.0 above MSL. The Reactor Building base-of-foundation response spectra are derived from the surface control motion by deconvolving the Regulatory Guide 1.60 spectra to obtain the appropriate vibratory ground motions at the base-of-foundation elevation, and then converting these base-of-foundation motions into base-of-foundation response spectra. In the de-convolution process, for certain sites, the foundation zero period 48 acceleration will be reduced from the surf ace value, and a maximum re- N130.6 duction of 40% is considered acceptable. ,m, For the ACNGS seismic soil-structure interaction analysis using the FLUSH code, the differences in the s.ero period accelerations at ground sur-f ace and Reactor Building mat level are negligible (0.12g vs 0.127g). Throughout the soil profile, the zero period accelerations are within 10-15% of the ground surf ace zero period acceleration. (Refer to Table 3.7.A-8) Thus, for the ACNGS site, the deconvolution process results in very small variations in the zero period accelerations. 9.0 ADDITIONAL CONSIDERATIONS Wi th respect to the overall seismic analysis procedures employed for the Allens Creek Project, there are certain aspects which provide additional design conservatism. 9.1 DYNAMIC MODELING The dynami c models for the various structures are prepared using three dimensional finite element analyses in order to establish the effective shear area and moment of inertia for the various members. 9.2 PARAMETRIC STUDIES Prior to performing the FLUSH finite element soil-structure interaction analyses, numerous parametric studies are carefully conducted in order to y establish structural and overall model adequacy, mesh size, the boundary [ conditions and many others. v 3.7.A-5 Am No. 48, 12/4/78
ACNGS-PSAR , 9.3 TORSIONAL ANALYSES 1 N For structures where the shear and mass center do not coincide, decoupled 48 dynamic torsional analyses are performed using translational and rocking time N130.6 1 histories obtained from the FLUSH analyses. In ' addition, a 5% eccentricity will be considered even for symmetrical structures. \ f i
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3.7.A-6 Am. No. 48, 12/4/78
ACNGS-PSAR APPENDIX 3.7. A REFERENCES
- 1. H W Coulter, H H Waldron, J F Devine, " Seismic and Geologic Siting Considerations ' for Nuclear Facilities , "Fif th World 48 Conference on Earthquake Engineering, Rome 1973. N130.6
- 2. " Analyses for Soil-Structure Interaction Ef fects for Nuclear 4 Power Plants" Draft Report by the Ad Hoc Group on Soil Structure Interaction, Nuclear Structu'res and Materials Committee of the Structural Division of ASCE, October 1978.
- 3. Berger, E. , Lysmer, J' and Seed , H B (1975) " Comparison of' Plane Strain and Axisymmetric Soil-Structure Interaction.
Analysis," Proc. Second ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, Decanber 1975.
- 4. Hwang , R. , Lysmer, J and Berger, E (1975)'"A Simplified Three-Dimensional Soil-Structure Interaction Study," Proc. Second
' ASCE Specialty Conference on Structural Design of Nuclear - ,
Plant Facilities, New Orleans, December 1975. l O O 3.7.A-7 Am. No. 48, 12/4/78
ACNGS-PSAR (&\ TABLE 3.7.A-1 IDENTIFICATION OF THE VARIOUS ANALYSES PERFORMED FLUSH - a Finite element analyses using the FLUSH Code and Standard Review Plan method ology. FortheACNGSN-Scrogs-sec5in(Refertogigure3.7.A-1),. three analyses are performed using AVE, AVE *l.5, and AVE /1.5, O where AVE represents the shear modulus vs strain curves established for the various layers as shown on PSAR Figures 2.5.4-7A,7B,7D,71, and 7K. Elastic Half-Space Spring - a Elastic half-space analyses for the ACNGS Reactor Building using spring and damping parameters as indicated in Table' 3.7. A-2. The input time 48 histories are consistent vith Regulatory Guide 1.60 response spectra. N130.6 l Elastic Half-Space Spring - b Elastic half-space analyses for the ACNGS Reactor Building using spring and damping parameters as indicated in Table 3.7.A-2. The input time histories were obtained from the FLUSH - an analyses at the base of the mat corresponding to the shear modulus considered. Elastic Half-Space Spring - c Elastic half-space analyses for the ACNGS Reactor Building usgng the spring and damping parameters as indicated in Table 3.7. A-2 for tne AVE case. The input time history is consistent with Regulatory Guide 1.60 response spectra. FLUSH - b Finite element analyses using the FLUSH Code. Design time histories (consistent with Regulatory Guide 1.60 respogse spectra) were applied at the bottom of the Reactor Building mat. AVE shear modulus values were used for the N-S cross-section. FLUSH - c Finite element analyses using the FLUSH Code. The ACNGS Reactor Building model is used and no embedment or other structures are gonsidered. The input motion is defined at the bottom of the mat. The AVE curves were used in the analysis. O 3.7 A-8 Am. No. 48, 12/4/78 l 9 Y *F'
-P W' F*- gp y p- gy w -.m y -w-e
, ACNGS-PSAR' TABLE 3.7.A-2 INPUT'PARAKETERS FOR ELASTIC HALF-SPACE APPROACH Rocking Tr ansl ation Rocking Translation CASE SOIL G Spring Spring Damping Damping (KSF) (FT-K/ RAD) (K/FT) (% CRITICAL) (% CRITICAL)
AVE 3132 4.04 x'109 1.033 x 10 6 8.43 58.3 48 G 9 5 N130.6 1 AVE /1.5 1830 2.36 x 10 6.03 x 10 8.89 61.6 l 6 A VE
- l . 5 5280 6.81 x 10 1.74 x 10 8.2' 51.2 1
l l I NOTES:
- 1. Values were established in accordance with " Structural Analysis and Design of Nuclear Plant Facilities ," ASCE, Supplement (DRAFT), 1976.
- 2. Shear Modulus (G values were obtained from Flush - a results at ;
theReactorBuilh[!g) mat (see Table 3.7.A-1). ! 1 1 5 t C i 3.7.A-9 Am. No. 48, 12/4/78' -e-- - .w , .,-m - ,%m,----,.r-,+--o.- .-,ww --,~i.wp.,w.. ..%.,w.e. . ..we, ,,,,w.,1 w w . .m..~-w.,,-rewe.c m.$,w..,-.,.,. ,es,-,, - ,y rw o r r + =, r r w - w ee e w .~ v
.ACNGS-PSAR
~ TABLE 3.7.A-3 Os COMPARISON OF MAXIMUM HORIZONTAL ACCELERATIONS FLUSH - a vs FLUSH - b LOCATION FLUSH - a FLUSH - b TOP OF MAT 0.12g 0.12g TOP OF DRYWELL (OPERATING FLOOR) 0.18 0.23 TOP OF RPV 0.22 0.28 48-N130.6 TOP OF PEDESTAL 0.12 0,136 TOP OF STEEL CONT, VESSEL 0.22 0.20 CRANE SUPPORT 0.19 0.166 NOTES:
- 1. Refer to Table 3.7. A-1 for identification of analyses.
- 2. Refer to Figures 3.7. A-3 through -8 for corresponding response spectra.
3.7.A-10 Am. No. 48, 12/4/78
ACNGS-PSAR TABLE 3.7.A-4 sO COMPARISON OF MAXIMUM HORIZONTAL ACCELERATIONS l l FLUSH - c vs SPRING - c LOCATION- FLUSH - c SPRING - c TOP OF MAT. .O.10g 0,13g TOP OF DRYWELL (OPERATING FLOOR) 0.193 0.I8 TOP OF RPV 0.237 0.23 TOP OF PEDESTAL 0.104 0.I1 48 TOP OF STEEL N130.6 CONT. VESSEL 0.22 0.22 CRANE SUPPORT 0.18 0.17 ! NOTES:
- 1. Refer to Table 3.7. A-1 for. identification of analyses.
- 2. Refer to Figures 3.7. A-9 through -14 for corresponding response spectra.
I i \ 3.7.A-11 Am. No. 48, 12/4/78
ACNGS-PSAR s . TABLE 3.7.A-5 \s_/! COMPARISON' 0F MAXIMUM HORIZONTAL ACCELERATIONS
-FLUSH - a vs SPRING - b-FLUSH - a SPRING - b ;
U U 0 LOCATION. AVE AVE /1.5 AVE *1.5 AVE AVE /1.5 AVE *1. 5 . TOP OF MAT- 0.11g- _0.10g 0.12g- 0.12g 0.11g 0.11g
. TOP 1DF DRYWELL (OPERATING FLOOR) 0.15' O.13 0.18 0.17 0.13 0.18 TOP 0F RPV 0.18 0.16 0.22 -0.18 0.15 0.19 TOP OF PEDESTAL 0.11 0.10. 0.12- 0.12 0.11 0.12 48 N130.6 TOP OF' STEEL-CONT.-VESSEL- '0.19 0.16 0.22 0.24 0.16 0.23 CRANE SUPPORT 0.17 0.13 0.19 0.19 0.13 0.20 O- ~ NOTES:
- 1. Refer to Table 3.7. A-1 for identification of analyses.
- 2. -Refer to Figures 3.7. A-15 through -20 for corresponding response spectra.
i
'l l
1 l l L l O O 3.7.A-12 Am. No. 48, 12/4/78
w ACNGS-PSAR TABLE 3.7.A-6 COMPARISON OF MAXIMUM HORIZONTAL ACCELERATIONS i FLUSH - a 'vs SPRING - a LOCATION FLUSH - a SPRING - a TOP OF MAT 0.12g 0.14g
?
TOP OF DRYWELL (OPERATING FLOOR) 0.18 0.!9 TOP OF RPV 0.22 0.31 TOP OF PEDESTAL 0,12 0.15 TOP OF STEEL 48 CONT VESSEL 0.22 0.26 N130.6 CRANE SUPPORT 0.19 0.20 NOTES: s
- 1. Refer to Table 3.7. A-1 for identification of analyser,.
- 2. Refer to Figures 3.7. A-21 through -26 for corresponding response spectra.
l l h
-\
3.7.A-13 Am. No. 48,12/4/78 1 e sw--ww-,+w-e,+-wwwee'+- gw --
.-. - . . . .. . .-. . . . - . - . . _ . . . ~ . - - . . , . . - . .
- ACNGS~'SAR' f
TABLE ' 3.7. A-7A i COMPARISON OF SHEARS AND MOMENTS REACTOR SHIELD BUILDING r
-Spring - a Flush - a LOCATION '
POINT ELEVATION SHEAR BENDING MDMENT SHEAR BENDING MOMENT (FEET) (KIPS) (FT - KIPS) (KIPS) (FT - KIPS) l' 295 740- 10700 590 8220 2 235 3900 230000 - 3300 180000 48 N130 3 176 - 550L 575000 5300 470000 .6 4 136 5900 840000 6200 720000 k I 4 I NOTES: I 48
- 1. All comparisons are for the 1.5 GAVE . case (which results in the N130 highest responses). .6 O 2. Bending moments and shears are obtained from the beam element right above the point indicated. ,
t 3.7.A-14 Am. No. 48, 12/4/78
--. ., - -,-,..., , . ..rm...
- mn.,-e- --%, . ---=.wi.-~ *. r...v, e,w.,+ . . . , ..-.w-w,.,#-r..~me----..-- ,+,,,-,,,-w-.,-we-p. - , . . . . . . , -.w.r--+-.-r-
. - _ _ _ . _ . _ _ _ _ _ . _ _ _ -_ _ . _ _ _ _ _ _ _ . . . . _. m._, _.m._..m__m. m._..._ .
O O O
~k ACNGS-PSAR TABLE 't.7.A-7B -
COMPARISON OF SHEARS AND MOMENTS '{; RPV AND RPV PEDESTAL. ) f Sprin:t - a Flush - a f POINT LOCATION ELEVATION SliEAR BENDING MDMENT . SHEAR BENDING MOMENT
'fj (FEET) (KIPS) (FT - TIPS) (KIPS) (FT ~ KIPS) .
RPV PEDESTAL i 139 1220 45000 107C 14000 l
'. }
RPV PEDESTAL 2 132 1300 55000 1150 42000 - 48 .) N130.6 'I ' RPV PEDESTAL 3 116 1370 62000 1230. 48000 RPY 1 208 20 100 .15 74.
't l
.t
- )
. l y
?
C; i i i
?
l i Y i E l ? NOTES: , w .i Q 1. All conparisons are for the 1.5 G case (which result s in the. 48 [
- highest responses). '
N130.6 ' 1
- 2. Bending moments and shears are obtained from the beau element right j above the point indicated.
I j 3. Design bending mmnents and shears for. the RPV wi1' be obtained
- following a detailed analysis to be performed by 6.eneral Electric #
i- Refer to PSAR Section 3.7.2.1.2.7.3. L _ _ . - ,_ _ _ . . _ _ _ ._ __ _ - _ . _ _ _. .. _ __._ _a . J
ACNGS-PSAR-ACNGS-PSAR
-[ TABLE 3.7.A-7C i
COMPARISON OF SHEARS AND MOMENTS CONTAINMENT j Spring - a Flush - a LOCATION , POINT ELEVATION SHEAR BENDING MOMENT SHEAR BENDING MOMENT )' (FEET) (KIPS) (FT - KIPS) (KIPS) (FT - KIPS) 1 290 80 1100 70 880 48 N130.6 2 260 310 10000 260 8200 , 3 209 800 47000 640 40000 4 178 980 75000 780 63000 5 121 1400 144000 1130 120000 i NOTES: 48
- 1. All comparisons are for the 1.5 G VE case (which results in the N130.6 highest responses).
Bending moments and shears are obtained from the beam element right O' 2. above-the' point indicated. 3.7 A-16 Am. No. 48, 12/4/78
1 . ACNGS-PSAR TABLE 3.7. A-7D COMPARISON OF SHEARS AND MOMENTS DRYWELL Spring - a Flush - a LOCATION POINT ELEVATION SHEAR BENDING M0 MENT SHEAR BENDING MOMENT (FEET) (KIPS) (FT - KIPS) (KIPS) (FT - KIPS) I I 208 1060 32000 920 29000 j 48 1 2 182 3500 125000 3!00 103000 N130 I
.6 3 138 5300 350000 5000 300000 4 116 5600 470000 5400 400000 I
i l l NOTES:
- 1. ~ All comparisons are for the 1.5 G AVE case (which results in the 48 highest responses). N130
.6 l 2. Bending moments and shears are obtained from the beam element right
\ above the point indicated.
3,7.A-17 Am. No. 48, 12/4/78
1 i
'l 1
ACNGS-PSAK~ i
- TABLE 3.7.A-8 DECONVOLUTED ZERO PERIOD' ACCELERATION ,
I HORIZONTAL SSE ELEVATION "g" VALUE- , 4 (FEET-MSL) 142. 0.12 140 0.119 ' 136 0.117 132 0.115 128 0.115 48 124 0.114 N130.6. 120 0.115 116 0.121 112 0.122 1 108 0.123 102.5 0.127 3 97 0.124 91.5 0.120 i
.86 0.116 80.5 0.109 -
75 0.107 , 69.5 0.100. _ 64 0.098 58.5 0.099-53 0.099 47.5 0.099 42 0.102 16.5 0.102 31 0.106 25.5 0.114 20 0.118 1 14.5 0.121 9 0.121 l l l 1 i l i 4 3.7.A-18 Am. No. 48, 12/4/78 2
-S I
d N-S-2 I I EL 166.50'
/
^
EL 212.50' EL 246. O N] TGB I N S-1 - - EL 294.04' l e EL 160.00' SWGR RM
'THROUGH NON-SEISMIC-CATEGORY l STRUCTURES.
KEY PL AN ROOF ELEVATit 4 1
- - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - . . _ _ - . _ _ - - - - - - - _ _ - - - -- -- -J
ACNGS - PSAR Q RCB l EL 176.00' 1 DGB EL 220.00' EL 235.00' _ _ 8 / l
.RWB W
EL 197.00'
'd EL 219.00' EL 227.00' EL 156.84' I l l EL 222.00'-+ PAB
~
h__ _. l s CB ! d EL 17.84' EL +207.00' , I RAB EL 259.17' EL 259.17'
@3' , /--
EL 202.33' RCB V} ' m I TOP OF SPRING C DOME LINE F
+ - &g EL 311.75' EL 271.83' / b
! # 5 FHB g._ i i -
/
W EL 229.50' h22.50' , / l / k d l EW 2* EL 182.33' d EL 207.50' ! EW-1 l Am. No. 48,12/4/78 l HOUSTON LIGHTING & POWER COMPANY aN Allen Creek Nuclear Generating Str. tion Unit 1 NORTH SOUTH CROSS-SECTIOrl FOR ANALYSES FIGURE 3.7.A_1 1 k
ACNGS - PSAR TOP OF SHIELD BUILDING TOP OF STEEL CONTAINMENT EL. 310 CRANE SUPPORT EL.271 OPERATING FLOOR d TOP OF RPV EL. 235 7 Y _ k/ EL.194 .
' ~
TOP OF PEDESTAL _ EL.155\
~ -
TOP O F MAT
\ -
\
EL.115 ) l ie V i 1 11111 isi Am. No. 48,12/4/78 HOUSTON LIGHTlHG & POWER COMPANY p Allens Creek Huclear Generating Station Unit 1 REPRESENTATIVE NODAL POINTS
' REACTOR BUILDING FIGURE 3.7.A-2 . . . _ . ._ _ - , _ _ _ . _ . . , - - . _ . . . _ ~ _ _ - . _ , , _ . _ . . . _ . . . . . _ . .-
(m (. ALLENS CREEK - REACTOR BUILDING 5.00 - TOP OF MAT CASE 0 BASE CAS CASE 4 ------ MODIFIE 4.00 - 3.00 - 6 a O G 2.00 - 1.00 - d Ny I l 0.50 1.00 2.00 {)N
l l ACNGS PSAR
# - FLUSH ENVELOPE (FLUSH - a)
FLUSH 1G CASE (FLUSH - b) t l l { r l [ f ! #w 1 l % L/ %
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I l l ! l 5.00 10.00 20.00 50.00 . FREQUENCIES (CPS) i ! Am. No. 48, 12/4/78 l HOUSTON LIGHTING & POWER COMPANY
- Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMP ARISON FLUSH-a VS FLUSH-b FIGURE 3.7.A-3
- ALLENS CREEK - REACTOR BUILDING OPERATING FLOOR i
CASEO . CASE 4 -- - - 4.00 - l 1 3.00 - l 6_ i E l O ,.,, _ l l i 1.00 - 1 y fd
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i I I I ! 5.00 10.00 20.00 50.00 ! FREQUENCIES (CPS) I Am, No. 48,12/4/78 , HOUSTON LIGHTING & POWER COMPANY r Allens Creek Nuclear Generating Station t Unit 1 l SPECTRA COMPARISON . FLUSH o VS FLUSH-b FIGURE 3.7.A-4 :
1 i 5.00 - ALLENS CREEK - REACTOR BUILDING j
}
TOP OF RPV i J t, CASE 0 , i CASE 4 --- - - J 4.00 l i l L P 3.00 -
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l l I I I 1 l 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMP ARISON FLUSH.a VS FLUSH-b FIGURE 3.7. A-5 i
ALLENS CREEK - REACTOR BUILDING 5.00 - TOP OF PEDESTAL CASEO C AS E 4 - - - - l l 3.00 -
' I.
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,-- . gH- ;
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- BASE CASE - FLUSH ENVELOPE (FLUSH - a)
MODIFIED FLUSH - 1G CASE (FLUSH - b)
/\
/ \
/ \
f \ e w l l 1 1 I f 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH a VS FLUSH-b FIGURE 3.7.A-6 i
1 ALLENS CREEK - REACTOR BUILDING TOP OF STEEL CONT. CASEO B. A00 - CASE 4 ----- M 3.00 - 1 3 W < O 2.00 - i l l l I I l 1.00 - r#(
/
)
d
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ACNGS - PSAR l pSE CASE - FLUSH ENVELOPE (FLUSH - a) @DIFIED FLUSH 1G CASE (FLUSH - b) I I l l l l l h '\- l
\
\
\
l 'A l l 1 1 I l 5.00 10.00 20.00 50.00 PREQUENCIES (CPS) 1 l Am. No. 48,12/4/78 l HOUSTON LIGHTING & POWER COMPANY l Allens Creek Nuclear Generating Station Unit 1 . SPECTRA COMPARISON FLUSH-a VS FLUSH-b l FIGURE 3.7.A-7
I 1 1 I 5.00 - ALLENS CREEK - REACTOR BUILDING ! CRANE SUPPORT i i 4 CASE 0 ,
^ ~ ~ " " - ~
4.00 - i l 3.00 f
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l ACNGS - PSAR l 1 BASE CASE - FLUSH ENVELOPE (FLUSH - a)
' MODIFIED FLUSH 16 CASE (FLUSH - b) l "'N N
wg
. - - -- - % _ _ - _ a i I I i 5.00 10.00 20.00 50.00 FREQUENCIES (CPS)
Am. No. 48, 12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH o VS FLUSH-b FIGURE 3.7.A-8
/%
5.00 - ALLENS CREEK - REA 4.50 - TOP OF MAT - CASE 5 CASE 3.d I 4.00 -
)
I 3.50 - i 3.00 - 6 l
- i d 2.50 O
4 i I 2.00 l i 1.50 - l 1.00 - 0.50 - d
~
l 1 0.50 1.00 2.00 l i l l i i
ACNGS - PSAR TOR BUILDING FLUSH - C
--- - SPRING - C l
I N^%Ag j %* - _ _ _ _ _ _ _ _ _ _ _ _ _ 0.0 2 i i l i I ( 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am No. 48,12/4/78 I HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit I l SPECTRA COMPARISON FLUSH-C vs SPRING-C l FIGURE 3.7.A-9
i 5.00 ALLENS CREEK - l 4.50 - OPER. FL. - CASE 5 ' CASE 3i 4.00 - 3.50 - l 1 3.00 - O 2.50 - l 1 2.00 - j i 1.50 l l 1.00 !
*s !
s
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l
ACi. 35 - PSAR l i i REACTOR BUILDING ) l FLUSH - C l [ .--- - SPRING - C J l d i + l g'%,
"Vm -_ _ _ ;# -
0.02 I I I I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 I HOUSTON LIGHTING & POWER COMPANY l Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH-C vs SPRING-C FIGURE 3.7.A-10
-+y v , ,.y - --... mew , r .__m, ,,,,,v.-, ,y ,,,,,.__.,,,,g,_
4 e ALLENS CREEK - REA 4.50 - TOP OF RPV - CASE 5 4 CASE 3 % l 4.00 - 1 3.50 - I l l I 3.00 - f a
@ 2.50 W
O 2.00 - I 1.50 1.00 - 0.50 -
***=g s.-
,j l I 0.50 1.00 2.00 Io
l ACNGS - PSAR TOR BUILDING FLUSH - C
--- SPRING - C 4
l A 4 I I
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\
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/ T e/ l p/ \
N2Li sO . - - - - - - - - 0.02 I I I I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH-C vs SPRING-C FIGURE 3.7.A-ll l
I l (O d 5.00 -
)
i ALLENS CREEK - READ l l 4.50
- TOP OF PED. - CASE 5 CASE 3 q 4.00 -
3.50 -
~
3.00 1 m g 2.50 - O 4 ( 2.00 - 1.50 - l 1.00 - 0.50 - -
~ g.
-x
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)
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ACNGS - PSAR TOR BUILDING FLUSH - C l
---- SPRING - C !
l l i a f 5 k.
'~~
0.02
= - _ q .i. -
5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON l FLUSH C vs SPRING-C l FIGURE 3.7.A-12 1
5.00 _ ( ALLENS CREEK - - 4.M - TOP CONT. - CA CASE ( 4.00 - 3.50 3.00 , g 2.50 - O nv 2.00 t 1.50 - J '\ 1.00
,- m \, ,
,/
f 0.50 -
/,7 l l
( 0.50 1,00 2.00 m
ACNGS - PSAP. 9EACTOR BUILDING i FLUSH - C s ----SPRING - C
/^^f\ \
r A r - _ CZY 0.02 1 1 1 1 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH-C vs SPRING-C FIGURE 3.7.A-13
5.00 - ALLENS CREEK-REACTOR f 4.50 - TOP OF CRANE - CASE 5 CASE 3 -c-4.00 3.50 - g 3.00 - O 4 2.50 - 2.00 - 1.50 - 1.00 - f *\
,# %me%
0.50 -
#I#~ be" l 1 0.50 1.00 2.00
I
- l. ACMGS - PSAR bulLDING h FLUSH - C
-- SPRING - C l
l l l l l l l t I l l t t % c n 0.02 I I I I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48,12/4/78 i _
- HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SP ECTRA COMPARISON FLUSH-C vs SPRING C FIGURE 3.7.A-14
,jrh k v 5.00 - ALLENS CREEK - REACTOR BUILDING TOP OF MAT - ENVELOP OF G AVE G* 1.5 G/1.5 Bs - - - - FLt 4.00 - 3.00 - U U O 2.00 - 1.00 -
#~
1 i 0.50 1.00 2.00 O i,m)
i l ACNGS - PS AR l SH - a
-LNG - b
, s r 1 p----. l t I I I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) '[
~
Am No. 48, 12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Genercting Station Unit 1 , SPECTRA COMPARISON FLUSH.A vs SPRING.B FIGURE 3.7.A-15
k l G ALLENS CREEK - REACTOR JUILDING OPERATING FLOOR - ENVELOPE OF G AVE : G
- 1.5 )
G/1.5 l CASE BASE ----- 2 4.00 - 1 3.00 _ O l O '
<C 6 t
2.00 - i I i
~
i 1.00
/%
PW '
/
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j amo em #
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l l l 0.50 1.00 2.00 l 4
.c--+r,==-.-. .--w,-- - . - - - _ --_------J
l ACNGS - PSAR , I l l l l l . l 0 (FLUSH - a) (SPRING - b) 1 l l l l l
\
f% rN* N gw^= w % - - ,,, , , _ _ _ - - . ., I I f I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48, 12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH-A vs SPRING-B / FIGURE 3.7.A-16
O 5.00 - ALLENS CREEK - REACTOR BUILDING TOP OF RPV - ENVELOPE OF G AVE; G* 1.5 ' G/1.5 CASE BASE ~ ~ ~ ~ < 2 4.00 - 1 3.00 - 4 w i O U ; O < 2.00 - r I. Y . t
+
5 1.00 - g% r#l
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l l
)
. , . . . _ - - _ . _ . . _ , . _ _ . , _ . . - - , _ . . _ , _ _ . . . - - , . . - - _ . , _ _ . _ . . . . ~ - - . . - - . . , _ _ . _ .
l ACNGS - PSAR 1 l
-- (FLUSH - a)
- (SPRING - b)
I d l I 5
\ /
/"\ \
\/ \
- j. / '/ *%\
# \
"""'"' % * #e s \ == ,
I l l 1 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) Am. No. 48, 12/4/78 HOUSTON LIGHTING & POWER COMPANY > Allens Creek Huclear Generating Station Unit 1 SPECTRA COMPARISON FLUSH A vs SPRING-B FIGURE 3.7.A-17 _ . _ . . __ __ _ . _ . . ._. . . . _ . . , . _ . - _ . _ _ _ _ _ . . _ _ , _ _ _ _ . _ , _ . _ . . . _ . , ~ . _ . . ._
j\/
\ ALLENS CREEK - REACTOR BUILDING 5.00 -
TOP OF PEDESTAL - ENVELOPE OF G AVE G*1.5 G/1.5 CASE B.AS E -- - - - d ' 2 4.00 - 3.00 - 6 O k 2.00 - a 1.00 -
,e. *%
,e / _.
-- D I I 0.50 1.00 2.00
/O V
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'- (FLUSH - a) - (SPRING - b)
I 4 i I
' ~ "'k __ _ m%% --
l f 1
-I 5.00 10.00 20.00 50.00 FREQUENCIES (CPS) i Am. No. 48,12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMP ARISON FLUSH-A vs SP RING.B FIGURE 3.7.A-18 L._ . _ . _ _ _ _ . _ - _ . . _ _ . . . _ . _ _ . _ _ . . . _ , _ __ . . -_ _ _ _ ... .. ._ , __,. _ , _ _ ,_
n 1 4 ( ALLENS CREEK - REACTOR BUILDING 5.00 TOP OF STEEL CONTAINMENT ENVELOPE OF G AVE G
- 1.5 G/1.5 CASE BASE ~ ~~~"
2 4.00 - 3.00 - 3 \ $ 2.00 1.00
/*
- A/
/
sl
# p.
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f f 0.50 1.00 2.00 (s_- i I i.
i Q ACNGS - PSAR i F - (FLUSH - a) - (SPRING - b) l l l l l l l 1 l l 4
'N
\
\
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Am. No. 48, 12/4/78 HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 SPECTRA COMP ARISON FLUSH-A vs SPRING-B FIGURE 3.7.A-19 s
i l I ALLENS CREEK - REACTOR BUILDING 5.00 - CRANE SUPPORT - ENVELOPE OF G AVE G
- 1.5 G/1.5 CASE BASE ~~~~ ~ ~'
2 4.00 3,00 - 3 w 9 U s' 2.00 - l 1.00 - pl %<
/,s*#
som #a e -
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1 1 0.50 1.00 2.00 l l ( i 4
ACNGS - PSAR
- (FLUSH - a)
- (SPRING - b)
) N
\
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.AXISYMMETRIC AND PL ANE STRAIN FINITE ELEMENT MODELS OF SOIL STRUCTURE SYSTEM FIGURE 3.7.A 27
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HOUSTON LIGHTING & POWER COMPANY l O Allens Creek Nuclear Generating Station i t Unit 1 FINITE ELEMENT MESH FOR AXISYMMETRIC AND PLANE STRAIN AN ALYSES l F1GURE 3.7.A 29 l
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- COMPARISON OF HORIZONT AL. AND
- VERTICAL RESPONSE SPECTRA FROM AXISYMMETRIC AND PL ANE STRAIN AN ALYSES ,
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ACNGS-PSAR TABLE 3.9-8 STRESS CATEGORIES AND STRESS LIMIT FACTORS FOR CLASS 1, 2, 3 AND MC LINEAR TYPE SL'PPORTS DESIGNED BY ELASTIC ANA'" STRESS LIMIT FACTORS FOR LOADING CATEGORIES 1 STRESS TEST CATEGORY DESIGN NORMAL (3) UPSET EMERCENCY LOADINCS 5 Primary Stresses K =1.0 K = 1.0 K = 1.0 K = 1.33 K = 1.33 s s s s s 43 plus 3 tresses Q110 2 Induced by Re- K = 1.0 K = 1.0 K = 1.0 K = 1.33 (4) K = 1.33 .7 straint of Free End Displace-se nt (3) K = 1.0 K = 1.0 K = 1.0 K = 1.33 bk ( 4. ) K.o k= 1,33 bk bk bk K $ 2/3 critical K $ 2/3 critical K $ 2/3 critical ( 5 2/3 critical g Q110.7-- y Primary plus 4 y Secondary EVALUATION NOT REQUIRED r N Stresses Y Peak Stresses EVALUATION NOT REQUIRED 4 i NOTES: (1) Control of deformation is not insured by these stress limits and must be considered separately d en required by the Design Specification. (2) F shall not exceed 0.425 . (3) SIresses induced by restrUint of free end displacement and anchor motions shall be considered as primary stresses except for the design loading condition and faulted condition. (4) Faulted limi t s comply wi th ASME Section 111, Appendix F. The faulted buckling limit complies with Appendix F.1370 (c). l NOMENCLATURE: *f00 Q1 0 K *7
= Stress limi t factor applicable to the Design allowable tensile and benling stresses.
K, = Stress limit factor applicable to the Design allowable shear stresses. ' K
,N bk = Stress limi t factor applicable to the Design allowable axial compressive and bending stresses to determine buckling limit.
2
,o This table applies to linear type support s only. The design criteria for standard and plate and shell type supports in the BOP 3 scope of supply are in compliance with ASME Section III, Subsection NF. There are no ASME Class 2 or 3 supporta in NSSS scope of supply. 48 1
m The correlation between service limit and allowable stresses for plate and shell type supports are in accordance with Subsection NF as 0110 f Ilows: k Stress Category Normal Service Limit *I-( A g Upset B l Emergency C Faulted D , .i
, _ _ _ _ . . . _ , m_.
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ACNGS-PSAR TABLE 3.11-3 (Cont'd) i Operating gg' Relative Radia- Rate D**iE B#*i" 5 Pressure Temp Humid ity tion Plant Sys' Intgated Dose (10) Accident C9) Dose Rate ( } Area (As Noted) ( F) (%) Type Opr g Normal Accident Tvpe .; 35 ;
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VII. Auxiliary Building -0.10 in. to 104 40 Nor Gamma 1.5x10 5. 3 v. 10 1.7x10 lI4CA . Equipment Ares +0.10 in. Nor Water gage 120 Max (6) 20 Min 45(U) ! ral Tim r 60 Min static + pressure ' 35 3 LPCS and RIIR -0.10 in. to 104 Nor ame as = 3.0x10' 5.3.x 10 4.5x10 LOCA (U) Equipment Area -0.25 in. 150 Max (6) above Water gage 60 Min ; static pressure HPCS and RCIC Same as 104 Nor ame a uma 2.0x10- 5.3 x 10 4.5x10 LOCA RER and CRDilS abee 150 Mag) (6) above ,w area 60 Min 48(U) l E i i 2$ f 5 E
- Superscript numbers refer to comments in Table 3.11-5 s '
R e ~l
P , ACNGS-PSAR EFFECTIVE PAGES LISTING ; CHAPTER 4 REACTOR Uge ' Amendment j 1*' 48 la* 42 12*.- 48 7 3* 46 4* .41 - 5*. 40 ; 6* - 41 7* 45 8* 40 9* 40 10* 39 11* 34 12* 34 i 34 ii- 34 iii 34 iv 34 v 34
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- 1 Am. No. 48, 12/4/78
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ACNGS-PSAR I EFFECTIVE PAGES. LISTING (Cont'd) t CHAPTER 4 O REACTOR Page- Amendment { 4.2-14h ;41 - 4.2-14h(1)' 34
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ACNCS-PSAR 44 O Check valves are checked for leakage as part of the Hydraulic Con- q .Q trol Unit's qualification tests prior to installation. 211.2 I48 4.2.3.1.4 Testing and Inspection (g) The tests' performed on control rods plus their related surveillance program are covered in Section 4.3.4, " Testing and Inspection." 4.2.3.1.5 Instrumentation
' The instrumentation for both the control rods and control rods drives is de-fined by that given for the manual control system. The objective of the Reactor Manual Control System is to provide the operator with the means to make changes in nuclear reactivity so that reactor power level and power distribution can be controlled. The system allows the operator to'manipu-late control rods.
(U)-Update 4.2-36a Am. No. 48, 12/4/78
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1' ACNGS-PSAR' EFFECTIVE PAGES' LISTING CHAPTER - REACTOR COOLANT SYSTEM
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ACNGS-PSAR a) 'The lowest safety valve be set at or below vessel design pressure b) The highest safety valve be set so that total accumulated pressure does not exceed 110 percent of the design pressure. 34 (G) The safety / relief valves are set to open automatically (relief function) (G}484g
- in the range from 1115 to 1155 Table 5.2-6. The safety function of the 2113 safety / relief valves is set to operate in the range from 1175 to 1215 psig. l48 This satisfies the ASME Code specifications for safety valves, because all l 30 (G) valves open at less than 1250 psig (nuclear system design pressure).
There are two major transients, the closure of all main steam line isola-tion valves and a turbine trip with a coincident closure of the turbine steam bypass system valves that represent the most severe abnormal opera-tional transients resulting in a nuclear system pressure rise. _ The transient produced by the closure of all main steam line isolation valves represents the most severe abnormal operational transient resulting in a nuclear system pressure rise when safety grade scrams are hypot hetically assumed to fail . The required safety valve capacity is determined by analy-zing the pressure rise from such a transient. The plant is assumed to be 34(G) operating at 105% of Nuclear Boiler rated steam flow conditions at a maximum vessel dome pressure of 1045 psig. The analysis hypothetically assumes the failure of the direct safety grade isolation valve position scram. The 34(G) reactor is shut down by the backup, indirect, high neutron flux, scram. 42 For the analysis, the relief, function setpoints are assumed to be . in the ( q range of 1115 to 1155 psig (upper boundary of the set points), while the r 2W b self actuated setpoints (safety function) of the safety / relief valves are assumed to be in the range of 1175 to 1215 psig. One half of the safety / ( relief valves is assumed to operate in power-actuated pressure relief mode; the other half is assumed to operate in spring action safety mode. The M({42 analysis indicates that the design valve capacity is capable of maintaining the vessel pressure below the peak ASME code allowable pressure in the nuclear system (1375 psig). Figure 5.2-1 shows the curve produced by this analysis. The peak pressure at the bottom of the reactor vessel is 1294 psig. There fore , the overpressure transient ef fect is well below the Reactor Coolant Pressure Boundary design. i l l O 5.2-14 (G)-GESSAR Am. No. 48,12/4/78
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1 l ACNCS - PSAR l
,-s , SECTION 5.2.10 REFERENCES l 48 (U) i )
\s_ / 5.2-1 W.L. Williams, Corrosion, Vol 13, 1957, p. 539t 5.2-2 J.M. Skarpelos and J.W. Bagg, " Chloride Control in BWR Coolants," June'1973, NEDO-10899. 5.2-3 W.L. Walker and J.P. Hi ggi ns , " Performance of 304 Stainless Steel Structural Components in General Electric Company Boiling 34(G) Water Reactors." Paper No. 103 presented at the NACE Corrosion Forum, Anaheim, Cali forni a , March 19 to 23, 1973. 5.2-4 S.H. Bush and R.L. Dillon, " Stress Corrosion in Nuclear Reactors," March 8, 1973, Presented at The International Conference on Stress Corrosion Cracking and Hydrogen Embrittlement of Iron Base Alloys, held in Unieux-Firminy, France, on June 12-16, 1973 A i
\m-7s x.-]
\ (U)-Update 5.2-34a (G)-GESSAR Am. No. 48, 12/4/78
1 ACNGS-PSAR EFFECTIVE PAGE LIST CHAPTER 7 l INSTRUMENTATION AND CONTROLS _ PAGE NO. AMENDMENT NO. 1* 48 la* 43 2* 46 3* 44 4* 40 5* 40 6* 48 6a ., 43 7* 44 8*- 37 9* 43 10* 46 11* 37 12* 40 13* 37 14* 37 15* 44 16* 37 17* 39 18* 37 19* 37 20* 39 21* 39 22* 37 23* 37 24* 37 1 44 11 37 iii 37 iv 37 v 37 vi 37 vii 37 viii 40 ix 37 x 37 xi 37 x11 37 xiii 37 xiv 37 xv 37 xvi 37 xvii 37 O xviii xix
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ACNGS-PSAR' However, the undervoltage protection scheme. will be reinstated if offsite 48. l 1 power becomes available and the loads are transferred to this power source. (U) . i P b d i - l . 4 l. f j 1 I
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i If the generator is synchronized at the time of bus tie breaker opening, no j , loads will be shed. See Auxiliary one line' diagram Figure 8.3-1 and Tables
- 8. 3-1 through 8. 3-3.
On opening of the bus tie breaker with the generator breaker open, the I 143 safety related bus will automatically shed all safety related load except
'(D) I those in the first load block. l I
The automatic starting of subsequent loads will be delayed by timing ) relays with five second intervals between them. The starting sequence for each safety related bus . is shown on Tables 8.3-1 through 8.3-3. If a design basis accident occurs with a bus tie breaker open, the required 37' safety related loads will be connected to the bus automatically in proper k3 sequences as in the preceding paragraph. (D) The diesel generator may be periodically run under load by operator action. Should normal ac power be lost during such a condition, the safety related bus tie breaker will trip but no loads will be shed. If a design basis accident precedes or follows this loss of normal ac power, the safety relatedl43 bus automatic loading, sequence will begin simultaneously. (D) Should a design basis accident occur and of f aite power is available3 the 48 required safety related loads will be sequentially connected to the safety (U) related bus.
- 7. 3.1.1. 6.1. 5 Diesel Generator Support Systems a) Fuel Oil System The three diesels will be provided with oil from three seven day 22 37 fuel oil storage tanks, each storage tank feeds a four hour day tank for each diesel. The feed of oil is accomplished by gravity. (D)
Control of oil flow to day tanks will be operated from level switches U) on the day tanks. When the fuel oli level falls to the three hour level in any day tank its oil supply valve will open and the oil will flow from the storage tank to its associated day tank. When the fuel oil level rises to the four hour level in any day tank, its oil supply valve will close. The range of operation is held between the 1,10 four hour and three hour levels to keep oil volume high to minimize Q1-condensation due to breathing. 8.40 The day tanks will be provided with level switches for high-high and low-low level annunciation in the Control Room. Oil flow indication will be provided for each oil supply at the storage and day tanks to be able to detect any oil 37 leak in the piping. The storage tanks will be provided with a level switch (} for low level annunciation in the Control Room and level indication for Pt.M. All valves and instruments for each diesel oil supply line, will be powered from that diesel bus. The diesel engine fuel pump will be engine driven and will take suction from the day tank at a point a few inches above the bottom of the gage glass which C (D)- Design (U)-Update 7.3-41 Am. No. 48, 12/4/78
ACNGS-PSAR t' 7 will allow maintenance personnel to see accumulations of water and drain The fuel pump will discharge to the fuel injectors. () it from the tank. engine will be stopped by tripping the injectors to the no load position by The nechanical means (overspeed trip) or by energizing a trip solenoid. Reset 1 O ( v 7.3-41a Am. No. 48, 12/4/78
ACNGS-PSAR EFFECTIVE PAGE LISTING
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EFFECTIVE PAGE LISTING CHAPTER 9 . AUXILN.RY SYSTEMS O PAGE NO. AMENDMENT No. 9.1- 1 37 9.1-la - 37 9.1 37 9.1-2a 37-9.1- 3 37 9.1- 3a 37
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1 ACNGS-PSAR 9 .1. 3 . 2 System Description The Spent Fuel Pool Cooling and Cleanup System will serve the following pools: In the Reactor Building, the reactor well and the Fuel Transfer and Storage Pool; in the Fuel llandling Building, the Fuel llandling Pool', 37(D) the Spent Fuel Pool and the Cask Pool. The system will consist of one Skimmer Surge Tank, two fuel pool circulating pumps , two Fuel Pool heat 148(D) I exchangers, two filter-demineralizers, and associated pumps and equipment, valves, necessary instruments and piping (see Figures 9.1-3 and 9.1-4, Spent Fuel Pool Cooling and Cleanup System Flow Diagram and Spent Fuel Pool Filter /Demineralizer P&ID). System equipment will be located in the Fuel 37(C) llandling Building. The pool water, either from the reactor we'l or the Fuel Trans fer and Storage Pool, overflows into skimmers and scuppers around the pe rime te rs of the pools and flows into a stainless steel skimmer surge tank. Water f8(D) either from the spent fuel pool, fuel handling pool, or cask loading pool overflows into skimmers and scuppers around the perimeters of the pools and is collected in the skimmer surge tank. From the skimmer surge tank, the water is pumped through the fuel pool heat exchangers in which the decay heat of spent fuel assemblies is removed. The water then goes through the filter demineralizer, where particulates are removed, and back to the pools. Because the particulates are radioactive and the demineralize rs may become radioactive, the filter demineralizers will be located in shielded cells. The design of the cooling system will be for a heat load of spent fuel stored in the Spent Fuel Pool resulting from 10 years ' normal batch ' discharges from the reactor. The heat load is taken as the sum of the decay heat released by one quarter of the core which has undergone 30 hours decay af ter 4 years of irradiation at full power conditions plus successive annual bat ch discharges which have undergone similar irradiation and have cooled for periods ranging from 1-9 years. The Fuel Pool Cooling and Clean- 37(D) up System will be designed to maintain the Spent Fuel Pool water temperature below 125 F while removing the design heat load from the pool with the com- l ponent cooling water temperature at its maximum. The Containment Pool will not be included in the determination of the design heat load since the same heat load will result from the described quantity of spent fuel regardless of its distribution amongst the various pools cooled by the system. I To provide additional cooling capacity, piping will be provided to permit 16 the use of the Residual lleat Removal Systems in parallel with the Fuel Pool Cooling and Cleanup Syetem to remove abnormal heat loads. The RilR System Ql-will not be initiated unless the reactor is in a cold shutdown condition 8.10 and in the refueling mode. To insure adequate spent fuel water cover during RilR backup operation, the suction line in the pool will be installed with its inlet above the 91- ) minimum level required for water coverage of the fuel during storage and 6.39 ' handling. (See Figure 9.1-3). (C)-Consistency l (D)-Design 9.1- 7 Am. No. 48, 12/4/78 _. __ _._, _. _ _._ _ _ _ _ _ _ _ _ . . . ._ _. _____ _ .D
ACNGS-PS AR w 22 ( 9.2.1.2.4 Essential HVAC Service Water System ; b l37(U) f The Essential HVAC SWS will have 2 trains. Each train will consist of a ; chiller and associated piping, valves and instrumentation. The essential i j HVAC SWS will be seismic Category I and Safety Class 3. 99,7 i During nonnal operation, the Essential HVAC SWS will not be operating. , In the event of loss of of f aite power and/or a LOCA the HVAC SWS will 48 be supplied with cooling water from the RHR Service Water System. (D) l l l l \ l Q(D l 1 l V) (D)- Design (U)-Update 9,2-5b Am. No. 48, 12/4/78
e ACNGS-PSAR dent condition will be initiated automatically and will require no operator 1
\ action. System operation for normal shutdown and reactor isolation will (37 D) Q9 be manually initiated from the Control Room.
In the event of a LOCA, the RHR SWS, HPCS SWS and Fuel Pool SWS pumps l37 22 10 will start. Should.a loss.of offsite power occur during a LOCA , these (C) Q8 pumps will momentarily stop until transfer to the Staadby Power System ' .13
-is completed and will be restarted automatically according to the diesel generator loading sequence.
In the event of loss of offsite ' power and/or a LOCA the HVAC SWS will b supplied with cooling water from the RHR Service Water System. 48 Each pair of RHR and Fuel Pool heat exchangers will be provided with a l 22 radiation monitor to detect any radioactive contamination resulting from a tube leak. When contamination is detected, an alarm will sound in the Control Room. Pressure and temperature indicators located at various points throughout the system will infonn the Contro1 ~ Room operator of pump performance and/or line integrity. 9.2.2 REACTOR AUXILIARIES C00LIi4G WATER SYSTEM 9.2.2.1 Design Bases D The Reactor Luxiliaries Cooling Water System (RACWS) will be designed to cool certain reactor and plant auxiliary equipment required for normal operation of the plant. The system will be in operation during all modes of normal operation including refueling. The system is not safety related, will not be required ~ to operate after a LOCA and will not be required for safe shutdown of the reactor. The system may operate under a loss of offsite power condition but with a reduced cooling load. (C)-Consistency (D)-Design 9.2-8 Am. No. 48, 12/4/78
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ACNGS-PSAR p The mixing box dampers will be arranged to satisfy the following condi- 22 t tions: Q214. N 17 a) To ' allow a. fixed quantity of outside air for ventilation and a fixed: quantity-of return air.
.b) Complete recirculation of return air without any outside air during 'l 37' (C) containment isolation operation.
9.'4.6.2.3 Drywell Cooling System The Drywell Cooling System, shown on Figure 9.4-7, will consist of 6 units,
. each of which will include air cooling coils, a fan and a gravity damper to e prevent recirculation through the deenergized unit. 42 During normal operation and all conditions when offsite power is available )
four of the six units will be operating with two acting as. standby. On 46 loss of of fsite power without LOCA, three units will be operated from (C) > the non-nuclear safety. Balance of Plant Diesel Generator power and on LOCA - 46 all units will be inoperative. (D)' 48(U) The air cooling coils will be supplied with water from the Equipment Pro-tectionClosedLoopgoolingWaterSystem. Based on a normal drywell cool-ing load of 4.5 x 10 Bru/hr and a 25 F rise in temperature, each of . the cooling units will have an' air flow of 30,000 cfm. Design data for 3 principal system components are presented in Table 9.4-6. Air will be supplied at high velocity, but not exceeding 360 ft/ min over
, exposed vessel parts, to all sections of the drywell subject to thermal pocketing or stratification, and it will then return to the unducted dry-well cooling unit inlets. >
A 50 percent standby cooling capacity will be provided to assure system re-liability and to prevent ambient temperatures in the control rod drive under the Reactor Vessel from exceeding 185 F. l' All drywell cooling units will be continuously monitored and controlled l from the Control Room, l 9.4.6.2.4 Drywell Purge System The Drywell Purging System will function'in two (2) modes as described below. The pre-entry drywell purge mode and the refueling operation mode, shown 42 schematically on Figures 9.4-5 and 9.4-7, will consist of an exhaust system (C) and a makeup air system. For.the pre-entry drywell purge mode.the exhaust system will utilize one train of the SGTS with a capacity of 5,000 cfm. Air will be exhausted from the drywell and from the containment operating floor level. An exhaust duct will be connected to the top of the drywell with pneumati- 37(D)
, cally-operated fail-closed (F022), and motor-operated (F021) isolation i valves, arranged in series. The duct , on its downstream run, will pene-t (C)-Consistency (D)- De sign Am. No. 48, 12/4/78 9g
ACNGS-PSAR ( 9.4.6.4 Tests and Inspection k Preoperational and pe-iodic tests of all system functions will be performed in accordance with normal plant operating procedures. Specifically, the tests numbered I, 2, and 3 delineated in Table 6.2-11 l37 will apply. Refer to Section 6.2.3.4 for a detailed discussion of testing (p) provisions and procedures. The systems will be proven operable by their use during normal plant opera-tions. Suitabla in .rumentation (Section 9.4.6.5) will be provided to con-tinuously monitor system performance. 9.4.6.5' Instrumentation Application Each fan will be monitored for flow failure with low limit annunciation at j the local HVCP. A single alarm point in the Control Room will annunciate the 32 existence of a problem in the Containment Cooling and Ventilation Systems. (C) All filters will be provided with local indication of pressure drop. All air cooling and heating coils will be provided with local indication of air inlet and outlet temperatures. Radiation monitors will be provided in the plant vent stack and also in the 42 exhaust duct (before isolation valves) to monitor the discharge of the (C) , Containment Exhaust System. The mo, tor will annunciate in the Control Room b upon receipt of high radiation level. In the event of a fire in any area served by Cont ainment Ventilation System, 46 a smoke detect or , located in the common exhaust air duct , will actuate an (C) alarm at The Fire Detection Panel. 9.4.7 ANNULUS VACUUM MAINTENANCE SYSTEM (AVMS) 9.4.7.1 Design Bases The Annulus Vacuum Maintenance System (AVMS) will be designed to maintain a negative pressure in the Shield Building annulus to control annulus pres-sure whenever the Standby Gas Treatment System (SGTS) i s not operating. The annulus pressure maintained by the AVMS will be suf ficiently negative 46(C) to prevent the post-accident annulus negative pressure from becoming less negative than 2 in, wg. assuming operation of the SGTS. 48 (U) 9.4.7.2 System Description The AVMS is shown on Figure 9.4-5 and will consist of two 100 percent ca-pacity exhaust fans connected in parallel. Each will draw air from the 7 annulus through two pneumatic f ail-closed, spring opposed isolation valves. 6(C) A 5 in, wg negative pressure will be maintained in the annulus by the modulation of a pneumatic damper located in a bypass inlet duct of the fan 46 (Oj -assembly. The AVMS will discharge through a vent located on the Auxiliary Building roof. (C) ' (C)-Consistency (D)- Design (U)-Update 9.4-18a Am. No. 48,12/4/78
ACNGS-PSAR
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CNGS-PSAR EFFECTIVE FIGURES LISTING CHAPTER 13 CONDUCT OF OPERATIONS -
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ACNGS - PSAR and operation of the nuclear power ' plant. Three major areas of environ-f( mental protection respor.sibilities include input .of factors af fecting the determination of the plant site location, assessment of the impact that the plant will- have on the environment during normal and abnormal operation and , the responsibility for acquisition of all local, state and Federal permits and approvals exclusive of NRC licensing, The Energy Production Department shown on Figure 13.1-2, . is responsible for h8 (U) the operation and maintenance of the fossil and nuclear power plants. The , testing and operation of the ACNGS will be under the direction of the ACNGS Plant Superintendent. He will report to the Operation Manager in the Energy Production Department. Engineering Department . The Engineering Department shown on Figure 13.1-3, is responsible for the b review of the engineering, design and development of the' nuclear power plant electrical auxiliary systems, switchyard and civil engineering of the power plant site, structures and buildings. 13.1.l'2.1-
. Executive Vice President Primary Responsibility 33 (U)
The Executive Vice President is responsible for and directs the engineering and construction of all generating facilities; the operation and main-tenance of generating facilities; the control and dispatching of system electric energy operations consisting of generation units, substations and high voltage transmission lines, environmental affairs, and QA. Reporting Function The Executive Vice President reports directly to the President for all , technical and administrative matters. I 13.1.1.2.2 Vice President - Power Plant Construction & Technical Services g Primary Responsibility @) The Vice President - Power Plant Construction & Technical Services is responsible for power plant engineering and construction, the quality assurance program, and environmental protection activities. Reporting Function The Vice President - Power Plant Construction & Technical Services reports : to the Executive Vice President 13.1.1.2.3 General Manager of the Power Plant Engineering & 46 Construction Department ;U) O U (U)-Update ; 13.1-3 Am. No. 48, 12/4/78
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ACNCS - PSAR
/^ 13.1.1.2.20 Projects Quality Assurance Manager, Prnjects Division, '( Quality Assurance Department PrimarLR espansibility 48 The Projects Quality Assurance Manager, Projects Divisinn, is respnnaible (U) for the development, implementation, coordination and administratinn of the project . quality assurance activities, including site activities, during the design, engineering and const ruction phases of the power plant prnjects.
Reporting Function i The Projects Quality Assurance Manager, Projects Division, reparts to the' Manager, Quality Assurance. 13.1.1.2.21. Project Quality Assurance Supervisor, Allens . Creek, Quality i Assurance Department Primary Responsibility- 33 The Project Quality Assurance Supervisor, Allens Creek, is responsible for the development, impl ement ation, conrdination, and administration nf the ae- 48 tivities, including site activities, performed by Quality Assurance personnel (U) for Allens Creek Nuclear GeneratinF Station, Unit 1. Reporting Function T 48 j The Proj tet Quality Assurance Supervisnr, Allens Creek reports to the Pro- kU) ject s Osality Assurance Manager. 1.' 1.s.2.22 Site Quality Assurance Supervisor (46 8 ' Primary Respnnaibility The Site Quality Assurance Supervisnr is responsible for the site quality assurance surveillance of the activities for Allens Creek Nuclear Generating Statinn, Unit 1. Reporting Function 48 The Site Quality Assurance Supervisor reports to the Project Quality Assur- (y) ance Supervisor, Allens Creek. i 13.1.1.2.23 Supervisor, Support Division, Quality Assurance Department l46, ; 48 Primary Responsibility (U) The Supervisor, Support Division, is responsible for the activities invnived 48 in audit planning, vendor surveillance, quality assurance records systems' (U) codes and standards, specification and engineering document review, t raining, licensing support and the development and maintenance nf quality assurance documents performed by Quality Assurance personnel for the varinus projects. (U)-Update 13.1-9 Am. No. 48, 12/4/78
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ACNGS - PSAR-Reporting ~ Function 33 (U) k The. Supervisor, Support Division, reports to the Manager, Quality Assurance. bO' (U) 13.1.1.2.24 General Supervisor, Records Management Divisinn, Quality Assurance Department Primary Responsibility The General Supervisnr, Records _ Management Division, is responsible far the design, develnpment, implementation, management and surveillance of the HL&P 48 records management system for the various project's. This respan'aibility ex- (U)_ tends into the receipt, etnrage, ret rie' val and dissemination af reenrds in-cluding those. generated during engineering, design, procurement, ennstruc-t ici. , operation and maintenance. Repnrting Functinn The General Supervisor, Reenrds Management Divisinn, reports to the Manager, Quality Assurance. 13.1.1.2.25 Supervisnr, Operations Divisinn, Quality Assurance Depart ment Primary Responsibility The Supervisor, Operations Division, is respansible for the develnpment and implementation of the Operations Quality Assurance' and Nuclear Fuel Quality Assurance Programs, Start-Up Program, Inservice Inspectinns and Nondestruc-tive Testing for the power plant s. Reporting Function The Supervisor, Operations Division, repnrts to the Manager, Quality Assur-ance. 13.1.1.2.26 Manager, Environment al Prot ectinn Department 46, 48 Primary Responsibility (U) The Manager, Environmental Prniection, is responsible for accumul ation af the various technical ennsiderations nf the envirnnment as generated by staff assistants. These environmental considerations invnive material 33(U) covered in the areas of site selection criteria, radinactive dispersion, t hermal ef fects, air and water quality considerations and environmental surveillance, including meteoroingical monitoring, geophysical testing, hydrological evaluations and all affsite nperatinnal effects of the nuclear power plant. He is responsible for the preparation of the Envirnnmental Report and environmental consideratinns required in support nf all Safety Analysis Reports. His responsibility alan includes acquisition of all local, State and federal permits and approvals exclusive of NRC licensing.
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(U)-Update 13.1-10 Am. No. 48, 12/4/78 -e .. v g,m n---, , , , , ,,,-,-,.en-- ,,,w.,,,,.n-.~~ ~ , . - ,e u rn.,---m, +-,,--~-e v., -n... ~n-- -- --,- - ~ - - -
I l 1 l ACNGS - PSAR l 1 / G Reporting Function 46 The Manager, Environmental Protection, reports to the Vice President - Power (U) Plant Construction and Technical Services. 13.1.1.2.27 Principal Engineer, Nuclear Quality, Environmental 46, Protection Department l48 (U) Primary Responsibility s46 Teh Principal Engineer, Environmental Planning & Assessment , is responsible l(U) for environmental licensing, permits and approvals and environmental sur-veillance of the Allens Creek Nuclear Generating Station Unit 1. He is responsible for preparation of the Environmental Report and all studies necessary for its completion. Reporting Function 46 The Principal Engineer, Environmental Planning & Assessment, re port s to the (U) Manager, Environmental Protection Department. 146, 13.1.1.2.28 Vice President, Operations 1 48(U) Primary Responsibility The Vice President , Operations, is responsible for activities related to j operation and maintenance of the HL&P generating stations. G . . Reporting Functton The Vice President , Operations, reports to the Executive Vice President. 13.1.1.2.29 General Manager, Energy Productic, Department g46, 48 Primary Responsibility (U) The General Manager, Energy Production, is responsible for the operation and maintenance of the fossil and nuclear power plants. Reporting Function The General Manager, Energy Production, reports to the Vice President , Operations. 13.1.1.2.30 Mechanical riaintenance Manager, Energy Production Department l(48 U) l Primary Responsibility The Mechanical Maintenance Manager la responsible for the preventative and (U) corrective maintenance for the fossil and nuclear power plants except the n transmission switchyard. (U)-Update 13.1-11 Am. No. 48, 12/4/78
ACNGS - PSAR 46 p}
'v Reporting Responsibility The Mechanical Maintenance Manager reports to the General Manager of the (U)
Energy Production Depertment. 13.1.1.2.31 Electrical Maintenance Manager, Energy Production l46, 48 ! Department. (U) Primary Responsibility The Electrical Maintenance Manager is responsible for the preventative and l 46(U) corrective maintenance for the fossil and nuclear pnwer plant s except the t ransmission switchyard. Reporting Responsibility The Electrical Maintenance Manager reports to the General Manager of the l 46(U) Energy Productinn Department. 13.1.1.2.32 46' Operation Manager, Energy Prnduction Department l48 Primary Responsibility The Operation Manager is responsible for the nperatinn af the fossil and nuclear power plant s. Reporting Function h)i ( The Operation Manager reportr. to the General Manager of the Energy 46 (U) Production Department. l46, 13.1.1.2.33 Vice President, Engineering 14 8 Primary Responsibility The Vice President, Enrineering, is responsible for enntractual activities as related to the design and engineering af the power plant electrical auxiliary sytems and the design, engineering and ennstruction of the switchyard at the nuclear power plant prnjects. This respnnsibility includes the civil engineering aspects of the site, buildings and structures associated with the project. Reporting Function The Vice President, Engineering, reports to the Executive Vice President . 13.1.1.2.34 Manager, Engineering Dasign and Develnpment l 46, l 48 Primary Responsibility (g) The Manager, Engineering Design and Development, Engineering Department is responsible for the engineering z.nd design of the nuclear pnwer plant l ep) electrical auxiliary systems and the overall electrical prot ectinn of k ,/ the switchyard and for the studies and specification of generator (U)-Update 13.1-12 Am. No. 48, 12/4/78
ACNGS - PSAR characteristics as they pertain to proper integration of the unit into the - (q) system. The electrical auxiliary systems include the normal, alternate and
'K./ emergency power supplies to electrical components and systems for the nuclear power plant projects.
Reporting Function The ManaFer, Engineering Design and Development, reports to the Vice President, Engineering. 13.1.1.2.35 Principal Engineer, Systems Division, Engineering O' g Department Primary Responsibility The Principal Engineer, Systems Division, is responsible for the engineering and design of the nuclear power plant electrical auxiliary systems, the overall electrical protection of the switchyard and for the studies and specifications of generator characteristics as they pertain to proper integration of the unit into the system. The electrical auxiliary systems include the normal, alternate and emergency power supplies to elect rical components and systems for Allens Creek Nuclear Generating Station, Unit 1. Reporting Function O The Principal Engineer, Systems Division reports to the Manager, k Engineering Design and Development, Engineering Department. 13.1.1.2.36 Manage r , Civil Engineer, Engineering Department 46, 48 Primary Responsibility (U) The Manager, Civil Engineering, Engineering Department is responsible for the civil engineering of the site, buildings and structures as well as coordinating the Engineering Department 's activities in support of the design and construction of all HL&P's generating units. Reporting Function The Manager, Civil Engineering, reports to the Vice President, Engineering. 13.1.1.2.37 46. Principal Engineer, Civil Division, Engineering 48 Depart ment U) Primary Responsibility The Principal Engineer, Civil Division, is responsible for site improvements and site investigations for Allens Creek Nuclear Generating Station, Unit 1, including the coordination of soil investigations, groundwater studies, potential pipeline relocation, highway modifications, review and evaluation of all pertinent civil engineering studies and A/E
,/ \ recommendations for plant grade, UHS, design criteria for dam and dikes, i / %/
(U)-Update Am. No. 48, 12/4/78
,13.1-12a
ACNGS - PSAR railroad spur layout, cooling lake operating level, and design earthquake e criteria. He is also responsible for the review and evaluation of all 33 (U) l (U)-Update 13.1-12b Am. No. 48, 12/4/78
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l PROJECTS OA MANAGER { W N PHILLIPS I O ALLENS CREEK i SOUTH TEXAS PROJECT i FOSSf L UNITS ; I PRO [ECTOA PROJECT OA SUPERVISING SUPERVISOR SUPERVISOR ENGINEER _ F W STOERKEL T D STANLEY D C TYLLICK ;
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SITE SURVElLLANCE SITE SURVEILLANCE -FOS $1L OA ] i I
. - HL&P/DE SIGNER / VENDOR /
SUPERVISOR SUPERVISOR OA COORDINATION l
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- PROJECT OA PLAN -PROJECT OA PL AN -AUDITING
- HL&P/EBASCO/GE OA - HL&P/B/R/ WESTINGHOUSE -SITE OA SURVEILLANCE COORDINATION - OA COORDINATION
- AUDITING - AUDITING
- REGULATORY LIAISON -REGULATORY LIAISON ,
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l ACNGS - PSAR VICE PRESIDENT E A TURNER MANAGER R A FRAZAR I SUPPORT OPERATIONS RECORDS MAN AGEMENT GENERAL SUPERVISOR P A SWEARINGEN I I RECORDS CENTER RMS DATA CENTER SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR C A McCLURE R C HENSON E D YOUNG J A MARKS l
- VENDOR SURVEILLANCE - OPERATIONS OA - RECORD SYSTEMS -NUCLE AR FUEL OA - TR AINING -INSERVICE INSPECTION & NDT - LICENSING SUPPORT - QA DOCUMENT DEVELOPMENT - AUDITS - CODE AND STANDARDS - SPECIFICATION REVIEW - MATERIALS APPLICATIONS AM,NO.48, 12/4/78 (U) - UPDATE l HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 QUALITY ASSURANCE DEPARTMENT FIGURE 13.1 ,
ACNCS-PSAR / l l l l l l O Figure 13.1-8 has been intentionally deleted 48(U) (U)-Update Am. No. 48, 12/4/78
ACNGS-PSAR i r~'si 13.lA.1 INTRODUCTION N' The primary responsibilities and reporting functions of the key Engineering personnel on the Allens Creek Nuclear Generating Station are provided in Section 13.1.1.2 of the PSAR. HL&P organizational charts are provided as figures to Section 13.1 of the PSAR. This Appendix 13.lA provides resumes of selected engineering personnel who work on Allens Creek. 13.lA.2 ACNGS PROJECT PERSONNEL RESUMES (U) 13.lA.2.1 C.W. Oprea, Jr. Mr. George 'W. 0prea, Jr. is Executive Vice President. He is a 1952 grad-uate of Rice University with Bachelor of Arts and Bachelor of Science Degrees in Electrical Engineering. He joined HL&P that year in the Dis-tribution Planning Section of the Engineering Department. With engineering design, planning and system operations experience at progre s sive levels of responsibility, he was elected Vice-President-Operations in November 1971, Group Vice President - Operations in January 1973, and Executive Vice , President in December 1974. Mr. Oprea is a Regiatered Professional Engineer in Texas, a member of the Atomic Industrial Forum and American Nuclear Society (ANS) and a senior member of the Institute of Electrical & Electronic Engineers (IEEE). 13.1A.2.2 E. A. Turner l48 t' ) (U) ( ,/ Edward A. Turner, Vice-President of Power Plant Construction & Technical Services, earned a Bachelor of Science degree in Civil Engineering from the University of Houston in 1952. That same year, he joined the John D. Trilsch Company as a Project Engineer responsible for the design and con-struction of large communication structures throughout the southeastern United States. In 1954, he joined HL&P as a Civil Engineer responsible for all phases of transmission and distribution design work as related to soils, foundations, and structures. In 1963, he was appointed Assistant Superinten-dent of the' Civil Engineering Division, in 1966 Assistant Superintendent of 46 the Design Engineering Division, and in 1968 Superintendent of Design (U) Engineering with overall responsibility for the design and construction coordination of all HL&P switchyard and vault projects. Responsibility for the design and implementation of all transmission, distribution, civil, and architectural projects was assigned to him in 1970 when he was appointed Manager of Project Engineering. He was named General Manager of Power Plant Engineering and Construction (PPE&C) in 1972 and General Manager of Trans-mission and Distribution in October 1976. He was appointed to his present position in April 1978. Mr. Turner has attended the Nuclear Operations Short Coarse for Utility Management conducted by Babcock and Wilcox and has participated in a number of management and executive development courses sponsored by various colleges and universities. He has also visited the manuf acturing f acilities and general of fices of all five domestic reactor manufcuturers and attended
,s a videotape course on nuclear power produced by NUS Corporation. He is a
( ; registered Professional Engineer in the State of Texas, v (U)-Update 13.1 A- 2 Am, No. 48, 12/4/78 s --
? I i l ACNCS-PS AR 4 _ TABLE 13. l A-1 (Cont'd) i NAME TITLE EDUCATION APPLICABLE EXPERIENCE (} i POWER PLANT ENGINEERING & CONSTRUCTION (Cont'd) Con s t ruction D. C. Barker Manager BSCE 10 Years 48(U) F. D. Asbeck tonstruction Supervisor BSCE 7 Years 33(U) t E. A. Pe ar son Construction Supervisor B Arch. Design & * { u Construction 23 Years Engineering 46(U) j j R. E. Fulghum Manager BSEE 11 Years 'i i
- 5. C. Schae f fer Principal Engineer, Electrical BSEE 8 Years
) R. T. Beauboue f Principal Engineer, r -[ Mechanical BSME, Ph . D . ME g 15 Years i > h6(U) t i, b R. D. Ellerman Supervising Engineer, l Electrical BSME 11 Years l 46(U) i j C. H. Gr i f fin Supervising Engineer, , Electrical BSEE 13 Years 46(U) ; E. M. Insall supervising Engineer B Sche 10 Years j Hechanical I J. E. Bouvier Senior Engineer Mechanical BSME, MSME 11 Years
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l 'I l ACNCS-PS AR TABLE 13.lA-1 (Cont'd) ! -4t l
- i. NAME TITLE EDUCATION APPLICABLE EXPERIENCE ;
i POVER PLANT CONSTRUCTION & TECHNICAL SERVICES QUALITY ASSURANCE 46(U) ,j t De par tment Management R. A. Frazar Manager BSCHE 10 Years 33 (U) Allens Creek l- g 48(U) ; j W. N. Willips Su pe rvi sor U. S. Navy Nuclear I i Power School 12 Years i i F. W. Stoerket Supervising Engineer BSME 10.5 years - ENGINEERING ASSURANCE DIVISION 46(U' g T. D. Stanley Supervising Engineer BS Metallurgy 9 Years . 48(U) g MS Material Science
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. Department Management 33(U) 48(U).
R. L. Evans vice President, Operations BA, Math 40 Years
'6(U) 4 i W. B. Little General Manager BSME 22 Years i
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ACNGS-PSAR EFFECTIVE PAGES LISTING ; CHAPTER 17 q
/n 1 QUALITY ASSURANCE l
\M Page Amendment l
'1 1* 48 ;
2* 48 3* 48 i 33 11 45 111 45 17.0-1 33 17.0-2 33 17.0-3 46 17.1-1 45 17.1-2 48 17.1-3 45 17.1-4 48 17.1-5 48 17.1-6 48 17.1-7 42 17.1-8 45 17.1-9 45 17.1-10 33
/ 17.1-11 33 17.1 12 33 17.1-13 33 17.1-14 33 17.1-15 33 17.1-16(Amendment Number not shown on page) 33 17.1-17 33 17.1-18 33 17.1-19 45 17.1-20 45 17.1-21 33 17.1-22 45 17.1-23 45 17.1-24 46 17.1-25 33 17.1-25a 33 17.1-25b 33 17.1-26 33 17.1-27 33 l 17.1-28 33,
! 17.1-29 33 17.1'-30 33 17.1-31 33 17.1-32 33 17.1-33 33 1 17.1-34 33 ( 17.1-35 45 ( 17.1-36 17.1-37 33 33 17.1-38 33 1 Am. No. 48, 12/4/78
ACNGS-PSAR EFFECTIVE PAGES LISTING CHAPTER 17
' QUALITY ASSURANCE Amendment- l Pane 17.1-39 33 17.1-40 33 17.1-41 33 17.1-42 33 17.1-43 33 17.1-44 -45 17.1-44a 45
.17.1-45' 33 17.1-46 46 17.1-47 46 17.1-48 45 17.1-49 45 17.1-50 45 17.1-51 45 17.1-52 33 17.1-53 33 17.1-54 33 17.1-55 45 17.1-56 33 17.1-57 33 17.1-58 45 17.1-59 33 17.1-60 33 17.1-61 33 4 17.1-62 45 j
17.1-63 45 17.1-64 33 17.1-65 33 17.1-66 42 17.1-67 33 17.1-68 33 17.1-69 33 17.1-70 33 17.1-70s 45 17.1-70b 48 17.1-70c 45 17.1-70d 45 17.1-70e 45 17.1-70f 45
'17.1-70g 45 17,,1-70h 45
-17.1-71 33 17.1 72 33 s
2 Am. No. 48, 12/4/78
.l ACNGS-PSAR EFFECTIVE FIGURES LISTING
~ CHAPTER 17-QUALITY ASSURANCE Figure Amendment 17.1.1A-1 48-17.1.1A-2 48 17.1.1A-3 48.
17.1'1B-1
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'17.1.1B-2 '33' 17.1.2B-1 45 17.1.3B-1 33 17.1.7B 33 17.1.10B-1 33 17.1.10-1 33(deleted) 17.1.1C-2 33(deleted) 17.1.10 33(deleted) i l
l i. l 3 Am. No. 48,12/4/78
I ACNGS-PSAR QA Plan and Departmental Procedures. The QA Program Manual delineates
} the policies and objectives of the QA Program Manual with the Departmental Procedures by describing the quality related activities to be performed by HL&P on the project, as well as those to be performed by the project's 33(U) prime contractors. The Departmental Procedures control individual depart-ment or division activities af fecting quality.
Disputes over differences of opinion between HL&P QA personnel and others are resolved at the lowest possible level; resolution to disputes that 45 cannot be handled at the lower levels will progress for resolution to the I (g) next higher tier, e .g. , Project QA Supervisor, Projects QA Manager, Manager l48 of QA, Vice President-Power Plant Construction and Technical Services,.up (U) to the Executive Vice President, who directs resolution to disputes which are not resolved at lower levels. The HL&P QA Department was established to provide the effective control of quality activities related to nuclear power plants. For the ACNGS the provision of this control will apply to organizations performing quality related services during the engineering, design, procurement, construction, and operating phases. In addition to the existing QA staff, HL&P utilizes the Design Review Committee and the QA Program Evaluation Committee to accomplish the QA requirements of the pr oj ec t . Intradepartmental and interdepartmental QA relationships are shown in Figure 17.1.1A-3. f 17.1.1A.1 Quality Assurance Department l17 The QA Department is responsible for the development, review, im plemen t a- 33(U) tion, and surveillance of the HL&P QA Program and the ACNGS QA Plan and fo r QA De partment pr oced ur es . This responsibility extends into project activities including engineering, design, procurement, construction, and 45 operation. The Manager - QA, reports on technical,and administrative (U) matters ' to the Vice President-Power Plant Construction and Technical Services. This reporting arrangement provides isolation of cost and scheduling influences from. activities performed by the Manager - QA. The Manager - QA has the duty and authority to identify quality problems; to initiate, recommend or provide solutions; and to verify the implemen-tation and effectiveness of solutions. To enforce this, he has authority to "Stop Work " for cause in the engineering, design, procurements, construction and operation phases of HL&P nuclear power plant projects through appropriate management channels. His principal duties and respon-sibilities include the following : a) Develops HL&P QA Program and individual Project QA Plans and 33(U) approves QA Department procedures, b) Establishes means for implementing the Program and the Plans including personnel indoctrination and training, definition t (U)-Update 17.1 Am. No. 48, 12/4/78
ACNGS-PSAR Assurance, are periodically. reviewed by the Vice President-Power Plant Construction and Technical' Services. 33 45 O' The Vice. President-Power Plant Construction and Technical Services also reviews and approves the HL&P QA Program Manual and Project QA Plans and revisions thereto. WI 41e retaining overall' responsibility. for the Quality Assurance Program, , HL&P has delegated portions of the Program, as listed in this chapter, to ! Ebasco and General Electric for implementation. The Projects QA ' Manager, who' reports to the Manager - QA, has' the responsi-48 (U) bility for. the development, impl ement a t ion , coordination and administration ! of the project quality assurance activities, including site activities, during the design,' engineering. and construction phases of the power plant pr oj e c t s . A Project QA Supervisor, who reports 'to the Projects QA Manager shall be 33 !48 (U) assigned to the ACNGS. This person will have the responsibility of im- (U) plementing the ACNGS QA. Plan and will- interface directly with the Project Manager and line organizations. His principal duties and responsibilities include the following: r a) Assist in developing, implementing and maintaining the ACNGS QA l 33 Plan. Maintain a continuing review of the Plan and recommend (U) revisions to assure viability of the program and compliance with regulatory requirements. b) Perform routine monitoring and surveillance activities for the project, c) Represent the Manager - QA, during regular project activities. l3(U) d) Interf ace and coordinate with other participating organizations on the project activities which affect quality. Provide planning assistance during the ' project for QA requirements. p5(U) e) in phases of engineering, procurement, construction, operations and l33(U) modifications f) Meet with NRC and other regulatory bodies at the home office and at 33(U) the site. g) Maintain skills and knowledge of the industry's QA standards. 33(U) 5 h) Design documents, specifications and purchase orders for systems, structures, components and material covered by the QA Program are 91' II- 7 reviewed for inclusion of adequate QA requirements.
'i) Prepare training outlines for attaining and maintaining proficiency 33 of project QA personnel. (U) p j) Revia and recommend improvements in maintaining complete and 5 retrievable records supporting the Project QA Plan.
(U)-Update 17.1-4 Am. No. 48, 12/4/78
s ACNGS-PSAR = k) Perform routine administrative work for the project quality assurance 33(U) O effort. I
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- 1) Direct the HL&P site QA group through supervision of the HL&P site 33(U)
QA' Supe rvisor. The Project QA Supervisor has the authority to "Stop work" for cause. in engineering, design and procurement by notification in . writing, detailing the reason and the recommended corrective action. The . following are some of the activities affecting quality that may require a work stoppage: a) In engineering, a program with incorrect input parameters. 5 b) In engineering, design or design review being performed by unquali-fied personnel or personnel not authorized to perform the activity. t c)' In engineering, a design based on incorrect assumption or based on an approved design basis which has been subsequently proven inade- t quate. d) In. engineering, engineering- being accomplished with inadequate preparation. e) .In procurement , procurement of equipment from vendors not capable of meeting the required quality levels, p f) In procurement, manufacture of a piece of equipment or material (3 performed in violation of an approved specification. g) In procurement, manufacture or processing of equipment or material in accordance with procedures or processes which do not have the required approval. Activities af fecting quality during construction are stopped through the HL&l' Site ' QA grou p, reporting directly to the Project QA Supervisor. This l48 "stop work" ord^r will be given in writing identifying the reason and the '33 (U)
-recommended corrective action.
HL&P-QA Department will have personnel at the ACNGS site performing a continuing surveillance of the constructor's QA/QC Program. The site QA group will be under the direction of the Project QA Supervisor as shown in l48 (U: 33 Figure 17.1. l A-2. The QA group will be supervised by the site QA Super- (U) visor who is assigned to the ACNGS Site during construction. 5 01-11-7 17.1.lA.2 Other HL6P Departments Having Quality Related Functions
.The design of the Allens Creek Project is delegated to Ebasco and General !
Electric and the Construction is delegated to Ebasco. However , HL& P per- l33(U)
' forms reviews and other quality related functions through various HL&P
' departments as stated below. 17 02-11.2 b i 17.1-5 (U)-Update '
Am. No. 48, 12/4/78
1 i ACNGS-PSAR j l l 17.1.lA.2.1 Po.cer Plant Engineering And Construction Department 17 k l Q2- 11. 2 ( . The Power Plant Engineering and Construction Department organization is shown in Figure 17.1.lA-2. The Project Manager is responsible for the coordination and scheduling for the nuclear power plant project. The Nuclear Division is responsible for the design, engineering, safety 33(U) analysis, nuclear licensing, health physics and plant security of the I nuclear system of the nuclear power plant project. The Engineering division is responsible for the coordination of the engineering, design 42(U) and review of the balance of plant mechanical systems and their asso-ciated instrumentation and control systems and electrical equipment. i The Construction division is responsible for the administration, I scheduling, and coordination of the construction of the nuclear power plant project. 17.1.lA.2.2 Engineering Department The Engineering Department is responsible for the engineering design and development of the nuclear power plant generator characteristics, major 42(U) transformers, communication system, electrical protection system for the generatore and switchyard, and the civil engineering of the power plant site, structures, and buildings. The Engineering Department organization is shown in Figure 17.1.lA-2. i l 17.1.lA.2.3 Environmental Protection Department l l The Environmental Protection Department is responsible for the environ-l pl mental protection of the environs of the plant during both planning ld and operation of the nuclear power plant. Three major areas of environmental protection responsibilities include input of f actors af fect-ing the determination of the plant site location, assessment of the impact l that the plant will have on the environment during normal and abnormal l operation, and the responsibility for acquisition of all local, state, I and federal permits and approvals exclusive of NRC licensing. The 33(U) Environmental Protection Department organization is shown on Figure 17.1.lA-2. 17.1.lA.2.4 Purchasing Department The Purchasing Department organization is shown on Figure 17.1.lA-2. The Purchasing Department is responsible for the procurement of all equipment, material, and services for the nuclear power plant projects. The procurement responsibility includes, when necessary, specification and inquiry review and approval, bidder list approval, bid evaluation and contract and purchase order approval. The procurement of equipment, material, and services required by on-site personnel is accomplished under procedures and instructions provided by the Purchasing Department. 17.1.lA.2.5 Fuel Resources Department The Fuel Resources Department is responsible for the acquisition and 45(U) management of the nuclear fuel for the nuclear power plants. The Fuel A Resources Department organization is shown on Figure 17.1.1A-2. (U)-Update 17.1-6 Am, No. 48, 12/4/78
ACNGS-PSAR f T.ABLE 17.1.2B-3 (Cont'd) ( 2)' a satisfactory quality assurance audit by Purchaser of the construct-ion contractors Quality Assurance Program. If the' manual review and/ or audit are unsatisfactory, and if, in the opinion of the Purchaser there is' no hope of successful corrective actions, the terms of the contract will permit Purchaser to absolve himself of the contract. b) A visit will be made to the home of fice of the contractor to discuss the
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techniques they intend to use in implementing their program ~at the Construction site. I c) Records of past or similer jobs shall be examined at the contractors ' office to verify implementation of constructors quality assurance program , or at least make an evaluation of the contractors qualifications and c apabili t y. d) The results of the proceeding review will be forwarded to HL&P for their 45 concurrence. (U)
- 5. Section QA-I-6 Quality Assurance Records 7
The requirement of this section is modified to impose upon the File , Custodian the responsibility for the review of records that will become. l part of Ebasco's permanent files. The documents will be inspected for ' legibility, proper identification and microfilmability. The require-ment for vendor quality assurance records to be legible and microfilm-able will be imposed on vendors through Ebasco's Purchase Order Speci-fication.
- 6. Section QA-II-5 Supplier Surveillance After issuance of a purchase order and prior to start of fabrication, the Project Quality Assurance Engineer prepares the Vendor Quality As-surance Plan for approval by HL&P QA. Af ter approval, the PQAE for-wards the Vendor Quality Assurance Plan to the Chief Vendor Quality i Assurance Representative for use by the Vendor Quality Assurance
~ Representatives.
F
- 7. Section QA-III-l Instructions, Procedures and Drawings (Section QA-III-
.l 2.
The following' changes are hereby established so that- Section QA-III-l 4 is compatible with Section QA-I-2 of the Ebasco Nuclear Quality As- 48 surance Program Manual - Allens Creek. Project. (U) 7.1 Paragraph 3.6 is replaced by the following: To. assure.that all Quality Control procedures and instructions comply with this manual,
~
applicable codes and regulatory requirements, they shall be submitted for review and acceptance to Quality Assurance prior to implementation. t V-17.1-70b .(U)-Update Am. No. 48, 12/4/78
~~ s -. 2 i
r CI TEj GENER AL MANAGER POWER PLANT ENGINEERING AND ) CONSTRUCTION MANAGER OF PROJECTS O PR >Ec1 MANAGER 1 a i G E-NED MANAGEMENT G E-APED MANAGEMENT 1 k OUALITY OUALITY PROJEM ASSURANCE ASSURANCE R MANAGER MANAGER l O i i tj $ $ JL
ACNGS - PSAR EXECUTIVE ICE PRESIDENT l (' VICE PRESIDENT
- POWER PLANT -
,NSTRUCTION AND "HNICAL SERVICES g i I l l MANAGER QUALITY ASSURANCE PROJECTS QA MANAGER i PROJECT OA EBASCO SUPERVISOR MANAGEMENT l 1 l i CHIEF QUALITY p
^
MANAGER ENGINEER b bN N l l GE NFD fEANAGEMENT QA LINES OF COMMUNICATION l C PROJECT LINES OF COMMUNICATION U - PROJECT AM, NO. 48,12/4/78 (U) Update MANAGER HOUSTON LIGHTING & P7WER COMPANY Allens Creek Nuclear Generating Station Unit 1 l EXTERNAL QA RELATIONSHIPS FIG.17.1.1 A-1
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- DE81GN REVIEW
= QA PROGRAM EVALUATION
( I I GENERAL MANAGER
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MANAGER POWER PLANT QUALITV ENGINEERING AND ASSURANCE CONSTRUCTION I CONSULTING ; ENGINEER PPE&C HOME OFFICE l l l H PROJECTS QA MANAGER MANAGER MANAGER CON ENGINEERING MANAGER ENVIRONMENTAL OF NUCLEAR PROTE CTION PROJECTS DIVISION MANAGER $ i PROJECT PROJE CT OA MANAGER SUPE RVISOR ALLENS CREEK i _ _ _ _ _ _ _ _ . . ____________ j l t :
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i ACNGS - PSAR i
. CHAIRMAN OF THE BOARD AND CHIE F
! EMECUTIVE OFFICER , PRESIDENT i I l EXECUTIVE VICE PR SIDENT VICE PRESIDENT ADMINISTRATIVE l l-I I VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT DPERATIDNS ENGINEERING , $RI S A ENERAL MANAGER GENERAL MANAGER FUEL RESOURCES ENERGY PRODUCTION L l I I I I I I
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l \ I l 1 VICE PRESIDENT POWER PLANT CONSTRUCTION AND TECHNICAL SERVICES SPE CI A L COMMITT E E S
- DESIGN REVIEW
-OA PROGRAM EVALUATION GENER AL MANAGER POWER PLANT ENGINEERING AND CONSTRUCTION CONSULTING ENGINEER PPE&C I
I I I I
^ E NGINE E RING CONSTRUCTION OF C PROJE CTS DIVISION MANAGER MANAGER EN VIRONMENT AL l I IL PROTECTION PROJE CT MANAGER j(
ALLENS CREE K Ik
=
OA LINES OF COMMUNICATION llNTE RN ALI
+ ACNGS - PSAR DOARD OF DIRECTORS jL CHAIRMAN OF THE BOARD - AND CHIEF EXECUTIVE OF FIC E R I PRESIDENT I GROUP EXECUTIVE VICE PRESIDENT VICE PRESIDENT ADMINISTRATIVE I I I VICE PRESIDENT VICE PRES 10ENT ViCE RES ENT p OPE R ATIONS ENGINEERING AND SERVICES GENE RAL MANAGER ENERAL MANAGER ENERGY FUEL PRODUCTION R E SOURCE S I _ l l I I I MANAGER MANAGER ^ OPE R ATIONS ENG E ING MECHANICAL E LE CT RICA L M AN AGE R DESIGN & PURCHA$ LNG STORES MAINTENONCE MAINT E NANCE DEVELOPMENT Jk lL lL )L jk jk l a u MANAGER
?- OUALITY ASSUH ANCE HOME OFFICE
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HOUSTON LIGHTING & POWER COMPANY
' Allens Creek Nuclear Generating Station pno g[- -
ouAuTY ASSURANCE Unit 1 SUPT RVISOR
------I INTERNAL QA RELATIONSHIPS FIG.17.1.1 A-3 oVAuTYinANCE GROUP f
ACNGS-PSAR EFFECTIVE PAGES LISTING APPENDIX C , eb h Amendment No.
-U -1* - 48 r
2* .46 1 3* '48 i 42 11- 42 iii- 42 iv 42 v 42 : vi 42 vii 42 viii 42 C1.1-1 17 C1.2-1 17- ! C1.3-1 35 C1.4 17 C1.5-1 35 C1.6-1 35 C1.7-1 35 C1.8-1 42 C1.9-1 35 C1.10- 1 35-C1.11-1 17 C1.12-1 35
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C1.13-1 35 C1.14-1 17 C1.15-1 17 C1.16-1 '35 C1.17-1 35 C1.18-1 35 C1.19-1 17 C1.20-1 42 C1.21-1 35 ! C1.22-1 31 C1.23-1 35 C1.24-1 17 C1.25-1 35 C1.26-1 42 C1.27-1 42 C1.28-1 46 C1.29-1 42 C1.29-2 42 C1.30- 1 45 C1.31-1 42 C1.31-2 (delet6d) 42 C1.31-3 (deleted) 42 C1.32-1 35 C1.33-1 42 C1.34-1 17 g C1.35 35 1 Am, No. 48, 12/4/78
- Effective Pages/ Figures List
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ACNGS-PSAR-EFFECTIVE PAGES LISTING APPENDIX C l
.j Page Amendment No.
5 C1.78-l' 35 C1.80 35
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C1.85-1 42' C1.86 35 C1.88-1 47 / C1.89-1 35 C1.91-1 42 , C1.92-1 42
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C1.96-1 35 C1.97-1 42 - 01.98-1 -42 C1.98-2 42 C1.99 42 C1.100-1 42
~C1.101-1 42 C1.102-1 35 C1.104-1 42 C1.105-1 - 42 :
C1.106-1 42 C1.108-1 42-C1.109-1 42 , 01.110-1 42: I C1.111-1 42 , C1.112-1 42 C1.113-1 42 , C1.114-1 35 C1.115-1 42 C1.116-1 46
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C1.127-1 42 - C1.130- 1 48 3 Am. No. 48,12/4/78 e erer *www-w ' wra w-w w er-we,-wwe-ew--
ACNGS-PSAR l ,n REGULATORY GUIDE 1.124 47 8 i (Rev. 1, 1/78 lQ110.15U) 42(U) DESIGN LIMITS AND LOADING COMBINATIONS FOR CLASS 1 LINEAR-TYPE COMPONENT SUPPORTS Applicant's Position: The applicant will comply with the requirement of this Regulatory 1 0.15 e s
\ (U)-Update C1.124-1 Am. No. 48,12/4/78
ACNGS-PSAR REGULATORY GUIDE I.130 (Rev. O, 7/77 47l48 (U) DESIGN LIMITS AND LOADING COMBINATIONS FOR CLASS 1 PLATE-AND- 110,15 SHELL-TYPE COMPONENT SUPPORTS Applicant's Por:ition: The applicant will comply with the requirement of this Regulatory Guide. b ( ( C1.130- 1 (U)-Update Am. No. 48,12/4/78
ACNCS-PSAR .m only those valves required for tight shut off will require seat leak-(\s_-} ing test qualification. (2) ~ Re fer to the response to 0110.1. Postulated pipe break loads will not be considered for active pump operability qualification. The justification for this is that if the pipe connected to a pump nozzle is postulated to break the pump would not be available. SinFle failure analysis has taken the loss of the pump into consideration. Postulated pipe break loads will be considered for operability qualification of active valves only for those valves which would be required to close following a postulated pipe break in the pipe in which that valve is located. (i) This response is incorporated into revised PSAR Appendix 3.9.B Section 2.2. (ii) Seismically generated nozzle loads will be considered for operability qualification of active pmnps as stated in Section 2.1 of PSAR Appendix 3.9.B. Seismically generated end loads will be considered for active valve operability qualification as stated in the response to Q110.1. Postulated pipe break loads will not be considered to prove active pump operability as justified above. (O/"'N) Postulated pipe break loads will be considered for operability qualification only for those active valves which are required to close after a postulated pipe break in that pipe in which the valve is located. (iii) The increase in flow, due to a postulated pipe break, vill be specified to the valve manuf acturer for those valves which are required to close following the postulated break. The valve manufacturer will be required to design the valve and valve / actuator combination to close with the increased flow. (iv) Postulated pipe break loads will not be applied for active pump operability qualification. The response for active valves has been incorporated into amended PSAR Appendix 3.9.B Section 2.2. (v) Refer to responses to questions 110.6 (3) and 130.20. Q 48 N 110.17 O 1110.17-2 Am. No. 48, 12/4/78
. - _ . . . ~ - _ . .
ACNGS-PSAR ~ !
. r /~ Item'No.
- 130.18 The anti-corrosion measure as described in Section 3.8.3.1.3 which is the same as in GESSAR-238 may not be-adequate,-and it '
- is our understanding that the GESSAR-238 application is being revised to include new anti-corrosion measures. Provide in-formation to. show that your proposed measures are adequate.
- The. following questions are related to 'the Containment Struc-
- tures Design Report , dated July,1977 and prepared by EBASCO Services Inc.
RESPONSES , The. response to this item is located in revised Sections 3.8.3.1.3 and 48 6.2.1.6. (U) x
'i r
p: ! ( I (U)-Update F130.18' Am. No. - 48,12/4/78
ACNGS-PSAR o Open Item No. j -{
~110;3 You have committed to performing a preoperational vibration and test program during startup and initial operating conditions 110.4 on' all safety-related ASME Class 1, 2, and 3 piping and associated supports and restraints. The staff position is that you should also commit to extend the test program to include (a) other high energy piping systems within seismic Category I structures, (b) high energy portions of systems whose failure could reduce the functioning of any seismic Category I plant feature to an unacceptable safety level, and .
(c) all seismic Category I moderate energy lines as required l by Regulatory Guide 1.68 Revision 2, " Initial Test Programs' for Water-Cooled Reactor Power Plants."
RESPONSE
The applicant will perform preoperational vibration testing during startup and initial operating conditions on all safety-related ASME Class 1, 2, and 3 piping and associated supports and restraints. This position is in accor-dance with current NRC criteria and prior commitments. The applicant does not recognize the use of Regulatory Guide 1.68 Revision 2 criteria, as this revision has not been formally issued for industry use. Failure ef fect i analysis will confirm that the rupture of non-ASME Class 1, . 2, and 3 piping will not reduce the functioning of 'any seismic Category I plant feature to p) ('d an unacceptable safety level. All non-ASME Class 1, 2, and 3 piping systems will he visually inspected for excessive vibration during the preoperational vibration test program. 48 Q 3110.3 l 110.4 l l i I I i ! ( M110.3&4 Am. No. 48,.12/4/78
ACNGS-PSAR m Open Item No. ['
'~~
110.6(3) In your response, you stated that for ASME Class 1, 2, and 3 components and supports that P.4 peaks of dynamic loads associated with plant Faulted Conditions will be combined by the Square Root of the Sum of Squares Method (SRSS). In the absence of acceptable technical justification for the use of the SRSS method, our position is that you should commit to combine dynamic loads by the method of absolute summation until and if the staff concludes that adequate technical justification has been provided for the use of the SRSS method.
RESPONSE
The applicant commits to apply the generic resolution of this issue to the design of Allens Creek. However, for cases where the generic resolution cannot be practically implemented such as steel plate structures within the containment boundary the applicant will justify the acceptability of the design to the satisfaction of the NRC Staff. 48 For each such case construction or installation will not be completed Q until NRC staff approval of the justification has been obtained. N110.6 (3) A V)s IA)
*. J M110.6(3)-1 Am. No. 48, 12/4/78
.. ~
t ACNGS-PSAR
}
. Open Item No'. ,
130.'20'- The information provided does -not' justify ~ your proposed method tof combining loads.- Unless- justification is provided and found acceptable, the staff's position is that' the combination of l loads should.be done by the absolute sum method.
RESPONSE
The applicant commits .to apply the' generic resolution of this issue 'to the However, for cases'where the generic resolution design of A11 ens' Creek. cannot-be practically implemented such as steel plate structures within the containment boundary the applicant.will justify the acceptability.of the design to.the satisfaction of the NRC Staff. For each'such case 48-construction or. installation will not be completed until NRC staff approval of the justification has been obtained. 9 N130.20 9 h b i i i 1 h i
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M130.20-1 Am. No. 48, 12/4/78
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ACNGS-PSAR Oprn Item No.
,m 211.2 With regard to the postulated loss of a CRD pump (Item 211.2-( ') 02/17/78), our position is that:
'w) (1) You should provide the bases that will be used to deter-mine that unacceptable impairment of control rod scram capability has occurred, (2) You should provide automatic protection or demonstrate that 20 minutes is available for operator action, and (3) You should describe initial and periodic test programs that will be used to demonstrate that this capability is maintained for plant lifetime. You should also describe how the leak performance of the check valves will be continuously monitored via monitoring of ac-cumulator pressure, as stated in the response.
RESPONSE
The event postulated by this question suggests a series of postulated con-ditions which are extremely improbable. As stated in the original response to this question, assuming one CRD pump is unavailable and the other one fails, the following two cases could occur: (X lo) If the accumulator pressure is maintained no operator action is O necessary 48 H, (b) If any one of the 177 check valves leak, the operator will be informed 211 of low pressure by an annunciation in the control room (see Figures .2 4.2-19a and 4.2-20) and the operator can then scram the reactor using reactor pressure and/or the Hydraulic Control Units with accumulator pressure. Each HCU (177) is safety grade and has its own accumulator. Assuming the accumulator check valves leak at the maximum allowable rate against which they had been designed, the operator has in excess of 20 minutes to initiate a scram after the loss of the CRD pumps has been alarmed in the control room. The two cases indicated above respond to this cuestion. However, during informal conversations with staff members, additional failures have been postulated. The applicants position is that it is unreasonable and un-justified to postulate simultaneously the loss of both CRD pumps, the common mode failure of accumulator check valves, and reactor pressure too low to drive the control rode into the reactor. The events postulated utilize accident assumptions applied to normal operational events. Accumulator pressure is continuously monitored. The leak pedomance of the accumulator check valves cannot be determined when either one of the CRD pumps is operating since the check valve may be open. However, if the CRD pumps are not operating (i.e. a plant outage), low pres sure in any of the accumulators would be alarmed and indicated by local pressure indicators. []) ( As shown on page 16.3/16.4-18, the proposed technical specifications call for the accumulators to be checked for level and pressure alarms once per shift. Hence, any gross accumulator check valve leakage would be detected. M211.2 Am. No. 48, /4/78
ACNGS-PSAR ;
,C Open Item No.- -
211.3' 'With: regard to the values assumed for, safety / relief valve openings in the' overpressure protection analysis, response to . , Item 211.3 (2/17/78) presents values which are lower than GESSAR and are inconsistent with ACNGS PSAR (page 5.2-14a) which. states that the: assumed values are typically 1 to 2 per- ; cent above the actual nominal values. Correct this inconsis-tency by using theLhigher values which account for serpoint errors and any instrument setpoint drift that may occur during , operation. T
. RESPONSE ,
The response to this item is located in revised Section 5.2.2.4 and 40( ) PSAR, Figure 5.2-6 , i m t
.- )
l. (U)-Update M211.3 Am. No. 48,12/4/78
'} ACNCS- PS AR
.Open Item No. ,
211.22' Your response to Item 211.22 M/17/7P) does not adequately demonstrate that safety criteria would not be exceeded if credit were not allowed for. non-safety-grade-equipment operation during anticipated transients. This area is currently being pursued with CE on a generic basis. Either provide additional analysis for Allens Creek showing that no credit is taken for non- -! safety-grade equipment, or provide a ec,mmitment to. comply with the generic resolution. 'I PESPONSE A GE Licensing Report, " Additional Generic Information Relative to' Plant I Operational Transient Analysis in BWR-4/BWR-5/BWP-6 Safety Analysis reports", ! forwarded to the'NRC on 8/11/78 addresses this subject. The generic findings ; cited in this document essentially apply per se to Allens Creek, as discussed
~
on page 1.5-1 of the report.
'I The applicant will adopt the generic resolution of this issue for the' ACNGS. ~ The applicant reserves the right to provide an acceptable alter-native at s'later date.
48 Any alternatives will be provided 'for NBC staff review and will not be constructed or installed until staff approval is obtained. M. 211.22 L
/
k I i M211.22 Am. No. 48, 12/4/78
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ACFGS-PSAP Open Item No. 5 211.26:(1): Your: response.does not consider leakage across' normally closed: valves into systems such as. the .LPCS and RllP. [ Identify all the systems connected to the PCPB where - such leakage could occur and describe how the leakage ' detection system will include the capability to detect leakage through all such valves and provide an ' alarm. l
.7 EESPONSE F The detection of intersystem leakage will be accomplished by the use' of :
existing. plant systems and equipment. 'This detection will be monitored ; from the reactor coolant pressure boundary to connect low pressure' systems, a) LPCS System: Two. pressure transmitters are available to monitor leakage through the containment (and system) isolation valves. Pressure transmitter E21-t!050 provides indication of leakage through the (inner containment' isolation) testable check valve 1E21-F006. Pressure transmitter E21-N054 provides indication of leakage through , the (outer containment isolation) motor-operated gate valve 1E21-F005 and system pressure downstream of the LPCS Pump disch'arge check valve. Both transmitters have control room indication and transmitter E21-N054 has a high pressure alarm function with a set- ! point of 525 psi. I A backup method 'of monitoring intersystem leakage is wi th relief valve 2E21-F018, which also protects the LPCS piping from overpres-- > sure. This relief valve is' set at 600 psig. Its capacity is 100 gpm and 'will lif t only af ter the pressure accumulation reaches the-valve setpoint. The discharge line is led back into containment'and into the suppression pool. Leakage into the suppression pool-(at the 100 gpm rate) would be detected by the suppression pool level monitoring system within approximately 15 minutes. b) Bl!R System (3 Loops): Two pressure transmi t ters are available to monitor leakage through the containment (and system) isolation valves. Pressure transmitter E12-N058 (A,B,sC) provides indication of leakage through the (inner containment isolation) testable check valve 1E12-F041 ( A,B ,6C) . Pressure transmitter E12-N053 (A,E6C) provides indication of leakage through the (outer containment iso-lation) motor operated gate valve 1E12-F042 and system pressure downstream of the RilR Pump discharge check valve. Both transmitters have control room indication and transmitter E12-F053 has a high-pressure alarm function with a setpoint of 500 psi. The shutdown cooling function suction line leakage is detected by pressure transmitter E12-N057, which is set to alarm in the Main Control Room at a setpoint of 180 psi. Pelie f valve 1E12-F005 . protects this. portion of the system (set at 200 psig). l
#~' / c) .
IIPCS System: This system is designed to a pressure rating consis- ; ( tent with the PCPB, therefore leakage into the system is of no great M211.26(1)-1 -Am. No. 48, 12/4/78 l l - ., - - . . , _ . , - - , - , , _ , . .--,. - - __ , - , . - . . _ . ~ . , ,e , ~
ACFCS-PSAR (j\ / consequence. However, relief valve IE22-F035 is provided for backup protection. For the unlikely case that leakage through both isolation 4g valves and the HPCS pump should occur, pressure sensors and relief valves are provided on the suction side of the HPCS pump as shown on 9 pressure alarms on high and low HPCS suction side pressure, and relief 211.26 valve F014 is provided for backup protection. d) R_CIC System: This system (on both steam and water lines) is designed consistent with the RCPB, there fore , leakage is of no great consequence. For the unlikely case that Icakage through both isolation valves and the RCIC pump should occur, pressure seasors and relief valves are 48 provided on the suction side of the RCIC pump as shown on PSAR Q Figure 5.5-7a. Pressure sensor PS-N021 alarms on high RCIC suction 211.26 side pressure and sensor PS-F006 alarms on low PCIC suction pressure. Felief valve F017 is provided for backup protection. Leakage from the FCPB in all cases will require passage throur.h a contain-ment isolation valve (PCIC, HPCS, RHR, LPCS). These valves undergo period-ic testing as described in Section 6.2.1.4.4.3. A typical detection system schematic is shown in the attached figure. The additional dose to plant operators from there relief valve discharges will be negligible considering the bounding case of the MS-SPV's. M211.26(1)-2 Am. No. 48, 12/4/78
ACNCS-PSAR.
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Open Item No. 361.4 Following its review of responses to Item 361.4 that you provided in Amendments 42 and 44, the Corps of Engineers in its letter of June 23,1978 (copy attached) provided ; additional comments about compaction requirements for l Class I-a Fill and the effect of slickensided surfaces ! on the design shear strength. Provide the clarification outlined by the Corps of Engineers in its letter of June 23,.1978. RESP 0NSE' Class 1-a Fill material will be compacted to a minimum relative density of 80 percent. The field control will be determined using the Modified Proctor test. The referenced Corps of. Engineers' letter comments that Class I. slope i stability analyses be performed using residual shear strength parameters. Houston Lighting. & Power Company will perform an effective stress stability l analysis, on Class I slopes using residual strengths for natural clays, remolded strengths for compacted clays and a safety factor of 1.25. 48 Preliminary testing indicates residual strengths of = 8. 5 F wi th . Q-C ='100 psf and = 16 with C = 200 psf. Additional' testing will be 361.4 performed on presplit samples. The results of this testing will be used to establish the final residual strength. A discussion of the analysis and tests results and their applicability to ACNCS Unit I Class I slopes will be submitted to the NBC in an appropri-ate manner when completed. In any event, use of the residual shear strength should not be applicable to Allens Creek site clays since the slickensided clays present are the result of drying and shrinkage , not large scale movements. There are no large scale slickensided surfaces in the Allens Creek clays; slickensides are approxi-mately 1/2 in size, irregular and nonplanar. Only large scale movements could result in the reduction to residual shear strength values, llIAP's slope stability analyses (see PSAR Section 2.5.5.2) indicates this will not occur in Allens Creek clays. M361.4-1 Am. No. 48, 12/4/78
l ACNGS-PSAR Open Item No. l ( 010.5 In Section 3.1.2.4.17.1 of the PSAR, " Evaluation Against ., l Criterion 46," you state "The Essential Services Cooling l Water System will be designed to permit testing of system l l operability with simulation of emergency reactor shutdown or LOCA conditions and transfer between normal and emer-gency power sources." Provide clarification that this commitment includes consideration of the coincident loss of the cooling lake as included in the design basis events - (Section 9.2.5.3.1.2 of the PSAR) in the evaluation against I the requirements of Criterion 46 of the General Design Criteria.
RESPONSE
The loss of the cooling lake will be taken into account in ACNGs eval . untion against criterion 46 in .that the testing will include those valves associated with directing discharge flow from the Essential Service Water Discharge Structure to the Ultimate Hest Sink Discharge Structure. These valves are shown in FSAR Figure 9.2-la. O o V N010.5 Am No. 48, 12/4/78
i ACNGS-PSAR I (M Open item No.
- ) i
, 'w/ l 110.2 For the PSAR to be in agreement with the staff l position as discussed in Standard Review Plan 3.6.1 and 3.6.2 and Appendix F to the Safety Evaluation Report, the first sentence in Section 3.6.2.2.4 (g) (3) must be removed. That sentence now states, " Piping welds subject to 100 percent volumetric inspection will be those short sections of process pipe themselves which serve as a part of containment, i.e. no guard pipe exists to contain the rupture in this section.
RESPONSE
The above sentence has been deleted from PSAR Section 3.6.2.2.4(g)(3), f\ v
/ 'v N110.2 Am. No. 48,12/4/78
t ACNGS-PSAR t Open Ite.n No. 110.3 The staff position remains as stated in the and 110.4 , enclosure to our letter of July 21, 1978.
RESPONSE
The response to this item is included in the revised response to Open Items 110.3 and 110.4 in PSAR Appendix M. , b
\
. O
\ "
N110.3 & 4 Am. No. 48, 12/4/78
ACNGS-PSAR .b -' ( Open Item No. 110.6(3) In you response'to Item 130.20, provided by Amendment 110.17 No. 47 to your PSAR,-you stated, "The applicant commits 130.20 to apply the generic resolution of this issue to the design of Allens Creek. However, for cases where the generic resolution cannot be practically implemented
.such as steel plate structures within the containment boundary the applicant will justify the acceptability of the design to the satisfaction of the NRC staff."
Provide clarification that for each such case con-struction~or installation will not be completed until NRC staf f approval of the justification has' been ob-tained.
RESPONSE
The' requested clarification is provided in the revised responses to Open Items 110.6(3) and 130.20 in Appendix M and the response to Item 110.17 in Appendix 1. b\ .k N110.6(3) Am. No. 48, 12/4/78
ACNCS-PSAR ,
, Open Item No.
110.7 You have not provided sufficient infomation in the PSAR to enable us to complete our l review of the design criteria to be used for supports for ASME Class 2 and 3 components. Specifically, design criteria and loading combinations have nc,t been provided for standard and plate and shell type supports in the balance-of-plant scope of supply and for all ASME Class 2 and 3 supports in the nuclear steam supply system scope of supply. For ASME Class 2 and 3 linear supports in the balance-of-plant scope of supply the stress limits and the methods used to combine response as described in the PSAR are not completely acceptable. The information is not sufficient to enable us to complete our review of the design criteria proposed for these supports. In particular and for the information in Table 3.9-8 of the PSAR (1) the factor of 1.2 under an upset condition does not exist in NF, (2) no faulted limits are given, and (3) verification that the faulted buckling p limit complies with F.1370(c) should be provided.
RESPONSE
The respsonse to this item is provided in PSAR Table 3.9-8. r o N110 7 Am. No. 48,12/4/78 M_
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ACNGS-PSAR l; Open Item No.
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110.18 'The . staff in its generic review has not completed its review of recommendations 5,_6,-and 7 of the report ORNL/SUB/2913.8. Therefore, for Allens Creek, reliance on these recommendations is not acceptable at this time. Revise your commitment to omit a dependency on those recommendations.' A commitment to the generic resolution which result from the ongoing discussions between the NRC and the BWR Mark II Owner's Group would also be acceptable.
RESPONSE
The' applicant commits to the generic resolution of this issue. The.appli-cant reserves the right to provide an acceptable. alternative at a' later - date. Any proposal alternative will not be implemented until it has been reviewed and approved by the NRC staff. . l Q' l
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i l l- < N110.18 Am. No. 48,12/4/78
i ACNGS-PSAR-l l Open Item No. l l 130.6 On the basis of discussions with your representetivas in l a meeting in Bethesda, Maryland on October 30, 1978, we ! sgreed to further review your propoaed analysis methodology l with modifications that you proposed to delineate in an I amendment to your Preliminary Safety Analysis Report. Gen- l erally, we understand that the methodology would be modified for frequencies less than 8 Hertz to utilize accelerations calculated using your methodology modified to acconinodate a free-field Regulatory Guide 1.60 response spectra control motion at an elevation corresponding to the bottom of the reactor building mat. In addition to a description of the modifications to the methodology and the bases as des-cribed in the meeting, provide the following additional information for our review: (1) For systems and components in the Reactor Building, Auxiliary Building, and Fuel Handling Building with natural frequencies less than 8 hz, we understand that you will increase the amplitude of the floor response spectra for design by a multiplier determined by the ratio of the Flush b/ Flush a response spectra within designated frequency ranges. Provide the explicit criteria to be used to establish these multipliers. Ihe Flush b response spectra should be based on the use of the envelope of G G *1.5, and G /1 .5, O or the response spectra hNe,d hhC AVE should h broadened by plus or minus 157.. (2) In order to justify the use of the Flush a analysis, provide a comparison of the design shears and moments as computed by the Flush a analysis and the Spring an analysis for the (Reactor Building, including the Shield Building, Steel Containment, Dryvell, and the RPV pedestal).
RESPONSE
The response to this item is located in new Appendix 3.7A. 130.6(1) - Section 2.3 of Appendix 3.7A 130.6(2) - Section 3.0 and Table 3.7A-7 of Appendix 3.7A
' N130.6 Am. No. 48, 12/4/78
' ACNGS- PSAR , '
Open Item No. J 211.2 With regard to the postulated loss of a CRD pump (Item 211.2-- 02/17/78), our position is that: (1) You should provide the bases that will be used to determine that 20 minutes is available for operator action, and (2) . You should provide automatic protection or demonstrate.that this capability is maintained for plant lifetime.. You should also describe how the leak performance of the check valves will be continuously monitored via monitoring of accumulator pressure, as stated in the response.
RESPONSE
The response to this item is located in revised Open Item No. 211.2 which ; is located in Appendix M. , I ( l i ( N211.2 Am No. 48, 12/4/78
ACNGS-PSAR Open Item No. 211.3 Your ' responsewe In particular, to note Item 211.3 requires that this breaksupplemental size (approx. discussiog).
.02 ft produce a peak cladding temperature in excess of the tempera-ture produced by a large break DBA'previously analyzed. The following additional information should be provided. q (1) Justify that the system provided for diversion of LPCI l flow meets single failure criteria so that diversion- !
P before 10 minutes need not be considered. (2) Provide' further justification that a diesel failure. causing loss of the LPCS is more limiting than. a loss of ! the LPCI for core cooling. It is not apparent in your discussion that CCFL effects on reflood times were-included. Discuss the relative effects of these low pressure systems upon the parameter in the LOCA calcu-lations, e.g. , reflood, core heat transfer. (3) Provide a-sensitivity study showing peak clad temperature as a function of break size for small break LOCA's assuming diversion will % initiated at 10 minutes. Perfom this study for FJCS and recirculation line breaks. For the most limi ting break, provide the q following figures: (a) Water level inside the shroud as a function of time during the LOCA , (b) Reactor vessel pressure vs. time. (c) Convective heat transfer coefficient vs. time (d) Peak clad temperature vs. time (e) ECCS flow rate vs. time. t (4) Justify that diversion at times greater than 10 minutes will have less severe consequences than diversion at 10 minutes (considering appropriate break sizes for later diversion). (5) Provide a discussion diversion whichsize for this break balances (approx. the need
.02 ft{or LPCI
) with .
the need for abundant core cooling (GDC 35). For example, this discussion could relate to Figure 6.2 with regard to the likelihood of LPCI diversion for this size break. O - N211.3-1 Am. No. 48.12/4/78
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ACNCS-PSAR p
RESPONSE
( Open Item 211.3 is the same as the concern identified in a letter dated October 11, 1978 to T. N. Ewing (Public Service Co. of Oklahoma) from Steven A. Varga, NRC. Analysis responding to this letter is being sub-mitted on the Black Fox application (docket Nos. STN 50-556 and STN 50-557). The response shows that the worst single f ailure break combina-tion, assuming LPCI diversion at 10 minutes to the containment spray mode, is the HPCS line break with failure of the LPCS diesel generator resulting in the loss of the LPCS pump and one LPCI pump. The PCT for this break combination is below that for the design basis recirculation line break, and hence, below the Appendix K limit of 2200 F. The conclusions reached in the Black Fox response are applicable to Allens Creek. Allens Creek Open Item 211.3 is responded to by referencing the Black Fox response. L N211.3-2 Am. No. 48, 12/4/78
ACNGS-PSAR l Open Item No. 211.22 In your additional response in Amendment 47 to the PSAR you stated, "The applicant will adopt the generic resolution ; of this issue for the ACNGS. The applicant reserves the right ! to provide an acceptable alternative at a later date". Provide ; a commitment that alternatives will be provided for staff ) review and will not be constructed or installed until staff approval is obtained.' ] RESPONSE , The response to this . item is provided in the revised reponse to Open Item i 211.22 in PSAR Appendix M. l I i 0 I t B
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ACNGS-PSAR i [ Open Item No. i 1211.26- Additional information is needed. relative to detection of l 1eakage into the HPCS and the RCIC systems to either show conformance with the position of Regulatory Guide 1.45 or q to demonstrate that such leakage does not need to be i considered. .i
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RESPONSE
The -response to this item is .provided in the. updated response to Open Item ! 211.26 in PSAR' Appendix M. t 5 t-N211.26 Am. No. 48,12/4/78
ACNGS-PSAR j] Open Item No. l 361.4 Following its review of responses to Item 361.4 that you ; provided in Amendments 42 and 44, the Corps of Engineers ' in its letter of June 23, 1978 (copy attached) provided additional comments about compaction requirements for Class I-a Fill and the effect of slickensided surf aces on the design shear strength. Provide the clarification outlined by the Corps of Engineers in its letter of June 23, 1978.
RESPONSE
The response to this item is located in the revised Open Item No. located in App. M. i i l i e i
-s LU N361.4 Am. No. 48, 12/4/78
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ACNGS-PSAR A I % Open
- en No.
O 361.5 In Section 9.2.5.3.2 of the PSAR you state, "In the event that the rate of sediment accumulation is such that it appears that the allowable level of accumulation will be exceeded during the life if the plant, the sediment will be removed before that allowable limit is reached. In addition to level of sediment accumulation, limits on slope of the surface of the accumulated sediments should be considered to assure that unacceptable consequences will not result from sediment flow into pump intake during design basis events. State the allowable configurations for accumulated sediments within the cooling lake and provide a preliminary description of the technical specifications that will be used to assure maintenance of acceptable sediment configurations. Include criteria, procedures, and technical specifications for maintaining sediment configura-tions.
RESPONSE
The response to this item will be provided in the FSAR. The Applicant will ensure that the effects of accumulated sediment do not prevent the UHS from performing its intended function. l A U l l , N361.5 Am. No. 48,12/4/78 _ - - ._ _ .- _ _ . . - _ , _ __ . _ . _ _ _}}