ML19347F263

Amend 58 to Psar,Providing Updated Info Addressing Contentions Re Charcoal Adsorber Fires,Turbine Missiles & Blockage of Ultimate Heat Sink Intake Structure

ML19347F263
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	05/15/1981
From:	HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML19347F262	List: ML19347F261 ML19347F263
References
NUDOCS 8105180130
Download: ML19347F263 (79)

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ACNGS-PSAR I HOUSTON LIGHTING & POWER CMPANY ALLENS CREEK NUCLEAR GENERATING STATION - UNIT NO.1 '

'O PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT NO. 58 INSTRUCTION SHEET ,

This amendment contains infornation pertaining to PSAR Update. i Each revised page bears the notatien Am. No. 58, (5/81) at the bottom of the page. Vertical bars with t'ie number 53 representing Amendment No. 58 have been used in the margins of the revised pages to indicate the location of the .

revision on the page. [

f The following page removals and insertions should be made to incorporate ,

Amendment No. 58 into the PSAR. (

i REMOVE INSE RT {

(EXISTING PAGES) (AMENDMENT No. 58 PAGES)

Chapter 2 Chspter 2 l 25*

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l ACNGS.PSAR REMOVE INSERT (EXISTING PAGES) (AMENINENT NO. 58 PAGES) lC Chapter 6 Chapter 6 1* 1* j 2* 2*  ;

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ACNGS-PSAR i EFFECTIVE FIGURES LISTING (Cont'd)

CHAPTER 2 r- SITE CHARACTERISTICS .

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M1 (Appendix M, Section 2.5) 58 M2 58 M2A 58 M3a 50

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M32 50 M33 50 M34 50 l M35 & 36 50 48 Am. No. 58, (5/81)

ACNGS-PSAR

,_s applicable, the horizontal (0.lg) anu ve:tical (0.067g) components of the design basis earthquake are input.

.[v} To find the worst possible radius and center of rotation yielding the circle with the lowest factor of safety, a search routine is built into the program by which a trial center of rotation is selected. The program will investigate different radii from that center of rotation computing and recording the safety factor for each radius. It then moves the cen-ter of rotation at a prescribed increment to a different trial location and the above process is repeated until the lowest safety -factor is reached.

The simplified Bishop solution yields results that are conservative in 50 that shear resistance between slices, which would tend to raise the fac- 361.4 tor of safety against sliding, is neglected. When the simplified Bishop solution is used to compute a factor of safety under dynamic loading ad- '

ditional conservatism is built into the program in that the computed safety factor is calculated assuming the components ut the design earth-quake accclaration act only in one direction, neglecting any back and forth motion, and the magnitude of the acceleration of the design earth-quake is taken to be a constant over the entire slope for an infinite i length of time.

In performing the sliding wedge method the Ebasco computer program was also used. The sliding wedge method consists of an active wedge being mobilized against a neutral harizontal block and a passive resisting wedge. The eg factor of safety is calculated as the ratio of the sum of the resisting forces in the horizontal direction to the sum of the driving forces in the

(\ -) horizontal direction. In applying the sliding wedge method to the two cross-sections the input data and search routine is similar to that of the slip circle analysis previously discussed. This method also includes a seismic loading in the analyses. This was done by including the product of the ,

weights of the wedges and the neutral block with the horizontal acceleration _

f actor of 0.lg. This force was then considered to act in the direction of the postulated slide as a driving force. The vertical component of the l seismic loading is also incorporated into the solution tending to reduce frictional resistance between the sliding wedges. This vertical setsmic force is computed as the product of the weights of the neutral block and t

the wedges with the vertical acceleration f actor of 0.067g.

The results of each of these aaalyses are presented on the tables on Figure M2. In all cases the actual safety factor exceeds the recommended minimum safety factor from Table MS, indicating that the slopes are safe.

l 58 The above described detailed investigation has accurately established the soil conditions in the area of the ultimate heat sink at Allens Creek. The continuous sampling in the upper soils and careful undisturbed sampling of clays and sand establishes a sound basis for the selection of lower bound ,

strength senples. Selection of design strength parameters incorporated the use of lower bound strength parameters from the test results, using scry conservative test procedures. Results of the analyses indicat ed s -

U 2.5-M7 . Am. No. 58, (5/81)

ACNGS-PSAR satisfactory safety factors. Reflected in the analyses are the changes -

's required to obtain the required safety factors. In order to maintain the 1 vertical to 3 horizontal slope of the causeway it was necessary to excavate the surface c]1ys from beneath the causeway. Additionally the slopes of the ultimate heat sink basin have been flattened to I vertical 50 to 8 horizontal from the original I vertical to 3 horizontal. These 361.4 changes are the result ot using the p=9 from the consolidated drained repeateo direct shear tests.

Mb UHS CAbbEWAY SLOPE S1AblLlIY In accordance with the design parameters and loading criteria outlined in Figure h-2, studies were performed to verify the design and the validity of tne slope stability of the Uhs Causeway. Directly underneath the Causeway there exists an approximate 50 foot thick layer of dense sand. This dense sand layer will act as a firm foundation and provide adequate support and sufficient stability for the Causeway. An analysis simulating a potential tailure was conducted and postulated that the potential f ailure plane would occur within the recompacted material, i.e. the man-made Causeway. For this failure a minimum safety factor of 1.46 against seismic loading conditions as shown in Table M6 would result.

A parametric study was also performed using the Seismic Wedge Method, con-sidering side ftiction alor. the potential failure plane. The failure plane was allowed to cut through 50 feet of dense sand, and to penetrate intu the under lying clay layer. This study yielded a minimum safety

,- s, tactor oi 1.34 and approximated the results obtained from the Simplified 58

; hisnop Method. Either method is considered acceptable and adequate for the

\,/ Causeway design.

It should be noted that the constant excitation forces utilized in both wed ge and Simplified bishop Method slope stability analyses provide an extremely conservative approach in the case of an instantaneously back and fortn earthquake motion of 10 second duration. This conservatism wa also coupled witn an unrealistic large amount of driving mass in the Wedge hethod as indicated in Figure M-2. A comparison of these parametric study results is presented in lable M6.

In summary, tne slope stability results as presented in Figure M-2 are extremely conservative, since they neglect all side friction forces along tne tailure planes, are coupled with unrealistically large amounts of driving mass, and neglect the low average soil design parameters. The mini- t mum safety factors for the Causeway design are well within acceptable limits for different loading conditions as established by U.S. Army Corps of Engineers. (bee PSAR Section 2.5 Appendix M Table M5.)

Furthermore, in accordance with the NRC Regulatory Guide 1.27, an analysis ,

postulating tqe f ailure of the man-made Causeway was performed. A cross .

section taken at the lake iront area was investigated for the potential blockage of the waterway should the Causeway be postulated to fail in this area. The results of the slope stability analysis along with postulated tailure plane are presented in Figure M-la.

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-s 2.5-M8 Am. Nn. 58 (5/81)

f ACNGS-PSAR The movement of the postulated failure plane was examined using Newmark's 7 method (1) with a safety factor equal to 1, i.e. an impending failure of

[\ the Causeway slope. This yielded a conservative soil movement of less than 4 inches. This soil movement would produce a very minor bulging type of

\s_/ deformation which would be experienced in the Causeway along the lake front '

area (see Figure M-2a). As an additional conservatism and in order to 58 provide a positive stoppage 01 soil movement, a concrete retaining wing wall structure will be provided at the Causeway / lake front area (see t Figure M-1) to assure a continuous passage of cooling water into the fore-bay canal of tne Ultimate lieat Sink Intake Structure, t

[

E : bedded piping and electric cabling within the Causeway embankment has i Deen designed to ensure physical independence and redundancy following postulated failure of the Causeway. Figares M-2 and M-2a contain Causeway  ;

cross sections illustrating piping and electric cabling corridors. j E

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4 (1) "Ef fect s of Earth' :e on Dams and Embankment", Fif th Rankine Lecture 58 s

by N.M. Newmark, Geotechnique London, England, 1965.

L 2.5-M8a Am. No. 58, (5/81)

1 1 ACNGS-FMiR i i

a T ABLE 'M6 .

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, COMPARISON OF THE RESULTS OF THE PARAMETRIC STUDY j

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l Pbthods and Conditions Factor of-Safety Remark [

Seismic Simplified Bishop 1.35 l

Slip Circle l Seismic Wedge Method using 1.46 Failure plane occurs within f average low test results causeway l.

Seismic Wedge Method with 1.34 Failure plane cutting  !

friction resistance along through 50 ft dense sand and ,

failure plane penetrating into under clay 58  ;

Seismic Wedge Method, using 1.15 Failure plane cutting average low test result and through 50 ft dense sand and j neglecting friction penetrating into under clay resistance j i

Design parameters: C = 1000 psf for clay, 9 = 38 for sand -

Loading Criteria: SSE and full reservoir WL = EL 118 f t i

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I U.H.S. CAUSEWAY - STABILITY AN ALYSES FIGURE M-2A

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ACNGS-PSAR l i

LIST OF EFFECTIVE PAGES I CHAPTER 3 DESIGN OF STRUCTURES,, COMs4JNENTS, EQUIPMENT AND SYSTEML Page Amendment 1* 58 ,

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• f- CHAPTER 3 DESIGN OF STRUCTURES. COMPONEffrS. EQUIPMENT AND SYSTEMS Figure No. Amen hent No.

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a f- x LIST OF TABLES

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3.2-1 Squipment Classification 3.2-15 3.2-2 Design Requirements for Quality Group A and Safety Class I Systems, Structures, and Components 3.2-35 3.2-3 Design Requirements for Quality Group B and Safety Class 2 Systems, Structures, and Components 3.2-36 3.2-4 Design Requirements for Safety Class 2 and 3 Systems and Components 3.2-37

, 3.2-5 Design Requirements for Quality Group C and Safety Class 3 Systems, Structures and Components 3.2-38 3.2-6 NOTES FOR TABLES 3.2-2, 3.2-3, 3.2-4, 3.2-5 3.2-39 3.2-7 Correlation of Quality Group Designations with Industry Codes and Standards for '

[' ' Mechanical Components 3.2-40 1

3.2-8 Conditions of Design for Safety Class 1, 2 and 3 components 3.2-41 I

3.2-9 Summary of Safety C1 css Design Requirements 3.2-43 3.2-10 Active Valves in Seismic Category I Systems 3.2-44 3.3-1 Vertical Wind Speed and Loading Distribution for Plant Structures 3.3-7 3.4-1 Flood Protection Measures 3.4-4 3.4-2 Ultimate Heat Sink Intake Structure Net Hydrostatic and Hydrodynamic forces 3.4-5 I 3.5-1 Structures Designed for Missiles and Missile Types 3.5-17 3.5-2 Intentionally Deleted 3.5-18 3.5-3 Turbine Missile Characteristics (Typical, for 38" Wheel) 3.5-19 N

3.5-4 Probability of a Turbine Missile Impacting

) Plant Safety Related Structures 3.5-20 3.5-5 Characteristics of Tornado Generated Missiles 3.5-21 xxxii Am. No. 58, (5/81)

ACNGS-PSAR

'V LIST OF FIGURES (CONI'D)

Figure Title  !

i 3.8-1 Containment Vessel Structural Features '

3.8-2 Deleted 54 l l 3.8-3 Reactor Containment Building Internal Structures Base Details 3.8-4 Reactor Building Piping Penetrations 3.8-5 Reactor Building Steel Plate RPV Pedestsi l54 3.8-6 Reactor Building Reactor Shield Wall 3.8-7 Typical Equipment Foundation 3.8-8 Typical Reinforcement Detail 3.8-9 Dryweli Base Detail 3.8-11 Reactor Euilding Dome - M&R l 54 3.8-12 Containment Vacuum Breaker AA/if vs. Maximum Containment Negative Pressure 3.8-13 Small Line Break Inside the Containment 3.8-14 Containment Response After Drywell and Containment Vacuum Breaker Initiation 3.8-15 Drywell Negative Pressure vs. Time f<

Steam Condensation Following Small Primary System 3.8-16 Relative Effects of Heat Sinks and Spray Depressurization 3.8-17 Inadvertent Spray Activatien 3.8-18 Three Foot Containment 2enetration Dedicated For 58 Degraded Core Rule Making 3.9-1 RPV and Internals Vertical Dynamic Model 3.9-2 The Amplification Factor p as a Function of the Frequency Ratic w for Various Amounts of Viscous Damping xxxviii (U)-Update l

Am. No. 58, (5/81) l l

l ACNGS-PSAR f'~N. from turbining of the driven end of the equipment due to blowdown of the f system pressure upon rupture of the system pressure boundary.

(V) l The most substantial piece of rotating equipment is the recirculatton pump ,

and motor which, in the event of a major tecirculation line break, and  ;

under certain system blowdown conditions, can theoretically reach overspeed beyond practical design limitations and result in ejection of various parts of the pump and motor. This hypothetical situation is cur rently the topic of discussions between CE and the NRC. The Applicant will implement 49 the generic resolution of these discussions. ,

3.5.2.2 Turbine taissiles 3.5.2.2.1 in t r.).t uc t ion The potential tot damage to safety related structures, systems and compo-nents due to turbine failure has been evaluated to determine whether addi-

. tional protection, beyond that inherently provided by plant building orien-tation and existing structural shielding, need be provided to further re-  ;

duce the probability of damage.

The probability of damage was calculated for the vital plant structures by ,

evaluating the product of the probability of missile generation, the prob- l ability of impact on the structure and the probability of damage to the 58 l s t ru c t ure s.

'~'} The evaluation of the individual probability components and a summary of y ,,/ the overall damage probability is discussed in the following sections.

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3. 5- 3 Am. No. 58, (5/81)

ACNGS-PSAR testing techniques, are now better able to discover surface and in-y' s ternal defects. In addition, a thorough inspection program will be u;ed to inspect the critical bore and keyway regions to minimize the (N- /) possibility of stress corrosion cracking. An ultrasonic test has 58 been developed to detect stress corrosion cracks in these regions well before cracks would grow to critical size necessary to cause turbira wheel failure. Laboratory investigation has revealed some of the basic relationships among structure strength, material sleength, FATT, and defect size and location, so that the reliabil- r ity of the rotor as a structure has been significantly improved during the past few years.

New starting and loading equipment and instructions reduce the severity of surface and bore thermal cycles incurred during service.

The laprovements include: better temperature sensors; better guid-ance for station operators in the control of speed, acceleration, and loading rates to minimize rotor stresses.

Progress in design, better material's and quality control, more ri-gorous acceptance criteria, and improved machine operation have sub-stantially reduced the likelihood of burst failures of turbine-generator rotors operating near rated speed.

d) Turbine-Generator Overspeed Protection The improvements of rotor quality discussed above reduce the chance of failures at operating speed, but they do tend to increase the f~'S hazard level associated with unlimited overspeed, because of the (N- greater missile energy associated with higher bursting speed.

) Therefore, it is pertinent to examine the Turbine Overspeed Protec-tion Systems. For condensing units, the devices to cor. trol the flow cf steam into the turbine are discussed in the following paragraphs.

e) Main and Secondary Steam Inlets Main and secondary steam inlets have valves in series. These valves are:

1) Control valves, or throttle valves, controlled by the speed l 35(U) governor and tripped closed by emergency governor and backup overspeed trip.

2) Stop valves or trip valves, actuated by the emergency governor l 35 (U) and backup overspeed trip.

Emergency main stop valves of the steam sealed design have been used l 17 on General Electric steam turbines of 10,000 KW and larger since 1948. More than 650 turbines have been shipped and placed in service during this period, and there has been no report of the main stop valve failing to close when required to protect the turbine.

Impending sticking is disclosed by the full closed test feature so that a planned shutdown could be made to permit the necessary cor-rection. Such correction almost always requirea removal of the i

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i ACNGS-PSAR j j oxide layer that builds up in the stem and bushing, which would not I occur on a low temperature nuclear application.  !

l Wr j Staan-driven auxiliary turbines, like the main units, include two l complete lines of defense: control and stop valves, speed and emer- 35 (U) ,

I gency governors against destructive overspeed. j

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(s have less stored energy than corresponding parts of the low-presrure turbine. %ese turbines are totally enclosed in a (V) shielded compartment on upper me?.zanine floor of the Turbine 35(U)

Building.

i

4) High-pressure turbine rotor [

This rotor would not be expected to fail at runaway speed. l35(U)

But, even if failure did occur, the fragments should be re-

, tained by the heavy-section, bo' died high-pressure shells.

5) Couplings }

T Couplings are designed to withstand overspeed higher than l the maximum speed at runaway. [

f The following components could produce high energy missiles:

I

1) low pressure turbine wheels f The wheel capable of producing the most dangerous missile is the last stage. Using the analysis techniques described in j Reference 3.5-3, it has been shown that a 120 degree frag-  ;

c.ta t is the most dangerous in terms of a concrete slab that i can be perforated af ter lesving the turbine casing. r

2) Bucke t vane

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The last stage bucket is both the heaviest and most ener-getic of the vanes capable of escape from the turbine  :

casing. Favorably oriented, i.e., head-on, the vane could conceivably penetrate approximately 10 inches of steel.

Such orientation is unlikely, as the tip of the bucket would ['

quickly strike the outer diaphragm ring tsogentially. In-terference with the other buckets could also be expected.

In the ensuing tangle, the inner casing and hor.d structure would probably contain the blade.

At worst, the vano could ricochet and escape through the 1-1/4 '

l inch-thick plate toward the end of the hood. It is judged that in any case no more than half of the initial energy is retained.  !

I i) Turbine Missiles - Probability Analysis

! e The present analysis utilizes the historical turbine failure pro-bability value of 4X10 per yea r per unit for destructive over-speed failures (see Reference 3.5-22). Analyses have indicated that for destruct.ive overspeed failure, the initial egergy imparted to 53 postylated last stage turbir missiles is 41X10 ft-lb . The missile energy outside the turbine asing is 20.5X10 f -lb 6 f 6

the turbine casing absorbs 20.5X10 ft-lb f(seeTable3.5-3f.and

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ACNGS-PSAR i Studies of design overspeed failure when applied to a turbine having

(a N

38" last stage buckets, indicate that the initial egergy imparted to a postulated turbine missile is approximately 16X10 ft-lb g.

This energy is not sufficient to penetrate the turbine casing. and consequently, all' missiles associated with a design overspeed failure would be contained. Alterrd.tively, if missile should perforate the 58 turbine casiing, their resLiual energy would be so low that they could not damage any safety-related structure.

. 3.5.2.2.3 Probability of Impact with Safety Related Structures Figure 3.5-1 shows the relative location of low pressure elements of the !35(U) i turbines and the plant structures. Missiles may be ejected at any angle of the 360 degrees about the turbine axis. The missile ejection angles and 3 53 directions are illustrated in Figure 3.5-2.

l Tests have indicated (Reference 3.5-5) that deflection angles (6 ) "I 2

turbine generated missiles will be close to zero (f 5 degrees) for interior discs and 0 to 25 degrees for the last stage disc. A uniform probability l 58  ;

distribution is assumed within this range of angles. To calculate the im- '

pact probability it was also assumed that a missile would be ejected with equal probability in the 360 degrees around the rotor axis (6 g).

Table 3.5-3 shows the characteristics of a typical turbine missile from the 1 58 last stage wheel.

s a) liigh Trajectory Turbine Missile Probabilities A convenient means for estimating the probability of strikes frot. high

trajectory missiles lies in calculating the overall extent of the region diich the missiles can reach, Because of the plant arrange-ment, the striking of all the vital plant structures with I igh tra- .

58 jectory missiles re these missiles be ejected at angles bounded by 85 (0,( 90. quires that Figure 3.5-3, which is based on information contained in References 3.5-5 and 3.5-16, is a map of the x y ground plane locations into i 35(U) which the missile indicated in Table 3.5-3 would fall. The turbine 1 58 wheel is shown at the intersection of the x and y-axis.

O is the angle between the velocity vectors' projection onto the YgZ plane (the plane of the turbine wheel) and the Y direction measured in the Y-Z plane.

0 is the angle between the velocity vectors' projection onto the ,

i Y2X plane (the horizontal plane) and the Y direction measured in 03.7  :

the Y-X plane.

+ is the true angle between the velocity vector and the Y-X plane (the horizontal plane) measured in the plane normal to the horizon-tal plane in Wiich the vector lies, j

' O See Figures 3,5-1 through 3.5-4 for a graphical representation of (j these angles, i

j 3.5-10 (U)-Update Am. No. 58, (5/81)

AGNGS-PSAR The maximum range achievable ne;glecting air resistance (air resis-(~~

(

V )s tance changes range probability distribution but its impact is no; significant) is with 4> equal to 45 degrees. Locations were cal-35(U)

I 58 culated by varying 6,from 0 to 90 degrees. A mirror image plot l 37(U) arocnd the x-axis would result for values from 90 to 180 degrees.

Values from 180 to 360 degrees would be meaningless, for these l missiles are ejected below the ground. Calculations were made for values of 6 frca 2

0 to 25 degress nich encompasses the maximum postulated turbine missile deflection angle.

The curves in the lower half of Figure 3.5-3 show landing zones for '

missiles released at values of 6 up to 45 degrees. The curves 1

above the y-axis define landing zones for high trajectory missiles,  ;

for which Sg lies between 45 and 90 degrees. The two patterns  ;

are actually coincident in each of the four quadrants of the x y plane. The dotted curves and arrowheads in the lower half of Fig- 1  !

ure 3.5-3 represent an overlay of the high trajectory curves on the 03,7 low trajectory solid curves. The dotted curves and arrowheads in  ;

the upper half represent an overlay of the low trajectory curves on  !

the high trajectory solid curves.

The significance of this figure lies in the uniform distribution of the angles 6g and 6 2 Because of this, and because the lines are calculated at uniform subdivisions of those distributions, each landing zone defined by a pair of S g lines and a pair of 6 lines has an equal probability of being struck. Further,focations p) h V

outside of the region bounded by r = 6,900 feet and the appropr# ate limiting value of 6 cannot be reached by the missile.

2 37(g)

All of the vital plant structures lie within the region bounded by 85 ( 6, ( 90 and 0( 6 (52 degrees. Thus the probability of striRing any particular plant structure is only related to the plane (horizontal) area of the structure, and is given by the ratio of that ,irea to the area of the region which can be struck by the missile. The latter in turn is conservatively estimated to be the ,

area of half the ellipse having a major axis equal to the maximum ,

range of the missile ejected with 6 = 0 degree and a minor axis 58  ;

with e g, = 90 degrees and any specified 6 . The probability of impact from high trajectory missiles on the safety related structures is given in Table 3.5-4 together with the aggregate total damage probability. .

b) Low Trajectory Missiles Due to the plant arrangement, orientation of the turbine generators,  ;

and the relative elevations of the turbine operating deck and the ,

other structures, no safety related structure is exposed to impact damage by low trajectory missiles. As shown on Figure 3.5-1 only the Radwaste Building is located within the 25 degrees deflection [

limit. .

35(U)  ;

Thr. Radwaste Euilding is located below the turbine operating deck 1rvel. In order to impact directly on the roof of thi- structure, a

3.5-11 (U)-Update Am. No. 58, (5/81) ,

i ACNGS-PSAR

,} it would be necessary to penetrate the operating floor at an impact 35(U) angle of five to fourteei degrees from horizontal. The operating ficor is composed of reinforced concrete three feet thick. (See Figure 1.2-22). At these angles of impact, the missile would not penetrate the operating floor.

To impact the walls of the Radwaste Building adjacent to the Turbine l35(U)

Building, it would be necessary for the missile to penetrate the reinforced concrete turbine pedestal and several reinforced concrete internal walls at the mezzanine level. The re fore , the missile would not have suf ficient energy to reach these walls.

2 Summacy and Conclusion Tne high trajectory' turbine missiles are characterized by their nearly vertical trajectories. The total damage probability of a high trajectoJV turbine missile strikieg the safety related structures is less than 10 per unit year as listed in Table 3.5-4. In addition, the. vulnerable safety related equipment area which is exposed to the potential turbine missile is redundant sci physically well separated. Consequently the risk from high r trajectory turbine missiles is insignificant.

t 4

The ACNGS turbine generator has been arranged in a peninsula orientation.

With the exception of the Radwaste Building, this configuration excludes all major systems important to safety from the low trajectory turbine missile s trike zones. The Radwaste Building does not contain any essential systems

/ required for safe shutdown and is located below the turbine operating deck

( level. The location of the Radwaste System components relative to the tur-bine is such that they are adequately protected by the presence of the re- 58 infurced concrete pedestal, internal building walls and the turbine opera-i ting deck floor. Thus, the plant configuration complies with the guide-4 lines of Regulatory Guide -1.115. " Protection Against Low Trajectory Turbine Missiles."

i In addition to the above, due to the redundancy and testing features of the j turbine overspeed protection, quality control manciacturing processes, materials, and inspection program, the hypothetical turbine missiles are considered very remote. Consequently the risk of potential turbine missile damage to safety related plant structures , sys teras, and components for the

facility is acceptably low.

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0 3.5-11a (U)-Update l Am. No. 58, (5/81) i i

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1 SECTION 3.5 DEFERENCE 3 (Cont'd) i i 1

3.5-15 L: ELE"t EC 58 i

. 3.5-16 DELETED  :

i 1 t l 3.5-17 DELETEu  ;

j 3.5-18 D. R. Miller and W. A. killiams, " Tornado Protection for the l Spent Fuel Storage Pool" APED 5696 of General Electric,  ;

l Novembe r , 1968. i f

l 3,5-19 W. F. luughes and J. A. Brighton, "Thecry and Problems of 21 Fluid Dynamics," Schaum's Outline Series, Schaum Publishing j l Co., 1967.

j

' 3.5-20 H. A. Luarez, " Missile Generation by Fluid Propulbion," '

presented' at ASCE, New York lietropolitan Chapter, March 26, .

1974. ,

3.5-21 Electric Powe r Research Institute , " Full-Scale Tornado-Hissile f 35 (D) l

impact Tests," NP-148, Interim Report, April, 1976.  !

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3.5-22 U.1 Nuc. lear Regulatory Commission Standard Review Plan [

! Section 3.5.1.3. 58 +

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, TABLE 3.5-3 I 2

}~ TURBINE MISSILE CHARACTERISTICS l (TYPICALt_FOR 38" WHEEL) l i

, r l

Fragment Angle, Deg. 120  ;

i i j Fragment Weight, Lb. 5944 l Radius of CG, ft. 2.093 Polar Inertia, Ib-ft-sec 2 462  !

t  :

l Min. Proj. Area, ft.2 3.657 2

i Max. Proj. Area, ft. 8.368 1

Failure Speed, Percent of 1800 RPM 169 i '

i Initial Velocity, ft/sec. 666.8

! Energies. Million Ft-lb.  !

I 1 Initial, Translation 41.0 [

Init ial . Rotation 23.5 [

t

Outside Turbine Casing, Maximum Translation 20.5  !

! 58 i Af ter Air Drag, (Vertical Trajectory), Maximum 16.3 I

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TABLE 3.5*4 [

! PROBABILITY OP A TURBINE MISSILE IMPACTING Pl. ANT t

[. SAFETY REI.ATED STR11CTt1RES [

~

)' i High Trajectory Total Damage (2) j Missile Impact Probability i l'. Structurn Probability (1) (yr-I) l Reactor Building 2.5 x 10-4 1.0 F. 10~8 f

< k Control Building 3.5 x 10-4 1.4 x 10-8 j I

4 Fuel Handling Building 4.3 x 10-4 1.7 x 10-8  ;

J l DG Building 1.7 x 10-4 6.9 x 10-9  !

58  !

! Reactor Auxiliary Buildtog 5.3 x 10-4 2.1 x 10-8 l

,i I UllS Intake Structure 1.6 x 10-4 6.4 x 10-9  !

f 7.5 x 10-8 1 "

j 1. Based on missile deflection angle (0 2 ) of + 25 and air resistance la j neglected.

t

2. B.ased on missile generation probability of 4.0 x 10-5 per unit per year.

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xy PLANE - TURBINE PLANE X

V,: INITI AL VE LOCITY 6:j ANGLE FROM y AXIS TO Vy ,

6:2 ANGLE FROM yz PLANE TO V, 03 : ANGLE ON THE GROUND

& : ANGLE FROM GROUND TO V, AM.NO. 68, (5/81)

HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huclear Generating Station O Unit 1 ILLUSTRATION OF VARIABLES USED IN THE TURBINE MISSILE ANALYSIS FIGURE 3.5-2

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CHAPTER 6 l 1

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TABLE OF CONTENTS CHAPTER 6 ENGINEERED SAFE 1'l IEATURES u

Section Title Page 6.1 GENE RAL 6.1-1 6.2 00NTAINNENT SYSTEMS 6.2-1 6.2.1 00NIAINMENT FUNCTIONAL DESIGN 6.2-1 6.2.1.1 Design Bases 6.2-1 l 6.2.1.2 System Design 6.2-2 6.2.1.2.1 General 6.2-2 6.2.1.2.2 Shield Building 6.2-4 6.2.1.2.3 Containment Vessel 6.2-5 6.2.1.2.4 Drywell and Components 6.2-6

6.2.1.2.5 Pedestal and Reactor Shield Wall 6.2-8 6.2.1.2.6 Weir Wall and Horizontal Vents 6.2-9 y ,/ 6.2.1.2.7 Upper Pool 6.2-10 1 -6.2.1.2.8 Pressure Suppression Pool 6.2-11 6.2.1.2.9 Containment Floors, Platforms and Rooms 6.2-13 d

6.2.1.3 Design Evaluation 6.2-15

)

6.2.1.3.1 Containment Transient Analysis 6.2-15 6.2.1.3.2 Subcompartment Transient Analysis 6.2-30 6.2.1.3.3 Shield Building Transient Analysis 6.2-31

6.2.1.3.4 Post-Accident Containment Pressure Calculation 6.2-32 i

6.2.1.3.5 Containment Pressure Calculation Innivertent Actuation 6.2-32 6.2.1.4 Testing and Inspection 6.2-32 6.2.1.4.1 Provisions for Testing 6.2-32 6.2.1.4.2 Pre-Operational Leak Rate Testing 6.2-33 4

6.2.1.4.3 Initial Integrated Leak Rate Test u.2-35a

( ,)' 6.2.1.4.4 Post-Operational Leak Rate Tests 6.2-37 l i Am. No. 58, (5/81) l l

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ACNGS-PSAR LIST OF TABLES U[ N Table Title Page l

6.2-la Design Criteria for Containment System 6.2-82 6.2-lb Supplementary Information for Water Pool Pressure 6.2-86 Suppression Containment 6.2 2 Accident Chronology Steamline Break 6.2-89 r 6.2-3 Primary System Energy Distribution 6.2-90 i i At the Time a Steam Line Break Occurs  !

6.2-4 Containment Subcompartment Analysis 6.2-91  !

6.2-5 Description of Assumptions Used in Shield Building 6.2-92 -[

Annulus Transient Analysis f i

6.2-6 Containment Response to a Loss-of-Coolant Accident 6.2-94 {

6.2-7 open 6.2-95 l 6.2-8 Design Data for Standby Gas Treatment System Components 6.2-96 .

6.2-9 Properties of Filter Media-Activated Coconut Shell 6.2-99 l

Charcoal U

6.2-10 Single Failure Analysis - Standby Gas Treatment System 6.2-100 I q

6.2-11 Summary of Tests 6.2-102 i

6.2-12 Containment Penetration and Isolation Valve Information 6.2-103 t

l 6.2-13 Intentionally Deleted 6.2-110 6.2-14 Comparison of Safety Related Air Filtration Systems with 6.2-111 ,

Regulatory Position of Regulatory Guide 1.52 6.2-15 Summary of Containment Analysis Results 6.2-114 l d

6.2-16 Deleted 6.2-115 I

6.2-17 Energy Sources and Sinks for DBA Recirc and Main Steam 6.2-116 Line Break Accident 6.2-18 Approximate Vent Flow Parameters 6.2-118 6.2-19 Loss of Coefficients Used in Vent Flow Model 6.2-119 6.2-20 Equivalent Loss Coefficients 6.2-119a  !

6.2-21 RWCU System Double Ended Line Break 6.2-120 f 4

i ix Am. No. 58, (5/81)  !

ACNGS-PSAR LIST OF FIGURES A

\ Figure Title

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6.2-23 Shield Building Annulus Transient Analysis Pressure vs. Time (Lineal Scale) 6.2-24 Shield Building Annulus Transient Analysis Temperature vs. Time 6.2-25 Standby Cas Treatment System 6.2-26 Containment Isolation Valve Arrengements - Sheet 1 6.2-26 Containment Isolation Valve Arrangements - Sheet 2

. 6.2-26 Containment Isolation Valve Arrangements - Sheet 3 1

6.2-26 Containment Isolation Valve Arrangements - Sheet 4 6.2-26 Containment Isolation Valve Arrangements - Sheet 5 6.2-26 Containment Isolation Valve Arrangements - Sheet 6 6.2-26 Containment Isolation Valve Arrangements - Sheet 7 6.2-26 Containment Isolation Valve Arrangements - Sheet 8 6.2-26 Containment Isolation Valve Arrangements - Sheet 9 p) i

\d 6.2-26 Containmenc Isolation Valve Arrangements - Sheet 10 6.2-26 Containment Isolation Valve Arrangements - Sheet 11 6.2-26 Containment Isolation Valve Arrangements - Sheet 12 6.2-26 Containment Isolation Valve Arrangements - Sheet 13 6.2-26 Containment Isolation Valve Arrangements - Sheet 14 6.2-26 Containment Isolation Valve Arrangements - Sheet 15 6.2-26 Containment Isolation Valve Arrangements - Sheet 16 6.2-26 Containment Isolation Valve Arrangements - Sheet 17 i 6.2-26 Containment Isolation Valve Arrangements - Sheet 18 i 6.2-26 Containment Isolation Valve Arrangements - Sheet 19 4 6.2-26 Containment Isolation Valve Arrangements - Sheet 20 6.2-26 Containment Isolation Valve Arrangements - Sheet 21 6.2-26 Containment Isolation Valve Arrangements - Sheet 22 6.2-26 Containment Isola tion Valve Arrangements - Sheet 23 6.2-26 Containment Isolation Valve Arrangements - Sheet 24 xiii Am. No. 58, (5/81)

ACNGS-PSAR  !

LIST OF FIGURES r

-~s Figure Titles j

, t L N- ' 6.2-26 Containment Isolation Valve Arrangements - Sheet 25 j 6.2-26 Containment Isolation Valve Arrangements - Sheet 26 l

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6.2-27b Intentionally Deleted l

6.2-28 Time Required to Reach 4 Percent by Volume Hydrogen in Drywell l vs. Peak Cladding Temperature - Conservative Tempe ca tare  ;

Distribution (Reference 6) l 6.2-28a Hydrogen Concentration in the Drywell (Without Blower Activation) t

-0.00023" Metal-Water Reaction i

! /'~'N  !

( ,) 6.2-29 Hydrogen Concentration in the Drywell (Following Blower Activation) 6.2-30 Long Term Post Accident Containment and Drywell H2 Concentra tion 4

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ACNGS- PSAR TABLE 6.2-14 COMPARISON OF SAFETY RELAIED AIR FILTRATION SYSTEMS 41TH REGULATORY POSIIION OF REGULATORY GUIDE 1.52,

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Regulatory ,

Control Room Position Standoy Gas ECCS Area Filtered Exhaust Emergency Filtration Item Treatment System System System la e All systems will comply with regulatory System complies with regulatory position. positions except:

2a System complies with regulatory position. System complies with 1) System does not include demisters.

regulatory position .Within this system there is no source of entrained water droplets

. and therefore demisters are not required.

1

2) Electric heating coil is provided as for humidity control.

58 n.

- 2b Physical separation will be provided between redundant components of all systems.

m 2c All systems will be designed as seimic Category I.

2d Not applicable. All systems are located outside Containment and therefore not subject to accident pressure surges.

1 2e All systems will comply with regulatory pos tion and radiation levels given in Table 3.11-3.

gp 2f All systems will comply with regulatory positica. Flow rates are less than 30,000 c fm. HEPA filter

. atranger. 6s will be limited to three high.

,z

j. 2g Pressure dif ferential indicators will be provided locally for the demister, medium efficiency filter, pre-HEPA, I

on after HEPA and across each ESF filtration train. Remotely located pressure differential indicators for the

(" s pre-HEPA and across each ESF filtration traic will . be provided in the Control Room. In addition, high limit alarm across the medium ef ficiency filter, pre-HEPA and across each ESF filtration train will be provided in the l! 10 Control Room.

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. - , , . - , . - , , - ,-,.-,-,-.--r----,-mwr r -- nee o* e ,e w ,,,.r-....w.,-m-w, . rw - wa.,en,,,- w--.-nm,re,+-,-t --,>w...e.- y .-c=,m..sm,-w,wm * --%3-w.-w-,,2.-,---,s-w-n-,w-.,-----=--.en-.,w,- .m n1.---.-. -, ,-n , , - -

! \ / 4

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~_ ./ v w' ACNGS- PSAR IABLE 6.2-14 (Cont'd)

Regulatory Control Room Position Stand 5y Gas ECCS Area Filtered Exhaust Emergency Filtration Item Treatment System System System Each ESF filtration train will be provided with a low ficw indicator and low flow a' arm in the Control Room.

Temperature indicators will be provided locally at the inlet of the electric heating c il and the charcoal adsorcer. These temperatures will be indicated in the Control Room. High charcoal adsorber bed temperature will be indicated and alarmed in the Control Room.

2n Systems will comply with applicable IEEE standards (see Section 7.1).

2i Systems will be automatically activateu upon the occurrence of a DBA by a redundant ESF signal.

2j-1 All systems will comply with regulatory position.

SE c.

L* 3a-j All systems will comply with regulatory positions.

R; 3R System design includes provisions for preventing adsorber fires by ensuring continued cooling of adsorbers by air flow even in the event of a single failu re . This is accomplis >.ed by providing cross connections between redundant fans and filter trains. In addition, system design includes provisions to apply water to the adsorber section, from the plant Fire Protection System, in the event of a fire in the adsorber beds.

31 p All systems will comply with regulatory position.

Aa-e All systems will comply with regulatory position.

g: Sa-d All systems will comply with regulatory positions.

g 63-b All systems will comply with regulatory positions.

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TABLE OF CONTENTS CHAPTER 7 (CONT'D)

, -~s Section Title Pare

\s / 7.5.1.5.1 Loss of Habitability of Control Room 7.5-26 7.5.1.5.1.1 Criteria 7.5-26 7.5.1.5.1.2 Conditions Assumed to Exist as the Control Room Becomes Inaccessible 7.5-26 7.5.1.5.1.3 Description 7.5-27 7.5.1.5.1.4 Procedure for Reactor Shutdown from Outside Control Room 7.5-28

7. 5.1. 5.1. 5 Controls and Instrumentation 7.5-29 7.5.1.5.1.5.1 Reactor Core Isolation Cooling (RCIC) System -7.5-29 7.5.1.5.1.5.2 Residual Heat Removal (RHR) System 7.5-30 7.5.1.5.1.5.3 Nuclear Boiler and Control Rod Drive System 7.5-31 7.5.1.5.1.5.4 Recirculation Flow Control System 7.5-33 7.5.1.5.1.5.5 Balance of Flant Systems 7.5-33 7.5.1.6 Safety Parameter Display System 7.5-32 7.5.2 ANALYSIS 7.5-33

'~' 7.5-33 715.2.1 General 7.5.2.2 Normal Operation 7.5-33 7.5.2.3 Abnormal Transient Occurrences 7.5-34 7.5.2.3.1 Shutdown and Isolation 7.5-34 7.5.2.4 Accident Conditions 7.5-34 7.5.2.4.1 Initial Accident Event 7.5-34 7.5.2.4.2 Post-Accident Tracking 7.5-34 I

7.5.2.4.2.1 Reactor Water Level and Pressure 7.5-34 7.5.2.4.2.2 Emergency Core Cooling 7.5-34 7.5.2.4.2.3 Containment and Reactor Vessel Isolation I control System 7.5-35 7.5.2.4.2.4 Standby Gas Treatment System 7.5-35 7.5.2.4.2.5 ECCS Area Filtered Exhaust System 7.5-37 7.5.2.4.2.6 Standby Power System 7.5-37

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xxi Am. Nn. 58, (5/81)

ACNGS-PSAR-LIST OF TABLES (CONT' D) e Table Title Page f

\

'- ') 7.3-17 ECCS Area Fan Coolers FMEA 7.3-187 7.3-18 Containment Vacuum Relief System FMEA 7.3-188 7.3-19 Control Room Alarms for Conditions that Render the 7.3-190 43(U)

Diesel Generator Unable to Respond to an Emergency Aeto Start 7.4-1 Reactor Core Isolation Cooling Instrument Specification 7.4-23 7.4-2 Reactor Shutdown Cooling Bypasses _and Interlocks 7.4-24 7.5-0 Post-Accident Monitoring Instrumentation- 7.5-44a l58 t 7.5-1 Containment and Reactor Vessel Isolation Control 7.5-45 System Control Room I & C I 7.5-2 Standby Gas Treatment System Control Room I & C 7.5-48 7.5-3 ECCS Area Filtered Exhaust System Control Room I & C 7.5-50 7.5-4 Standby Power System Control Room I & C 7.5-52 7.5-5 Essential Services Cooling W' ter a System Control Room I & C 7.5-54 7.5-6 Control Room Air Conditioning Control Room I & C 7.5-55 7.5-7 ECCS Area Fan Coolers Control Room I & C 7.5-61 7.5-8 Panel Arrangement for Nucienet 1000 Control Console 7.5-63

' 7.5-9 Specific Regulatory Design Requirements 7.5-69 7.5-10 Containment Vacuum Relief System 7.5-70 7.5-11 Nuclenet Control Panel Inserts 7.5-71 4

7.5-12 Standby Information Panel Inserts 7.5-72

, 7.5-13 Reactor Core Cooling Benchboard Inserts 7.5-73 7.6-1 Refueling Interlock Effectiveness 7.6-69 7.6-2 Process Radiation Monitoring Systems Characteristics 7.6-70 7.6-3 SRM System Trips 7.6-71 7.6-4 IRM Trips 7.6-72 7.6-5 LPRM System Trips 7.6-73 7.6-6 APRM System Trips 7.6-74 7.6-7 Reactor Whter Cleanup Annunciators 7.6-75 7.7-1 Rod Control and Information System Instrument Specifications 7.7-39

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ACNGS-PSAR O EFFECT1'JE FIGURES LISTING l CHAPTER 9 AUXILI ARY SYSTEMS All figures, whether labelled " Unit 1" or " Units 1 and 2," are to be considered applicable to Unit No. 1 i Figure No. Amendment No. l 9.1-1 -

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9.2-13 58 9.2-14 58 9 Am. No. 58, (5/81)

ACNGS-PSAR The submerged evaporative pond is located to the south of the plant area with a diversion dike perpendicular to the 4,800 acre cooling lake shore.

/N The submerged pond serves as the UHS. The normal source of essential

( . services cooling water is the 4,800 acre cooling lake since water will be

supplied from this heat sink whenever it is available. When using the 4,800 acre lake as the heat sink for essential services cooling the essen-tial services, cooling water will be discharged from a seismic Category I structure located at the lake edge south of the UHS. The 50 acre submerged pond will perform its function as the Ultimate Heat Sink in the unlikely event of a loss of water from the 4,800 acre cooling lake. Switchover from the normal ESCWS mode of operation to the UHS mode of operation will be 37(U) manual.

A minimum level of 8 ft will be maintained in the 4,800 acre cooling lake at all times by creek inflows and by pumping from the Brazos River. In the event of a total loss of cooling water in the lake, the submerged pond will be more than adequate to permit emergency shutdown and cooldown for 4 months or in the event of an accident, to permit control of the accident for 4 months.

The intak2 is situated approximately 500 feet from the lake shore. The in-take canal starts at the UHS bottom elevation 92 feet sloping downward con-tinuously to the foot of the sill on UHS intake structure at elevation 58 86 feet. This corresponds to a 6 foot drop over a 139 foot horizontal run (approximately a 1:23 slope). The one foot sill will be provided as a barrier to limit the amount of silt which might enter UHS intake structure.

A causeway will provide access from the plant area to the intake structure. l 37(U) j The_ crest of the UHS Causeway is established at EL 145.5 ft. The Causeway

.,m,/ is designed for the Brazos River Probable Maximum Flood (PMF) at EL 135.4 ft, coupled with a wind set up of 0.7 ft and a wave run up of 8.2 ft generated by a 52 MPH Wind Velocity overwater (see PSAR Section 2.4 Appendix D, Table D4)

This design flood consideration conservatively yields a design flood elevation of EL 144.3 ft, which will provide an adequate safe margin for the UHS Causeway.

j Ihe Causeway slopes are protected against wave action by the placement of two (2) foot thick layers of soil-cement, measured perpendicular to the slope. The causeway slope stability analysis is described in Section 2.5 Appendix M Subsection M8. The design and placement of soil-cement will 58 be in accordance with the ACNGS specifications, and subject to a strict construction quality control program.

9.2.5.3 Safety Evaluation 9.2.5.3.1 Consumptive Use 9.2.5.3.1.1 Methodology 1

In order to select an UHS design and to evaluate its expected performance, the following process has been used:

26 I. Selection of Design Basis Accident (Section 9.2.5.3.1.2)

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9.2-15 (U)-Update Am. No. 58, (5/81)

ACNGS-PSAR

11. Selection of Design Meteorological Conditions (Section (N 9.2.5.3.1.3)

V)

(

a.

b.

Maximum water temperature conditions Maximum evaporation rate conditions 26 111. Simulation of the Design Basis Accident (Section 9.2.5.3.1.4) l37(C),

a. Determination of the maximum Essential Services Cooling Water System (ESCWS) intake temperature.

b. Determination of the adequacy of a submerged 50 acre pond l37(D) to provide four months supply of ESCWS water.

9.2.5.3.1.2 Selection of Design Basis Accident The Ultimate Heat Sink was evaluated for the following events: 26

1. Safe shutdown i

2. LOCA 37(U)

'lhese events were assumed to occur coincidently with the loss of the cool-ing lake.

A comparison ot these alternatives is presented in the following:

'lable 9.2-5 indicates that the maximum instantaneous heat rejected to the

,O Essential Services Cooling kater System for safe shutdown conditions with-out Fuel Fool Cooling is 286 x 106 Btu /hr. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of cooldown (D) operation, the total heat rejection would be reduced from 286 x 106 btu /hr to approximately 100 x 106 btu /hr.

F igure 9.2-11 presents the ins tantaneous heat rejection rate to the ESCkS as a function of time for safe shutdown. l 37(U)

Iable 9.2-6 indicates that the maximum instantaneous heat rejected to the Essential Services Cooling kater System during and following a Design Basis LOCA is 230 x 106 Btu /hr without Fuel Pool Cooling.

i l37(D)

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i 9.2-15a (C)-Consistency (D)-Design (U)-Update Am. No. 58, (5/81)

b ACNGS-PSAR ,

l 58 C

g Under emergency shutdown conditions with the 4,800 acre lake available, the essential services cooling water discharge will be at the seismic Category I l37(D) discharge structure located to the south of the UHS.

t Access to the UHS Intake Structure is via a manmade earth causeway extending 37(D) f i

from the main plant area and abutting the UHS Intake Structure. The causeway will be designed to be stable following a Safe Shutdown Earthquake or any l other severe natural phenomena. 58 i I

Buried piping and conduit leading from the UHS Intake Structure, the UHS discharge structure and the ESCWS discharge structure will be designed to withstand the effects of the SSE and, in addition, will be designed t 37(C)

! withstand a dif ferential ground settlement of 3 inches per 1000 feet. 1 An investigation was conducted to evaluate the potential sediment build up in l l the Ultimate Heat Sink. This supposition being that if an unusually large I sediment deposit is postulated and is allowed to go uncorrected that it might  ;

adversely affect the cooling capacity of the URS for plant shutdown. l t  ;

The principal sources of sediment deposits are identified as Allens Creek and  !

the makeup water for the Cooling Lake derived from Brazos River after prior

passage through a 190 acre sedimentation basin. In order for sediment to '

enter the Essential Service Cooling Water System pumps it must passover the i one f oot sill and then pass through trash bars and traveling water screens.

The UHSIS has a flat floor at elevation 86 feet and the pumps are set back approximately 60 feet from the sill. The ptsap suction bells are 53 inches of f f the floor. This arrangement eliminates the possibility of degrading the pumps d

by the provision of suf ficient clearance around pump suction areas. l Based upon the sediment deposit analysis and conservative engineering judgement, the sediments would most likely unitormly distribute over the

entire cooling lake including the UHS. For sucn a uniform distribution, the  :

I study indicates that the total accumulated sediments in the UHS will most likely be less than one-half inch for 40 years of plant life. This is well 58 within the 1 foot maximum deposition considered in the plant design.  ;

A 1 foot sediment deposit will have no significant effect on the total  ;

inventory of required cooling water for an emergency cool down for 4 months, l in the event of a total loss of the Cooling Lake. Since the total consumptive l

< use for an emergency cooldown, including seepage and evaporation, will be in

the order of 4 feet of a total of 8 feet water depth from a net storage of 404 l acre-ft UHS.

Consideration was given to any slumping ef fect from the sediment deposit along the Heat Sink slopes. It has been recognized by the U.S. Army Corps of Engineers that for a slope having a 1 vertical to 20 horizontal or flatter, j sediment will remain stable and no slumping or sliding effect will take place. Per PSAR Section 9.2.5.2 the intake canal will have 1:23 slope. As a I

9.2-18 (C)-Consistency (D)-Design

( Am. No. 58, (5/81) i

. - , , - , . __ _ ,_ _ _ _ ,,_,,._,__=m, _ _ , _

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ACNGS-PSAR m

, - (' - preventive measure, '.rovisions will be made to flatten the upper 1 foot of the  !

UHS side slopes and .he UHS intake forebay car.11 bottom floor to 1 vertical

-and 20 horizontal or flatter.

In addition, in order to safeguard all the above, a monitoring program will be ,

designed and implemented to check on any potential sediment accumulation in the UHS. The sediment monitoring program will be comprised of:

i (1) Establishing monument plates on the UHS fluor. These plates will be located to provide representative indication of sediment deposition on flat and sloping surfaces in the Ultimate Heat Sink.

1 (2) Probing will be conducted once per quarter for the first year of {

plant operation, and at least annually thereafter or until the l monitoring program results indicate probing at a lesser frequency would not cause a safety hazard. 58 .

(3) Probing operation will be performed in conjunction with the actual t visual observation.  ;

I (4) Sediment removal frequency will be determined as required according to the actual monitoring results during plant operation. Howeve r, ,

the upper bound of the allowable limit of sediment build up will not exceed 1 foot on the bottom floor within the confinement of the  ;

Ultimate Heat Sink in the vicinity of the embankment and 6 inches in l

, [ \ the intake canal proper.

'; \v/

In conclusion, based upon the sediment analysis results, the conservative  ;

considerations for the design and the sediment monitoring program, sediment i deposition will not contribute any adverse ef fect en the functional capability of the Ultimate Heat Sink.

9.2.5.3.3 Capability to Withstand a Single Failure ,

The UHS is designed to withstand the single failure of a manmade earthen  !

structure or a single f ailure of any UHS active component. -

In the event of f ailure of the cooling lake dam and complete loss of lake water, essential services cooling water will be drawn from the submerged pond. l37(D)j i l

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9.2-18oa (D)-Design Am. No. 58, (5/81) t

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4,800 ACRE COOLING LAKE k

o W

t COOLING LAKE _

DIVERSION DIKE N -= l ULTIMATE HEAT SINK INTAKE STRUCTURE s

CIRC. WATER INTAKE STRUCTURE s

O I

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v is l

ACNGS - PSAR

~

FLOW

'Lf !l??

i i i ;? i!;!,! ils!j,j,i FOR EASE OF VIEWING, THE SUPPLY LINE ACTUALLY REPRESENTS FOUR SERVICE WATER LINES. AN RHR "A", AN RHR "B",

AN HPCS, AND AN FPCS SERVICE WATER-LIN E.

/ THE DISCHARGE LINE ACTUALLY REPRESENTS p THREE SERVICE WATER LINES, AN RHR "A",

AN RHR "B" AND AN FPCS SERVICE WATER-g g LIN E.

e S

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%4 (l ESSENTIAL SERVICE WATER ,

DISCHARGE STRUCTURE i

P r ,

l AM. NO. 58, (5/81)

HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station g Unit 1 ULTIMATE HE AT SINK SCHEMATIC  :

(SHEET 1)

FIGURE 9.2-13

ACNGS - PSAR ULTittATE HEAT SINK - EXCAVATION 4,000 ACRE COOLING LAKE F r t GRADE EL 100* itTYP.)  %,

j I

I W

5 4

5 i

a N

i E

j , ,

TOP OF DIKE 9

- /

EL 103' 3

0 ULT. HEAT SINK f '

DIVERSION DIKE 8

\ 5 d

8 O ULTIMATE HEAT W ULTIMATE HEAT SINK INTAKE STRUCTURE SINK DISCH.

RETAINING WING WALL

< \ . u ,u la 2

\ // l -l N

/, ) ,

=

=.

TO UN T 1 4 FROM UNIT 1  ;

AM. NO. SS, (5/31) f .

HOUSTON LIGHTlHG 8 i;0WER COMPANY Allens Creek Huclear - snerating Station -

Unit 1 ULTIMATE HEAT SINK SCHEMATIC (SHEET 2)

FIGURE 9.214

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ACNGS-PSAR (v ) LIST OF TABLES Table Titic Page 12.2-2 Containment Inhalation Dose 12.2-16 12.2-3 Radweste Building Inhalation Dose 12.2-17 12.2-4 Maximum Annual Inhalation Dose to Plant Personnel 12.2-18 12.2-5 Airborne Radiation Monitoring System Characteristics 12.2-19 12.2-Sa Accident Release Monitoring Points 12.2-19a 12.2-6 Steam and Liquid Leaks for Determining Airborne Activity and Ventilation Release 12.2-20 12.2-7 Noble Gas /Off-Gas Release Comparison of Calculational Model 12.2-21 12.2-8 Radionuclide Reactor Water Concentration

• Comparison of Calculational Model With 1971 GE Source Terms and Monticello 12.2-22 12.2-9 Pressure Transient Comparison of Calculational Model With Monticello Measurements 12.2-23 12.2-10 Suppression Pool Partitioning 12.2-24 (N_,)~xi 12.2-11 Summary of Activity in the Containment Building With Time After Isolation Scram 12.2-25 12.2-12 Source Strength in Containment Building Ventilation Air With Time Af ter Isolation Scram Summation of Gamula Energies 12.2-27 b

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v Section Title Page 13.0 ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1-1 13.1 ORGANIZATIONAL STRUCTURE OF APPLICANT 13.1-1 13.1.1 MANAGEMSNT AND TECHNICAL SUPPORT ORGANIZATION 13.1-1 13.1.1.1 Design and Operating Responsibilities 13.1-1 13.1.1.2 Organizational Arrangements 13.1-5 13.1.1.3 Qualifications 13.1-8 33(U) p 13.1.1.4 Qualifications of Corporate Personnel 13.1-8 ,

l 13.1.1.5 Project Management Organization 13.1-12 13.1.2 OPERATING ORGANIZATION 13.1-43 13.1.2.1 Plant Organization 13.1-43 ,

13.1.2.2 Personnel Functions, Responsibilities and <

Authorities 13.1-43

\' - 13.1.2.3 Shift Crew Composition 13.1-48 13.1.3 QUALIFICATION REQUIREMENTS FOR NUCLEAR PLANT PERSONN EL 13.1-48 13.1.3.1 Minimum Qualification Requirements 13.1-48 13.1.3.2 Qualifications of Plant Personnel 13.1-48 '

APPENDIX 13.lA HL&P ACNGS PERSONNEL RESUMES 13.lA-1 33(U) 13.2 TRAINING PROGRAM 13.2-1 13.2.1 PROGRAM DESCRIPTION 13.2-1 13.2.1.1 Program Content 13.2-1 ,

13.2.1.2 Coordination with Preoperational Tests and [

Fuel Loading 13.2-3 l l

13.2.1.3 Practical Reactor Operation 13.2-3 33(U) 13.2.1.4 Reactor Simulation Training 13.2-3 r

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,m APPENDIX C (J) APPLICANT'S REGULATORY CUIDE POSITIONS 17 TABLE OF CONTEhTS (Cont'd) humber Title 1.46 Protection Against Pipe Whip Inside Containment (Rev. O, 5/73) l 35(C) 1.47 bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems (Rev. O, 5/73) l 35(C) 1.46 Design Limits ano Loading Combinations for Seismic Category I Fluid System Components (Rev. O, 5/73) l 35(C) 1.49 Power Levels of Water-Cooled huclear Power Plants (Rev 1,12/73) l 35(C) 1.50 Control of Preheat Temperature for Welding of Low-Alloy Steel (Rev. O, 5/73) l 35(C) 1.51 Inservice Inspection of AShE Code Class 2 and 3 Nuclear Power Plant Cemponents (withdrawn) l 35(C) i 1.52 Design, Testing and Maintenance Criteria for Engineered-Safety - 35(C)

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l.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (Rev. O, 6/73) l 35(C) 1.54 Quality Assurance Requirements for Protection Coatings Applied to "

Water-Cooled huclear Power Plants (Rev. O, 6/73) 1.55 Concrete Placement in Category 1 Str tures (Rev. O, 6/73) 35 (C) l 1,56 Maintenance of Water Purity in Boiling kater Reactors (Rev. O, 6/73) 1.57 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components (Rev. O, 6/73) l35(C) ,

1.5E Qualification of huclear Power Plant Inspection, Examination and Testing Personnel (Rev. O, 8/73) l 35(C) 1.59 Design Basis Floods for Nuclear Power Plants (Rev. 2, 8/77) l 42(U) 1.60 Design hesponse Spectra for Seismic Design of Nuclear Power Plants (Revision 1, 12/73) 1.61 Damping Values for Seitnic Design of Nuclear Power Plant s l35(C) p)

\v 1.62 (Rev. O, 10/73)

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DESIGN, TESTING AND MAINTENANCE CRITERIA FOR ENGINEERED-SAFETY-FEATURE AIMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT- (

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