ML19208A971

Psar,Amend 53

ML19208A971
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	09/17/1979
From:	Eric Turner HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML19208A968	List: ML19208A967 ML19208A971
References
NUDOCS 7909180368
Download: ML19208A971 (38)

Text

Before the UNITED STATES NUCLEAR REGULATORY COMMISSION Docket No. 50-466 Allens Creek Nuclear Generating Station Unit 1 Amendment 53 to the PSAR llouston Lighting 6 Power Company, applicant in the above captioned proceeding, hereby files Amendment 53 to the Preliminary Safety Analysis Report filed in connection with its application.

Amendment 53 consists of additional PSAR information updating the PSAR since the issuance of the Allens Creek Safety Evaluation Report, Supplement 2.

Respectfully submitted 110U STON L I Gilt IN G S POh'ER COMPANY O

o/ %

E. A. Turner Vice President Power Plant Construction 6 Technical Services 00U003.

790918o3eg

STATE OF TEXAS COUNTY OF liARRIS E. A. TURNER, being first duly sworn, deposes and says:

That he is Vice President of liOUSTON LIGilTING G POWER COMPANY, an Applicant herein; that the foregoing emendment to the application has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said documents and the facts contained therein are true and correct.

DATED: This[f day - , 1979.

V Signed: O h E. A. Turner Subscribed and sworn bef re me this / 7 ~

day of ' , 1979.

U C-iAE.- ajasawn NOTtary Public in and for the

~

County of liarris, State of Texas My commission expires:

V-SO'S) 360034

Houston j i[ Lighting

i & PcnAner company Electric Tower PO Box 1700 m Houston. Texas 77001 September 17, 1979 A C -IIL - A E - 3 5 7 Mr. liarold R. Denton, Director Office of Nuclear Reactor Regulation Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

Allens Creek Nuclear Generating Station Unit 1 Docket No. 50-466 Amendment 53 Please find under separate cover sixty (60) copies of Amendment 53 to the Houston Lighting 6 Power Company Allens Creek Nuclear Generating Station Unit 1 PSAR.

A copy of this transmittal letter is attached to each amendment copy.

Amendment 53 consists of additional PSAR information updating the PSAR since the issuance of the Allens Creek Safety Evaluation Report, Supplement 2.

Very truly yours,

\sq - rp O L. / h E. A. Turner Vice President Power Plant Construc. ion 6 Technical Servic s LDR/ngb cc: J. G. Copeland (Baker 6 Botts)

R. Gordon Gooch (Baker G Botts)

J. R. Newman (Lowenstein, Newman, Reis, Axelrad S Toll)

P. A. llo r n All Parties ISOOhbab

Before the UNITED STATES NUCLEAR REGULATORY COMMISSION Docket No. 50-466 Allens Creek Nuclear Generating Station Unit 1 Amendment 53 to the PSAR Houston Lighting 6 Power Company, applicant in the above captioned proceeding, hereby files Amendment 53 to the Preliminary Safety Analysis Report filed in connection with its application.

Amendment 53 consists of additional PSAR information updating the PSAR since the issuance of the Allens Creek Safety Evaluation Report, Supplement 2.

Respectfully submitted 110V ST ON L I GliT I N G 6 POWER COMPANY CL & ' :s_

E. A. Turner Vice President Power Plant Construction S Technical Services 360036

STATE OF TEXAS COUNT) OF flARRIS E. A. TURNER, being first duly sworn, deposes and says:

That he is Vice President of liOUSTON LIGliTING f POWER C05fPANY, an Applicant herein; that the foregoing amendment to the application has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said documents and the facts contained therein are true and correct.

DATED: This /9 M day . , 1979.

i Signed: -

CL /

E. A. Turner Subscribed and sworn to before me this / 7-// day of hjfL M , 1979.

g-j' ',_ ,'?

bg Notary Public in and for the County of llarris, State of Texas h!y commission ex ires:

U~_86- ?

360097

ACNGS-PSAR Remove Insert (Existing Pages) Amendment 53 Pages Chapter 1 1* 1*

2a* 2a*

1.3-14 1-3-14 Chapter 3 1* 1*

4* 4*

3.1-55 3.1-55 3.2-28 3.2-28 Chapter 6 1* 1*

4* 4*

10* 10*

6.2-73 6.2-73 6.2-73a 6.2-77 6.2-77 6.2-78 6.2-78 Fig. 6.2-28a Fig. 6.2-28a Fig. 6.2-29 Fig. 6.2-29 Chapter 9 1* 1*

2* 2*

9* 9*

i i xvii xvii 9.1-3 9.1-3 9.1-3a 9.1-3a (Deleted) 9.1-4 9.1-4 9.1-5 9.1-5 9.1-Sa 9.1-Sa 9.1-6 9.1-6 (Deleted) 9.1-6a 9.1-6a 9.1-6b Fig. 9.1-2 9.1-2 (Deleted)

Fig. 9.1-2a Fig. 9.1-12a Fig. 9.1-12a (Deleted)

Fig. 9.1-12b Fig. 9.1-12b (Deleted)

Fig. 9.1-12c Fig. 9.1-12c (Deleted) 360098 Am. No. 53, 9/17/79

ACNGS-PSAR EFFECTIVE PAGES LISTING CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Page Amendment 1* 53 2* 46 2a* 53 3* 37 4* 39 i 37 ii 33 iii 7 iv 37 v 37 vi 37 v11 37 1.1-1 37 1.1-2 33 1.1-3 33 1.1-4 33 1.1-5 33 1.1-6 33 1.1-7 33 1.1-8 33 1.1-9 33 1.1-10 s3 1.2-1 39 1.2-2 39 1.2-3 37 1.2-3a 37 1.2-4 -

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• Effective Pages/ Figures Listings SG0099 1 Am. Nc. 53, 9/17/79

ACNG S- PS AR ACNCS-PSAR EFFECTIVE PAGES LISTING (Cont'd)

CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Pace A. :e n d me n t 1.3-1 -

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1.6-7 33 360100 1.7-1 -

2a Am . ' o . 53, 9/17/79

ACNCS-PSAR TABLE 1.3-5 CO'cARISON OF STRUCTUPAL DESIGN REQUIFD'ENTS Allens Creek Susquehanna Bailly Limeric Seismic Design Reference PSAR Section 3.7 PSAR Appendix C 2.5.3 2.5.3 50% Safe Shutdown Eartnquake (horizontal g) 0.05 0.05* 0.10* 0.06*

Safe Shutdown Earthquake **

(horizontal g) 0.10 0.20 0.12 Earthquake vertical shock

(% of horizontal) 67 60 67 66 b'ind Design Reference PSAR PSAR Appendix C Section 3.3 Section C.2 12.2 2.3 Maxir.ur sustained wind (cph) 157*** 80 90 90 Tornadoes ( ph) 290 rotational 300 tang. 300

+70 trans. 300 + 60 trans. 53 (C) 360 tangential

• previously called Operating Basis Zarthquake

• ' previcusly called Design Basis Earthquake

• ' <

• i minute average at 30 foot level 360101 1.3-14 An. No. 53, 9/17/79 (C) Consistency

ACNGS-PSAR .

EFFECTIVE PAGES LISTING CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EOUIPMENT AND SYSTEMS PAGE AM ENDM ENT 1* 53 la* 48 2* 39 3* 53 4* 53 5* 49 6* 48 7* 49 8* 50 9* 46 10* 45 11* 49 12* 48 13* 42 14* 47 15* 48 16* 44 16a* 39 17* 50 18* 48 i 35 ii 35 iii 35 iv 35 v 35 vi 35 vii 35 viii 35 ix 35 x 35 xi 35 xii 35 xiii 37 xiv 35 xv 35 xvi 44 xvii 44 xviii 44 xix 48 xx 35 xxi 44 xxii 35 xxiii 35 xxiv 35 xxv 35

• Effective Pagas/ Figures Listing 3601.02 I Am. .No. 53, 9/17/79

ACNGS-PSAR EFFECTIVE PACES LISTING CHAPTER 3 DESIGN OF STRUCTURES, COMPOSENTS, EQl'IPMEST AND SYSTEMS Pace Ame ndme n t 3.1-24 35 3.1-25 35 3.1-26 35 3.1-27 35 3.1-28 35 3.1-29 35 3.1-30 35 3.1-31 35 3.1-32 -

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:1  : 36010a 3.2-13 21 3 Am. No. 53, 9/17/79

ACNGS-PSAR EFFECTIVE PACES LISTING CllAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS Page No. Anendment No.

3.2-14 35 3.2-14a 35 3.2-14b 35 3.2-14c 35 3.2-15 35 3.2-16 35 3.2-17 3.2-18 /

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a, 3.2-51 47 3.2-52 360104 4 Am. No. 53 9/17/79

ACNGS-PSAR Spent fuel is stored under water in the Spent Fuel Pool. The racks in which spent fuel assemblies are placed are designed and arranged to ensure suberiticality in the storage pool. Spent fuel is maintained at a sub-critical multiplication factor Keff of less than or equal to 0.95 for all h3(D) conditions.

Refueling interlocks include circuitry which senses conditions of the re-fueling equipment and the control rods. These interlocks reinforce operational procedures that prohibit making the reactor critical. The Fuel Handling Syst2m is desicned to provide a safe, effective means of trans-porting and handling fuel and is designed to minimize the possibility of mishandling or maloperation.

The use of geometrically safe configurations for new and spent fuel storage and the design of fuel handling systems precludes accidental criticality in accord with Criterion 62.

For further discussion, see the following sections:

a. All other Instrumentation Systems Required for Safety 7.9

b. Fuel Storage and Handling 9.1 3.1.2.6.4 Criterion 63 - Monitoring Fuel and Waste Storage Appropriate systems shall be provided in fuel storace and radioactive waste systems and associated handling areas, (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels, and (2) to initiate appropriate safety actions.

3.1.2.6.4.1 Evaluation Acainst Criterion 63 Appropriate systems will be provided to meet the requirements of this criterion. A malfunction of the Fuel Pool Cooline and Cleanup System which could result in loss of residual heat removal capability and excessive radiation levels is alarmed in the Control Room. Alarmed conditions include high/ low fuel pool cooling water pump discharge pressure and high/ low level in the f uel s tora ge pool and drain tanks. System tem-perature is also continuously monitored in the Control Room. Radiation Monitor continuously monitors radioactivity in this area and initiates 35(C) an alarm in the Control Room on abnormal radiation levels.

Radiation and tank and sump levels are monitored and alarmed to give indication of conditions which may result in excessive radiation levels in radioactive waste system areas. These systems satisfy the requirements of Criterion 63.

(C) Consistency (D) Design 3.1-55 Am.No. 53 9/17/79 360105

T ABl! 3.2-1 (Cont'd)

INVIRON*ftNTAL CAPABII.ITY I I Nh I'U Quality OI Scorc of Safety I'I Quality remponent I'I Seismic III ratreme(EI Tornado /

I c icM II Assuranar Francipal Ceepenent Supply Class broup locataon rat ego ry hand Massaic Protection Program Comen t s

3. fuel flandlang Building, i n c l ud a r.g :

Base Slab P 3 N\ M i b b a B St ructural ha lls P 3 N% R I a a b B l 22 Structural floors P 3 Nt R I b b c B Spent-fuel pool P 2 NA R I b b c B luel cask storage pool P 3 NA R I b b c B 5 pent fuel-storage racks P 3 NA R I b b c B Temporary f uel-stor.1ge racks 01 N% I b b c B 53

% C fuel speciality racks (J 3 NA R I b b c B N luel llandlang Crane 00 3 NA R I b b c B All Sessmic Category I

• w ,,

,, equipment support s P 3 Nt R (u) b b c B s e v>

4. Control Building  %

Base Stab P 2 Nt M  ! b b a B St ructural walls P 2 NA B I a a b B l22 St ruc t ura l th> ors P 2 Nt B I b b c B All Seismic Category I equipment support s P 1,2,3 NA B I b b c B (v)

Sk 5. Diesel Generator Building yi 9 Base Slab P 3 Nt M I b b a B l72

{, Structural halls Structural Floors P

P 3

3 Nt NA S

S I

I a

b a

b b

c B

B a

.," All Seasmic Category I g3 D equipment support s P 3 N4 S I b b c B U

M us .

W C

H C

C

ACNGS- PSAR EFFECTIVE PAGE LISTING CHAPTER 6 ENGINEERED SAFETY FEATURES Pace Amendment 1* 53 2* 40 3* 46 4* 53 5* 47 6* 46 7* 46 8* 46 9* 46 10* 53 11* 46 12* 46 1 37 11 37 iii 46 iv 37 v 37 vi 45 vii 37

/ viii 37 ix 37 x 37 xi 37 xii 37 xiii 37 xiv 37 xv 42 xvi 42 xvii 37 xviii 37

• Ef fective Pages/ Figures Listings 1 Am. No. 53, 9/17/79 300107

ACNGS-PSAR EFFECTIVE PAGE LISTING CilAPTER 6 ENGINEERED SAFETY FEATURES Page Amendment No.

6.2-51 40 6.2-51a 10 6.2-52 -

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6.2-54 37 6.2-55 46 6.2-55a 37 6.2-56 -

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6.2-70a 23 6.2-71 23 6.2-71a 17 6.2-71b 8 6.2-72 28 6.2-72a 8 6.2-73 53 6.2-73a 53 6.2-74 37 6.2-75 46 6.2-76 46 6.2-77 53 6.2-77a 37 6.2-78 53 6.2-78a 37 6.2-78b 37 6.2-78c 37 6.2-78d 37 6.2-78e 39 6.2-78f 37 6.2-78f(i) 39 6.2-78g 39 6.2-78h 39 6.2-79 39 6.2-80 37 6.2-80a 23 E0108 4 Am. No. 53, 9/17/79

ACNGS-PSAR EFFECTIVE FIGURES LISTING CFJ.PTER 6 Figure No. Amendment No.

6.2-27b 26 6.2-28 -

6.2-28a 53 6.2-29 53 6.2-30 (deleted) 37 6.2-31 (deleted) 37 6.2-32 (deleted) 37 6.2-33 23 6.2-34 23 6.2-35 37 6.2-36 5 6.2-37 5 6.2-38 5 6.2-39 5 6.2-40 5 6.2-41 5 6.2-42 5 6.2-43 5 6.2-44 37 6.2-45 5 6.2-46 23 6.2-47 37 6.2-48 5 6.2-49 23 6.2-50 23 6.2-51 23 6.2-52 37 6.2-53 5 6.2-54 5 6.2-55 5 6.2-56 5 6.2-57 5 6.2-58 5 6.2-59 5 6.2-60 8 6.2-61 8 6.2-62 21 6.2-63 37 6.2-64 (deleted) 37 6.2-65 37 6.2-66 23 6.2-67 30 6.2-68 30 360109 Am. No. 53, 9/17/79

ACNGS-PSAR 6.2.5.1 Design Bases The following are the criteria used to design the Combustible Gas Control System:

a) The system will be designed in accordance with Regulatory Guide 1.7 Revision 1, September 1976 and General Design Criterion 41 of 10 CFR 26 50 Appendix A.

b) The hydrogen resulting from metal-water reaction is assumed to be the larger of the amounts which would evolve based on two criteria i) A core wide reaction of the cladding to a depth of 0.00023 inch ii) Five times the percentage of total cladding mass reacted based on ECCS evaluation In either case, the hydrogen is assumed to evolve in the first two minutes following the postulated LOCA.

c) The system will have the capability of sampling and measuring the hydrogen concentration throughout the drywell and containment during all modes of operation.

d) The system will have the capability of mixing the atmosphere in the containment and drywell following a LOCA.

e) The system will have the capability of controlling combustible gas concentrations in the containment atmosphere without reliance on purging and without the release of radioactive material to the environment.

f) The Combustible Gas Control System and the equipment for mixing, measuring and sampling will meet the design, quality assurance, re-dundancy, energy source, and instrumentation requirements compatible with the safety of the syster g) The system will not introduce safety problems that would affect con-tainment integrity.

h) As a backup to the hydrogen recombiner subsystem of the Combustible Gas Control System, capability will be provided to control gas con-centrations in the containment by purging the containment through the Standby Gas Treatment System.

i)

Contribution of combustible gases from secondary sources such as de-composition of coating or corrosion will be included in sizing the Combustible Gas Control System and backup purge system.

360110 6.2-73 (U)-Update Am. No. 53, 9/17/79

j) Regulatory Guide 1.7 indicates that the hydrogen concentrction limit is four percent by volume if more than five percent by volume oxygen is present. Since the Containment will not be inerted (oxygen - de-ficient) the oxygen concentration will always be above 5 percent by 37(C) v uume; hence, hydrogen concentration will be the control limit.

Control measures will be taken to assure that the c.ydrogen concen-tration will not exceed four percent by volume in either the drywell or the Containment.

(C)-Cons i st ency 6 . 3 a A"' N"' 15' 9/I? 79 360111

ACNGS-PS AR c omponent s are shown in Table 6.2-25.

Th e pur ge f an discharges into the recirculation branch of the SCTS at a point downstream of the check valve, preventing an unfiltered discharge of drywell purge air to the outside.

Radioactive materials in the purge discharge airstream are in this way directed into the Shield Building annulus where mixing and holdup occur prior to treatment by the SGTS.

6.2.5.3 Design Evaluation In evalaating the Combustible Gas Control System design, it was found to be necessary to make calculations regarding:

a) Hydrogen generated in the post-LOCA environment.

b) Resultant drywell and Containment concentrations.

c) The consequent functional requirements of the Combustible Gas Con-trol System 6.2.5.3.1 Sources of Hydrogen 6.2.5.3.1.1 Short Term Hydrogen Generation In the period of immediately af ter the LOCA, hydrogen would be generated by both radiolysis and cetal-water reaction. "ewever, in evaluating short-term hydrogen generation, the contribution from radiolysis is insignificant in comparison with the hydrogen generated by a postulated one percent cetal-water reaction.

The generation of hydrogen by metal-water reaction is dependent upon the t empera t ure of the cladding and the time at temperature. Based on loss of coolant calculational procedures established by the AEC in the " Interim Acceptance Criteria for Emergency Core Cooling System," the extent of metal-water reaction in the BWR/6 core is negligible (see S'-tion 6.3 of the PSAR and CE Topical Report NEDO-10569, 11013-77 and NEDO 10723). The design of the B'w'R/6 ECCS is such that the real zirconium temperature is 1500 F; at this temperature, virtually no metal-water reaction occurs, and therefore hydrogen production by this means is insignificant.

Figure 6.2-28a shows drywell hydrogen concentration as a function of time l37(G' based on the assumptions listed below. Based on these assumptions, it is ,

calculated that the control limit (4%) would be reached in the drywell 33(U) approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the LOCA.

The transient hydrogen concentration curve shown in Figure 6.2-28a is based 37(G) upon a non-condensible air pressure of 14.7 psia and a drywell air volume reduced by water flooding up to the top of the weir wall. This qua si steady state condition is conservatively assumed to exist immediately following t he hypot hesized me tal-wa ter reaction at LOC A + 2 minu t e s . This takes no credit for steam purging of the drywell or steam dilution due to pr e s s ur e increase in t he drywell . In a more realistic sequence, the (U ) -Upda t e .

6.2-77 (G)-GESSAR SG0112 Am. No. 53, 9/17/79

ACNGS-PSAR 6.2.5.3.2 Analysis The loss-of-coolant accident is described in Section 6.3. As previously 28 noted, the occurrence of a one percent metal-water reaction cannot be re-lated to any credible physical process during a LOCA. Therefore the fol-lowing assumptions were made with respect to the sequence of events occur-ring after a LOCA.

a) As mentioned above, it is assumed that the metal-water reaction begins immediately after the LOCA and proceeds for two minutes at a rate that results in hydrogen generation equal to the larger amount determined by cne of the two methods previously mentioned. 53(U)

In this case the second method described (five times the calculated value) results in the larger value (.85*. vs .67*.) .

b) Radiolysis begins immediately and proceeds at a rate consistent with the assumptions stated in Section 6.2.5.3.1.2.

Based on these assumptions, it is calculated that the hydrogen concentra-tion in the drywell will reach four volume percent approximately eight hours after LOCA. To calculate the redistribution of hydrogen between 46(C) drywell and Containment due to liydrogen Mixing System operation, a compu-ter model of the system is used. The model predicts the volume concen-tration of each atmospheric constituent in each region as a function of time.

The calculations indicate that actuation of the Drywell flydrogen Control l 26 System causes an immediate decrease in the drywell hydrogen concentra-tion. (See Figure 6.2-29). The recirculation of the Containment atmos-phere causes the hydrogen concentration to become uniform throughout the Containment and drywell.

Eventually, radiolytic generation may cause the hydrogen to again approach four volume percent. If this were to occur the liydrogen Recombiner Sub-system would be manually activated. This system is designed to keep the hydrogen concentration below four volume percent. l37(G)

Both the mixing system and the recombiner subsystem are manually activated ,

~6 systems. Ilowever, the mixing system would not be activated until the new hydrogen monitoring instrumentation indicates that the four percent limit l of Regulatory Guide 1.7 is being approached. A realistic analysis indi- I 26 cates that the ECCS would operate to limit the amount of metal-water reaction, and therefore, the amount of metal-water reaction generated hy-drogen, to undetectabic amounts. ECCS operation will also prevent fuel rod perforations; thus, no significant fission products would be released, and there would be no significant radiolysis in the pressure Suppression Pool.

The only significant source of hydrogen, then, would be radiolysis of the coolant in the core region. Based on this realistic analysis, no action i to control hydrogen buildup would be required until weeks after the LOCA. l 37(G)

(U)-Update 6.2-78 (C)-Consistency (G)-GESSAR SGO.u3 Am. No. 53, 9/17/79

ACNGS PSAR 5

4 -

4% H2 CONCENTRATION IN DRYWELL 3 -

u i 5

c-l 2 -

I i

i i

1 l l

l l

960114 0 I II Ill I I I II II I III Il I I i!!!!!

0.01 0.1 1 10 100 TIME (HR) A51. N O . 53, 9/17/79 (U)-UPDATE HOUSTON LIGHTING & POWER COMPANY Al! ens Creek Nuclear Generating Statien Unit i HYDROGEN CONCENTRATION IN THE DRYWELL(WITHOUT BLOWER ACTIVAT ON

. 8 5 ". MET AL WATER RE ACTION FIGURE 6.2-28a

ACNGS- PSAR EFFECTIVE PAGE LISTING CHAPTER 9 AUXILIARY SYSTEMS Page No. Amendment No.

1* 53 2* 53 3* 46 4* 46 5* 47 6* 48 7* 46 8* 43 8a* 40 8b* 40 9* 53 10* 46 11* 40 1 37 ii 37 111 37 iv 37 v 46 vi 37 vii 37 viii 37 ix 37 x 46 xi 42 xii 46 xiii 42 xiv 40 xv 37 xvi 37 xvia 37 xvii 37 xviii 37 xix 40 xx 40 xxi 40 960115

• Effective Pages/ Figures Listings 1 Am. :s o . 53, 9/17/79

ACNGS - PSAR 4

ASSUMPTIONS

• 0. 359 METAL WATER REACTION

• 500 SCFM BLOWER

• BLOWER START TIME = 1 HR

• RECOMBINER START TIME = 24 HRS.

• 100 SCFM RECOMBINER 3 -

DRYWELL

[

2 2 -

u 5

s.

1 T

LCONTAINMENT I I I 0

0.1 10 100 1000 0.01 1 TIME (HR) AM. NO. 53. 9/17/79 (U)-UPDATE HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Genereting Station Unit 1 HYDROGEN CONCENTRATION IN THE DRYWELL (FOLLOwlNG BLOWER ACTIVATION)

FIGURE 6.2-29 fj(j()11 6

ACNGS-PSAR EFFECTIVE PAGE LISTING Chapter 9 AUXILIARY SYSTEMS Page No. Amendment No.

9.1-1 37 9.1-la 37 9.1-2 37 9.1-2a 37 9.1-3 53 9.1-3a 53 9.1-4 53 9.1-5 53 9.1-Sa 53 9.1-6 (Deleted) 53 9.1-6a 53 9.1-6b 53 9.1-7 48 9.1-7a 23 9.1-8 39 9.1-9 39 9.1-9a 37 9.1-10 37 9.1-10a 41 9.1-11 37 9.1-12 37 9.1-13 37 9.1-13a 37 9.1-14 37 9.1-15 37 9.1-16 37 9.1-17 41 9.1-17a 41 9.1-18 37 9.1-19 37 9.1-20 39 9.1-21 39 9.1-22 37 9.1-23 37 9.2-1 22 9.2-2 22 9.2-3 37 9.2-3a 52 9.2-4 52 9.2-5 46 9.2-5a 46 9.2-5b 48 9.2-6 37 9.2-6a 37 9.2-7 46 9.2-8 48 9 2-8 9.2-9 (amendment number not shown on page) E at SG0117 9.2-10 37 2 Am. No. 53, 9/17/79

ACNGS-PSAR EFFECTIVE FIGURES LISTING Chapter 9 AUXILIARY SYSTEMS All figures, whether labelled " Unit l' or " Units 1 and 2," are to be considered applicable to Unit No. 1.

Figure No. Amendment No.

9.1-1 -

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9.1-10 39 9.1-11 39 9.1-12a (Deleted) 53 9.1-12b (Deleted) 53 9.1-12c (Deleted) 53 9.1-13 37 9.2-1 37 9.2-la 37 9.2-1b 37 9.2-2 37 9.2-2a 37 9.2-2b 37 9.2-2c 37 9.2-3 37 9.2-4 37 9.2-5 37 9.2-6 37

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9.2-13 37 9.2-14 37 360118 9 Am. No. 53, 9/17/79

ACNGS-PSAR TABLE OF CONTENTS CHAPTER 9 AUXILIARY SYSTEMS Section Title Page 9.0 AUXILIARY SYSTEMS 9.1-1 9.1 FUEL STORAGE AND RANDLING 91-1 9.1.1 NEW FUEL STORAGE 9.1- 1 9.1.1.1 Desien Bases 9 .1- 1 9.1.1.1.1 Safety Design Bases 9.1- 1 9.1.1.1.2 Power Generation Design Bases 9.1-1 9.1.1.2 Facilities De s c r i p t i on 9.1-1 9.1.1.3 Safety Evaluation 9.1-2 9.1.1.4 Testine and Inspection o.1-2a 9.1.2 SPENT FUEL STORAGE 9.1- 3 9.1.2.1 Desien Bases 9.1- 3 9.1.2.1.1 Safety Design Bases 9.1- 3 9.1.2.1.2 Power Generation Design Bases 9.1- 3 9.1.2.2 Facilities Description o .1- 3 9.1.2.3 _Sa fety Evaluation 9.1-4 Testine and Inspection 9.1-5a 1 9.1.2.4 33 9.1.2.5 Summary of Radiolocical Considerations 9.1-5a E")

9.1.3 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM 9.1-6a 9.1.3.1 Desien Bases 9.1-6a 9.1.3.2 System Description o .1- 7 9.1.3.3 Sa fety Eva lua t i en o.1 G 9.1.3.4 Inspection and Testin_c Recuirements 9.1-10 9.1.4 FUEL HANDLING SYSTEM 9.1-10 9.1.4.1 Desien Basis 9.1-10 300119 Am. No. 53, 9/17/79 i

ACNGS-PSAR LIST OF FIGURES Figure Title 9.1-1 New Fuel Storage Rack 9.1-2 Spent Fuel Storage Rack (Deleted) 53 9.1-2a High Density Spent Fuel Storsge Racks (9) 9.1-3 Not Used 9.1-3a Flow Diagram Fuel Pool Cooling anj Cleanup System - Fuel Handling Building 9.1-3b Flow Diagram Puel Pool Cooling and Cleanup System -

Reactor Building 9.1-4 P&ID Fuel Pool Filter /Demineralizer 9.1-5 Fuel Preparation Pachine 9.1-6 Fuel Preparation Machine (showing channeling procedure) 9.1-7 Fuel Inspection Stand 9.1-8 Jib Crane 9.1-9 Re fueling Platform 9.1-10 Dry Cask Handling Isometric 9.1-11 General Arrangement Cask Handling System 9.1-12a Eccentric Positioning (Deleted) 3 9.1-12b Fuel Stored in Control Rod Racks (Deleted) Q9.1 9.1-12c Abnormal Fuel Storage Conditions (Deleted) 9.1-13 Inclined Fuel Transfer Tube 9.2-1 Not Used 9.2-la Flow Diagram Essential Service Cooling Water System (Sheet 1 of 2) 9.2-lb Flow Diagram Essential Service Cooling Water System (Sheet 2 of 2) 9.2-2 Not Used 9.2-2a Flow Diagram Circulating Water and Au).iliary Cooling Water Sys tems 9.2-2b Flow Diagram Circulating Water and Auxiliary Cooling Water Systems 360120 xvii Am No. 53, 9/17/79

ACNGS-PSAR 9.1.2 SPENT FUEL STORAGE 9.1.2.1 Design Bases 9.1.2.1.1 Safety Design Bases a) The fuel array in the fully loaded spent fuel racks shall be sub- ';3 (D) critical with a keff of less than or equal to 0.95 for all conditions.

b) Each spent fuel storage rack containing its full complement of fuel will be designed to withstand "specified loads" to minimize distor-tion of the fuel storage arrangement.

c) Shielding for the spent fuel storage arrangement will be sufficient to protect plant personnel from exposure to radiation in excess of published guideline values.

d) The spent fuel storage facility will be designed to prevent missiles generated by high winds from causing damage to the fuel.

e) The spent fuel storage facility ventilation system will be designed to limit the potential offsite exposures in the event of significant I release of radioactivity from the fuel as detected by a high radia- q9.3 tion signal. The signal will shutdown normal ventilation systems and automatically place in operation the Standby Gas Treatment Sys-tem, filtering all exhaust air prior to its release to the environ-ment.

f) The spent fuel storage racks will be designed to seismic Category I requirements.

9.1.2.1.2 Power Generation Design Bases a) Spent fuel storage space will be supplied to initially accomodate 1710 fuel bundles in high-density spent fuel storage racks (5 years normal batch discharges plus a full core reserve), with the capability to add 53(D) high-density storage racks for up to 2790 fuel bundles (10 years nor-cal batch discharges plus a full core reserve).

b) Spent fuel storage racks will be designed and arranged so that the fuel assemblies can be handled efficiently during refueling operations.

9.1.2.2 Facilities Description Spent fuel storage racks will provide a place in the Fuel Pool for storing spent fuel received from the reactor vessel. These will be top-entry racks designed to maintain the spent fuel in a space geometry that precludes the 53(D)

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possibility of criticality under all conditions.

The loaction of the spent fuel storage facilities within the station com-plex is shown in Figure 1.2-17a. Spent fuel can be stored both in the 37(9)

Reactor Building and in tne Fuel liandling Building, liow eve r , fuel will not 9.1-3 (D)-Design Am. No. 53, 9/17/79 960121

ACNGS-PSAR This page has been deleted.

360122 9.1-3a Am. No. 53, 9/17/79

ACNGS-PS AR be stored in the Reactor Building except during periods of refuelini, on 37(9) a temporary basis. The Containment Fuel Pool will have a capac;ty for storing 25 percent of a core and spent fuel storage in the FilB can accommodate 53(D) 1710 fuel bundles. (with the capability to expand to 2790 fuel bundles)

The rack arrangement is designed to prevent accidental insertion of fuel bundles between adjacent racks. The storage rack structure is so designed that the upper tic plate casting cannot be lowered below the top of the _,

upper rack. This prevents any tendency of the fuel bundles jamming on inser- 3'(D) tion or removal from the rack. The rack spacing is such as to maintain mini-mum spacing of adj acent racks for geometric reactivity control.

9.1.2.3 Safety Evaluation The design of the FilB high-density spent fuel storage racks will provide for a suberitical multiplication factor (Leff) of 0.95 or less for both normal or ab- ;53 (D) normal storage conditions. Normal conditions exist when the fuel storage racks are located in the pool and are covered with a normal depth of water (about 25 feet above the stored fuel) for radiation shielding and with the maximum number of fuel assemblies or bundles in their design storage position. The spent fuel will be covered with water at all times by a minimum depth required to provide 33(p) sufficient shielding. An abnormal condition may result from accidental dropping of equipment without first disengaging the fuel from the hoisting equipment.

To ensure that the design criteria are met, the following normal and abnormal spent fuel storage conditions are analy:cd:

a) Closest eccentric fuel positioning in the spent fuel storage array b) Pool water temperature increases to 212 F.

53 (D) c) Moving fuel bundle in regions peripheral to the storag=: rack array d) Fuel bundle drop on or through a rack Objects such as fuel handling fixtures that could ccnceivably fall into the fuel would not transfer energy amounts exceeding the specified limits of the fuel racks. The design of the Fuel Handling Building does not require that the spent fuel shipping cask be lif ted to the operating floor level.

This design precludes any movement of the cask over the storage pools.

Any connections to the Spent Fuel Pool will be designed such that the watertight integrity of the pool is not compromised (See Section 9.1.3.3).

An appropriate air handling system with radiation sensing instrumentation in the exhaust ducts will be provided to adequately control radiation leakage to the environment (See Section 9.4.5).

In case of a significantly high radiation level, the exhaust ducts will close, and the Standby Gas Treatment System (Section 6.2.3) will be activ-ated.

.. Mio.Ur2 (D) Design 9.1-4 Am.No.53, 9/17/79

ACNGS-PSAR For the foregoing analysis, it is concluded that the spent fuel storage arrangement meets its derign bases and satisfies the requirements of ReguJatory Guide 1.13.

Offsite exposure to release of radioactive products from damaged or failed fuel in the fuel building is dependent on three systems. The functions of these systems as described in Section 9.4.5 are:

a) The ventilation exhaust radiation monitoring system detects radio-activity in the fuel building atmosphere b) The Standby Gas Treatment System minimizes the release of contani-nated air to the environment c) The fuel building isolation control system automatically closes isolation dampers to block potential leakage of contaminat id air to the environment d) The fuel storage facilities will be designed to seismic Category I requirements to prevent earthquake damage to the stored fuel.

From the foregoing analyses, it is concluded that the spent fuel storage arrangement and design meet the safety design bases and satisfy the intent of Regulatory Guide 1.13.

The spent fuel storage racks will be designed to meet seismic Category I requirements (See Section 3.2). Stresses in a fully loaded rack will not exceed stresses specified by the ASME BSPV Code,Section III, Subsection NF-3400. 53(D)

The spent fuel storage racks will be made from type 304 stainless steel.

Materials used for construction will be specified in accordance with the latest issue of applicable ASTM specifications (A-240, A-479).

Control of weights handled within the fuel building will be administra-tive control of crane movement and location of spent fuel bundles in the 3 fuel storage pool. The 15 ton general purpose crane will traverse the full q9.1 length of the fuel building. Q9.2 Administrative controls will specify the fuel storage pool area to be filled with spent fuel bundles for decay storage, in order to assure that if required, the fuel storage pool could be traversed by the loaded 15 ton general purpose building crane without moving a load over stored spent fuel.

Ilandling operations involving the 15 ton general purpose building crane will avoid the area of the new fuel vault by moving loads over the cask 37(c) pool.

Transfer of fuel assemblies between the new fuel vault and the water filled storage pool and also within the storage pool, transfer pool and cask vault is performed with the fuel handling platform. The fuel grapple or the auxiliary fuel hoist is used depending on the transfer operation. The grapple and hoist provided with load sensing and limiting devices designed to the following load limits:

(C) Consistency (D) Design Am.No.53, 9/17/79 9.1-5 360124

ACNGS-PSAR Fuel Auxiliary Grapple Hoist (1bs) (1bs) load limiting switch 1200 1000 .

s load sensing switch 485 485 Q9.2 stall torque of hoist system 3000 3000 The load limiting features of the platform fuel grapple and auxiliary hoist will prevent damage to the fuel racks during fuel transfer operations.

These load limits provide a redundant safety feature since the fuel handling grapples are not lowered below the upper fuel rack and they will interface with the fuel bail, only.

9.1.2.4 Testing and Inspection The spent fucI storage racks require no periodic special testing or inspec-tion for nuclear safety purposes.

9.1.2.5 Summary of Radiological Considerations By adequate design and careful operation procedures, the safety design bases of the spent fuel storage arrangement are satisfied. Thus, the ex-posure of plant personnel to radiation is maintained well below published guideline values. Further details of radiological considerations including those for the spent fuel storage arrangement are presented in Chapter 12.

SG0125 9.1-Sa (D) L sign Am. No. 53, 9/17/79

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3G0126 9.1-6 Am. No. 53, 9/17/79

ACNGS-PSAR 9.1.3 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM 9.1.3.1 Design Bases The Spent Fuel Pool Cooling System will be Safety Class 3, seismic Category I in compliance with Regulatory Guides 1.13 and 1.26. The cleanup portion of the Spent Fuel Pool Cooling and Cleanup System will be Safety Class 23 "other" and "non-seismic", and can be isolated with redundant isolation valves which are part of the safety related fuel pool cooling system.

The objective of the Spent Fuel Pool Cooling and Cleanup System will be to (1) remove the decay heat from the fuel assemblies, (2) control water clarity, (3) maintain upper pool water temperature and clarity to permit normal reactor refueling and servicing and maintain fuel pool water level 37(U) such that fuel is not uncovered.

The Spent Fuel Pool Cooling and Cleanup Systen wil? be designed to:

a) Minimize corrosion product buildup and control water clarity, so that the fuel assemblies can be efficiently handled underwater.

b) Minimize fission product concentration in the water which could be released from the pool to the fuel handling area environment.

c) Monitor Spent Fuel Pool and containment fuel transfer pool water level and maintain a water level above the fuel sufficient to pro-vide shielding for normal building occupancy.

d) Maintain the pool water temperature below 125 F under normal oper-ating conditions. The maximum normal heat load in the spent fuel pool consists of one quarter of the reactor core fuel assemblies having decayed 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (the minimum time required to transfer spent fuel from the reactor to the storage pool) and plus successive 37(D) annual batch discharges (which have also been irradiated for 4 full power years ) and have cooled for periods of I-9 years . 53 (D) e) Remove drywell beat transferred to the containment pt al above the 37(D) drywell.

f) Cooling portion will be Safety Class 3, seismic Catecory I, and 23 cleanup portion will be Safety Class "other" and "non-seismic" The reactor will be loaded with a fuel charge of 732 fuel bundles. The Spent Fuel Pool storage capacity in the Fuel Handling Building will be 1710 fuel bundles. (with the capability to expand to 2790 fuel oundles) 37(D) 53(D)

SGOU?7 The heat sources are based on full power operation for 4 years prior to re-moval of fuel assemblies fran the reactor using the methodology of USNRC 53(D)

(U) Update (D) Design 9.1-6a Am. No. 53, 9/17/79

ACNGS-PSAR Standard Review Plan Branch Technical Position APCSB 9-2 and the fission 53 (D) product decay heat rates of ANS Standard 5.1, October 1973.

dG0128 9.1-6b

( D) Design A.n.No.53 9/17/79

ACNGS - PSAR This figure intentionally deleted.

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SG0133 Am. No. 53, 9/17/79 (D) Desien

- HOUSTON LIGHTING & POWER COMP ANY Allens Creek Nuclear Generating Station Units 1 & 2 FIGUTE 9.1-12C