ML20147J414

Amend 49 to Preliminary Safety Analysis Rept for Subj Facil Providing Addl Info in Response to Telephone Conversations Between Hlp & NRC

ML20147J414
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	12/21/1978
From:	HOUSTON LIGHTING & POWER CO.
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Houston Q ComJany r

Electric Tower PO Box 1700

1. Houston. Texas 77001 December 21, 1978 AC-HL-AE-271 Mr. Harold R. Dent on , 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 49 Please find under separate cover sixty (60) copies

/r] of Amendment 49 to the Houston Lighting & Power Company Allens Creek Nuclear Generating Station Unit 1 PSAR.

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

Amendment 49 consists of additional PSAR information related to issues identified in telephone conversations between Houston Lighting S Power Company and the Nuclear Regulatory Commission.

Very truly yours, f 4. /%

E. A. Turner Vice President Power Plant Construction S Technical Services LDR/bkl cc: J. G. Copeland (Baker S Botts)

R. Gordon Gooch (Baker S Botts)

J. R. Newman (Lowenstein, Newman, Reis, I Axelrad S Toll) l P. A. Horn O

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s (m / Before the i

1 UNITED STATES NUCLEAR REGULATORY COMMISSION l

l Docket No. 50-466 Allens Creek Nuclear Generating Station Unit 1 Amendment 49 to the PSAR Houston Lighting & Power Company, applicant in the above captioned proceeding, hereby files Amendment

(T 49 to the Preliminary Safety Analysis Report filed in

( ) connection with its application.

Amendment 49 consists of additional PSAR information related to issues identified in telephone conversations between Houston Lighting & Power Company and the Nuclear Regulatory Commission.

Respectfully submitted HOUSTON LIGHTING & POWER COMPANY

. Q.. +m E. A. Turner Vice President Power Plant Construction

& Technical Services O

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STATE OF TEXAS COUtn't OF RARRIS E. A. TURNER, being first duly sworn, deposes and says:

That he is Vice President of HOUSTON LIGHTING S POWER COMPANY, 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 J/M day /s d u 1978.

1 Signed: a /w E. A. Turner Subscribed and sworn to before me l this J/M day of .A cam M t., 1978. '

t4 i8- cv Notary Public 4W and for the County of Harris, State of Texas  ;

l My lconmission t) L /S. / FC.expires G

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ACNGS-PSAR HOUSTON LIGHTING S POWER COMPANY ALLENS CREEK NUCLEAR GENERATING STATION - UNIT NO. 1 PRELIMINARY SAFETY f.NALYSIS REPORT

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j AMENDMENT NO. 49 INSTRUCTION SHEET This amendment contains additional information which is submitted in response to telephone conversations between HLSP and NRC personnel. Each revised page bears the notation Am. No. 49, 12/21/78 at the bottom of the page.

Vertical bars with the number 49 representing Amendment No. 49 have been used in the margin of the revised pages to indicate the location of the revision on the page. The revised pages have the question number (eg QO10.5) next to the appropriate information which rcsponds to the question.

The following page removals and insertions should be made to incorporate Amendment 49 into the PSAR.

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ACNGS-PSAR

, - ~ from turbining of the driven end of the equipment due to blowdown ,f the

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system pressure upon rupture of the system pressure boundary.

D) The most substantial piece of rotating equipment is the recirculation pump and motor which, in the event of a major recirculation line break, and under certain rystem 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 currently the topic of discussions between GE and the NRC. The Applicant will implement 49 the generic resoletion of these discussions.

3. 5. 2. 2 Turbine Missiles 3.5.2.2.1 Introduction The potential for 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 orientation and existing structural shielding, need be provided to further reduce the probability of damage.

l The probability of damage was calculated for each Category I structure by evaluating the product of the probability for missile generation and the probability of impact on the structure.

The evaluation of the individual probability components and a summary of g' the overall damage probability is discussed in the following sections.

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\j 3.5-3 Am. No. 49, 12/21/78

ACNGS-PSAR SECTION 3.5: REFERENCES

-( / 3.5-1 P. W. Marriott, et. al., "The Loss-of-Coolant Accident and the Environment - A Probabilistic View", ASME Publication 72-WA/NE-9 3.5-2 Reference 3.5-2 has been deleted. 49 3.5-3 E. E. Zwicky, Jr. , TR-675L 211 GE, "An Analysis of Turbine Missiles Resulting from Last Stage Wheel Failure", October 3, 1967.

3.5-4 J. E. Downs, General ~ Electric Company, " Hypothetical Turbine  :

Missiles - Probability of Occurrence", March 14, 1973.

3.5-5 R. Salvatori, Westinghouse, " Failure Angles of Discs", May 24, 1971.

3.5-6 J. V. Rotz, "Results of. Missile Impact Tests on Reinforced 35  !

Concrete Panels," Vol. 1A, pp. 720-738, Second Specialty (C)

Conference on Structural Design of Nuclear Plant Facilities, 1975, New Orleans, LA.

3.5-7 R. C. Gwaltney, " Missile Generation and Protection in Light-Water-Cooled Power Reactor Plants", USAEC Report ORNL-NSIC-22, September, 1968.

3.5-8 " Structural Analysis and Design of Nuclear Plant Facilities," 35 Draft, Trial Use and Comment, Prepared by The. Committee on (C)

Nuclear Structures and Materials of the Structural Division of the ASCE, 1976.

3.5-9 C. V. Moore, Nuclear Engineering and Design, "The Design of Barricades for Hazardous Pressure Systems",1967.

3.5-10 C. V. Chelapti, R. P. Kennedy and I. B. Wall, "Probabilistic Assessment of Aircraft Hazard for Nuclear Power Plants" NED-19 (1972), pp. 358-360, North-Holland Publishing Company.

3.5-11 J. M. Biggs, " Introduction To Structural Dynamics", (1964) pp. 43-49 and 76-79, McGraw Hill Book Company.

3.5-12 R. A. Williamson and R. R. Alvy, " Impact Effect of Fragments Striking Structural Elements", of Halmen and Narver, Inc.,

NP-6516 (1957).

3.5-13 W. B. Cottrell and A. W. Savolinen, "U. S. Reactor Containment Technology", ORNL-NSIC-5, Vol, I, Chapter 6.

3.5-14 A. E. Long, "A Two-Phase Approach to the Prediction of the ' 35 Punching Strength of Slabs," ACI Journal, February,1975. (C) ,

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_ 3.5-16 Am $ 9 12/21/78

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ACNGS-PSAR 3.7 SEISMIC DESIGN 3.7.1 SEISMIC INPUT h 3.7.1.1 Design Response Spectra Design response spectra for the SSE and OBE for 1, 2, 4, 7 and 10 percent structural damping are shown on Figures 3.7-1 and 3.7-2 for the horizontal l21 earthquakes and on Figures 3.7-3 and 3.7-4 for the vertical earthquakes, 35 respectively. The maximum ground acceleration for the SSE and OBE was (U) selected equal to 0.10g and 0.05g respectively as described in Section 2.5.2.10.

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The design response spectra presented in Figures 3.7-1 through 3.7-4 were 130.3 developed according to the criteria' set forth in Reference 3.7-1. The design response spectra are applicable to the level of the foundation of 49 each Seismic Category I structure. (U) 1 3.7.1.2 Design Response Spectra Derivation 03.15 1

Acceleration time history records for the horizontal and vertical SSE (see l3 Figures 3.7-15 and 3.7-16, respectively) have been generated to envelop the 1 design response spectra presented in Figures 3.7-1 and 3.7-3, respectively. l 35 Figures 3.7-5 through 3.7-9 show the match between the time history response (C) ,

spectra and the design response spectra for the horizontal earthquake, and i Figures 3.7-10 through 3.7-14 show the match for the vertical earthquake. I3p5 '

The OBE motions can be obtained by scaling the respective SSE time histories g1;(C) 1 I

by 0.5.

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i' The time history records were generated from actual earthquake acceleration motions by selective amplification and phasing of their Fourier components.

The time histories are discretizec at a time step of 0.005 seconds and have a duration of 10 seconds. l35 i (C) l Thc response spectra were computed from the generated time histories using 75 points in the frequency range 0.2 - 34 cps. The frequency intervals of these points within the frequency range is as follows:

Frequency Range (Hertz)

Increment (Hertz) l 0.2 - 3.0 0.10 1 03.1, 3.0 - 3.6 0.15 35(U) 3.6 - 5.0 0.20 5.0 - 8.0 0.25 8.0 - 15.0 0'50 15.0 - 18.0 1.0 (C)-Consistency O

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3. 7- 1 Am. No. 49,12/21/78

ACNGS-PSAR APPENDIX 3.7.A t SEISMIC DESIGN CONSIDERATIONS L.))

1.0 INTRODUCTION

This appendix presents an in-depth discussion of the soil structure interaction analysis methodology employed to design Allens Creek NGS -

Unit No. I for earthquakes. It demonstrates that the analytical methods presented herein result in a conservative treatment of the seismic design of those structures, systems and components important to safety.

Table 3.7. A-1 provides a summary listing of the various analyses performed.

'2 . 0 INPUT MOTION 2.1 INTENSITY As indicated in PSAR Section 2.5.2.9, the highest horizontal site acceler- 48 ation is obtained by conservatively assuming the nearby largest event N130.6 (Bonham 1882 VII) could migrate to the closest approach of an associated or similar structure. This would place an 1882 Bonham-type event at the southern limi t of the Wichita /Ouachita tectonic province, 205 miles from the site. Attenuation of this event using iso-seismal data would result in an earthquake of Modified Mercalli Intensity IV at the site. Using Figure 1 of Reference 1, the horizontal SSE ground acceleration for the Allens Creek site would be, considering average foundation considerations, less than.0.018g. Thus, the SSE horizontal ground acceleration of 0.lg

"' selected for the site is considerably higher than the maximum horizontal ground acceleration that could be, even remotely, expected at the site.

2.2 DESIGN RESPONSE SPECTRA Design response spectra were obtained in accordance with guidelines provided in Regulatory Guide 1.60. The Regulatory Guide 1.60 spectral shape is considered conservative over certain frequency ranges when applied to the deep alluvium deposit Allens Creek site. As such, the use of the Regulatory Guide 1.60 response spectra rather than site-specific response spectra provides additional conservatism for the ACNGS seismic soil-structure interaction analyses.

2.3 CONTROL MOTION ELEVATION For Allens Creek, the design time histories (consistent with Regulatory Guide 1.60 response spectra) will be applied at the foundation level of each Category I structure.

49 During discussions with the NRC, analyses were performed comparing the effect of the location of the control motion with respect to the Reactor Building. A comparison of accelerations obtained at various points (refer to 4g i Figure 3.7. A-2) indicates a relatively close agreement between results ob-N130.6 -

tained with the control motion defined at the bottom of the Reactor Building  !

p\ mat (FLUSH-b) vs. at the ground surface (FLUSH-a). Maximum accelerations at t various points from the above two cases are presented in Table 3.7.A-3 and comparisons of floor response spectra are presented in Figures 3.7.A-3 1 through -8. I

3. 7. A- 1 Am. No. 49, 12/21/78

The design of' structures based on the maximum accelerations obtained at the various floor levels. These maximum accelerations are the peaks of the -

time histories obtained at the various floors and are identical to the n '

high frequency range accelerations shown in the corresponding spectra. The (b di fferences in maximum accelerations presented in Table 3.7. A-3 are in the range of 25% or less.

Systems and subsystems located on a certain floor are designed for the 48 corresponding . floor response spectra. The floor response spectra com- N130.6 parisons presented in Figures 3.7. A-3 through 8 indicate a more accentuated dif ference in floor spectral accelerations in the frequency range of 3 to approximately 7 cycles per second which is caused by di fferences in the

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control motion input in the 3 to 6 cycles per second frequency range.

Nevertheless, the ACNGS analysis methodology for the Reactor Building will reflect the horizontal floor response spectral values obtained when the control motion is defined at the bottom of the Reactor Building mat and a 49 T 15% peak broadening is performed. Similarly, the control motion will be defined at the foundation level for the other Seismic Category I structures.

3.0 EMBEDMENT EFFECTS AND STRUCTURE - STRUCTURE INTERACTION One of the reasons for selecting the FLUSH type finite element analyses to establish the soil-structure interaction effects is that the analysis comes closest to representing in a rational way all the important aspects 48 O of the problem. -

N130.6 While a FLUSH type finite element analysis allows for the adequate re-presentation of structure-structure interaction and embedment effects, an elastic half space type approach does not.

As an example, a FLUSH type analysis of the Reactor Building, where there is no embedment and no adjacent structures, was performed (FLUSH-c) and results were compared with responses obtained from an elastic half space solution (Spring-c). A comparison of maximum horizontal accelerations l at various pohts is provided in Table 3.7. A-4, and comparisons of response I spectra are pre.sented in Figures 3.7. A-9 through -14. The responses obtained from the two analyses are in excellent ag r eeme nt in terms of both maximum accelerations and response spectra.

3.7.A-2 Am. No. 49 12/21/78 O

l ACNGS-PSAR A However, when the embedment and adjacent structures are considered, more i i pronounced di fferences, caused by embedment and structure-structure ef fects b become present. A comparison between FLUSH type finite element analyses (FLUSH-a) versus a h lf space approach using as input motion the time history. Obtained as th= bottom of the Reactor Building mat for the three shear moduli consider, t Coring-b) indicate more accentuated di fferences both in terms of maximum acierations (refer to Table 3.7. A-5) and response spectra at the various points (refer _ to Figures 3.7. A-15 through -20). These pronounced differences are caused by not consider-ing, among others, the embedment and structure-structure effects.

Lastly, a comparison between FLUSH finite element analyses (FLUSH-a) and an elastic half-space solution (Spring-a) is presented in Table 3.7. A-6 and Figures 3.7. A-21 through -26 in terms of maximum accelerations and floor response spectra, respectively. The comparison indicates maximum acceleration di fferences in the range of 25-30% or less. A comparison of the design shears and moments resulting from the FLUSH-a/ Spring-a analyses is presented in Table 3.7. A-7. The di fferences in the design moments and shears obtained at various points in the Reactor Building from the tw 48 analyses are in the range of 25-30% or less and parallel closely the N130.6 di f ferences in the maximum accelerations.

4.0 SOIL LAYERING An appropriate seismic soil-structure interaction analytical procedure should be able to take into account, among others, the variations of soil O characteristics with depth. While a FLUSH type finite element analysis I is capable of adequately representing the soil layering, an elastic half space approach does not.

l For the Allens Creek site, the soil layering, as presented in PSAR Figure 2.5.4-7A, makes the use of a FLUSH type finite element analysis preferable.

5.0 TREATMENT OF THE THREE-DIMENSIONAL SOIL-STRUCTURE SYSTEM WITH A TWO-DIMENSIONAL MODEL l

A nuclear power plant represents a complex three-dimensional system which '

is very difficult to represent analytically. While an elastic half space approach considers three-dimensional effects, the lack of adequate repre-sentation of the structure-structure, embedment, and soil layering effects makes the approach undesirable for the Allens Creek site.

The finite element FLUSH type analysis selected for the ACNGS. (FLUSH-b) [ 49 is a two-dimensional idealization using plane-strain elements. While a three-dimensional finite element analysis is theoretically possible, its practical implementation is difficult and the benefits derived are minimal.

Theoretically, it should be possible to analyze with finite elements any arbitrary three-dimensional geometry. However, as discussed in Reference 2, restrictions of cost and available computer capacity and the lack of good 3-D stress-strain relationships for foundation materials make a full three-dimensional analysis impractical. Fortunately, indi-

, cations are that while 3-D effects might be significant for the above -

\. ground parts of structures they are relatively unimportant for the 3.7.A-3 Am. No. 49 12/21/78 1

ACNGS-PSAR O U p AVE, where AVEisrgpresentedbythgcurvesasindicatedinFigures 2.5.4-7B and 7D, 1.5 AVE, and 0.667 AVE. The envelope of the three V)

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floor response spectra obtained at a mass point shall be used for seismic qualification and' design purposes of systems and sgbsystems with the re-striction that the peak frequency shifts from the AVE case should be a 48 mi nimum of + 10% . Also, for the design of structures, the highest of the N130.6 three maximu_m accelerations obtained at a given mass point for each direction shall be used.

In view of the above, it is felt that the FLUSH type finite element analysis chosen for the ACNGS site is the most adequate. Furthermore, the implementation procedures used will significantly increase the con-servatism level of the ACNGS seismic analyses.

8.0 This section has been deleted.

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1 9.0 ADDITIONAL CONSIDERATIONS l l

Wi th respect to the overall seismic analysis procedures employed for the 48 Allens Creek Project, there are certain aspects which provide additional N130.6 design conservatism.

9.1 DYNAMIC MODELING The dynanic models for the various structures are prepared using three I dimensional finite element analyses in order to establish the effective l shear area and moment of inertia for the various members.

9.2 PARAMETRIC STUDIES Prior to perfoming the FLUSH finite element soil-structure interaction I analyses, numerous parametric studies are carefully conducted in order to establish structural and overall model adequacy, mesh size, the boundary p conditions and many others.

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3.7.A-5 Am. No. 49, 12/21/78 1

ACNGS-PSAR (a) Snubbers are subjected to force or displacement versus time loading at frequencies within the i range of significant modes of the piping system.

(b) Displacements are measured to determine the per- s formance characteristics specified. 1 (c) Tests are conducted at various temperatures to insure operability over the specified range.

(d) Peak test loads in both tension and compression will be equal to or higher than the rated load requirements.

(e) The snubbers are also tested for various abnormal 41 l environment conditions. Upon completion of the Q above abnormal environmental transient test, the 110.11 5

snubber shall be tested dynamically at a frequency with a specified frequency range. The snubber must operate normally during the dynamic test.

c) Snubber Installation Requirements As installation instruction manual is requirea by the t aspension design specification. This manual is required to contat in-structions for storage, handling, erection and adjus+ matt (if necessary) of snubbers. Each snubber has an installation location drawing, which contains the installation, location of the snubber on the pipe and structure, the hot and cold h)

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settings, and additional information needed to install the particular snubber. l The suspension design specification requires that snubbers be provided with position indicators to identify the rod posi-tion. This indicator facilitates the checking of hot and cold settings of the snubber, as specified in the installation manual, during plant preoperational and startup testing.

d) Inspection, Testing, Repair and/or Replacement of Snubbers The suspension design specification requires that the snubber supplier prepare an installation instruction manual. This manual is required to contain complete instructions for the testing, maintenance and repair of the snubber. It also con-tains inspection points and the period for inspection.

The suspension design specification requires that hydraulic snubbers be equipped with a fluid level indicator so that the level of fluid in the snubber can be ascertained easily.

3.9.2.9 Component Supports Piping component supports for ASME Code Class 1, 2, 3 or MC systems will 41 be designed in accordance with ASME Section III NF Load combinations for Q piping component supports are given in Table 3.9-7 for Ebasco. For GE (n

Class 1 supports see the loading combinations listed by Table 5.2-2a. 49 110.7 General Electric does not supply Class 2, 3, or MC supports. Allowable 110.8 stress limits are given in Table 3.9-8. In situations where a device which is not governed by the rules of the ASME B6PV Code is in the load path for the support of ASME 3.9-6b Am. No. 49, 12/21/78

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ACNGS-PSAR During reactor startup and hot standby, the dissolved oxygen content of p reactor water may be higher than during normal power operation. During

( this period more restrictive limits are established. After power operation

\- has been established boiling deaerates the reactor water reducing the influence of oxygen on potential chloride stress corrosion cracking. 8 The relationship of chloride consentration to specific conductance measured Ql-at 25 C for chloride compounds such as sodium chloride and hydrochloric 5.1 acid can be calculated. Values for these compounds essentially bracket values of other common chloride salts or mixtures at the same chloride concentration. Surveillance requirements are based on these relationships.

The sampling frequency when reactor water has a low specific conductance is adequate for calibration and routine audit purposes. When speci fic con-ductance increases, and higher chloride concentrations are possible, or when continuous conductivity monitoring is unavailable, increased sampling is provided.

Coolant Chemistry Requirements are summarized in Table 5.2-11. 34 5.2.4 (C)

/RACTURE TOUGRNESS 5.2.4.1 Compliance with Code Requirements

~

The reactor vessel pressure retaining components comply with tne require-ments of NS-2300 as written in the Summer 1972 Addenda of the ASME Code,Section III (Formerly Code Case 1514).

/O The reference temperature, RT will be established for all required pressure-retaining materials b d in the construction of Class I vessels.

Tnis includes plates, fo rg i ng s , weld material and heat affected zone. The RT NDT di f fers from the nil-ductility temperature, NDT, in that in addi-tion to passing the drop weight test, three Charpy-V Notch specimens (transverse) must exhibit 50 ft-lbs absorbed energy and 35 mil lateral expansion at 60 F above the RT temperature.

NDT

~

The ferritic materials used for piping, pumps and valves of the Reactor Coolant Pressure Boundary are 2-1/2 inches or less in thickness and are not subject to dropweight tests or the 35 mils lateral expansion; 50 f t-lb Charpy-V Notch test. Only those tests required by NB-2332 and NB-2333 for thickness 2-1/2 inches and less are performed.

5.2.4.2 Acceptable Fracture Energy Levels 7 Tne initial upper shelf fracture energy levels for core belt line material (base, weld, heat affected zone) is required to be 75 f t-lb (transverse) l49 nV 5.2-19 (C)-Consistency Am. No.49, 12/21/78

ACNGS-PSAR

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minimum to allow for degradation of the upper shelf due to neutron fluences  !

to 50 ft Ib during reactor service life. '

V 5.2.4.3 Operating Limitations Appendix G of the ASME Code,Section III, Protection Against Non-ductile 17 Failure, will be used in determining pressure / temperature limitations for Q2-all phases of plant operating and the additional requirements of 10 CFR 50 2.2-Appendix G for operation when the core is critical will be complied with as indicated in Section 5.2.4.6.5. The adjusted reference temperature used in determining operating limits will be based on integrated fast neutron flu-ence at 1/4 and 3/4 of the beltline shell thickness as appropriate. The expected shift in RT with neutron fluence is defined in Section  ;

NDT 5.2.4.6.8.

5.2.4.3.1 Operating Limitations During Startup and Shutdown 17 l

Appendix G of the ASME Code,Section III, Protection Against Non-ductile '

Failure, will be used in determining pressure / temperature limitation for all phases ~of plant operation.

49 i

5.2.4 4 Compliance with Reactor Vessel Material Surveillance Program Requirements Reactor vessel material surveillance specimens are provided in accordance 5 with ASTM-E-185-73.

\'

At least three sets of specimens will be provided. Each set will consist Q5.12 of 12 Charpy-V-Notch specimens of base material, weld material, and weld heat affected zone material so that withdrawal of samples can be per ASTM-E-185-73. Archive material per ASTM-E-185-73 will be furnished with the surveillance test specimens.

5.2.4.5 Reactor Vessel Annealing l 5

In-place annealing of the reactor vessel because of radiation embrittlement QS.11 is unnecessary because the predicted value in transition of adjusted refer-ence temperature will not exceed 200 F - see 10CFR50, Appendix G, Para-graph IV.C.

5.2.4.6 Compliance with 10CFR50 Appendices G 6 H The BWR-6 reactor design is in full compliance with Appendices G S H of'10CFR50 as follows:

5.2.4.6.1 Specimen Orientation for Original Qualification 46 '

Versus Surveillance (Reference Appendix G-IIIA) (G)

Allens Creek uses both transverse and longitudinal specimens for 49 qualification' testing. Allens Creek uses transverse specimens to QS.41 meet 10CFR50 Appendix H surveillance requirements. QS.12 f Charpy-V-Notch tests.as defined in NB-2321.2 are to be conducted on 5 both unirradiated and irradiated'ferritic materials; however, the special beltline' longitudinally oriented Charpy specimens required by 5.2-20 (G)-GESSAR Am. No. 49,12/21/78

ACNGS-PSAR f)

P Thin sections of nozzles and appurtenances where the thickness is 2.5 inches or less will meet the fracture toughness requirements of 5 Paragraph (c) of Code Case 1572.

QS.11 g) WRC-175 procedures will be used for other cases where Section III ~

is not definitive, such as when the ratio a/, exceeds 1.0 in accordance with Code Case 1572, Paragraph (d).

5.2.4.6.5 Fracture Toughness Margins in the Control of Reactivity Reference G-IV.A.2.c. The specific requirements for operation when the 17 core is critical which are stated in the second sentence of Paragraph Q2-G-IV.A.2.c are considered unncessarily conservative to account for the 2.2 design basis reactivity accident and impose operational inconveniences to the normal method of BWR reactor startup. A study and topical report are being prepared by GE to support a request for change in the Regulation. In the interim, the specific temperature limits for operation when the core is ,

critical stated in the Regulation will be adhered to based on an analysis of the vessel heads and shell areas remote from discontinuities as required by the cross reference to Paragraph IV.A.2 a in G-IV.A.c.

5.2.4.6.6 Bolting Materials (Ref. G-IV A 4) 49 1

Q5.11  ;

Bolting will meet this added requirement. Q5.12 5.2.4.6.7 Upper Shelf Energy for beltline (Ref. Appendix G-IV B) 49

{

Q121.2 l For the ACNGS reactor vessel, 75 ft. Ibs, minimum upper shelf Cy energy )

will be specified for beltline material.

1 5.2.4.6.8 '

Predicted Shift in RINDI (Ref. G-II G and WB)

For design purposes the adjusted reference temperagure for BWR vessels will be predicted using the curve, " Upper Limit for 550 F GE-BWR Operating Experience", Figure 5. 2-2. Although this curve is based on 30 ft. Ibs.

Iongitudinal Charpy-V data, it is considered appropriate for the following reasons:

5 a) It is the most applicable curve available for GE-BWR's. Q1-5.9 b) GE data shows that the shift at 30 ft. Ibs, is not substantially different for transverse versus longitudinal orientation.

c) GE specifications will restrict the copper and phosphorus content of beltline materials as follows: q1, 5.10 Base Metal: Cu - 0.12 percent maximum P - 0.015 percent maximum i l

Weld Material: Cu - 0.1 percent maximum P - 0.025 percent maximum 5.2-20c Am. No. 49, 12/21/78

ACNGS-PSAR Calculations performed in accordance with the rules of Regulatory Guide 1.99 p (Figure 1) using specification limits on copper agd phosphorus show that the '42

( maximum shift in RT g ase metal wouM h 77 F. N madmum sWt for Q NDT weld metal would be 97 F. 121.2 Actual shifts are expected to be considerably lower than those calculated using maximum ~ specification limits of copper and phosphorus. This is especially true of the weld metal. . Chemical and analyses on weld metal con-sistently show copper below 0.08% and phosphorus below 0.02%. Therefore, it is expected that the actual RT shifts calculated in accordance with the rules of Regulatory Guide 1.9 Nill be in the range of 50 to 70 for both base metal and weld metal. Experiencehasshownthattheinigial reference tempergture for beltline materials should not exceed +10 F for base metal and 0 F for weld metal so9the adjusted reference temperature at EOL is predicted to be less than 100 F, hence three capsules are used. The capsule withdrawal schedule is: 46 (G)

First capsule - one fourth service life.

Second capsule .three fourths service life.

Third capsule . standby.

The decrease in upper shelf energy predicted in accordance with Figure 2 ' 42 ofRegulatoryGuidef.99sgowsamaximumdecreaseof20%atapeakEOL Q fluence of 4.5 x 10 n/cm . Since the initial upper shelf energy is 121.2 required to be 75 foot pounds, the end of life upper shelf energy will remain above 50 foot pounds.

5.2.4.6.9 , Positioning of Surveillance Capsules (Ref. H II C (2)) 49

( Q1-Surveillance capsules are placed such that the neutron flux received by 5.9 the capsules is not more than three times the vessel beltline surface. 5.10 Typirelly the fluence is 1.1-1.5 times this amount. l Capsules are not attached to the vessel but are in seal welded capsule i holders. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding as shown in Figure 5.2-8. Since re-actor vessel specifications require that all low alloy steel pressure ves-sel boundary material be produced to fine-grain practice, underclad crack- 34 ing is of no concern. The capsule holder brackets allow the removal and (G) reinsertion of capsule holders. These brackets are designed, fabricated, and analyzed to the requirements of Section III ASME Code.

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TABLE 5.2 -2 (Cont'd) 34(G)

^

+.C No. of Cycles Emergency Conditions (Cont'd)

13. Improper Start of Cold Recirculation Loop 1***

14. Sudden Start of Pump in Cold Recirculation Loop I 1*** l

15. Improper Startup with Reactor Drain Shut Off

~

l Followed by Turbine Roll'and Increase to Rated Power 1***

Faulted Condition

16. Pipe Rupture and Blowdown ** *** l 1*** '

17. Safe Shutdown Earthquake at Rated Operating Conditions 1*** l l

l I

l l

• Applies to reactor pressure vessel only

• • Bulk average vessel coolant temperaturechangeinany1-hourperjod

• • • The annual encounter p{obability of the one cycle events is (10 for energency and (10 for faulted events

• • • • Includes 10 maximum load cycles per event.

• • • • • All the dynamic effects of a postulated pipe rupture, including RPV l

. cavity annulus pressurization, will be taken into account. 49 5.2-38a (G)-CESSAR Am. No. 49,12/21/78

ACNGS-PSAR T c insulation is cither the all-metal reflective type or thc conventional ser e s type. It is prefabricated into components for fielc insta11atien.

'te :.cyc;_e insulation is provided at various locations to permit period:c inspection cf the equipment.

cravisions taken to control those factors that contribute to stress corro-sica c cking are discussed in Section 5.2. e 5.5.1.4 Safety Evaluation Reactor Recirculation System malfunctions that pose threats of damare to the fuel barrier are described and evaluated in Chapter 15, " Accident Analysis". It is shown in Chapter 15 that none of the malfunctions result in fual damage. The recirculation system has sufficient flow coastdown charscteristics to maintain fuel thermal margins during abnormal opera-tional transients.

.igure 5.5-4 shows the core flooding capability of the recirculation sys t er.. The core flooding capability of a jet pump design plant is dis- ,

I cussed in detail in the Emergency Core Cooling Systems document filed wi . l c.e AEC as n General Electric topical report (see Reference 5.7-4). Tc.s ability to ref'. cod the BWR core to the top of the jet pumps as shown sche-  ;;-5  !

matica;1y on Figure 5.5-4 and discussed in Reference 4 applies to all jet .24 l pump BK2's and dces not depend on the plant size or product line.  !

l

?iping and pump design pressures for the Reactor Recirculation System are ]

( based on peak steam pressure in-the reactor dome, appropriate pump head j allowances, and the elevation head above the lowest point in the recircu-

.ation loop. Piping and related equipment pressure parts are chosen in cecordance with applicable codes. Use of the listed code design criteria assures that a system designed, built, and operated within design limits ans an extremely low-probability of failure caused by any known failure mechanism.

-General Elcetric Purchase Snecifications require that the recirculation pumps first critical speed Ishall not be less than 130 percent of operating speed. Calculation submittal is required and verified by General Elcetric Design Engineering.

General Electric Purchase Specifications require that integrity of the pump cc.se be maintained through all transients and that the pump remain operatic thrcugh all normal and upset transients. The design of the pump and - rsr bearings are required to be such that dynamic load capability at rated c,perating conditions is not exceeded during the Safe Shutdown Earthquake.

Calculation submittal to General Electric is required.

Analyses performed to determine if the recirculation pump can become 1 missile indicates that for the postulated full double-ended pipe breax

-(t.0CA) in either recirculation pump suction or discharge line, destructive pump and motor overspeed enuld occur. This hypothetical situation is currently the topic of discussions between GE and the NRC. The Applicant 49 will implement the generic resolution of these discussions.

l 5.5-5 Am. No. 49, 12/21/78

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/ 7.3.1.1.6 ESF Support Systems Instrumentation and Controls l1 7.3.1.1.6.1 Standby Power System Inst rument at ion and Control The Onsite Standby Power System will consist of three completely independen:

power source s, any two of wnich will be required for safe shutdown or acci-nent mitiga: ion. The power sources will be diesel generators and related equipment rrcr;9c in threa divisions, a) Division 1 - Diesel Generator No. IA wi th it s auxiliaries and safety related bus No. lA 4KV ESF SWGR Inter-Tie System

3) Division 2 - Diesel Generator No. 1B with its auxiliaries and safety related bus No. IB 4KV ESF SWGR Inter-Tie System i

c) Division 3 - Diesel Generator No. 1 HPCS with its auxiliaries and safety related bus No. 1 KV HPCS SWGR A :mnplete physica*. description of the diesel generators, auxiliary equipment ano Bus systems is given in Section 8.3.1.1.10, 7  ?.1.. 6.1.1 Diesel Generctor Starting System f~'\

s_,)

Each diesel generator will have a manual and an automatic starting mode. A cer. trol switch on the control board will allow the operator to manually start 37 the diesel generator. Momentary closure of the " start" contact of the con-trol switch will cause starting to proceed to completion. Operator starts 3) will be used for testing and for anticipating abnormal conditions within the unit.

In the manual mode, control can be cecomplished from the Control Room or 43 locally. Control switches and pushbuttons are provided at both locations Q for start and stop functions. A selector switch is also provided in the 040.5 Control Room to select between local and Control Room controls.

Automatic starts of each diesel generator will be initiated by low voltage on its associated ous and/or actuation of an Engineered Safety Feature signal.

The undervoltage protection for the safety- related buses for ACNGS will include two levels of redundant, coincident logic with a time delay. The time delay shall be determined from an analysis of the voltage requirements of the safety-related loads at the onsite system distribution IcVe13.

The time delay shall not exceed the maximum time delay asst..~d in PSAR accident analyses and shall minimize the effect of short du- "ir-disturbances from reducing the availability of the offis 49 surces.

Voltage sensors shall automatically initiate the discon of offsite power sources whenever the voltage set point and time limits have been exceeded. Voltage sensors shall be designed to satisiy IEEE 279-1971.

.imitint. conditions for operation of this system will be included in the plant Technical Specifications.

(D)-Design 7.3-3F Am. No. 49, 12/21/78

ACMGS-PSAR ne generator is synchronized at the time of bus tie breaker opening, no

..s e , vill ne shed. See Auxiliary one line diagram Figure 8.3-1 and Tabica E.3-1 cugh 8.3-3.

On oper.in; c.c the bus tie breaker with the generator breaker open, the

',g3 saf aty related bus will at.:omatically shed all safety related load except ~

nose in the first load block. '

~he .

neacic star:..n;; of subsequent loads sill be delayed by timing c . lay:. . i:: fiv . :.ccond intervals between them. The starting sequence fcr u.:h s.af e, raia:ed bus is shown on Tables 8.3-1 through 8.3-3.

If a design basis accident occurs with a bus tie breaker open, the required 37' g

aafety related loacs will be connected co the bus automatically in proper sequences as .n the preceding paragraph. lf

.ac :..esel generator may be periodically run under load by operator act ion.

c.ould norta; ac power be lost during such a condition, the safety related bus tie braaker will trip but no loads will be shed. If a design basis acciLnt ;,rece dec ar follows this loss of normal ac power, the safety related f g3 g)

s autenatic loading, sequence will begin simultaneously.

^ncJ.d a design basis accident occur and of f aite power is available3 the 48 recuired safety related leads will be sequentially connected to the safety (U) re b ed bus.

\'

Ne have not yet specified a type of sequencer for automatic loading of the safety buses. It will meet the following criteria:

1. Designed to meet applicable NRC requirements including IEEE 323-1974, IEEE 344-197S, IEEE 279-1971 and others.

2. There will be only one sequencer provided for each train. It will sequence the loads to both on and offsite power for its associated train. No single failure will prevent safe shut down since the safety trains are redundant.

?

The sequencerwill be provided with a continuous (several minute interval) 49 auto-tester feature with Control Room alarm or detection of failure.

4. In addition, the applicant will provide a comparative probabalistic analysis addressing the availability of ECCS given sequencing on off-site power versus not sequencing offsite power during the operating license stage which will show an equal or better reliability than other design alternatives.

5. Redundant Class IE Power supplies (two AC circuits with one circuit backed up by the Class IE Battery System) will be provided, O

V (U)-Update (D)-Design 7.3-41 Am. No. 49, 12/21/78

7.3.1.1.6.1.5 .

Diese1' Generator Support Systems O

'y/ a) Fuel Oil System The three diesels will be provided with oil from three seven day g 37' fuel cil storage tanks, each storage tank f eeds a four hour day ,

.1.q tank for each diesel. The feed of oil is accomplished by gravity.

I:

Control r,f 4 flow to day tanks will be operated f rocn level switches '

on the -y tanks. When the fuel oil level f alls to the three hour j level :.n any day tank its oil supply valve will open ar.6 the oil will flow from the storage tank to its associated day tank. When the fuel oil level rises to the four hour level in any day tank, its oil supply valve will close. The range of operation is held between the 1,10 i

i four hour and three hour levels to keep oil volume high to minimize Ql-condensation due to breathing. 8.40

, . .; tanks di be provided with level switches for high-high and low-low

..ei annunciation in the Control Room. Oil flow indication will be provided 3

fer e ch oil supply at the storage and day tanks to be able to detect any oil (

le k in the piping. The storage tanks will be provided with a level switch Cor low level annunciation in the Control Room and level indication for PAM.

All valves and instruments for each diesel oil supply line, will be powered fre,a that diesel bus.

' diesel engine fuel pump will be engine driven and will take suction from c / tank at a point a few inches above the bottom of the gage glass which

> ili allow maintenance personnel to see accumulations of water and drain it from the tank. The fuel pump will discharge to the fuel injectors. The engine will be stopped by tripping the inj ectors to the no load position by mechanical means (overspeed trip) or by energizing a trip solenoid. Reset 1

b)

(/ (ll)-Updat e (D)-Design 7.3-41a Am. No. 49, 12/21/78

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-1 j CHAPTER 8 ELECTRIC POWER Page Amendment No.

1* 49 2* 45 3* 45 i 35 ii 35 iii .

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,_E_ m - ,r ACNGS-PSAR Uron a loss of offsite power, the tie breakers between the normal and ESF buses will automat ically open and the standby diesel generators will auto-mat ically st art . A fter reaching rated s peed and volt age, they will auto-mat ically be connected to the ESF buses. The diesel generator automatic st art ing and loading sequence is discussed further in Section 8.3.1.1.10.

The undervoltage protection for the safety-related buses for ACNGS will include two levels of redundant, coincident logic with a time delay. The time delay shall be determined from an analysis of the voltage requirements of the safety-related loads at the onsite system distribution levels.

The time delay shall not exceed the maximum time delay assumed in PSAR 49 accident analyses and shall minimize the effect of short duration disturbances from reducing the availability of the offsite power sources.

Voltage sensors shall automatically initiate the disconnection of offsite power sources whenever the voltage set point and time delay limits have been exceeded. Voltage sensors shall be designed to satisfy IEEE 279-1971.

1.imiting conditions for operation of this system will be included in the plant Technical Specifications.

The two 4 KV non-safety related station service t rans formers will be rated [35(C) 15 mva, FOA at 55 C rise, with a 112 percent supplementary rat ing at 65 C risa. These t rans formers shall be connected delt a-delt a with a 7 percent impedance (approximately) on a 15 mva base.

l35(C)

) The 4 KV normal buses I A and IB will be rated 3000 amps and will be pro- 2-7.39 vided with 3000 amp incoming breaker of 350 mva interrupt ing capacity.

ESF buses S-1, S-2, and HPCS will be rated 1200 amps. The incoming feeder 35(D?35 breakers of these buses will be rated at 1200 amps , 350 mva IC and all out- 5 (C) going breakers of all 4 KV buses will be 1200 amps, 350 mva IC. Buses will Q1-7, be protected by time overcurrent relays. The motors fed from these buses 14 will be protected by t ime overcurrent and instantaneous overcurrent relay devices, The t ime overcurrent relays will be set to trip on overloads of  ;

125 to 140 percent norma l current s. Instant aneous overcurrent relays will t rip for fault magnitude current s. A t ime overcurrent relay will be used l to alarm for currents of 100 to 125 percent full load current s. Relay settings will be determined on an individual basis to selectively trip loads and feeders as required to insure safe plant o pe rat ion . Safety related loads will have the capability of being controlled only from the cont rol 5(D) l (C)-Consistency (D)-Design 8.3-3 Am. No. 49, 12/21/78

4 EFFECTIVE PAGES LISTING CHAPTER 17 O* QUALITY ASSURANCE h Amendment 1* 49 2* 49 3* 48 i 33 11 45 iii 45 17.0-1 33 l 17.0-2 33 17.0-3 49 17.1-1 45 17.1-2 48 17.1-3 45 17.1-4 48 17.1-5 48 17.1-6 48 17.1-7 42 17.1-8 45 17.1-9 49 O 17.1-10 17.1-11 17.1-12 33 33 33 17.1-13 33 17.1-14 33 17.1-15 33 17.1-16(Amendment Number not shown on page) 33 17.1-17 33 17.1-18 33 17.1-19 45 17.1-20 45 17.1-21 33 17.1-22 45 17.1-25 45 17.1-24 49 17.1-25 33 17.1-25a 33 17.1-25b 33 17.1-26 33 17.1-27 33 17.1-28 33 17.1-29 33 17.1-30 33 17.1-31 33 17.1-32 33 17.1-33 33

+0 17.1-34 17.1-35 33 45 17.1-36 33 17.1-37 33 17.1-38 33

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EFFECTIVE PAGES LISTING O CHAPTER 17 QUALITY ASSURANCE h Amendment 17.1-39 33 i 17.1-40 33 17.1-41 33 17.1-42 33 17.1-43 33 l 17.1-44 45-17.1-44a 45 17.1-45 33  ;

17.1-46 46 17.1-47 46 17.1-48 45 17.1-49 45 17.1-50 45 17.1-51 45 17.1-52 33 17.1-53 33 17.1-54 33 17.1-55 45 '

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17.1-70 33 17.1-70s 49 17.1-70b 48 17.1-70c 45 17.1-70d 45 17.1-70e 45 17.1-70f 49 17.1-70g 45 17.1-70h 49 17.1-70'i 49 17.1-71 33 17.1-72 33 O  :

2 Am, tio. 49,12/21/78 t

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p ACNGS - PSAR V Site Const ruction Quality Control activities will be under the administrative and technical control of the Ebasco Quality Assurance 45 Engineering Department instead of the Construction Quality Control (U)

Department. This organizational structure is described in Table 17.1.2B-4. All Ebasco quality assurance related activities performed prior to January 1,1977 were done in accordance with the program described in this chapter. All Ebasco quality assurance activities subsequent to January 1, 1977 will be performed in accordance with the latest HL&P and NRC accepted revision of Ebasco's Topical Report 49 No. ETR-1001, which at present is Rev. 7 , except for the site (g)

Construction Quality Control organizational changes described above and other approved modifications listed in Table 17 .1. 2 B-3 . Later 45 )

NRC approved revisions to ETR-1001 may be incorporated when deemed necessary. (U)

If necessary to define any additional clarifications, or modifications

.o the project Nuclear Quality Assurance Program Manual because of HL6P 45(U) contract requirements or to suit the unique Project conditions, they will be submitted for NRC acceptance in accordance with established pro- l 46 vi sions which require execution of an authorization form involving (U) approval of specified authorities to assure, among other things, that safety and/or quality ire not sacrificed or compromised. Approved changes will be incorporated in above referenced Tables, as required.

The Ebasco Quality Program defined herein assures that structures, O systems, and components important to saf e ty as defined in Section V 3.2 of this PSAR, are reliable and possess a high degree of quality.

This objective is achieved by the implementation of the Ebasco Nuclear Quality Assurance Manual which defines the policy, proce-dures, and requirements by which Ebasco will design, purchase and erect the Allens Creek Nuclear Generating Station. Implementation of the Ebasco Nuclear Quality Assurance Manual provides a quality program which is in compliance with the requirements of the Code of Federa l . Regu lations , 10 CFR 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants," cnd ANSI N45-2-1971, " Quality Assurance Program Requirements for Nuclear Power Plants".

17.0.C GENERAL ELECTRIC The Quality Assurance Program for safety related activities and services for the Allens Creek Nuclear Generating Station is described in the 33(U)

General Electric Nuclear Energy Divisions BWR Quality Assurance Program Description, NED0-ll209-04A.

49 O

17.0-3 (U)-Update Am. No. 49, 12/21/78

ACNCS-PSAR (3 ]he OA Program Evaluation Committee has the following responsibilities: 33 (U)

C/ -

a)

Monitorine the effectiveness of the OA Program by reviewing reports of 42 audits, inspections, design reviews, etc. (U) b)

Re vie win g the activities of the QA Department to evaluate the per- 42 formance of the program, t (U) c)

Providing written recommendations for program improvements or 42 modification to the Executive Vice President. (U) 17.1.lA.4 Consultants ifL&P may utilize the services of qualified consultants to assist in the performance of appropriate quality tasks, such as audits, inspections, interpretations of test results, reviews, etc.

17.1.lA.5 Organizational Interfaces HL&P will formally establish with each of its prime contractors a division of responsibility for the project. This division of responsibility will become the basis for identifying specific external interfaces to the system, 45 structure and component level which HL&P must control. The division of g) ,

responsibility, both internal and external, for interface control and l verification is shown in Table l i. l . l A-1, I 2

O 17.1 lA.6 Summary U)

O The individual positions and ercups in the HL&P organization performing quality related activities, such as reviews, verification, checking, h3 au d i t i ne is and (U) nel; the ManaPer - OA and his staff; other cognizant HL&P person- 26 the Design Review Committee; the Ouality Assurance Program Evaluation Commit tee and qualified consultants.

Their authority, responsibilities, and designated functions are described in Sections 17.1.lA.1 through 17.1.lA.4 of the PSAR. Figure 17.1.lA-3 illustrates their independence to 33 carry out the required quality related functions. (U) 17.1.lB ORGANIZATION Tts Ebasco organization which includes the quality assurance organ: ation (

este olished for the ACNGS is shown on Figure 17.1.1B-2 as amended w ETR-1001, Figure 1.2-4, Revision 3. Reporting to Ebasco's President, 49 through the Senior Vice President of Engineering and Construction, are seven independent lines of authority; the Vice President of Construction; the Vice President of Nuclear Engineering (who is in charge of Materials 33 Engineering and Quality Compliance Department and the Licensing Depart- (U) ment); the Vice President of Engineering (who is in charge of engineering and design section); the Vice President of Purchasing; the Vice President of Plant Operation and Betterment; the Vice President of Consulting En-gineering; and the Vice President of Projects (who is in charge of the g project management and coordination). The Ebasco Project Manager for t

b ilLGP reports to the Vice President of Projects through the Manager of Projects. The Ebasco Project Manager is responsible for all matters relating to the overall execution of the project and is the primary contact with the Client.

17.1-9 (U)-Update Am. No. 49, 12/21/78

1 ACNGS-PSAR The Ebasco Quality Assurance Manual requires that quality related activ- 5 ities such as inspections and test s are performed under suitable conditions Ql-and using appropriate equipment. This is achieved by inclusion of neces- 11.;11 nary requirements in procurement documents. Vendors of safety related 17.2.11 systems, structures, components and services are required to submit to 17.2.16 Ebasco for review, their procedures for activities such as inspection and testing. Ebasco Vendor Quality Compliance Representatives monitor these act ivi t ies in Vendor shops to assure that the applicable Vendor procedures are implemented. The Ebasco Const ruction Quality Control Procedures are also reviewed by Materials Engineering and Quality Compliance for com- 3 pliance with the applicable codes and regulatory agency requirements. The quality control activities at the construction site are also monitored by g,g the site quality compliance personnel.

17.2J8 The Ebasco Quality Program is structured so that modifications can be made to comply with NRC regulations and industry standards as they are adopted.

The Quality Assurance program for safety related activities and services performed by Ebasco in the design, engineering, procurement, and construc-tion of the Allens Creek Nuclear Generating Station is now described in 33(U) the Ebasco Nuclear Quality Assurance Program Manual for the Allens Creek Project. This manual is a revision to Ebasco's Topical Report No. ETR-1001, which was accepted by the NRC on May 12, 1975. The revision will consist of nodifying the Ebasco Site organization. The Site Construction Quality Control activities will be under the administrative and technical control of the Ebasco Quality Assurance Engineering Department instead of the Con-h N./

t ruction Quality Control Department. This organizational structure is des-cribed in Table 17.1.28-4. All Ebasco quality assurance related activities 45 performed prior to January 1, 1977 were done in accordance with the program described in this chapter. All Ebasco quality assurance activities subse- 46(U) quent to January 1,1977 will be performed in accordance with the latest ilL6P and NRC accepted revision of Ebasco's Topical Report No. ETR-1001, '

which at present is Rev, 7 except for the site Const ruction Quality 45l49 Control organizational changes described above and other approved modifi- (U) cations listed in Table 17.1.2B-3. Later NRC approved revisions to ETR '

j46 1001 may be incorporated when deemed necessary ~ (U) if necessary to define any additional clarifications, or modifications to the project Nuclear Quality Assurance Program Manual because of HL&P con-tract requirements or to suit the unique Project conditions, they will be 45(U) submitted for approval in accordance with established provisions which ,

require execution of an authorization form involving approval of specified authorities to assure, among other things, that safety and/or quality are not sacrificed or compromised. Approved changes will be incorporated in above referenced table (s), as required. 45(U) j O

i l 17,1-24 (U)-Upda te Am. No. 49, 12/21/78

ACNCS-PSAR G

k) TABLE 17.1.2B -3 Pro j ec t -Re lat ed Clarificat ions , or Modificat ions To Ebasco Topical Report ETR-1001 Rev. 7 ,

49

1. General i Where t he word "c lient " appears within the appropriat e sect ions of EBASCO's Nuclear Quality Assurance Program Manual it shall be under- ,

st ood t o mean "Houst on Light ing 6 Power Company".  ;

2. Deleted I

49 45

3. Sect ion QA-1-4 Design cont rol (U)

Alt hough Figure I-4.1 in Sect ion QA-I-4, leaves t he required review up t o t he discret ion of t he Lead Discipline Engineer, t he Project Qualit y Assurance Engineer shall review all bidders list s, vendor proposal a and

, Ebasco purchase orders to Vendors.

( 4. Sect ion QA-I-5 Qualit y Assurance Evaluat ion of Suppliers / Cont ract or_s,_

4.1 - Paragraphs 3.1.2 and 5.1 are modified t o allow for alt ernat e methods of evaluat ion and qualificat ion of supplier's capabilit ies by met hods ot her t han audit by Ebasco. Such methods are detailed as fol-l lows:

a) Audit s of suppliers by HL&P or ot hers quali fied t o do so, b) Hist orical dat a is available subst ant iat ing the capabilit y of the supplier t o provide product s which have perforiced sat isf act orily in act ual use and were f abricat ed in accordance with an accept able qualit y assurance program. Such hist orical dat a shall only qualify suppliers l who have provided ident ical or similar product s in the past.

4.2 - Paragraphs 2.3, 4.1 and 5.1 are modified such t hat in t he event Const ruct ion Cont ract ors are awarded a cont ract before review and ap-proval of their quality assurance manual or t heir facility, but prior to st art of ar.y safet y-relat ed work, the following shall be complied with:

a) The " Terms and Condit ions" sect ion of t he Purchase Order will st ipulat e t hat t he award of t he cont ract is predicated on:

f}

\_ ,e

1) sobmit t al of const ruct ion cont ract ors qualit y assurance manual for review and comment by Purchaser, (U)-Update 17.1- 70a Am. No. 49, 12/21/78

ACNGS-PSAR TABLE 17.1.2B-4 a Organizat ional and Administ rat ive Changes to sbasco Topical Report ETR 1001 Rev. 7 49 Sit e Qualit y Cont rol Engineertng A Qualit y Cont rol Sit e Supervisor and st af f of engineers and specialistThe s are assigned t o each project const ruct ion sit e on a resident basis.

Quali,ty Cont rol Sit e Supervisor report s t o the Qualit y Program Site 19 Manager and is responsible for:

est ablishing and (a) Perf orming inspect ion in all areas of const ruct ion, enf oreing qua1it y cont rol document at ion requirement s, including pro-cedures, speci ficat ions , drawings and purchasing document s.

(b) Ident ifying and init iat ing correct ion of nonconformances to require-ment s indicat ed by t he drawings, specificat ions , codes or procedures for it ems, and reject ing nonconforming it ems and services or when necessary requiring the st oppage of work unt il such nonconformance is 45 corrected * (U)

(c ) Assisting in organizing and adminis't ering t raining seminars as re-quired t o assure proper level of qualit y cont rol .

(d) Preparing inspect ion requirement s based upon such document s as speci-ficalions, drawings, codes and st andards, as est ablished by the

- Engineering Depart ment .

(e) Supervision of Qualit y Cont rol Engineers who direct ly supervise Quali Cont rol Inspect ors / Specialist s for t he various const ruct ion disciplines (Soils, Concret e , Elect rical, flechanical, Mat erial Con-t rol, et c).

(1) Supervision of the NDE Group who are responsible for performance and/

or monit oring of nondest ruct ive t est ing act ivit ies, hecords Cent er A Quality Records Supervisor and st af f are assigned t o the project const ruc- l 49 t ion sit e on a resident basis. The funct ions of the Records Cent er Super-visor are as follows:

(a) Est abli sh

1. Project file indexing and locat ion, including ret ent ion and classificalion requirement s.

2. Filing and st orage inst ruct ions for special process records, f oilowing manuf act urer's recommendat ions and/or est ablished pract ices.

(b) Develop record audit checklist s ar.d process Records Deficiency Report s.

(U)-Update 17.1-70f Am. No. 49, 12/21/78

ACNGS-PSAR O TABLE 17.1.2B-4 (Cont 'd )  :

)

(a) Est ablish and/or int erpret NDE requirement s and accept ance crit eria l for fabricat ed and erected equipment as required.

(b) Review and comment on NDE procedures and radiographic filma submitted by manuf act urer8, sit e const ruction forces and/or Client s.

(c) Advise manufact urer and sit e const ruction forces as t o proper NDE pro-cedures, applications, t echniques, equipment and qualificat ions.

Const ruct ion Prirnary responsibilit y for const ruct ion rest s wit h t he Vice-President of Con-st ruct ion (refer t o Figure 17.1.2B-1 as amended by ETR-1001, Figure 1.2-6, i

l Revision 3 and Figure 1.2-7, Revision 3. 49 (a) Const ruct ion Managers report to the Vice-President of Const ruct ion 45 and are responsible for overali supervision and coordination of all (U) const ruct ion &ct ivit ies and services. 1 (b) The Manager of Const rurq ion Services report s t o t he Vice-President of Const ruct ion and i responsible for general supervision of t he Const ruct ion Engineer ing Group. j 1

The Manager of Const ruct ion Engineering report s t o t he Manager of Const ruct ion Services and is responsible for t he inclusion of qualit y requirement s in Construct ion Cont ract s and review of Engi-neered Document o as required by the Qualit y Assurance Program Manual.

( All const ruct ion cont ract s involving sa fet y-relat ed equipment are subject t o review by the Quality Assurance Engineering Department for compliance wit h the applicable code and regulat ory agency require-ment s and Qualit y Assurance Program requirement s).

.(c) For the individual projects, the Site Manager reports to a Construction Manager and has the responsibility for direction and coordination of all on-site activities associated with the construction of the plant.

(d) The Project Superintendent reports to the Site Manager and is responsibic for performing general site supervision of construction in accordance with drawings, specifications and contractual obligations. 49 (c) The Construction Superintendent reports to the Project a Superintendent and has the responsibility of assuring that jobsite fabrication and installation is in accordance with drawings, specifications and other prevailing documents.

(f) Area Superintendents report to the Proj ect Superintendent and are responsible for area planning and scheduling, area construction control engineering and area construction engineering.

l l 17.1-70h (U)-Update

, Am. No. 49 12/21/78

ACNGS-PSAR (g) The Senior Resident Engineer reports to the Project Superintendent and is responsible for all phases of field office and field en-gineering.

(h) The Administration Manager reports to the Site Manager, and is responsibic for management of site office services, including -

purchasing, materials administration, data processing and accounting.

49 (i) The Purchasing Administrator reports to the Administration Manager and is responsibic for the issuance and control of purchasing documents between vendors and personnel at the jobsite.

(j) The Material Administrator reports to the Administration Manager '

and is responsible for commercial receiving spection, storage and issue of materials at the site.

fh V

17.1-701 Am. No. 49 12/21/78

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i ACNGS-PSAR Applicant reserves the right to provide an acceptable alterna-tive at a later date.

C) Table 1-3, item 5, Environmental Qualification Program for l non GE supplied Class 1E equipment is described in Section 3,11 l of the Allens Creek PSAR. The Applicant will adopt the reso-lution with respect to GE supplied equipment for ACNGS as de-tailed in PSAR Section 3.11.2.3. The Applicant reserves the right to provide an acceptabic alternative at a later date.

J D) The Applicant will adopt the resolution with respect to GE i supplied equipment for ACNGS. The Applicant reserves the right to provide an acceptable alternative at a later date.

E) The Applicant will adopt the resolution with respect to GF supplied equipment for ACNGS. The Applicant reserves the right to provide an acceptabic alternative at a later date.

F) The Applicant will adopt the resolution for ACNGS. The differ-ences in arrangement of the ACNGS and GESSAR control room comp ,

plexes will not affect conformances of the former to the necessary l instrumentation and control requirements, including physical and cicctrical separation requirements to meet Regulatory

/"'N Guide 1.75, " Physical 'Indepenrience of Electric Systems".

(ms/ The Applicant. reserves the right to provide an acceptable alternative t' a later da+e l C) The Applicant commits to incorporating any design modification  !

that may be required by NRC to resolve the ATWS issue. The '

resolution of tnis issue may require implementing some or all s

of the plant modifications discussed in NUREG-0460. The plant modifications discussed in NUREG-0460 are:

1) ATWS Recirculation Pump Trip

2) liigh Capacity-liigh Flowrate Automatic Boron Injection 49 System

3) Feedwater Pump Trip

4) Alternate fligh Pressure Makeup System The ACNGS will be designed such that inclusion of these above four systems will not be foreclosed by the construction of Allens Creek.

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K-032.2 ,2 Am. No.219,_12/21/78__l

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I i ACNGS-pSAR'

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I' i

Open item No.

j. 110.6(3) In your response, you stated that for the ASME Class 1, 2, l and 3 ccmponents and supports that the peaks of dynamic j l

j -

loads associated with plant Faulted Conditions will be l

combined by the Square Root of the S,um of the Squares i Method (SRSS). In the absence of acceptable technical justification for the use of the SRSS method, our j

position is that.you should commit to combine dynamic i ' loads by' the method of absolute summation until and '

i

! if the staf f concludes that adequate technical justifi-cation has been provided for the use of the SRSS method. ,

i i l

RESPONSE

l-i l

l Applicant will apply the absolute summation method to ASME Class I, II, )

and iII components-and supports and to seismic Category I structures j except where NRC requirements permit use of SRSS method (e.g., NUREG-0484).

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M110. 6 (3)- 1 Am. No. 49, 12/21/78 L,- , , ..L_,-;._..-_..._,,__......_..._._._...__ . -..__..____._________u._________.__!

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l ACMGS-PSAR Rosponne l

This analysis was performed to investigate the effect on the ECCS l

nnalysis for the Allens Creek Nuclear Generating Station for diverting low ,

pressure coolant injection (LPCI) pumps to the containment spray mode  !

l ten minutes after a loss-of-coolant accident (LOCA) initiation. l l

l Automatic diversion of LPCI flow to containment spray has been provided  !

in response to an NRC requirement to assure containment integrity for postulated high steam flow bypassing the suppression pool. Such flow diversion would occur only if a high containment pressure (>9 psig) signal is present after ten minutes. The assumption of sufficient '

bypassing to cause such a pressure has been shown by GE to be extremely conservative and unrealistic . 2 CONCLUSION f

V The results show that the worst single failure / break typb combination is the high pressure core spray (HPCS) line break (approximately .02 ft2 )

assuming the failure of the low pressure core spray (LPCS) diesel generator (D/G) which powers one LPCS pump and one LPCI pump. This single failure / break type combination yields the highest tpeak cladding temperature (approximately 1985 F) of all the cases affected by LPCI e

diversion at ten minutes. The peak cladding temperatures experienced by the cases affected by LPCI diversion are below the limiti established in 10 CFR 50.46 (2200 F). This temperature is also below tie peak clad temperature (PCT) calculated for the break of a recirculation line (2038 F) which is not adversely affected by LPCI diversipn at ten minutes.

i p 1NED0-10977 Drywell Integrity Study: Investigation o,f Potential k Cracking in BWR/6 Mark III Containment N211.3-2 Am. No. 49, 12/21/78

ACNGS-PSAR ASSUMPTIONS

1) A maximum of two LPCI pumps (specifically LPCI "A" and LPCI "B")

can be fully diverted at ten minutes to the containment spray mode.

(NOTE: LPCI "A" shares an emergency diesel generator with the LPCS; LPCI "B" and "C" share an emergency diesel generator. The pump associated with LPCI "C" cannot be diverted to containment sprays.)

2) The standard SAR assumption of one automatic depressurization system (ADS) valve failure combined with the worst additional single failure was retained because this assumption is built into the present model. This bounding assumption yields conservatively higher calculated peak cladding temperatures (PCTs) by approxi-mately 100 F. The PCT reported on Page 1 does not include this assumption.

3) Approved Appendix K analysis models were used, except that some LPCI flow to the reactor vessel was stopped ten minutes after the accident.

GENERAL OBSERVATIONS FROM THE ANALYSES i

Only those accide'nt cases which are not reflooded to the hot node before ten minutes are affected by the assumed LPCI diversion. Once the core has been reflooded, only one ECCS pump is necessary to keep the core covered. Thus, the breaks affected include small breaks less than approximately 0.2 ft2 (depending on the break location) and outside steam line breaks (OSLB). The effect of the assumed LPCI diversion on the OSLB is small and is discussed in a later section of this report.

After reviewing the effect of diversion on the rest of the small breaks, l general statements can be made to describe the results in the area of

, interest:

N211.3-3 Am. No. 49, 12/21/78 f

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ACNGS-PSAR

1. The calculated PCTs (no LPCI diversion) in the small break regions -

O ,

affected by LPCI diversion generally decrease with decreasing break V

size. This follows from the fact that the core is uncovered for shorter periods and that the decay heat is lower at the time of uncovery as the break size decreases.

2. The maximum temperature for the assumed LPCI diversion case for any given break location occurs at approximately that break size where the LPCI system would normally inject flow into the vessel starting at 600 seconds (i.e. the assumed LPCI diversion time). Bigger breaks get some reflooding benefit from the LPCI pumps before diversion. Smaller breaks have the same ECC systems available as this maximum break, but the smaller break area has a lower calcu-lated PCT, as discussed previously. As an example, this worst break is indicated on Figure 1. A longer LPCI diversion time would have correspondingly smaller breaks where the maximum temperature would occur and hence lower calculated PCT.

3. Diverting LPCI from its ECCS flooding function does not always result in higher PCTs. When compared to no LPCI diversion, a reduction in PCT can be observed as a result of diverting LPCI if the LPCS is available. The reduction of subcooled LPCI water results in a reflooding mixture (due largely to LPCS flow) of steam and water which has higher voids. Thus, in the case where little LPCI flow is available for reflooding, even though less ECCS flow is entering the vessel, the swollen level inside the lower plenum is higher and reflooding can occur sooner. In such cases the calculated PCTs are extremely low and changes in PCT in either direction are insignificant.

4. Because this investigation is primarily concerned with small breaks, the failure of the HPCS, for non-core spray line breaks, is I

the worst single failure for this study. If the HPCS were operable, the break sizes being analyzed would reflood earlier than ten i minutes with the very small break sizes never uncovering. ,

N211.3-4 Am. No. 49, 12/21/78

ACNGS-PSAR The following break locations were considered: A) core spray line,

'v B) recirculation line, C) feedwater line, D) the steam line, and E) LPCI line. A brief summary of each analysis is provided below.

A. Core Spray Line Break (HPCS Line) - It is conservatively assumed that no flow enters the vessel through the broken line independent of the break size. For this case, the failure of the diesel generator associated with LPCS and LPCI "A" is the worst single failure since all credit for core spray cooling is eliminated. The ECC systems remaining before diversions are 2 LPCI + ADS and 1 LPCI + ADS after diversion at ten minutes. Becuase in both cases the reflooding time is based on only cooled LPCI flow reflooding the vessel, there is a longer reflooding time associated with the diverted case with reduced ECCS flow. The results of this investigation are shown in Figure 1. Because the temperature increase from the non-diverted case is a result of a loss of reflooding flow O from 1 LPCI pump, intermediate cases (loss of part of the b flow) will experience intermediate (lower) temperature increases.

This particular failure / break type combination was the most adversely affected by the assumed LPCI diversion. However,  !

the peak cladding temperatures are still below the limit of 2200 F.

1 B. Recirculation Line Break - For this break, the worst single failure is the HPCS failure, as described peviously. The ECCS remaining before diversion are 3 LPCI + LPCS + ADS and,

-after diversion, 1 LPCI + LPCS + ADS. Since in the diverted case the remaining LPCI flow is not enoegh to significantly quench the voids in the lower plenum, the mixture in the lower plenum will reflood with a higher voided mixture. This higher void fraction for the diverted case more than offsets the reduction in ECCS flow entering the vessel due to this diversion of LPCI. Hence, there is a net reduction in PCT due to a N211.3-5 Am. No. 49, 12/21/78

ACNGS-PSAR shorter reflooding' time and the recirculation line break D

(V without diversion which has already been reported is bounding relative to a line break with diversion. A representative break (.01 ft )2 was analyzed which confirmed these results.

The results of this investigation are shown in Table 1.

Intermediate cases (diversion of less than the full flow from two pumps) should result in smaller temperature decreases.

C. & D. Feedwater and Steam Line Breaks - For these breaks, the worst single failure is the HPCS failure, as described previously.

The rCCS remaining before diversion are 3 LPCI + LPCS + A05 and after diversion 1 LPCI + LPCS + ADS. For the diverted case, there will be a reduction in calculated PCT for the same reasons discussed for the recirculation line break. A representa-tive break (i.e. 01 ft ) was 2 again analyzed which confirmed the anticipated results. The results of this investigation are shown in Table 1. For both cases, insignificant decreases in calcu-g lated PCT result from LPCI diversion. The outside (isolated)

U steam line break was also considered with similar results.

E. LPCI Line Break - As in the case of the core spray line break, it is conservatively assumed that no flow enters the vessel through the broken line independent of the size. For this break, the worst single failure is the HPCS failure, as described previously. The ECCS remaining before diversion are 2 LPCI + LPCS + A05 and, after diversion, LPCS + A05 (if the break is in line "C") or LPCS + LPCI + A05 (if the break is in line "A"/or "B"). In either case there is insufficient LPCI flow to significantly quench the voids in the lower plenum.

Therefore, the core will reflood with a voided mixture. This higher void fraction more than offsets the reduction in ECCS flow entering the vessel due to diversion of LPCI, Hence, there is a net reduction in PCT due to a shorter reflooding time.

n As above, the .01 ft2 break was analyzed which confirmed the b anticipated rescits. The results of both diverted cases are shown in Table 1.

N211.3-6 Am. No. 49, 12/21/78

ACNGS-PSAR RESPONSE TO QUESTION (1)

(

The system provided for diversion of LPCI flow is a safety grade system.

' Consequently, it has a high reliability in performing its intended function. Postulation of a failure of this system to perform its function in combination with another single failure is not required under GDC 35 or 10 CFR 50.46.

RESPONSE TO QUESTION ON OPERATOR ACTION The operation of the ECC systems including diversion of LPCI to contain-ment sprays requires no operator action for at least 10 minutes following accident initiation. Ten minutes is the present licensing basis for operator manual action time following automatic actuation of the ECC system. There is no requirement either in 10CFR50.46 or GDC 35 for assuming no operator action 20 minutes after the initiation of the accident. Ten minutes continues to be the licensing basis used and supported by General Electric. It is also the basis for the containment performance evaluation as it has been for other BWR plants.

m d

i N211.3-7 Am. No. 49, 12/21/78 1.

ACNGS-PSAR TABLE 1 THE EFFECT ON THE PCT OF DIVERTING LPCI FLOW AT 2

10 MINUTES FOR VARIOUS .01 FT BREAK TYPES BREAK PCT PCT TYPE NO DIVERSION WITH DIVERSION Recirculation Line 948 F 877 F Feedwater Line 917 F 836 F Inside Steam Line 920 F 831 F LPCI Line 834*F 1) 804 964 FF( (2)

/'%

O NOTE: (1) PCT if break occurs in LPCI line "A" or "B" (2) PCT if break occurs in LPCI line "C" t

\ /

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e e O

ACNGS-PSAR PURPOSE This analysis was performed to investigate the ef fect on the ECCS analysis for the Allens Creek Nuclear Generating Station for diverting low pressure coolant I

injection (LPCI) pumps to the containment spray mode ten minutes after a f loss of-coolant accident (LOCA) initiation.

Automatic diversion of LPCI flow to containment spray has been provided in response to an NRC requirement to assure containment integrity for postulated high steam flow bypassing the suppression pool. Such flow diversion would occur only if a high containment pressure (>9 psig) signal is present after ten minutes. The assumption of sufficient t>ypassing to cause such a pressure has been shown by GE to be extremely conservative and unrealistic . The results of the drywell cracking 1

l study referenced in footnote 1 showed that for small breaks, which are l of interest in this study, no through wall cracking of the drywell would occur where the structure has only minor prior cracking damage. Therefore,

()

(

no bypass is expected to occur and the 9 psig containment pressure wi11 not be reached.

CONCLUSION The results show that the worst single failure / break type combination is the high pressure core spray (HPCS) line break (approximately .02 ft2 )

assuming the failure of the low pressure core spray (LPCS) diesel generator (D/G) which powers one LPCS pump and one LPCI pump. This single failure / break type combination yields the highest peak cladding temperature (approximately 2085 F) of all the cases affected by LPCI diversion at ten minutes. The peak cladding temperatures experienced by the cases af fected by LPCI diversion are below the limits established in 10 CFR 50.46 (2200 F).

The maximum cladding oxidation is less than 2%, well below the 17%

limit. The maximum hydrogen generation is less than 0.17%, well below the 1% limit.

3NE00-10977 Drywell Integrity Study: Investigation of Potential Cracking in BWR/6 Mark 111 Containment N211.3-15 Am. No. 49, 12/21/78

ACNGS-PSAR ASSUMPTIONS C

l l 1) A maximum of two LPCI pumps (specifically LPCI "A" and LPCI "B")

i can be fully diverted at ten minutes to the containment spray mode.

(NOTE: LPCI "A" shares an emergency diesel generator with the l LPCS; LPCI "B" and "C" share an emergency diesel generator. The pump associated with LPCI "C" cannot be diverted to containment i sprays.)

2) The standard SAR assumption of one automatic depressurization system (ADS) valve failure combined with the worst additional single failure was retained because this assumption is built into the present model. In addition, failure to account for this ADS valve failure would result in limitations on the operation of Black Fox plant which could affect plant availability. This bounding j assumption yields conservatively higher calculated peak cladding f

temperatures (PCTs) by approximately 100 F.

l 1

\

3) Approved Appendix K analysis models were used, except that some LPCI flow to the reactor vessel was stopped ten minutes after the accident.

RESPONSE TO SPECIFIC NRC CONCERNS Only those accident cases which are not reflooded to the hot node before ten minutes are affected by the assumed LPCI diversion. Once the core has been reflooded, only one ECCS pump is necessary to keep the core covered. Thus, the breaks affected include small breaks less than approximately .02 ft2 (depending on the break location) and outside steam line breaks (0SLB). The effect of the assumed LPCI diversion on the OSLB is small and is discussed in a later section of this report.

The following break locations were considered: A) core spray line, B) recirculation line, C) feedwater line, D) the steam line, and E) LPCI line. A brief summary of each analysis is provided below.

N211.3-16 Am. No. 49, 12/21/78 9

ACNGS-PSAR A.

Core Spray Line Break (HPCS Line) - It is conservatively assumed that no flow enters the vessel through the broken line independent of the break size. For this case, the failure of j

the diesel generator associated with LPCS and LPCI "A" is the l worst single failure since all credit for core spray cooling is eliminated. The ECC systems remaining before diversions are 2 LPCI + ADS and 1 LPCI + ADS after diversion at ten minutes. Because in both cases the reflooding time is based on only subcooled LPCI flow reflooding the vessel, there is a longer reflooding time associated with the diverted case with reduced ECCS flow. The results of this investigation are shown in Figure 1. Because the temperature increase from the i

i non-diverted case is a result of a loss of reflooding flow  !

from 1 LPCI pump, intermediate cases (loss of part of the l flow) will experience intermediate (lower) temperature increases, i

This particular failure / break type combination was the most adversely affected by the assumed LPCI diversion. However, the peak cladding temperatures are still below the limit of 2200 f.

B.

_ Recirculation Line Break - Because this investigation is primarily concerned with small breaks, the failure of the HPCS, for non-core spray line breaks, is the worst single failure for this study. If the HPCS were operable, the break sizes being analyzed would reflood earlier than ten minutes with the very small break sizes never uncovering. For a recirculation line break, the worst single failure is consequently the HPCS failure. The ECCS remaining before diversion are 3 LPCI + LPCS + ADS and, after diversion, 1 LPCI + LPCS + ADS.

Since in the diverted case the remaining LPCI flow is not enough to significantly quench the voids in the lower plenum, the mixture in the lower plenum will reflood with a higher voided mixture. This higher void fraction for the diverted d

N211.3-17 Am. No. 49, 12/21/78

ACNGS-PSAR l

l case more than offsets the reduction in ECCS flow entering the (3

\

/ vessel due to this diversion of LPCI. Hence, there is a net reduction in PCT due to a shorter reflooding time and the recirculation line break without diversion which has already been reported is bounding relative to a line break with diversion.

A representative break (.01 ft )2 was analyzed which confirmed these results. The results of this investigation are shown in Table 1. Intermediate e.ases (diversion of less than the full flow from two pumps) shou:d result in smaller temperature  ;

decreases.

1 C. & D. Feedwater and Steam Line Breaks - For these breaks, the worst single failure is the HPCS failure, as described under Item B.

The ECCS remaining before diversion are 3 LPCI + LPCS + ADS and after diversion 1 LPCI + LPCS + ADS. For the diverted case, there will be a reduction in calculated PCT for the same reasons discussed for the recirculation line break. A representa-tive break (i.e. 01 ft ) was 2

again analyzed which confirmed U,o the anticipated results. The results of this investigation are shown in Table 1. For both cases, insignificant decreases in calculated PCT result from LPCI diversion. The outside (isolated) steam line break was also considered with similar results.

E. LPCI Line Break - As in the case of the core spray line break, it is conservatively assumed that no flow enters the vessel through the broken line independent of the size. For this break, the worst single failure i.s the HPCS failure, as described previously. The ECCS remaining before diversion are 2 LPCI +

LPCS + ADS and, af ter diversion, LPCS + ADS (if the break is t in line "C") or LPCS + LPCI + ADS (if the break is in line i j

"A"/or "B"). In either case there is insufficient LPCI flow I to significantly quench the voids in the lower plenum. Therefore, j the core will reflood with a voided mixture. This higher void r~N fraction can more than offset the reduction in LCCS flow  !

entering the vessel due to diversion of LPCI~. In these cases l i

N2]l.3-18 Am. N o . 4 9 , 12/21/ 78

ACNGS-PSAR R

the calculated PCTs are extremely low and changes in PCT in I either direction would result in calculated PCT's which are still far below the limit. As above, the .01 ft2 break was analyzed and the results of both diverted cases are shown in Table 1.

RESPONSE TO QUESTION ON OPERATOR ACTION The operation of the ECC systems including diversion of LPCI to contain-ment sprays requires no operator action for at least 10 minutes following accident initiation. Ten minutes is the present licensing basis for operator manual action time following automatic actuation of the ECC system. There is no requirement either in 10CFR50.46 or GDC 35 for assuming no operator action 20 minutes after the initiation of the accident. Ten minutes continues to be the licensing basis used and supported by General Electric. It is also the basis for the containment performance evaluation as it has been for other BWR plants.

l RESPONSE TO QUESTION (1)

The system provided for diversion of LPCI flow is a safety grade system.

Consequently, it has a high reliability in performing its intended function. Postulation of a failure of this system to perform its function l in combination with another single failure is not required under GDC 35 or 100FR 50.46.

RESPONSE TO QUESTION (2)

As noted in items B, C, and D, for non-ECCS line breaks there is a decrease in the PCT in going from the non-diverted to the diverted case.

As noted in E, the PCT for the LPCI break is far below that for the limiting break. For LPCS line breaks, with the worst single failure of the HPCS, there are 3 LPCI pumps available before diversion and one LPCI available after diversion.

N211.3-19 Am. No. 49 12/21/78 l

- ACNGS-PSAR

')

for.the HPCS line break with fa' lure of the LPCS/LPCI diesel only 2 LPCI's .

are available before diversior and 1 LPCI is available af ter diversion.

Consequently, the HPCS line break represents the-most limiting break

• location when evaluating LOCA with diversion.

RESPONSE TO QUESTION (3)

The diesel generator failure for the LPCS/LPCI is more limiting than the diesel generator failure for the 2 LPCI's because the LPCS, as opposed to the LPCI, will rapidly reflood the core with a voided mixture. With a voided mixture, the swollen level inside the lower plenum is higher and reflooding can occur soorar. The effect of LPCI flow is to quench voids in the lower plenum therefore requiring more water to reflood the hot node. Consequently, the reflooding time with just LPCI available is longer than that with LPCS available, and the loss of LPCS relative to LPCI would have a more adverse impact on reflooding capability. An g analysis was performed for this-limiting break and a failure of the V LPCI/LPCI diesel generator and the calculated PCT's were substantially below those for the assumed worst single failure. It should be noted that CCFL effects on LPCS which have been included in this analysis are small for this break, and consequently, the LPCS diesel failure still represents the worst single failure.

RESPONSE TO QUESTION (4)

Figure 1 is a plot of peak cladding temperature vs. break area for an HPCS line break assuming an LPCS diesel generator failure. Both diverted and non-diverted cases are shown. The curve for flow diversion was generated using-data points at break areas of .005, .01, .016, .02, .03,

.04, .05, and .06 ft 2, r

Figures 4a through'4e are plots of water level inside the shroud, reactor 4

vessel. pressure, convective heat transfer coefficient, peak clad temperature, and.LPCI flow versus time.

N211.3-20 Am. No. 49, 12/21/78 l

4 ACNGS-PSAR RESPONSE TO QUESTION (5)

')

The maximum temperature for the assumed LPCI diversion case for any given break location occurs at approximately that break size where the LPCI system would normally inject flow into the vessel starting at 600 seconds (i.e. the assumed LPCI diversion time). Bigger breaks get some reflooding benefit from the LPCI pumps before diversion. Smaller breaks have the same ECC systems available as this maximum break, but the smaller break area has a lower calculated PCT. This follows from the fact that: (1) The core is uncovered for shorter periods for smaller breaks since less mass is lost through the break during the blowdown from the time the reactor water level trip setpoint is reached until the time ADS is actuated 120 seconds later allowing the LPCI to operate, and (2) the decay heat is lower at the time of uncovery for smaller breaks.

For breaks smaller than the critical break size of approximately .02 ft2 ,

the LPCI system would normally inject flow into the vessel at some time after 600 seconds. Therefore, a longer LPCI diversion time would have correspondingly smaller breaks where the maximum PCT would occur.

From the above argument two points arise: (a) Maximum PCT from diversion will occur at a smaller break for longer diversion times, (b) the smaller 1 the break the lower the calculated PCT as discussed previously. Consequently,  ;

diversion at times greater than 10 minutes will have less severe consequences  !

than diversion of 10 minutes.  !

It should also be noted that the dotted line curve of Figure 1 is bounding for small breaks (< approximately 0.02 ft )2 since only one LPCI pump (the minimum possible since LPCI pump "C" does not divert) is assumed to operate for these breaks, Consequently, if one assumed a new scenario whereby diversion occurred at times later than ten minutes, the Figure 1 curve would peak at a smaller break area and the PCT at that peak would be the same as that on the dotted line curve for that break area. '

i b)

N211.3-21 Am. No. 49, 12/21/78 l 1

l

ACNGS-PSAR TABLE 1 O

THE EFFECT ON THE PCT OF OIVERTING LPCI FLOW AT 2

10 MINUTES FOR VARIOUS .01 FT BREAK TYPES BREAK PCT PCT TYPE NO DIVERSION WITH DIVERSION Recirculation Line 948 F 877 F Feedwater Line 917 F 836 F Inside Steam Line 920 F 831 F LPCI Line 834 F 1) 804 964 FF( (2)

,}

l 1

NOTE: (1) PCT if break occurs in LPCI line "A" or "B" (2) PCT if break occurs in LPCI line "C" l

l l

l i

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O O O

ACNGS-PSAR

/ Open Item No.

()/ 361.5 In Section 9.2.5.3.2 of the PSAR you state, "In the event that the rate of sediment accumulation is such that it appears that the allowable level of accumulation will be exceeded during the life if the plant, the sediment will be removed before that allowable limit is reached." In addition to level of sediment accumulation, limits on slope of the surface of the accumulated sediments should be considered to assure that unacceptable consequences will not result from sediment flow into pump intake during design basis events. State the allowable configurations for accumulated sediments within the cooling lake and provide a preliminary description of the technical specifications that will be used to assure maintenance of acceptable sediment configurations. Include criteria, procedures, and technical specifications for maintaining sediment configura-tions.

R[iS PONSl!

The applicant will periodically inspect the Ulls to determine if unacceptable sediment buildup is occurring. Both depth of sedi-mentation and slope will be measured to determine buildup. If the limits to be established in the FSAR are exceeded, the f}

Ase Applicant will remove the excess sedimentation.

l l

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N361.5-1 Am. No. 49, 12/21/78 i

,