ML20041E683
| ML20041E683 | |
| Person / Time | |
|---|---|
| Site: | Skagit |
| Issue date: | 03/04/1982 |
| From: | Spangenberg F NORTHWEST ENERGY SERVICES CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NLN-22, NUDOCS 8203110241 | |
| Download: ML20041E683 (150) | |
Text
7 NORTHWEST ENERGY SERVICES COMPANY P O BOX 1090
- KIRKLANC WASHING ton i8033 March 4,19S2 NLN-22 t-Director of Nuclear Reactor Regulation g U. S. Nuclear Regulatory Commission s Washington, D.C. 20555 p C5fyED j
Subject:
Puget Sound Power & Light Company a D1982h '$1 Skagit/llanford Nuclear Project, Units 1 & 2 "
Y many%g Docket Nos. 50-522 and 50-523 rac 4 NRC PSAR Question Responses U Elinor G. Adensam letter to N
References:
(1)
Frank A. Spangenberg dated February 12, 1982 (2) Elinor G. Adensam letter to Frank A. Spangenberg, undated, (received February 16, 1982)
(3) Robert L. Tedesco letter to Frank A. Spangenberg, dated February 25, 1982 Gentlemen:
In reply to the above referenced letters, enclosed are our responses to the NRC's questions on the Skagit/Ilanford Nuclear Project Preliminary Safety Analysis Report Amendment 23. Formal responses to these questions will be submitted with Amendment 24 to the Skagit/IIanford Nuclear Project Preliminary Safety Analysis Report which will be submitted by April 19, 1982.
Very trul yours, 4.A. u F. A. Spangen erg Assistant Project Manager, Nuclear Encl.
ec: E. Adensam, NRC R. Tedesco, NRC M. Mallory, NRC y p\
8203110241 820304 PDR ADOCK 05000522 A PDR
S/HNP-PSAR 03/04/82 QUESTION 220.01 (3.5.3)
With respect to PSAR Section 3.5.4, missile barriers are s ~ designed as described in BC-TOP-9A, which shows the
-ductility ratio 20 for flexture in steel barriers. The staff had approved BC-TOP-9 in 1974 taking exception for ductility ratio over 10. The current staff position is
. included in SRP Section 3.5.3 Appendix "A" (Attachment 1). Confirm that you have not used ductility ratio more than 10 or justify deviation from above mentioned SRP criteria.
Also, for columns with slenderness ratio more than 10, the current SRP requires ductility ratio less than or equal to 1.0 while BC-TOP-9A mentions 1.3. Confirm that
- you will use all the ductility ratios per SRP criteria or justify the deviation, and revise the PSAR sections
'accordingly.
RESPONSE
See' revised PSAR Section 3.5.4 The use of a ductility ratio between 10.0 and 20.0 for steel beams (in cases where local and lateral buckling is
. precluded) is based on the ductilities given in the Air Amendment 24
S/HNP-PSAR 03/04/82 Force Design Manual, " Principles and Practices for Design of Hardened Structures", by Newmark and Haltiwanger (Technical Documentary Report Number AFSWC-TDR-62-138).
This publication, which is also the basis for BC-TOP-9A, gives a ductility of more than 20.0 for steel beams subjected to lateral loads.
s Amendment 24
) M .O \
093075
- b. For missiles which are generated by release of stored strain energy, the strain energy is equated to kinetic energy in determining missile velocity. The ultinatt stress of the material is used resulting in a larger amount of energy than would be present at fracture.
y Losses due to heating, friction, and the dhlaxation of the material are ignored.
c.
For missiles acted upon by a single phase liquid stream for a certain distance, kinetic energy for these missiles is determined by converting work enercy inte kinetic energy. Tha calculational techniqua used in obtaining missile velocity is found in ORNL-NSIC-22, subsection 4.1.1.
3.5.4 BAFRIEF DESIGN PROCEDUFES The tornado-generated missiles considered in the design of Seismic Category I structures are listed in Table 3.5-2.
The wall and roof thicknesses provided to resist the effects of tornado generated missiles are considered to be more than w adequate. At least 21 in. of reinforced concrete is provided
- for the missile resisting walls. At least 16 in. of reinforced concrete (f = 4000 psi) is provided for roof slabs with re- ?
w movable form;ing, or above the flutes where metal form deck is used. However, the design will consider the contribution of the permanent presence of the metal form deck to prevent spalling. If it is shown that the metal form deck is effective, the concrete thickness above the flutes will be reduced accor-dingly, but will not be less than 12 in. (Figure 3.5-4).
In general, protection for internal missiles is provided by barriers. The procedures and calculations employed in design o of missile-resistant barriers for both internal and external 5 missiles are described in Bechtel Topical Report " Design of L St uctures for Missile Impact" (BC-TOP-9A) (Ref 10)3 AdA JYa w l g gf ) W hh typical analysis to determine structural response due to kM prJP impact by an automobile (unconventional missile) will include the following steps: The force-time history of the automobile !_m 0 ggyndY ~
is obtained from Ref 10. The area on which the force acts is fhyd! assumed to be equal to the missile frontal area. The funda-w mental frequency of the combined wall and missile is calculated and the dynamic load factor is determined. An equivalent static
~
force (the dynamic load factor multiplied by the maximum force from the force-time history) is applied to the structure and the maximum reactions, such as moments and shears, are deter-mined. The capacity of the wall to resist punching shear caused by the automobile is analyzed by the conventional procedures e of the ACI Building Code ( ACI- 318 )', using an allowable punch- g ing shear stress of 4 vTT{.
- y 3.5-13
M .O ) l 1'
INSERT TO SECTION 3.5.4:
- a. In general the ductility ratio for flexure in steel will be limited to 10.0. However, values between 10.0 and 20.0 will be permitted, provided that the members are proportioned to preclude lateral and local buckling, thus insuring their ability to sustain fully plastic
- behavior. The design criteria to preclude buckling are as given i
in the AISC " Specification for the Design, Fabrication and Erection of Structural Steel Buildings," February 12, 1969.
I
- b. The ductility ratio of steel columns will be limited to 1.0 when the slenderness ratio is more than 20.0. For columns with a slenderness ratio equal to or less than 20.0, a ductility ratio less than or equal to 1.3 will be used.
a I!
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S/HNP-PSAR 03/04/82 QUESTION 220.02 (3.7.1)
Figure 3.7-3 and 3.7-4 show that more than 8 points of the time history response spectra fall below the design response spectra for 1%, 2%, 5%, and 7% damping. SRP Section 3.7.1 subsection II.1.b requires that no more than 5 points should fall below the design response spectra. Justify your deviation from the SRP position.
RESPONSE
Figures 3.7-3 and 3.7-4 are taken from Bechtel Topical Report BC-TOP-4A, Revision 3, Chapter 2, wherein the synthetic time-histories are described in more detail.
These time-histories have been extensively reviawad and approved by the NRC staff as part of the Topical Report.
Please note that the time-histories remain unchanged from the original Skagit PSAR.
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 220.03 (3.7.1)
With respect to PSAR Section 3.7.1.2 you did not demon-strate that the frequency interval for calculation of response spectra are small enough that further reduction does not result in more than 10% change in computed spectra. This is one of the requirements of SRP Section 3.7.1. Confirm that you will meet the SRP criteria and revise the PSAR accordingly.
RESPONSE
Please refer to the response to Question 220.02. The frequencies at which the response spectra have been computed are given in bC-TOP-4A, Revision 3, Chapter 2.
These frequencies are so chosen that most of the increments do not exceed 5 percent within the range of 1 to 15 cps, which is the usual range of power plant structure frequencies.
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 220.04 (3.7.2)
With respect to PSAR Section 3.7.2.11, you have mentioned that torsional effects will be considered in Category I
, structures using 3-dimensional models, or 2-dimensional models with static factors to account for torsional accelerations. Confirm that the dynamic analysis method you use, will also have the consideration of rocking, and translational responses of the structures and their foundations.
RESPONSE
See revised Section 3.7.2.1.
Amendment 24 l
ms S/HNP-PSAR 12/21/81 #~ E CN 2.24.06 either the time history method or the response spectrum 3p method. For the time history method the modal responses are combined algebraically. For the response spectrum method the modal responses are combined as described in Section (t,
3.7.2.7 of this PSAR. %P -
% 'l In the complex response method, the input forcing function e time history is separated into its frequency components by { thy .4 means of the Fourier transform. The structural responses ji are calculated in the frequency domain, and the inverse .
Fourier transform then gives the reponse time histories. 4 Some variations of this method use modal properties to f 'y describe portions of the model. 3 %
~
1 When a modal analysis is performed, the significant modes ik 5 a will be chosen on the basis of frequency, participation Q{j factor and generalized mass. Sufficient mass points will be used to M, adequatelytLU V define the mode shapes.h n es k y co- k 'lr & & t y f $ &j~ A
} 0-A Q sys A QM M42 ,. c e W ,. m a,y, m yW a ~Llo . 0 5 3.7.Z.2 Natural Frequencies and Respons@ Loads Seismic loads for Category I structures will be summarized in the FSAR. If modal superpos2' tion analysis is used, the significant mode shapes and frequencies w211 be given. If 23 complex response analysis is used, relevant transfer func-tions will be 92ven. In addition, the response spectra at major Category I equipment locations will be provided.
3.7.2.3 Procedures Used for Analytical Modeling Analytical models are developed for all Category I structures. The type of superstructure model used will be determined by the characteristics of the structure itself.
For symmetrical structures such as the Containment and Drywell, 2-dimensional lumped-mass stick models will be used. Structures which are highly complex or asymmetrical, i such as the Auxiliary-Fuel-Control Building complex, will be
! represented by 3-dimensional finite element models.
Sufficient refinement will be provided in the models to adequately define significant mode shapes.
Subsystems are assumed rigid and their masses lumped into the supporting structural system whenever significant coupling between the primary (supporting) system and the 3.7-4 Amendment 23
S/HNP-PSAR 03/04/82 QUESTION 220.05 (3.7.2)
In seismic analysis methods for Category I structures you have not mentioned the effect of differential support i movement. Confirm that your analysis methods will have i the consideration of maximum relative displacements among supports of Category I structures, systems, and components. Also discuss the extent to which your analysis method conforms with the applicable criteria of SRP Section 3.7.2.
1 RESPONSE:
See revised Section 3.7.2.1. The proposed analysis methods described in the S/HNP PSAR and BC-TOP-4A, Rev.
3, are in agreement with Section 3.7.2.1 of the SRP (NUREG 0800).
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. Amendment 24 I
l S/HNP-PSAR 03/04/82 OUESTION 220.06 ( 3. 7. 2)
In PSAR Section 3.7.2.4, it is stated that for soil-structure interaction elastic half space method will be used and a " simplified finite element analysis" will be performed as a confirmation. The words " simplified finite element analysis" are unclear to the staff.
Describe in detail the procedures, assumptions, and boundary conditions (specially affected by the site move amendment) that you intend to use in your finite element analysis. The current position of the staff about soil-structure interaction is outlined in REV 1 of SRP Section 3.7.2 issued in July 1981. Confirm that you will meet the SRP criteria, and revise the PSAR accordingly.
RESPONSE
Section 3.7.2.4 has been revised to comply with the referenced SRP criteria.
Amendment 24
S/ HHP-PSAR 13/21/81 g 4 secondary (supported) system does not occur. The decoupled subsystems are later analyzed using the response spectra generated at the supporting levels.
h 9 7.2.4 Soil-Structure Interaction The input motion, as given in Section 3.7.1, is defined at the surface level in the free field. Because the presence of the Plant structures modifies this motion, a soil-structure interaction analysis will be performed.
The ma]or Category I structures (the Auxiliary, Fuel, Control, and Reactor Buildings) will be constructed on a common basemat, approximately 20 ft thick. The soil-structure interaction analysis for these structures will be done with a combined model, using the " lumped parameter" approach. (This approach is also known as the "sub-structure," the " foundation impedance," and the "multistep" approach.) The decision to use this approach is based on the shallow embedment, relative to horizontal size, of the common basemat.
23 The analysis will consist of the following steps:
- 1) A free-field soil column analysis is performed, using the input acceleration time history defined in Section 3.7.1 as the surface control motion.
Strain dependency of stiffness and damping will be considered, using an iterative equivalent linear method.
- 2) The soil impedance functions are calculated, using the soil stiffness and damping derived from the free-field analysis. Soil layering will be
! explicitly considered.
- 3) The base and structural responses are calculated using substructuring techniques.
Soil parameter variations will be accounted for by multiple l analyses, using a realistic range of soil parameters.
A s.implified finite element analysis will be performed as a confirmation of the lumped-parameter approach. b lsO Ld' .tk< f~Lld ti fluy h q f % & M ad peakw J % %m.%g ULJ WJeg a w Q k< J, k CAce- S 0\Cto k eL udh. a%cLMA a:af A pcudove f 5 kJ po .: . k gw am su k %nkth A % W n1 p.~-
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- f 3.7-5 Amendment 23 i
S/HNP-PSAR 03/04/82 QUESTION 220.07 (3.7.2)
{ In seismic analysis methods for Category I structures you i
2 did not address the accidental torsion. Confirm that, in your seismic analysis, you will account for accidental torsion by taking an additional eccentricity of +5% of the maximum building dimension over and above the actual geometric eccentricity. This is the requirement of SRP Section 3.7.2 subsection II.11.
RESPONSE
See revised PSAR, Section 3.7.2.11.
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Amendment 24 l
4 i
210.07 u / n ie e - v b A n 12f;21/01 i
I i h 3.7.2.9 Effects of_Farameter Variations on Floor Response l Spectta_
The offects of parameter variations on floor response !
apectra shall be considered by widening the spectra, using the following procedure ' ',
Let f be the structural frequ2ncy, which is determined by
- using)the most probable material and section properties in formulating the structural model. The varsation in the .
t structural frequency is determined by evaluating the ,
individual frequency due to the most probable variation in ,'
each parameter that is of significant effect, such as soil modulus, mater 3al denalty, material stiffness, etc. The total frequency variation, 4dfj, is then determined by -
taking the square root of tee sum of squares of a minimum i variation of 0.0$fj and the individual frequency variation I (Af))n, that is: 6, i i
~
afj = (0.05fj)2 , y(jg))n 2
($_1) n 23 A value of 0.]fj is used if the actually computed valu of Af) is less than 0.10fj. i i 3.7.2.10 Use of Constant Vertical Static Factors -
i e
Constant. vertical static factors will not be used for I
Category I structures, lH5ttT NEH 3 .3.2. I l -
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S/HNP-PSAR 03/04/82 OUESTION 220.08 (3.7.4)
PSAR Section 3.7.4 shows that triaxial response spectrum recorders and triaxial time history monitors are provided at specific locations, meeting the requirement of SRP Section 3.7.4. However, for control room operator notifi-cation you have mentioned that, whenever an acceleration time history is being or has been recorded, a visual annunciation will be made in the control room.
SRP Section 3.7.4 requires that triaxial time history monitor will readout peak acceleration in the control room and the response spectrum recorder will readout values at discrete frequencies. Just the visual annunciation is not sufficient. Confirm that you will meet the SRP criteria or justify the deviation.
RESPONSE
Section 3.7.4.3 has been revised to comply with the referenced SRP criteria.
Amendment 24
A2C C?
O/EXP-PSAR -
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(3) one of the floors of the Aus111ery Buf1dinj
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where selssic Category I equipment $s .
supported -
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J.'[g i (p) 4 5.g ,a One of the Salamic Category I piping supports -
in the. Auxiliary Building t
(5) Containment saes, with indication ...
in the
.; '. .i I 4 control room. '* ',
y . . 1 33 -
C. Triavfal seismic switchea, with Indication in the ~
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control roost when OBE acceleration has been i
i exceeded, at the following locations: ; j ,]j l
r (1) Contalsvoent Base s ', .'
j
- .- {j ]
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Reactor equipment support (or piping support). a . -
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' Instruments will provide sufficiently accurate data for the iji i cubnoquent analyses of the Plant components. -
\ *j 13.'F.4.2 Loc _a tion and Dec :'ript ion of Tras trurtantation '
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5 I .f!
!The types, location, basis for selection of location, and I
',I .
cparational be capability of seismic instrumentation that will '
e t d ,4 installed for seismic Category I structures and compoa nanto oill be described in the FSAR. It is intended to 7;.,."19,I provide meismic instrumentation which, when used with the; .A-
.k Plant operating instrumentation, W111 Provide sufficient:
inforcotton to determine the Plant's capability for _ .a.
[
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, continued use following the occurrence of an earthquake. j:
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'3.7.4.3 Control _ Room operator motification .
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Whan the acceleration at the base of the Containment or at a L 2 Beactcr 3c.T.paroble equipment support OEt accelerat f on (orbothpiping support) an 'auditile 'a~ndexceeds ' visual' N the _ f ,,
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Dnnunciation will be made in the control room. In addition, ' ;
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S/HNP-PSAR 03/04/82 QUESTION 220.09 (3.7.4)
PSAR does not mention the inservice surveillance program for seismic instruments. SRP Section 3.7.4, subsection II.5 requires that each seismic instrument be demon-strated operable by the performance of test operations at specified intervals. Revise the PSAR to meet the SRP criteria or justify the deviation from same.
RESPONSE: '
See new Section 3.7.4.5.
Amendment 24
%Lo.09
- 3. 9. 9. 6 Ma. Sdua '
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S/HNP-PSAR 03/04/82 l
l QUESTION 220.10 (3.8.1) ,
l In your Amendment 23 of PSAR Section 3.8.1, concrete containment was not included. Please confirm that the site move does not affect your commitment in the PSAR including the design and analysis procedures of concrete containment, loads and loading combinations, structural acceptance procedures and the applicable codes, standards and specifications to comply with ASME Section III, Div.ision 2 code and the related Regulatory Guides. Also, please indicate that in the containment loads and loading combinations the LOCA/SRV related hydrodynamic loads in suppression pools manifested as jet loads and/or pressure loads will be considered.
RESPONSE
The site move does not affect the commitment in the PSAR.
As stated in Section 3.8.1.3, design of the containment will comply with the provisions of the ASME Section III, Division 2, Code and related Regulatory Guides. Refer to the response to Question 220.12 for applicable Regulatory Guides.
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Amendment 24 i
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l S/HNP-PSAR 03'/04/82 l 12.o . t o (Co n4 'd) l i LOCA~and SRV hydrodynamic' loads are considered: pool swell loads are included in the term "Pa", as defined in Section 3.8.1.4; SRV loads are included in the term,"L",
as defined also in Section 3.8.1.4.
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j Amendment 24 l
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S/HNP-PSAR 03/04/82 QUESTION 220.11 (3.8.1)
Provide an ultimate capacity analysis of the containment responding to the internal pressure build-up due to accidents. The guideline and the staff position on this subject is enclosed (Attachment 2).
RESPONSE
The NRC issued its Final Rule on Licensing Requirements for Pending Construction Permit and Manufacturing License Applications on January 12, 1982 (Federal Register at page 2286-2305 on Friday, January 15, 1982). The rule became effective on February 16, 1982, and is applicable to construction permit and manufacturing license appli-cations pending on that date.
S/HNP issued Appendix 1B to its PSAR in Amendments 21 and 22 responding to the requirements of NUREG-0718, Rev. 1.
Among these requirements was Item II.B.8 (4) (a) regarding the structural integrity of the containment under Factored Load or Service Level C conditions which is the same in the Final Rule as it was in NUREG-0718, Rev. 1.
The integrity of the containment was required to be maintained during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction and I
Amendment 24
S/HNP-PSAR 03/04/82 220.0 (Cont'J) that utilizes a hydrogen control system. The requirement further specified that as a minimum, containment integrity will be maintained for a combination of dead load and internal pressure of 45 psig under Factored Load or Service Load C conditions.
This requirement was met in Appendix 1B by: a) presenting the results of a preliminary structural analysis, b) committing to redesign the dome, base mat, hatches and penetrations to meet the factored or Service Level C conditions and, c) committing to a multi-faceted two-year post-CP program that would confirm the pressure capability of the S/HNP containment structure including hatches and penetrations.
It is S/HNP's position that the analysis and commitments made in Appendix 1B to demonstrate that the structural integrity of a containment designed for 15 psig (seismic plus additional loads) can also meet Factored and Service Level C conditions (dead load, internal pressure and thermal effects, only), although not meeting the specific l requirements of the ultimate capacity analysis set forth in Attachment 2 (SRP Section 3.8.1) , meets the intent of demonstrating the adequacy of the S/HNP containment beyond design basis conditions.
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 220.12 (3.8.1, 3.8.4)
Update and expand the list of Regulatory Guides that would be applied to all Category I structures (e.g., R.G.
1.94, 1.115, 1.117, 1.122, 1.136, 1.142, . . .). Address any exceptions and deviations from these Regulatory Guides and provide comments and explanations.
RESPONSE
Sections 3.8.1.3.3 and 3.8.3.2.3 have been updated to show S/HNP compliance with Regulatory Guide 1.94.
In addition, the design and analysis of S/HNP Category I structures will comply, as clarified, with the Regulatory Guides listed below.
Regulatory Guide 1.115, Protection Against Low-Trajectory Turbine Missiles, Rev. 1:
Project Position: The analysis presented in Sections 3.5 and 10.2 comply with the guidance contained in Regulatory Guide 1.115.
Regulatory Guide 1.117, Tornado Design Classification, Rev. 1:
Amendment 24 L
S/HNP-PSAR 03/04/82 220.41(Cant'd)
Project Position: Protection against the effects of a l Design Basis Tornado will be provided in accordance with Regulatory Guide 1.117, Rev. 1, with the following exception:
(1)
Reference:
Paragraph 3 of section B of the Regu-latory Guide. The Guide states that "the physical separation of redundant or alternative structures or components required for the safe shutdown of the plant is generally not considered acceptable by itself for protecting against tornado effects, including tornado-generated missiles." Separation and redundancy of the ultimate heat sink mech-anical draft cooling towers affords adequate protection to this system from the larger, less numerous vertical tornado missiles which could penetrate the fan discharge protection grating.
Regulatory Guide 1.122, Development of Floor Response 3
Spectra for Seismic Design of Floor-Supported Equipment or Components, Rev. 1:
Project Position: The requirements of Regulatory Guide 1.122 will be met. Section 3.7.2 describes a l
l method of development of floor response spectra for Amendment 24 t
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S/HNP-PSAR 03/04/82 Eto . il (Ca ni 'J)
S/HNP which complies with Regulatory Guide 1.122, Rev. 1.
1 Regulatory Guide 1.136, Materials, construction, and Testing of Concrete Containments, Rev. 2:
Project Position: Materials, Construction, and Testing of concrete containments comply with Regulatory Guide 1.136 except as follows:
(1) Regulatory Position C.8: The Summer 1980 Addenda to the ASME Code, CC-4333, 5.2, permits both production and sister splice samples. The requirement for all samples to be production splices does not ensure added integrity.
(2) Regulatory Position C.9: Staggering of splices is a design objective and will be specified wherever l possible. At certain locations, such as a penetration and blackouts, staggering is impractical.
l (3) Regulatory Position C.13: Design specifications will state that remedial measures shall be i
! undertaken or a retest shall be conducted.
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Amendment 24 i
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S/HNP-PSAR 03/04/82 1.w . n. C Co n i 'J )
Regulatory Guide 1.142, Safety-Melated Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments):
Project Position: Comply .
The PSAR will be revised to reflect these commitments within six months of issuance of a Construction Permit.
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l Amendment 24 i
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, S/HNP-PSAR 03/04/82 I
QUESTION 220.13 (3.8.1)
In PSAR you have mentioned the use of 10 CFR 50 Appendix A, GDC 2, 4, 16, 50, etc. Why has GDC 1 been omitted?
Include GDC 1 also and address its effect on your Quality i
Assurance Program.
RESPONSE
The Quality Assurance Program described in Chapter 17 of this PSAR complies with GDC 1.
See revised Sections 3.8.1.3.3 and 3.8.3.2.3.
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Amendment 24 i
,_m . %,._ _ , . . __ - . _ _-. -
D.o.13 042176 Criterion Jos .A 2, , 16, 50, 51, 52, 53, 54, 55, 56, 2A#U and 57. l
- b. NRC Regula Guides (compliance is discussed in Appendix 3A of this PSAF) . -
g Regulatory Guide 1.10 - Mechanical (Cadwefd) Splices in Reinforcing Bars of Category I Concrete Structures Regulatory Guide 1.15 - Testing of Peinforcing Ears for Category I Concrete Structures Regulatory Guide 1.18 - Structural Acceptance Test for Concrete Primary Reactor Containments Regulatory Guide 1.19 - Non-Destructive Examinatior, of Primt ry Containment Liner Welds Regulatory Guide 1.29 - Seismic Design Classification Regulatory Guide 1.46 - Protection against Pipe Whip Inside Containment Regulatory Guide 1.54 - Quality Assurance Pequirenents for Protective Coatings Applied to Water-cooled Nuclear Power Plants Reculatory Guide 1.55 - Concrete Placement in Category I Structures Regulatory Guide 1.57 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components.
Regulatory Guide 1.63 - Electric Penetration Assentlies in Containment Structures for Water-Cooled Nuclear Power Plants,
- c. Industry Standards Nationally recognized industry standards, such as those published by the American Society for Testing and Materials (A STM) , are used whenever possible to describe material properties, testing procedures, and fabrication and construction methods. The applicable ASTM specifications are listed in Section 3.8.1.7 of this PSAR.
- d. Bechtel Power Corporation Topical Reports BC-TOP-1(s3 Containment Building Liner Plate Design Report, Pevision One, December, 1972 with Supplement and Conditions per g letter dated February, 1974. w 3.8-5
042176 po.I 3 3.8.3.2.3 General Design Criteria, Regulatory Guides, Industry Standards, and Topical Reports
- a. 10 CFR 50, Appendix A - General Design Criteria for Nuclear Power lants.
y Section 3.1) (Conformance is discusse$ in I
go, r3 criterionifs./2 3, 4 and 16.
- b. NRC Regulat ry uides (conformance is discussed in Appendix 3A lC this PSAP). -
Regulatory Guide 1.10 - Mechanical (Caldweld)
Splices in Reinforcing Pars of Category I Concrete containments Regulatory Guide 1.15 - Testing Reinforcing Bars for Ostegory I Concrete Structures e
Regulatory Guide 1.29 - Seismic Design Classification e
Regulatory Guide 1.46 - Protection against Pipe Whip Inside Containment
. Regulatory Guide 1.55 - Concrete Placement in seismic Category I Structures Regulatory Guide 1.57 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components.
- c. Industry Standards o
Nationally recognized industry standards, such as those published by the American Society for Testing and Materials ( ASTM) , are used whenever possible to describe material properties, testing i procedures, fabrication methods, and construction methods. The ASTM specifications listed in Subsection 3. 8.1.7 of this PSAR are applicable to the internal structures.
t
- d. Bechtel Power Corporation Topical Reports
- BC-TOP-4-A Seismic Analysis of Structures and l
Equipment for Nuclear Power Plants, lZ Revision Three, November, 1974 l[
eBC-TOP-9-A Design of Structures for Missile Impact, ^
Revision Two, November, 1974 S
- BN-TOP-2(*) Design for Pipe Break Effects, Fevision _
Two, May 1974.
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3.8-28
S/HNP-PSAR 03/04/82 QUESTION 220.14 (3.8.1)
PSAR Table 3.8-1 shows load combinations and load factors. For service load at construction stage you have not included the wind loads. The current SRP refers to the Table CC-3230-1 of ASME Section III Division 2 Code for load combinations. The construction load in this table includes wind. Confirm that you will meet the SRP criteria and include the wind loads at construction stage and revise the PSAR section accordingly.
RESPONSE
See revised Table 3.8-1. Table 3.8-1 is now identical with ASME Table CC-3230-1.
Amendment 24 l
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\ (m TABL_t 3.0-1 *W IDAD CDsEBINATIONS AND IDAD FACTORS 4
CATFGORY D LIII Eg Pt Pa Tt To T. So Ese # 't 8o #e 8r Fe Pew Tow Servleer
/.0 Test 1. 1.0 1.0 --- 1.0 --- 1.0 --- --- --- --- --- --- --- --- --- --- ---
Construction 1. 1.0 1.0 --- --- --- ---
3.0 --- --- --- , --- --- --- --- --- --- --- l h M
- b,1 Nor mal 1. 1.0 1.0 --- --- --- ---
1.0 --- --- --- - ---
I.0 --- --- 1.0 --- ---
Severe 1. 1.0 1.0 --- --- --- ---
3.0 ---
1.0 --- --- --- I.0 --- --- 1.0 --- ---
Environmental 2. 1.0 1.0 --- --- --- ---
3.0 --- --- --- 1.0 --- 1.0 --- --- 1.0 --- ---
Factorods sever. 1. O)1.3 1.0 g# --- --- --- --- 3.0 1.5 Environmental 2. 3.0 1.3 --- --- --- ---
1.0 1.5 1.0 1.0 1.0 1.0 g.lj Estreme 1. 1.0 1.0 --- --- --- --- 3.0 --- --- 1.0 --- --- 1.0 --- ---
1.0 --- ---
Environmental 2. 1.0 1.0 --- --- --- --- 1.0 --- --- --- ---
1.0 1.0 --- ---
1.0 --- ---
UI N
Abnormal 1. 1.0 1.0*** --- --- 1.5 --- --- 3.0 --- --- --- --- --- 1.0 --- --- --- --- %
- 2. I.0 1.0 --- --- 1.0 --- --- 3.0 --- --- --- --- ---
1.25 --- --- --- --- Z
- 3. 1.0 --- ---
h 1.0 1.0 *O Abnormal /
Severe 1.
2.
1.0 I.0 1.0 1.0 1.25 ---
1.25 ---
1.0 1.0 1.23 1.23 ---
1.0 - -- b M 3.0 --- --- ---
p Environmental 3. 1.0 1.0 1.0 --- --- ---
1.0 ---
1.0 or 1.0 --- --- --- --- --- --- ---
y Abnormal /
Estreme 1. 1.0 1.0 --- --- 1.0 --- --- 1.0 --- 1.0 --- --- ---
1.0 1.0 --- --- ---
Environmental kneludes all temporary construction loading during and af ter cor.etruction of containment.
- Concrete tangential sheer not to esceed 40 pet for the containment structure described in Subsection 1.8.1
- Concrete tengential sheer not to enceed 60 pel for the containment structure described in sub=ection 3.O.1
- For Meln steen safety pellet valve 1,oede. a load factor of 1.25 shall be used. For all other live loads, a load factor of unity shall be used.
g k pu% go at M ;,o6 y y a 4 M A A d *d # . au.14 g a
w a m ya u+ e
=
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S/HNP-PSAR 03/04/82 '
QUESTION 220.15 ( 3. 8. 3)
In your Amendment 23 of PSAR Section 3.8.3, concrete and steel internal structures of steel or concrete containment, were not included. Please confirm that the site move would not affect your commitment in the PSAR including the design of containment internal structures, loads and loading combinations, structural acceptance procedures and the applicable codes, standards and specifications to comply with ASME Section III, Division 1 and 2, ACI 349, AISC and related Regulatory Guides.
For structures or structural components subject to hydrodynamic loads resulting from LOCA and/or SRV actuation, the consideration of such loads should be included. Please refer to the Appendix to NUREG-800 SRP Section 3.8.1.
RESPONSE
The site move does not affect the commitment in the PSAR to design the containment internal structures, loads and loading combinations, structural acceptance procedures and the applicable codes, standards and specifications to comply with ASME Section III, Division 1 and 2, Codes; ACI 349, AISC and related Regulatory Guides as described Amendment 24
S/HNP-PSAR 03/04/82 2.r.o.es (coni'd in the PSAR. Also, refer to the response to Question 220.12.
As stated in the PSAR, the internal structures are designed for the loads and loading combinations given in Section 3.8.6. LOCA and SRV hydrodynamic loads are considered: pool swell loads are included in the term "Pa", as defined in Section 3.8.6.1.4; SRV loads are included in the term "L", as defined in Section 3.8.6.1.1.
Load combinations and load factors are considered. See revised Sections 3.8.6.2.2 and 3.8.6.3.2.
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l Amendment 24
S/HNP-PSAR 03/ A/82 QUESTION 220.16 (3.8.4)
With respect to PShh Section 3.8.4.1.2, you only discussed wall separation between the Auxiliary Building and containment structure. There are other Category I structurer adjacent to containment. Please indicate whether there are any wall separations between them. If any, please specify how much is the wall separation.
Please provide this information and the technical bases to verify that adequate separation is provided.
RESPONSE
Wall separation between the containment and other Category I structures will be in accordance with Section 4.1 of BC-TOP-4A. See revised Sections 3.8.4.1.2 and 3.8.4.1.3 (the Auxiliary and Fuel Buildings are the only Category I structures adjacent to containment).
Amendment 24
S/HMP-PSAR 12/21/81 [)d.[h Auxiliary Building and the Containment structur.e will allow 12 free late d(corda nc e 23\2LD.lb w.% Sech,ral movement o 41 during a seismic eventg La of BC-top-4A.
Emergency core cooling systems (ECCS), Reactor Water Cleanup l System (RWCU) and Reactor Core Isolation Cooling System (RCIC) equipment are supported at the foundation level of the building in compartments with an elevated door. A controlled entrance for personnel is provided to this 12 level.
l Hatches in the concrete floor slab provide access to the RHR heat exchangers and to the secondary isolation valves for the main steam and reactor feedwater lines. The RHR heat exchangers are situated in vertical compartments on either side of the steam tunnel.
The only rooms in the Auxiliary Building which need to be designed to handle the consequences of high-energy pipe breaks are the RCIC and RWCU rooms and the main steam tunnel. The RCIC and RWCU rooms will be designed for high-energy pipe breaks including the associated pressure, temperature and Jet forces. Main steam and feedwater line 042.6 breaks are not postulated in the main steam tunnel. However the tunnel vent size has conservatively been based on the energy released from a non-mechanistic blowdown of a main steam line as discussed in Section 3.6.1.4.
The other rooms need not provide for high-energy pipe breaks l 23 since the RHR system contains high-energy steam less than 2 percent of the time the Plant is in operation. There are no other sources of internal pressurization in the Auxiliary Building and the highest pressure that can occur is atmos-pheric (except for RCIC & RWCU Room and the main steam tunnel discussed above).
As described in Section 6.5 and 9.4, the Standby Gas Treat-ment (SGTS) and Auxiliary Building HVAC Systems normally maintain the Auxiliary Building air pressure at 1/4 inch of l water vacuum. Following an accident, one of the Standby Gac 8 Treatment System (SGTS) fans controls the pressure in the Auxiliary Eailding at 1/4 inch of water vacuum by exhausting at a flow rate equal to the inleakage rate of one volume per day in the Auxiliary Building and the Enclosure Building.
In the unlikely event that the SGTS exhaust fan recircula-ting damper V005 (Figure 6.5-1) fails closed, one of the 042*65 SGTS fans will draw air from these buildings (without recir-culation) at a reduced fan flow rate due to higher system pressure drop. Consequently, a maximum vacuum of 1.7 inches of water in the Auxiliary Building will be produced using flow-pressure relationships:
3.8-35 Amendment 23
S/HNP-PSAR 12/21/81 2Ro.I4 Table 6.5-2 shows the operator action on the control failure mode. Except for the RCIC and RWCU room and the main steam tunnel discussed earlier, the maximum and minimum design pressures for the Auxiliary Building are 0 and 5 in, of 042.6 l water vacuum, respectively. This represents a margin of 300 l12 percent above the maximum vacuum that the SGTS can produce.
The Auxiliary Building HVAC System cannot produce as high a negative pressure as the SGTS.
3.8.4.1.3 Fuel Building The Fuel Building is located adjacent to the Containment and opposite to the Auxiliary Building (refer to the general arrangement drawings of Section 1.0). The building's prin-cipal function is housing the equipment and fscilities for receiving, storing, shielding, shipping and handling of fuel. The building is a Seismic Category I concrete struc-ture designed for tornado and missile protection. Figures 1.2-2 through 1.2-9 show the main structural features of the building.
The Fuel Building will be supported on the common power 23 block mat. The building is enclosed by reinforced concrete walls which support the floor framing. The central part of the building is occupied by the fuel pool and equipment com- 12 partments formed by concrete walls and slabs. Stainless steel liner plates seal the interior pool surfaces.
Liner plate joints are fitted with leak chases draining to a sump thus allowing testing and monitoring of leaktightness.
Concrete floors surrounding the pool are supported on steel beams which frame to concrete walls or to columns bearing on th he o 'oints will allow lateral 1smic movement be. tween the a' '
w ls I $c ab8o Dei +* m c. fiac t M So d^d Mc
'""^ + 5Wh &
- w. a llow oe. I em mw men t durim a seisodc vent % accuh. ace 4-
.6 5"
^
e y
- u in i n a e maintai water negative pressure by the H&V System. 12 i l of '
~
As described in Sections 6.5 and 9.4, the Standby Gas Treat-ment (SGTS) and Fuel Building HVAC Systems normally maintain 042.
the Fuel Building air pressure at 1/4 inch of water vacuum.
Following an accident, one of the SGTS fans controls the pressure in the Fuel Building at 1/4 inch of water vacuum by exhausting at a flow rate equal to the inleakage rate of one volume per day in the Fuel Building. In the unlikely event that the SGTS exhaust fan recirculating damper V005 (Figure 042*65 6.5-1) fails closed, one of the SGTS fans will draw air from this building (without recirculation) at a reduced fan flow rate due to higher system pressure drop. Consequently, a 3.8-36 Amendment 23
S/HNP-PSAR 03/04/82 QUESTION 220.17 (3.8.4)
With respect to PSAR Section 3.8.4, you didn't indicate whether masonry construction was utilized or not.
Enclosed is a copy of design criteria for safety-related masonry wall evaluation (Attachment 3). Identify any difference in requirements of materials, testing analysis, design and construction between SKAGIT/HANFORD design, and staff position. Provide justification for these differences or indicate your compliance with them.
If no Category I masonry wall construction is planned, please so indicate and neglect the technical portion of this question.
RESPONSE
In a generic letter dated April 21, 1980 to all Construc-tion Permit and Operating Licence Applicants regarding an information request on Category I masonry walls employed by plants under CP and OL review the NRC requested information as to whether concrete masonry walls will be used in any of the Category I structures of the pl' ant.
There will be no concrete masonry walls used in any of the S/HNP Category I structures. See revised Section 3.8.3.6 (which is referenced by Section 3.8.4.6).
Amendment 24
21:.r1
.. . 942175 J.e.1.4 Materiale._ Quality control. and sneetal construction Technfaces '
The internal structures are conetructed of concrete andbsteel Material properties and characteristics assuned]in design given in Table 3.8-3 Q
Ihkt W3 h 4 Sb ,.j soorg(ANS. 9d
- 3. 6. 3. 6.1 Concrete e M. j t
toncrete Migh density isconcrete the same as thatwhere aggregates t
d scribed in tubsection used, conform to ASTM C3.8.1.7.1.a. [
, 636-73, moescriptive Homenclature of Constituents of Aggregates for Aggregates for Radiation Shielding concrete."for Radiation S 3.8.3,.6.2 Reinforcing steel ;
%s Reinforcing 3.8.1.7.2. steel is the same as that described'in subsdetion '
J. 8. 3. 6. 3 structural steel .
Ctructural 3.8.1.7.3. steel is the same as that described in subsection ,
3.4.3.6.8 construction Procedures ; -
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Cubsection 3.8.1.7.5.The construction procedures are the son.e as those describ i
- 3. 8. 3. 6. 5 Quality control '
The quality control requirements are met an described ik '
- Cubsection 3.8.1.7.5.6 and Chapter 17 of this PSAR. -
3.3.3.7 : !
Testing and Inservice surveillance Requirenents With the exception of the drywell, a formal program of te' sting cnd inservice surveillance is not planned for the internal otructures. l The internal structures are not directly related .
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S/HNP-PSAR 03/04/82 9
QUESTION 220.18 (3.8.4)
~
With respect to PSAR Section 3.8.4.1.3, Fuel luilding, discuss, in detail, the design of spent fuel pool racks.
Enclosed is a copy of staff position on the " minimum requirements for design of spent fuel pool racks" (Attachment 4). Modify your analysis and design, if necessary, to agree with this position.
RESPONSE
The design of the spent fuel racks is described in PSAR Section 9.1.2 and has been accepted by the NRC in SER Section 9.1.2. The spent fuel rack and fuel pool designs have not changed.
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Amendment 24 l
S/HNP-PSAR 03/04/82 QUESTION 220.19 (3.8.3, 3.8.4, 3.8.5)
With respect to PSAR Sections 3.8.3.2, 3.8.4.2, and 3.8.5.2 applicable codes standards and specifications, it is the staff's position that ACI 349-76 code should be used in conjunction with Regulatory Guide 1.142.
Identify any deviations of Category I structural design from the requirements of the code and the Regulatory Guide and justify your deviations.
RESPONSE
ACI 349-76 code will be used in conjunction witn Regulatory Guide 1.142 for the design of safety related concrete structures.
See revised Sections 3.8.3.2.2, 3.8.3.4, 3.8.6.1.5, 3.8.6.2.1, 3.8.6.2.2, 3.8.6.2.3, 3.8.6.3.1, and 3.8.6.3.2.
Amendment 24 i
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$22dLl9 043176
- 3. 8. 3. 2 Applicable Codes, standards, and specificatione i
The following regulations, codes, standards, and specifications are used in the design of the containment internal getructures:
3.8.3.2.1 Regulations
- a. Code of Federal Regulations, Title 10-Atomic Energy Pa rt 50, " Licensing of P-oduction and Utilization i Facilities."
- b. Code of Federal Regulations, Title 29-Labor, Part 1910, soccupational Safety and Eealth Standards.'
l 3.8.3.2.2 codes and Standard Specifications Acceptance of the followine codes and standards for design or for design bases does not constitute full compliance with them.
Exceptions to these codes and standards are given in Subsection 3.8.1.7.5.
- a. Uniform Building Code (UDC) , 1973 edition, (applicatie and 1975 C ugeents (applicable portions) '
kE4- 'WE \
b' b tYkM
- c. American Institute for Steel Construction,
" Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," adopted Fe bruary 12, 1969 and Supplement Nos. 1, 2 and 3
- d. American welding Society, ' Structural welding Code" (AWS D1.1-74) .
- e. American Society of Mechanical Engineers, Boiler and Pressure vessel Code ' Nuclear Power Plant Components",
Section III, Division 2 (1975 Edition) Subsection O CC-3000. 5
- f. American National Standards Institute, "Supplemen-tary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants (ANSI M45.2.5).
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3.8-27 kl3
042176 2,10.IT 30 psig for a large pipe break and 5 psig for a small pipe break. For details of pressure and temperature transients, see Section 6.2.
3.8.3.4 Design and Analysis Procedures b
The sic techniques of analyzing the internal structures can be broadly classified into two groups:
- a. conventional methods involving simplifying assumptions such as those found in beam theory, and
- b. Those based on plate and shell theories of dif ferent degrees of approximation.
The strength methods given in the AcI M code are used for design. The internal structures are provided with connections D
~
capable of transmitting axial and lateral loads to the cortainment base slab. Table 3.8-2 lists the computer programs N used for analysis.
4 The following computer progrars may be used to evaluate the effect of radiation-generated heat on the shield structure of the i internals: !
- a. Grace II(*3 (NE- 34 8) - This program solves multigroup, multiregion, gamra ray attenuation problems for gamma ray heating and also dose rates in infinite or I semi-infinite slab shields with movable source regions.
- b. Heating II(?3 (ME- 611) - This program solves transient and/or steady state heat transf er problems in three dimensions. (cartesian, cylindrical, or arpherical coordinates system) .
In the final stages of design of the internal structures, the l proportioning of reinforcing steel in concrete structures is I based upon the specified codes of practice. The reinforcing l oteel is distributed according to common detailing methods.
I.ikewise, the selection of structural steel sections and the cethods of fabrication and connection are in accordance with ongineering codes and accepted industry practices.
_3. 8. 3. 5 structural _ Acceptance criteria Internal structures are designed for structural acceptance l criteria as outlined in subsection 3.8.3.2.
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_ _ _ _ _ _ d l L ---
, ;. S/ENP-PSAR 12/21/8A 220. t9 Ra = Pipe reactions under thernal conditions generated by the postulated break and including Ro.
Yr =
Load on a structure generated by thq reac-g tion on a ruptured high-energy pipe Eduring the postulated event. The time-dependent ,
nature of the load and the ability of the '
' structure to deform beyond yield is con- ,
sidered in establishing the structural capacity necessary to resist the ef fects of '
Y. g Yj =
Load on a structure generated by the jet impingement free a ruptured high-energy pipe during the postulated event. The time-dependent nature of the load and the ability of the structure to deform beyond yield is considered in establishing the structural capacity necessary to resist the impact.
Ya =
The energy resulting from the impact of a ruptured high-energy pipe on a structure or a pipe restraint during the postulated event. The type of impact, ie, plastic, elastic, etc., together with the ability of
' the structure to deform beyond yield is considered in establishing the structural capacity necessary to resist the impact.
3.8.6.1.5 other Definitions S - : - . :~ = . , _ a eksemmedesigerummendsaswatastemeeSamuseur .D M ai -
_ _ _ _ _ _ _ _ . . . ***=
For structural steel, S is the W required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC
' Specification for the Design, Fabrication and Erection of Structural Steel Build-ings," February 12, 1969.
U = For concrete structures, U is the section strength required to resist design loads, y
cased o}n methods described in ACI M omem, m 3%r76 .
6 de sfrcy& Afy> 4 g 3.8-45 4
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S/HNP-PSAR 12/21/81 1M I$
3.8.6.2.2 Load Combinations for Factored Load Conditions For these conditions, which represent extreme environmental, a bnorm al , abnormal / severe environmental, and abnormal /
extreme environmental conditions, respectively, the strength design method is used and the following load combinations are considered: l L4 s O + La 4 T.4 K. + CE33 or W/ or V) [3.t-9) ue D + L + To. + R., + 1.6 P u.3-io)
LA D + L o + b + R + 1.25 P., + (y,. 4 + M 2'10.19
+1.25(Eo o< W4 or d (3.g4D a , ou.e as e . (w q. p
+ ( E. 3 c< W or D (3.f l'2-)
aa us-n3 In combinations (3.8-M), (3.8- ) , 6, ll the maximum effects of P a, T,a R a* Y 1r Y, r and Ym are considered unless a time-history analysis is performed to justify otherwise.
R I *t.
For combinations (3.8-]G) to( 3 . 8 -Na-) , strains due to Ta 16 and due to the dynamic effects of Wt (tornado missile impact), Par Yrr Yj, and Y mmay exceed the allowable strains, provided there will be no loss of function of any safety-related system.
ItJ5EM >Whenever strains are permitted to exceed yield due to a f/ Log gg certain type of load, the structure is checked to satisfy that its ability to carry other loads is not jeopardized.
g The cases of L having its full value or being completely absent are both checked.
The effects of tornado-generated differential pressures and missiles are combined in accordance with BC-TOP-3-A (Ref 1). 12 3.8.6.2.3 Concrete Temperatures The limitations listed below are considered applicable only to concrete structural components:
3.8-47 Amendment 23
5/IDiP-PS AR 12/21/81 2.lC.li Y = For structural steel, Y is in the section strengt.h required to resist design loads, i based on plastic design methods described ;
I in Part 2 of AISC " Specification for the Design,FabricationandErectionofptruc-F tural Steel Buildings,' February 12, 1969. j l
'3.8.6.2 Load Combinations and Criteria for Seismic Category ;
I Concrete Structures The following presents a set of load combinations and allow-able design limits used for Seismic Category I concrete structures. To assure that the structural integrity will be ma in t a ine d, limits on the resulting stresses and the required strength capacities are considered for service loads, including earthqaake (OBE) and wind loads, and for factored loads, including earthquake (SSE), tornado, and pipe break ef f ects and various combinations thereof.
3.8.6.2.1 Load Combinations for Service Load Conditions The strength design method is used, and the following load combinations are considered: 1 ,
U = 1.4D + 1.7L y%-d/i (3.8-1) g U =
1.4D + 1.7Q + 1.9Eo $ 2 (3.8-2)
U = 1.4D + 1.7L + 1.7W (3.8-3) ,
If thermal stresses due to To and Rn are present, the 4 following combinations are also used:
U =
(0.75)(1.4D + 1.7L + 1.7To + 1.7R o ) (3.8-4)
U = 23
( 0.75) (1.4D + 1. 74 +1.9 Eo + 1. 7To + 1.7Ro)
(3.8-5)
U =
(0.75)(1.4D + 1.7L + 1.7W + 1.7To + 1.7Ro)
(3.8-6)
The cases of L having its full value or being completely absent are both checked and the following calculations are also satisfied:
0 =
1.2D + 1.9Eo (3.8-7)
U = 1.2D + 1.7W (3.8-8)
,, gg 3.8-46 Amendment 23
[ IA
xx. L9 INSE W TD 3.g 6,21 h CambmY G.8-Ib), fu QCcewk fSt. b fd 4 sav laA. ,n m/s ru, m a faA j L 644t k dev md 4 1.16 9
7 S/ENP-PSAR 12/21/61 220.11 If thernal stresses due to To and Ro are present, the following combinations are also used:
81 S:.D+L+Ro+To (3.8- M)
M I 5=D+In+En+Ro+To (3(8-M) (23 6
5 = D + L + W + Ro + To (3.8-W~i l No increase in allowable stress is raittedfodr ff h
load combinations (3.8- ), (3.8- ) and (3.8- ,
except as indicated below. l '3 If the thermal stresses due to To and Ro are secondary and self relieving, the value of S may be increased by 50 percent.
The cases of L having its full value or being completely absent are both checked.
- b. If plastic design methods are used, the following p load combinations are considered: p Y & = 1.7D + 1.7L 6 (3.8
- 7
) f y 4 3 = 1.7D + 1.71c 6 + 1.7Eo (3.8-M) 10 23 ]
y semos = 1.7D + 1.7L M + 1.7W (3.8- 5)
The cases of L having its full value or being completely absent are both checked.
If thermal stresses due to To and Ro are present, the f ollowing combinations are also to be satisfied:
$ Y Y p = 1.3(D + L + To + Ro) (3.8-W ) @
- D.
23 Y m -1.3(D+Lo+zo + To + Ro) ( 3 . 8 -= )
3.1
( 3. 8-M ) M y M = 1. 3 ( D + L + W + To + Ro) 3.4.6.3.2 Load Combinations for Factored Load Conditions The following load combinations are considered:
- a. If working stress design methods are used, the l23 applicable load combinations are
\
S/HNP-PSAR 12/21/81 ) Ad .I 9 l.6 5
- D + L.. + r. 4 R. 1 (Es, or w e o V) (3.t-zQ i .6 s : D + t. + Tu + R. + P (3 J, LU 8'N l . f. S - D + L. t T, t R. + p,. ., (y,. ., g +yj g (.3.g , u) l.'l S ' 3 + L. tT* + R $ 4 Pi + C1r + YJ +%)
+ (& 3, or w, o< V) (3.t-2,5)
- b. If plastic design methods are used, the applicable
. load combinations are: ,
N D + L. + T. t Ro + (E ss o< W .e o < V') ( 3.I- U.')
Y = .D + L + T. + R. + l . s P. (3.3 -11)
Y - D + L , +Ta + R + i.1.s P.
22 'O 1
+ ( Yr + Y3+ Y ,, ') 4 l. 2C Ea ( 3.Y -2. d l l ' D + L . + T. + R ,. + P. '
l + (Y + y', + b) + (E ss or % u v) (3.2 -29) ;
2.7. '2.9 i
In combinations (3.8-N) to (3.8-38e), thermal loads can be j g,,g I neglected when it can be shown that they are secondary and self-limiting in nature and where the material being designed for is ductile.
- A 2S '2.3 26 27 /
g,3 $
In combinati s (3.8- 5) through (3.8-9) and (3.8-3D) 16 through ( 3. 8 liha ) , the maximum effects of Pa, Ta,R, a Yj, Y r, and Y m are used unless a time-history analysis is performed to justify otherwise.
i L 22. M gq l
Qt For combinations (3.8-B) through ( 3. 8-N ), ath 49.-9 -SSP-- 1 Lu Ahr.awyh JQ,.JMh% strains due to Ta and the dynamic effects
- of W t (tornado missile impact), Pa' Yr' Yj, and Ym may 3 exceed the allowables provided there will be no loss of function of any safety-related system.
Whenever strains are permitted to exceed yield due to a
}
certain type of load, the structure is checked to satisfy l that its ability to carry other loads is not jeopardized.
l ri Crm Mw (3.I"Z'b, b M f M Y M h d 4 Asuh, m c.m tdw.mu:t M4 , 99- J e n d f e A h L--
I ,6O N bCAWd I'1
- 3.8-50 Amendment 23
./ 5/ENP-PSAR 220,l$
12/21/81 E
N- ..g...." m ..a _ . a; e * *e n y M .4+esseest t - '-- e2 14,;_--!~ -E
'r__<--e c t m = h u u - -e n p weeM9Tt h %
ThI ef fects of tornado-generated dif ferential pressures and missiles shall be combined in accordance with BC-TCP-3-A 2 (Ref 1).
I 3.8.6.3.3 Steel Temperatures I
For structural steel elements, the marinum temperatures are limited to 700*F and the allowable values are reduced by 5 percent for each 100'F increase in temperature using 100',F as the base for the allowables.
l 3.8.6.4 Procedures for Determination of the Ef fects of Missile Impact on Concrete and Steel Structures Missile barriers, whether of concrete or steel, are designed with suf ficient strength to stop the postulated missiles in accordance with BC-TOP-9-A section 3.5. To accoreplish this 12 objective a prediction of local and overall damage due to the missile impact is necessary.
Local damage prediction in the immediate vicinity of the impacted area includes estimation of the depth of penetra-tion and determination of secondary missiles that might be generated by spalling in the case of concrete targets.
Overall damage prediction includes estimation of structural response of the target to the missile impact, including structural stability and deformations.
In general, =issiles are characterised by impact velocity, missile mass, and impact area. Procedures used in determin-
) ing these parameters are discussed in Section 3.5, Missile iProtection.
3.8.6.4.1 local Damage Prediction f BC-70P-9-A is used to estimate missile penetration, perfora-tio'n, and spalling. _~
Rg A4. S St4b~ Y Y / )
]
Q (3.g. u) w g (3.s 5), k. & hi kcA L " "' M " " #
k AtSc crikrk V %@ SM- l k q' .t uu.H I
- O 1992 . 3.5-51 , , f, ,
S/HNP-PSAR 03/04/82 OUESTION 230.1 Provide a complete interpretation of all reflection and refraction lines within 5 miles of the site. Present the final processed section of the reflection data in a depth format next to your preferred interpretation. Also present the travel time curves of the refraction data from which you estimated velocity and thickness of the different layers. Compare the interpretation obtained from reflection and refraction lines covering the same area, and explain any discrepancy between the two sets if one exists and why.
RESPONSE
Available Data Seismic reflection data collected by Rockwell during fiscal years 1979 and 1980 along lines shown in Fig 2K-8 of the PSAR were reviewed by consultants to NESCO (Appendix 2L, Section 4.4.1.1, p 2L-10, S/HNP Amendment 23). Seismic refraction data were collected along lines shown in Figure 2K-3 of the PSAR by consultants to NESCO and were interpreted independently by them and correlated with data from drillholes and magnetic r.nd gravity surveys.
l l
Amendment 24
S/HNP-PSAR 03/04/82 Deficiencies of Processed Reflection Data There are several reasons why the processed seismic reflection data have not been interpreted in detail nor reprocessed into a depth format, but remain in the time domain:
- 1. The processed seismic reflection data in their present form are of poor or marginal quclity and only gross configurations of the basalt are suggested.
- 2. The processed data are unreliable because they show apparent reflections which are discontinuous and are not correlatable with the top of basalt or other geologic horizons recognized from drill holes and seismic refraction surveys.
- 3. Downhole velocity surveys and seismic refraction profiling indicate that lateral velocity changes in the post-basalt sediments are highly variable and not adequately defined to permit reliable reprocessing.
Therefore, reprocessing of the reflection data to incorporate all of the velocity variations would be a costly and time consuming operation with limited chance of significantly improving the processed records. Reducing the stack from 1200% to 300% or Amendment 24
S/HNP-PSAR 03/04/82 600% would reduce possibl.e interference between wide-angle reflections and refraction arrivals, but would probably not iaeasurably improve the quality of the reflections due to the reduction in stack.
Furthermore, as described below, the reflection data do not correlate with refraction and test boring data indicating that the reflection data as processed by Rockwell are unreliable for determining top of basalt or structure within the basalt.
A. Reflection Line 2 Near Southeast Anticline Reflection Line 2 crosses the Southeast Anticline in the vicinity of Station 300 to 340 and then extends southward towards Horn Rapids' as shown in Figure 230.1-1 (to be supplied under separate cover). The location of seismic refraction data and test borings in the vicinity of Line 2 are also shown in Figure 230.1-1. The seismic refraction data and the test borings in the vicinity of the northern portion of reflection line 2 are shown as an overlay on the reflection profile (Figure 230.1-2, to be supplied under separate cover). Although this figure l illustrates a general similarity in the configuration of the basalt surface as defined by reflection,
, refraction and test boring data, the l
d Amendment 24 j l
S/HNP-PSAR 03/04/82 reflecting horizons are not continuous and cannot be directly correlated with geologic horizons recognized by the other methods. Test Boring 4 and refraction line 1 generally confirm a rise in the basalt surface near Station 470, although the amplitude of the rise predicted by the reflection data cannot be determined because of a lack of continuity of the reflecting horizons.
B. Reflection Line 2, Horn Rapids Area The southern end of reflection line 2 is located along Horn Rapids Road (see Figure 230.1-1) where seismic refraction investigations were conducted during studies for the WNP 2 Site. The seismic reflection data (Figure 230.1-3, to be supplied under separate cover) show flat-lying reflectors between Stations 1240 and 1340, with reflectors rising to the south between Stations 1480 to 1540. Between Stations 1340 and 1480 the reflectors lack clarity and continuity, whereas the subsurface profile as determined by seismic refraction is continuous as shown on Figure 230.1-4 (to be supplied under separate cover) and an overlay to reflection line 2 (Figure 230.1-3) . The refraction profile shows a zone of complex lateral and vertical velocity changes between Stations 1340 and Amendment 24
S/HNP-PSAR 03/04/82 1480 where there are discontinuous reflecting horizons. These velocity changes render a proper interpretation of the processed reflection data difficult to impossible.
C. Reflection Line 3 with Refraction Line 8 A portion of Reflection Line 3 was specifically investigated by seismic refraction (Line 8), gravity and test borings in order to address the inter-pretation of faulting by Seismograph Service Company as described in Myers and Price (1979). As shown on Figure 2K-38 (S/HNP PSAR, Amendment 23, 1981) the seismic refraction, gravity and test boring data are consistent and define an anticlinal ridge. The seismic refraction profile and borings drilled along Line 8 are shown as an overlay on the reflection profile (Figure 230.1-5, to be supplied under separate cover). Also shown is the total bouguer gravity profile, which is consistent with the refraction profile. The refraction profile shown is descriptive of the lateral velocity changes, the greater of which occur where the reflections lack continuity. In summary, the refraction, test boring and gravity data provide a continuous profile of the basalt surface and define an anticlinal ridge, no faulting of the basalt Amendment 24
S/HNP-PSAR 03/04/82 surface is indicated and, as can be seen, the reflection profile does not accurately portray this ridge.
D. Reflection Line 10 with Refraction Line M Reflection Line 10 (Stations 120 to 430) and Refraction Line M (Stations 206 to 360) are approxinately the same location (Figure 230.1-1).
Borings S-8, S-9, S-10 and S-15 were located on or immediately adjacent to the seismic lines. A gravity profile was also obtained along refraction line M.
The seismic refraction, gravity and test boring data are in good agreement (Figure 2L-18, S/HNP PSAR, Amendment 23, 1981) and define a smoothly varying basalt surface as shown on the overlay to the reflection data (Figure 230.1-6, to be supplied under separate cover). Downhole velocity values below the near surface low velocity zone range from 7040 ft/sec in Hole S-8 to 7650 ft/sec in Hole S-10. Although there is no direct match between reflectors and the i
bedrock surface as shown on the refraction profile, the discontinuous reflectors at a time (two way) of 0.25 to 6.30 seconds show a configuration similar to l'
that indicated by the other techniques. The relatively strong reflecting horizon at 0.2 sec (two-i l
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[
S/HNP-PSAR 03/04/82 way time) between Stations 230 and 360 is apparently a horizon within the hingold Formation: however, there is no readily identifiable layer described on the boring logs that correlate with this reflector. The lack of continuous reflecting horizons, and the smoothly varying basalt profile defined by seismic refraction, gravity and test boring data, would appear to preclude any structure in the vicinity of Stations 410 to 415 as proposed by Rockwell in ST-14 (see Response 231. 3. b. 3) .
E. Reflection Line 11 with Borings and Gravity Data Although seismic refraction data have not been acquired along Reflection Line 11 (Figure 230.1-1),
gravity and test boring data define a very gently sloping basalt surface. The gravity profiles constructed from the contour map and the test borings are shown on the overlay to the reflection data (Figure 230.1-7, to be supplied under separate cover).
The depth.to rock as identified in the test borings has been converted to time using an average downhole velocity value of 7,100 ft/sec as measured in nearby boreholes. There is no continuity of the reflectors nor is there any correlation with recognizable geologic horizons between the reflection data and the Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 230.2 The shear velocity profile under WPS 1, 2, 4 (Figure 361.1, Supply System) shows a velocity contrast at a depth of 100 ft. while the shear velocity profile for your site in the same region (Figure 2.5.10 PSAR) does not show the contrast. Explain why there is a difference.
RESPONSE
Seismic refraction measurements throughout the Hanford Reservation have detected a persistent refracting horizon with a compressional velocity generally varying from 7,500 ft/see to 12,000 ft/sec between elevations 350 to 400 feet above sea level. This type of lateral variation in velocity is common for cemented layers and is probably associated with variations in porosity which existed prior to cementation. Therefore, within a layer charac-terized by a velocity range the lower or higher ends of the range are correlatable to fine or coarse material.
Compressional velocities of 10,000 to 12,000 ft/sec have been detected in the referenced refraction horizon in the WNP 1, 2 and 4 plant site area and northward towards Line
- 1. In the area north and west of Line 1, including the Amendment 24
S/HNP-PSAR 03/04/82 gravity and test boring data. The rise in the reflector at about 0.25 sec (two-way time) in the vicinity of Stations 440 to 450 amounts to approximately 125 feet (assuming an average downhole velocity of 7,100 ft/sec), which would correspond to nearly a one milligal positive gravity anomaly. Such a gravity anomaly is not present.
Seismic Refraction Travel Time Curves Seismic refraction data have been acquired over a wide area of the Hanford Site as part of S/HNP investigations (see Figure 2K-3) and previously on the UNP 1, 2 and 4 sites as documented in Appendix 2L and 2M of the WNP 1 and 4 PSAR (Amendment 9, 1974). Transmitted under separate cover are one copy each of the interpreted and uninterpreted refraction time-distance plots for those lines, all or part of which are within five miles of the S/HNP Site as shown on Figure 230.1-1. The refraction data have been obtained and processed in accordance with the procedures described in Section 4.1.1, page 2K-9 and 2K-10 of S/HNP, Amendment 23, and in Appendix 2L of Amendment 9 to the WNP 1 and 4 PSAR.
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S/HNP-PSAR 03/04/82 S/HtfP Site area, the velocity of this refracting horizon is generally in the range of 7,500 to 9,000 ft/sec. The depth to this horizon is approximately 100 feet (elevation 350 feet) in the WNP 1, 2 and 4 plant site area approximately 140 feet (elevation 390 feet) in the S/HNP Site area. The difference in elevation of 40 feet between the S/HNP site and the WPPSS site, a distance of 5 miles, is not significant.
The shear wave velocities of this refracting horizon have been measured by the crosshole technique at the S/HNP Site (Figure 2L .i, two other locations to the north and northeast of S/HNP (Figure 2L-2), and in the WNP 1, 2 and 4 plant site area. The shear wave velocity of this high compressional velocity refracting horizon is in the range of 2,500 to 3,000 ft/sec in the S/HNP area, 3,000 to 4,000 ft/sec north and northeast of S/HNP, and 4,000 to 5,000 ft/sec in the WNP 1, 2, and 4 site area. These differences, like the differences in compressional velocity, are probably due to normal lateral variations in cementation.
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i S/HNP-PSAR 03/04/82 QUESTION 230.3 The bedrock contour map based on refraction data, southeast anticline (Figure 2K-61, PSAR) does not show any faulting yet the structural contour map (Figure 2R-7, PSAR) based on boring indicc*es the existence of a fault.
Explain the reasons for the differences.
RESPONSE
The structural contour map shown on PSAR Figure 2R-7 is based on an interpretation of all of the geologic data reported in PSAR Appendix 2R (Stratigraphic Investigation of the Skagit/Hanford Nuclear Project). These data indicate the presence of a fault at depth on line 4A. In particular, core 125 intersected an anomalous thickness of Elephant Mountain basalt and several shear zones within this flow. In Appendix 2R, the fault is conservatively inferred to extend northwest to line 3 and southeast to line 4C.
The seismic refraction profiling across the Southeast Anticline did not identify any abrupt offset susjestive of faulting in the 16,000 ft/sec refracting horizon identified by test borings as Elephant Mountain basalt.
The maximum slope on the high-velocity basalt surface in Amendment 24 4
S/HNP-PSAR 03/04/82 the vicinity of the Southeast Anticline is less than 10 degrees (see Figure 2K-61 PSAR). The smooth and gentle slopes on the high-velocity basalt surface are not indicative of faulting in the basalt; accordingly, a fault is not shown on the Bedrock Contour Map Based on a
Refraction (Figure 2K-61).
i 1
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,- , . . , ,- . - - - - , . - . . - - - - - . - . . , . n. .,-.--,-_--,r., .-. _.-,-.-.,,-,,n .._,., -
S/HNP-PSAR 03/04/82
'f QUESTION 231.1 Among the numerous remote sensing surveys conducted in the Columbia Basin area are reports by Slemmons and O'Malley (1979, Revised 1980) and by Shannon and Wilson, Inc. (1980). Many of the lineaments identified in the reports are within the immediate vicinity of the 1
Skagit/Hanford site. For the most part, the above reports simply identify lineaments and do not discuss their possible origin. With respect to these reports perform an analysis, including field-truth as required, of the origin and possible safety significance of the lineaments within an approximate 5 mile radius of the Skagit/Hanford site. Include in your analysis (1) lineaments interpreted from the side looking airborne radar which was flown for the Seattle District of the Corps of Engineers in the Fall of 1979 and (2) additional lineaments not included in the above two reports, but orally described to the NRC staff at Richland, Washington on February 2, 1982.
RESPONSE
A photogeologic analysis has been made of the area within a 10-mile radius of the Skagit/Hanford Site. This analysis examined all lineaments within a 5-mile radius Amendment 24
S/HNP-PSAR 03/04/82 of the Site and all other linear features within a 10-mile radius that trend toward the Site. This analysis utilized the followir.g imagery:
Imagery Scale Source U-2, Color IR 1:135,000 NASA, 1978 U-2, Black & White 1:135,000 NASA, 1978 PSLG, Color 1:48,000 BPA, 1973 Radar Mosaic (SLAR-east 1:250,000 Dept. of Army
& west looking) Seattle D'_ strict, Corps of Engineers (Fall, 1979)
The analysis identified all linear features previously identified by Glass (1977) and Slemmons and O'Malley (1979 - 1980) within 10 miles of the Site, and several additional lineaments which had not been noted by
! previous workers (Figure 231.1-1).
l The origin of these features was determined by grouping lineaments having similar characteristics on the imagery and then field checking lineaments representative of each group. Four groups of features were identified and attributed to either previously mapped geologic structures or to one of the three folloving origins:
Amendment 24
S/HNP-PSAR 03/04/82 o Pleistocene glaciofluvial flood channeling or other glaciofluvial processes; o Eolian processes (e.g., they are the edges of dunes or the margins of stabilized dune fields);
or o Wildfire burns.
Lineaments which could not, by inspection of the imagery, be assigned either to previously mapped geologic structure or to one of the three origins noted above (Figure 231.1-2) were specifically field checked. The characteristics and origins of these lineaments are discussed below.
Feature 1 is a 2 1/4 mile-long, generally north-northeast trending curvilinear tonal alignment crossing stable and active dune areas approximately 1 1/2 miles southeast of the 200E area. Field inspection indicated tnat the northern end of the feature is the escarpment of a Pleistocene flood channel and is thus of glaciofluvial origin.
Feature 2 is a north-northeast trending, linear tonal-alignment 3 1/4 miles northwest of the Wye Barricade.
i j Amendment 24 l
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S/HNP-PSAR 03/04/82 The tonal alignment represents an irregular vegetation change that crosses active and stabilized dune fields.
At the northern end it coincides with a 15- to 20-ft high, west-facing embankment of dune sand. The feature is of eolian origin.
Feature 3 is a tonal contrast in an active dune field which trends northwest for approximately 2 miles south of the 200E area. The area to the northeast of the lineament is darker in tone than to the southwest. This feature is caused by a contrast in vegetation density in more and less active dune areas.
Feature 4 is a portion of the Cold Creek lineament defined in Shannon and Wilson (1980). The Washington -
Public Power Supply System has deteLmined that this lineament is a fortuitous alignment of various non-tectonic features.
Feature 5 is a 4 1/2 mile-long lineament that strikes approximately N350E from 3/4 mile southeast of the 200E area to just south of Gable Mountain. This feature was detected only on east-looking SLAR imagery (scale 1:250,000) where it is expressed as a narrow, subtle alignment of weaker signal return which contrasts with areas of stronger signal return on either side. The Amendment 24
S/HNP-PSAR 03/04/82 alignment does not appear on any conventional aerial photography and does not correspond to any topographic, geomorphic, vegetational, or tonal alignments.
Consequently, the feature is considered to be an artifact of SLAR instrumentation and processing.
In summary, the origin of all of the lineaments within a five-mile radius of the Site has been determined. None of the lineaments show any evidence of structural control. Field investigation has shown that they have been produced by glaciofluvial, eolian or other non-tectonic procccces. Consequently, they are of no safety significance to *he Site.
REFERENCES CITED Glass, Charles E., 1977, Remote sensing analysis of the Columbia Plateau: Amendment 23, Appendix 2R-K; Preli-minary Safety Analysis Report, WPPSS Nuclear Project
'No. 1.
Glass, Charles E., and Slemmons, David B., 1977, Imagery and topographic interpretation of geologic structures in central Washington: Amendment 23, Subappendix 2R-F, Preliminary Safety Analysis Report, WPPSS Nuclear Project No. 1.
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S/HNP-PSAR 03/04/82 Shannon and Wilson, Inc., 1980, Geologic evaluation of selected faults and linecments, Pasco and Walla Walla Basins - Washington, by Farooqui, S. M. and Thomas, R.
E.: Prepared for Washington Public Power Supply System.
Slemmons, D. B. and O'Malley, P., 1979 (Revised 1980) ,
Fault and earthquake hazard evaluation of five U.S.
Corps of Engineers dams in southeastern Washington:
Prepared for Seattle District U.S. Corps of Ent_ineers.
Amendment 24
COMPILATION OF LINEAMENTS DETERMINED Figure 231.1-1 FROM ALL IMAGERY x
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y EXPLANATION Q j.ineaments (ie. tonal, textural, and g vegetational contrasts vegetational e f alignments. )
rrTrrr topographic escarpment (hachures Indicating down hief alde)
V T.
E
- Golder Associates
LINEAMENTS NOT ATTRIBUTED TO GLACIOFLUVIAL l
.OR EOLIAN PROCESSES WILDFIRE BURN SCARS, Figure 231.1-2 i OR MAPPED GEOLOGIC 8TRUCTURE FROM '
l INSPECTION OF IMAGERY I
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EXPLANATION 4
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S/HNP-PSAR 03/04/82 QUESTION 231.2 A northwest-trending, northeast-dipping feature has been interpreted on bedrock surface (see PSAR Appendix 2R, Figure 2R-7) within approximately one mile southwest of the Skagit/Hanford Site. Weston Geophysical Corporation views this feature (see Appendix 2R, page 2R-32) as the steeper-dipping southwestern limb of the northwest trending Cold Creek syncline. Based upon a staff review of the seismic coverage (see PSAR Figures 2L-14 and 2K-3), Weston's refraction data does not extend sufficient distance to the southwest of the Skagit/Hanford site to define the geometry of the syncline. On this basis, present all geologic and geophysical data, including appropriate diagrams and stratigraphic / structural cross sections supporting your interpretation, of the synclinal origin and extent of the feature. Include with your response a discussion of why the feature cannot be of fault origin. The cross-sections should extend to the southwest beyond the Cold
, Creek lineament area.
l
RESPONSE
The geologic and geophysical data that bear on the origin and the extent of the northwest-trending slope southwest i
l Amendment 24
S/HNP-PSAR 03/04/82 of the Site consist of drilling data (Rockwell, 1979; Rockwell, 1981; S/HNP PSAR Appendix 2R), downhole geophysical surveys (S/HNP PSAR Appendix 2R), and gravity and magnetic surveys (S/HNP PSAR Appendices 2K and 2L) .
The most complete data coverage of the S/HNP study across this feature consists of a gravity survey (Figure 231.2-1). Contour maps of the gravity data in this area are included as Figure 2K-13 in Appendix 2K and Figures 2L-8, 2L-9, 2L-10, and 2L-ll in Appendix 2L of S/HNP PSAR, Amendment 23. Gravity profiles for Lines 1, 4B and 4D in the S/HNP Site area (Figures 231.2-1, 2, and 3 respectively to be provided under separate cover) show the location of the Cold Creek syncline and the southwestern limit of the Cold Creek syncline, defined by a small gravity high (probably a small anticline) on Line 4D at Station -355 and on Line 4B at Station -330. This gravity high becomes broader and less pronounced on Line
- 1. Gravity gradients indicate that dips on the southwestern limb of the Cold Creek syncline are approximately 5 degrees.
Although data collected for the S/HNP Project cover only a part of the feature, they are compatible with all of the data available from the region. These data collectively indicate that the northwesterly-trending bedrock slope is the southwestern limb of the Cold Creek Amendment 24
_._m
S/HNP-PSAR 03/04/82 syncline. In particular, the bedrock maps and structure cross sections included in reports by Rockwell (1979, Plates III-4 and III-5, 1981, Figures 8-3 and 8-7) show that the feature southwest of the Site conforms to regional trends in both strike and dip.
The gentle nature of the bedrock slope strongly suggests that the slope was produced by folding. No evidence suggests that the slope southwest of the Site may have been produced by faulting.
REFERENCES CITED Rockwell Hanford Operations, 1979, Geologic studies of the Columbia Plateau - A status report, RHO-BWI-ST-4, Myers, C. W., Price S. M., and Others: Prepared for the U.S. Department of Energy.
Rockwell Hanford Operations, 1981, Subsurface geology of the Cold Creek syncline, RHO-BWI-ST-14, Myers, C.
W., and Price, S. M., Editors: Prepared for the U.S.
Department of Energy.
Amendment 24
S/HNP-PSAR 03/04/82 OUESTION 231.3 Provide an assessment of the information included in a recently released report by the Department of Energyl on the geologic and geophysical interpretations within 5 miles of the Skagit/Hanford Site. The report suggests that several structures and geophysical anomalies within the Pasco Basin may be faults. For any structures and anomalies within 5 miles of the site determine if these features are faults. For any feature which is determined to be a fault, determine whether it is capable. Provide all bases for determination including geophysics (magnetics, gravity, reflection and refraction seismology) and geology (maps, cross-sections, borehole logs, borehole correlations, remote sensing, etc.).
RESPONSE
The applicant has not identified any information in the referenced report which might affect the geologic and geophysical interpretations within 5 miles of the S/HNP Site other than (1) those items specifically identified 1
Rockwell Hanford Operations, 1981, Subsurface geology of the Cold Creek syncline, RHO-BWI-ST-14, Myers, C. W., and Price, S. M., Editors: Prepared for the U.S. Department of Energy.
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S/HNP-PSAR 03/04/82 by the NRC staff in their informal comments providing additional detail to this question, received by the applicant on February 16, 1982; and (2) the dike-like geometric solutions determined by Werner deconvolution of.
aeromagnetic data presented in Figures B-ll and B-12 of the referenced report. The staff's informal comments have been addressed as Questions 231.3a and 231.3b. Both dike-like and fault-like Werner solutions from Figures B-ll and B-12 are addressed in the response to Question 231.3a.
l l
l P
Amendment 24 l
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S/HNP-PSAR 03/04/82 !
QUESTION 231.3a A number of possible normal faults, based upon an inter- '
pretation of aeromagnetic data, have been identified in Figure B-ll and B-12 of a report recently released by the Department of Energy (DOE).2 The location of one of the northwest-trending features (faults) is nearly coin-cident, both in trend and location, with the top of bedrock feature located about one mile southwest of the Skagit/Hanford Site. This bedrock feature is discussed in NRC Question 231.2 Provide an analysis of each of the aeromagnetically-interpreted faults (?) identified in the above report within at least a five mile radius of the Skagit/Hanford site. The analysis is to include a discussion, and the underlying bases, of the confirmation or rebuttal of the existence of each of the faults (?).
Particular emphasis should be directed toward discussing the coincidence of the relationship of the top-of-bedrock i
l feature (NRC question 231.2) and the aeromagnetic fault
(?) one mile southwest of the Skagit/Hanford site as j shown on DOE Figures B-ll and B-12. If your analysis l
l 2 Holmes, G. E. and Mitchell, T. H. Seismic-reflection and aeromagnetic surveys in the Cold Creek syncline area, Appendix B, in Myers, C. W., and Price, S. M, Editors,
! Subsurface geology of the Cold Creek syncline, R90-BWI-ST-14, prepared by Rockwell International for the
. U ited States Department of Energy, 1981.
l l Amendment 24 i
i
S/HNP-PSAR 03/04/82 confirms the existence of any of the aeromagnetically-identified f aults(?) demonstrate, by appropriate text and figures, the non-capability of the feature. Your response should include relevant portions of the appro-priate figures from the DOE report.
RESPONSE
The "possible normal faults" identified on Figure B-ll and B-12 of the referenced report are, in fact, not identified as possible faults but rather as geometric solutions determined by Werner deconvolution of aero-magnetic data. Chese solutions are presented graphically in three forms: (1) fault-like solutions, (2) dike-like solutions, and (3) structural disturbances. It is noted on page B-22 of the referenced report that:
"It should be emphasized that a fault-like solution does not necessarily mean that an actual fault is present. Rather, the fault-like solution indicates that a horizontal magnetic source terminates at a particular location. In the Cold Creek syncline, horizontal termination of magnetic sources (l ava' flows) can be caused by flow pinchout, possible abrupt changes in the magnetic properties of a flow, steep Amendment 24
S/HNP-PSAR 03/04/82 anticlinal /synelinal flanks, as well as fault displace-ment."
" Solutions of thin magnetic. layers dipping vertically
(+450) are mapped as " dikes" on the interpretive maps P
(Fig. B-ll and B-12) , although they may not represent true geologic dikes. A similar Werner deconvolution solution is obtained over anomalies caused by recog-nized anticlinal or synclinal structures. Structural disturbances, as noted on the interpretive maps, are generally solutions that do not meet the criteria for either dike-like or fault-like features. For example, structural disturbances may be mathematically resolved as a dike-like structure'on one survey level, and possibly as two fault-like solutions from another flight level which represents the two edges of a dike-like body." -
All of the Werner solutions within a 5-mile radius of the S/HNP Site are described along with other available geophysical data in Tables 231.3-1 and 231.3-2. All Werner solutions within 5 miles of the S/HNP Site which can be evaluated using other geophysical data have been interpreted as either the axes or gently-dipping flanks of folds or as not being expressed in the basalt surface and therefore not due to post-Elephant Mountain Basalt Amendment 24
S/HNP-PSAR 03/04/82 deformation. Consequently these Werner solutions do not affect the applicants geologic interpretation of the S/HNP Site vicinity. The following discussion concerns the interpretation of Werner solutions which were not used in the structural synthesis on Figure 8-8 of Rockwell (1981). Those Werner solutions are discussed in the response to Question 231.3.b.l.
Werner Solution N-53 This east-west trending Werner solution is not interpreted by Rockwell (1981). The 760-meter solution for N-53 trends normal to the Total Bouguer Gravity contours shown on Figure 2K-13 (Appendix 2K, S/HNP Amendment 23, 1981) and where N-53 intersects Line 1 (Station 80+00) the seismic refraction, gravity and land magnetic data are interpreted as indicative of a smooth
, bedrock surface (Figure 2K-54, Appendix 2K, S/HNP l Amendment 23, 1981). The 1220 meter solution for N-53 i also trends normal to the gravity contours and projects into seismic refraction Line 1 (Station 65+00). The gravity and seismic refraction data indicate a smooth bedrock surface (Figure 2K-54, Appendix 2K, S/HNP i
l Amendment 23, 1981).
Amendment 24 I
S/HNP-PSAR 03/04/82 l
i The geologic source for Werner Solution N-53 is not manifested in the bedrock surface.
Werner Solutions D-240 and SD-10 These Werner solutions are located approximately 4.5-5 miles northeast of the S/HNP Site and are not interpreted by Rockwell (1981). At Station 120+00 on seismic refraction Line 5, a slight rise is indicated south of where the projected sources for D-240 and SD-10 would be located (Figure 2L-A4, Appendix 2L, S/HNP Amendment 23, 1981). Both solutions trend perpendicular to the gravity contours (Figure 2K-13, Appendix 2K, S/HNP Amendment 23, 1981) and the bedrock contours derived from seismic refraction data (Figure 2K-14, Appendix 2L, S/HNP Amendment 23, 1981).
The geologic source for Solutions D-240 and SD-10 is not manifested in the bedrock surface.
Werner Solution N-54 Werner Solution N-54 is not interpreted by Rockwell (1981). The source for Solution N-54 is located in an area of gravity and land magnetic coverage and trends perpendicular to the gravity contours shown on Amendment 24
S/HNP-PSAR 03/04/82 Figure 2K-13 (Appendix 2K, S/HNP Amendment 23, 1981).
The potential field data are interpreted as indicative of a smoothly varying, low gradient bedrock surface. The individual geophysical profiles for Lines M, K, and 5 are compatible with a 30 southeastward-dipping bedrock surface.
This low gradient bedrock slope is a possible source for Solution N-54.
Werner Solutions N-246 and N-74 Although Solution N-246 (760 meter level) and Solution N-74 (1220 meter level) have different orientations, their locations are the same and probably have a common geologic source. Rockwell (1981, P. B-48) interprets the source of N-246 as the northern limb of the Yakima Ridge extension. Rockwell does not interpret Solution N-74.
The land magnetic and gravity data in the S/HNP Site vicinity (Figures 2L-13 and 2L-8, Appendix 2L, S/HNP Amendment 23, 1981) are indicative of a slight rise in the basalt just northeast of the source location for the northern one-half of N-74. There is no indication for a
, source for the east-trending N-246. To the southwest of Solutions N-246 and N-74, the gravity and magnetic data are indicative of a gently-sloping (~50 NE) bedrock Amendment 24
S/HNP-PSAR 03/04/82 surface which has been interpreted as the southwest flank of the Cold Creek syncline (Appendix 2L, S/HNP Amendment 23, 1981).
The geologic source for Werner Solutions N-246 and N-74 may be the 50 northeastward-dipping flank of the Cold Creek syncline or a very small bedrock high within the syncline.
Werner Solution N-88 This Werner solution is not interpreted by Rockwell (1981). The northeastern end of the source for Solution N-88 would be located at Station -380+00 on Line 4B. The land magnetic data for this line are indicative of a smooth bedrock surface at this location (Figure 2L-A9, Appendix 2L, S/HNP Amendment 23, 1981). Solution N-88 trends perpendicular to the gravity contours (Figure i 2L-8, Appendix 2L, S/HNP Amendment 23, 1981).
The geologic source for Werner Solution N-88 does not i affect the bedrock surface, l
l l Amendment 24 l
l
S/HNP-PSAR 03/04/82 Werner Solution D-23 Rockwell Seismic Reflection Line 2 and Seismic Refraction WNP Line 1 intersect the source for Werner Solution D-23.
These seismic data have been specifically discussed in the response to Question 230.1. As noted in that response, the refraction data show a flat lying basalt surface, no structural feature is indicated.
Other Werner Solutions Within 5 Miles The geologic sources of the above Werner solutions have been interpreted as being either gentle basalt slopes or as not being expressed in the basalt surface and therefore not due to post-Elephant Mountain Basalt deformation. Although there are no other geophysical data with which to interpret Werner Solutions N-245, D-227, N-248, N-246A, D-37 and N-63, their geologic sources are likely to be of similar origin.
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 231.3.b Provide an assessment of the impact (including positive and, if applicable, negative aspects) of the 1981 Department of Energy report 3 as it relates to the geologic interpretation of the Skagit/Hanford Site vicinity (within 5 miles of the site). Several specific aspects of the report to be incorporated within the assessment include:
- 1. Inferred bedrock structure (Figure 8-8, Chapter 8)
This figure shows several northwesterly-trending structures of undefined nature, one passing directly beneath the site.
Breccia, both tectonic and flow-top, has been noted at depths below 1,700 ft.
3Rockwell Hanford Operations, 1981, Subsurface geology of the Cold Creek syncline, RHO-BWI-ST-14, Myers, C. W., and Price, S. M., Editors: Prepared for the U.S. Department of Energy.
Amendment 24
S/HNP-PSAR 03/04/82
- 3. Seismic reflection anomalies (Figure B-5, Appendix 8)
Bedrock feature anomalies have been identified on seinmic reflection survey lines 10 and 11.
RESPONSE TO QUESTION 231.3.b.1:
The " structures" shown on Figure 8-8 of the referenced report have been tentatively interpreted by Rockwell (1981) not as structures but as the boundaries of "relatively intact volumes of bedrock." These boundaries are proposed on the bases of previously interpreted bedrock structure and geometric solutions determined by Werner deconvolutions of aeromagnetic data. Those Werner solutions within 5 miles of the S/HNP Site which have been used to define these boundaries have been interpreted hv Rockwell and the applicant as the axes or gently-dipping flanks of low amplitude folds.
Consequently, the " structures" shown on Figure 8-8 of the referenced report do not affect the applicants' previous geologic interpretation of the S/HNP Site vicinity.
Those " structures" tentatively interpreted by Rockwell within 5 miles of the S/HNP Site and designated as features "A" through "E" on Figure 231.3-1 (attached) and are discussed individually below.
Amendment 24
S/HNP-PSAR 03/04/82 Feature "A" The basis for this feature is Werner Solution D-22 which is shown on both Figures B-ll and B-12 (Rockwell, 1981).
Rockwell interpreted this anomaly as a deeply buried, asymnetric anticline (Rockwell, 1981, p. B-53). The gravity and seismic refraction data acquired for the S/HNP Site along Line 4A-1 (Figure 2L-16, Appendix 2L, ,
S/HNP Amendment 23, 1981) confirm the existence of a low amplitude bedrock rise with gently sloping flanks. The central and northern portions of D-22 are also coincident with a northwest-trending, low-amplitude gravity high in the vicinity of Station -40+00 on Line 1 and Station
-210+00 on Line 2 as illustrated on Figure 2L-8 (Appendix 2L, S/HNP Amendment 23, 1981).
Feature "A" is interpreted to be coincident with the crest of a low amplitude anticlinal high.
Feature "B" Feature "B" (Figure 231.3-1) is based upon an alignment of Werner Solutions D-28, N-73 and N-269 (Rockwell, 1981,
- p. 8-23). There is no additional geophysical data over Solutions D-28 and N-269, but Rockwell interpreted this to be a small fold (Rockwell, 1981, p. B-53).
Amendment 24
S/HNP-PSAR 03/04/82 Solution N-73 is only present on the 1220 meter level and is partially coincident with Solutions D-2L and N-269, therefore, Solution N-73 probably has a common source with Solutions D-28 and N-269. The Total Bouguer Gravity data for this portion of the Hanford Site (Figure 2L-8, Appendix 2L, S/HNP Amendment 23, 1981) only include the southeasternmost end of Solution N-73. This Werner solution is coincident with a very slight gradient on the southwest flank of a small gravity high. The ground magnetic data for Line C are compatible with a small bedrock rise.
The ground geophysical data confirm the Rockwell interpretation that Solutions D-28, N-73 and N-269 are due to a small, low amplitude fold.
Feature "C" The feature labeled "C" on Figure 231.3-1 is coincident with the May Junction Linear as first described by Myers and Price (1979). This bedrock structure was studied extensively during recent field studies for the S/HNP Project. Gravity, land magnetic and seismic refraction data (Appendix 2K, S/HNP Amendment 23, 1981) as were acquired to investigate the May Junction structure.
Amendment 24
S/HNP-PSAR 03/04/82 The results of the geophysical studies characterize Feature C as an eastward dipping monoclinal fold.
Feature "D" The bases for this feature are the interpreted contin-uation of the May Junction structure and Werner Solution N-242 (Rockwell, 1981, p. 8-23). Rockwell postulates a small fold at the source for Solution N-242 based upon a nearby reflection line (Rockwell, 1981, p. B-53) . A low amplitude gravity high is indicated one-half mile southeast of the location for Solution N-242 (Figure 2L-8, Appendix 2L, S/HNP Amendment 23, 1981). The land magnetic data collected between Stations -50+00 and
-90+00 on Line X and Stations -10+00 and -50+00 on Line 4 (Figure 2L-A12, Appendix 2L, S/HNP Amendment 23, 1981) are compatible with the interpretatior of a low amplitude, anticlinal fold.
The source for Solution N-242 is interpreted to be s northeast-trending, low amplitude fold, and appears to be a saddle in the Cold Creek syncline.
Amendment 24
S/HNP-PSAR 03/04/82 Feature "E" Feature "E" (Figure 231.3-1) coincides with the interpreted extension of the Yakima Ridge structure (Rockwell, 1981, p 8-22) and Werner Solutions D-29 and N-74 (Rockwell, 1981, p B-48). Solution D-29 is concident with a gravity high located approximately 2 miles west-southwest of the S/HNP Site (Figure 2L-8, Appendix 2L, S/HNP Amendment 23, 1981). The Total Bouguer Gravity contours delineate a northwest-trending high in this area (Figure 2L-8, Appendix 2L, S/HNP Amendment 23, 1981, and Solution N-75 is coincident with the projected crest of this gravity high.
The only other geophysical data in the vicinity of Solution N-75 is Rockwell Reflection Line 1 (Table 231.3-2) (see response to Question 230.1). The reflection horizons on Line 1 lack continuity in the area of Solution N-75.
The gravity data support the interpretation that Solutions D-29 and N-75 are the result of a subsurface
- anticlinal ridge that appears to be asymmetric to the northeast. The steeper northeast limb is the southwest flank of the Cold Creek syncline.
Amendment 24
S/HNP-PSAR 03/04/82 RESPONSE TO QUESTION 231.3.b.2:
Brecciated core noted in boreholes DC-8 and DC-12 are considered by Rockwell (1981) to be minor features unrelated to large displacements. The following excerpt from Chapter 6 of the Rockwell report (p. 6- 3 ) discusses these breccia zones:
In general, tectonic breccias are infrequent in the thousands of feet (meters) of core drilled in the Cold Creek syncline area and elsewhere in the Pasco Basin.
The breccia zones that were identified (Table 6-1) are generally intact and <4 in. (<10cm) in thickness, although some are slightly thicker (Figure 6- 2) . They appear in all deep boreholes within the Hanford Site and are principally in the Grande Ronde and Wanapum Basalts. As noted in the next chapter, such small tectonic breccia zones and their associated fractures are viewed as typical strain features of folded basalt and should be expected within the limbs of any of the Yakima folds, including the Cold Creek syncline. None of the tectonic breccias examined are judged as being associated with large displacements. This conclusion is based on comparisons with surface exposures of similar breccias and the lack of anomalously thick Amendment 24
S/HNP-PSAR 03/04/82 basalt flows that would be expected if the section were repeated.
These brecciated cores are, therefore, considered to have no implications regarding the interpreted geologic structure in the S/HNP Site vicinicy.
REFERENCED ITEMS Rockwell Hanford Operations, 1981, Subsurface Geology of the Cold Creek Syncline , RHO-BWI-ST-14, Myers, C. W.,
and Price, S. M., Editors: Prepared for the U.S.
Department of Energy.
RESPONSE 231.3.b.3:
As discussed in the response to Question 230.1, Rockwell seismic reflection records collected during fiscal years 1979 and 1980 are of marginal quality for any structural interpretation. Other data in the vicinity of anomalies identified on seismic reflection Lines 10 and 11 would appear to preclude the existence of structures such as those proposed by Rockwell at Line 10 station 410 to 415 -
and Line 11 Stations 440 to 450.
Amendment 24
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i PUGET SOUND POWER & UGHT COMPANY l SKAGli / HANFORD NUCLEAR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT FEATURES FROM RHO-BWi-ST-14 (FIG.8-8)
(AFTER ROCKWELL (1981) FIG. 8-8) FIGURE 231.3-1 l
l
S/HNP-PSAR 03/04/82 l
1 l
QUESTION 240.1 (2.4)
Throughout Section 2.4 you refer to the top of basemat elevation. The staff is interested in access openings and access floor levels in regard to potential flood levels and potential flooding of safety related equipment. Please explain the relationship between the top of basemat elevation and plant grade and access floor elevations.
RESPONSE
Top of basemat, elevation 527'-0", is the minimum floor elevation which has access openings into safety related structures, except for the diesel generator fuel oil tank vault, as noted in Table 2.4-1. Finished plant grade at safety related structures is elevation 526'-6" (see Figure 2.4-3) . S/HNP flood design considerations will be as described in PSAR Section 2.4.2.2.
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 240.2 Discuss the potential for volcanic ash accumulation in the safety-related Mechanical Draft Cooling Tower Basins and provide an estimate of the maximum depth,'and the basis for your estimate.
RESPONSE
As indicated in Section 2.5.1.2.6.1 (page 2.5-103) of the WNP-2 FSAR, Amendment 18, the maximum volcanic ashfall rate at the Hanford Reservation is established to be 3 inches in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The design basis ashfall scenario for the S/HNP is conservatively six inches in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
As discussed in Section 9.2.5.2 of the S/HNP PSAR, the volume of the Ultimate Heat Sink basin required for at least 30 days cooling of the reactor includes a one foot depth for sedimentation resulting from suspended solids.
Based on the conservative assumption that a 6" ash fall occurs in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the airborne particulate concentration would be approximately 750 mg/m3 which results in maximum cooling tower volcanic ash intake of l about 66,000 lbs during the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. With uniform distribution across the basin floor assumed, the height of the ash accumulation would be significantly less than l
Amendment 24 l
i l
S/HNP-PSAR 03/04/82
]
1 l
1 1 4 the one foot allowance for sedimentation. This is ebnsideredtobeanadequateallowanceforvolcanicash
! accumulation.
1 1
i i
i l
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Amendment 24
1 S/HNP-PSAR ,,_ 03/04/82 l
QUESTION 240.3 i
Provide a commitment to operationally test the safety-related Mechanical Draft Cooling Towers to verify the parameters used for preliminary estimates of maximum temperature and water use.
RESPO'ISE :
The design of the UHS cooling towers for S/HNP has not been finalized to date as the towers have not yet been specified or purchased.
Design parameters used to construct mechanical draft i'
cooling towers are very well understood today so that a
- high level of confidence exists that the towers' performance capability will meet or exceed cooling requirements for maximum conditions.
The S/HNP Standby Service Water System incorporates many conservative design assumptions, all of which will work to produce system components with extra margins of capability and a system that will generously exceed the requirements for handling the maximum heat loads at the maximum ambient conditions. Examples of these design conservatisms are listed below:
l Amendment 24 l
S/HNP-PSAR 03/04/82 (1) Fuel pool heat loads used for determining the cooling requirements have been conservatively assumed to be over one-third greater than that actually calculated as necessary for maximum cooling conditions.
(2) Fuel pool decay heat and core decay heat are based upon a core power level of 4100 MWt whereas maximum licensed power will be 3800 MWt.
(3) Fuel pool initial (maximum) heat rate is assumed to be constant over the 30 day period, whereas heat rate actually decreases with time.
(4) UHS basin water volume is determined assuming all heat will be dissipated by evaporation, whereas other mechanisms beside evaporation will actually dissipate the heat loads. For instance, some heat will be dissipated by heating the stored basin water.
(5) The volume of water in the standby service water pump sump located below the UHS basin bottom required for sump submergence, was not included in calculating the 30 day water volume.
Amendment 24
S/HNP-PSAR 03/04/82 (6) The UHS basin storage adequacy analysis shows a significant portion of the total water inventory will remain at the end of the 30 day period.
(7) The 30-day water requirement corresponds to a water volume of 9.17 x 106 gallons. The transient analysis was performed based on a basin water volume of 8.24 x 106 gallons. The actual water volume of 9.17 x 106 gallons provides a conservative margin over that required.
Further details on UHS basin sizing and the method of analysis are provided in PSAR Section 9.2.5.
Based on the above discussion on the high level of confidence in the safety-related mechanical draft cooling towers' performance capability and Standby Service Water System design parameters' conservatism, committing to operationally test the Standby Service Water System mechanical draft cooling towers to verify preliminary estimate parameters will be discussed at the FSAR stage, should testing remain a requirement at that time.
Amendment 24
l S/HNP-PSAR 03/04/82 ,
I I
QUESTION 271.1 Page 3.10-1 of Section 3.10, and page 3.10-3, state that the seismic qualification of electrical equipment supplied, other than by General Electric, will be in accordance with IEEE 341-1971 (and NRC Staff Technical Position EICSB No. 10). However, both revisions 1 and 2 to the Standard Review Plan (SRP) Section 3.10 state that, for plants for which the construction permit application was docketed after October 27, 1972, the qualification of electrical equipment should be in accordance with IEEE 344-1975 and Regulatory Guide 1.100.
Also, the 251 NSSS GESSAR, referenced in the Skagit/Hanford PSAR, references IEEE 344-1975. There-
RESPONSE
i l See revised Sections 3.2, 3.10, 7.1 and 7.3.
l Amendment 24
thel 2 3.2-1 '
esteet le of 39 .
N 931.1-1973 code for Power Piping ..
9W44 =
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i Ittt-300 e Ittt-100. Stenderd criterle for close It Electrical fretees for nuclear Power Generating Stettone - 1974.
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3.10 SEISMIC DESIGN OF CATEGOkY I INSTRUMINTATION AND CLASS IE ELECTRICAL EQUIPMINT
%W (See 251 NSSS GESSAR for complete discussion of NSSS equipment seisnic design.) (
(Tor electrical motors driving mechanical equipment, see
. Section 3.9.)
Seismic Category I instrumentation and Class IE electrical equipment will be designed to operate during and after an Operating Basis Earthquake (OP E) and a Safe Shutdown Earthquake (SSI) . This will be demonstrated by either one or a corbination of both of the following two methods:
- a. Prediction of the instrument or electrical equipment performance by mathematical analysis
- b. Test under Jimulated seismic conditions.
Scanumind few.htts An ev et ription of the methodology is provided in IEEE~ 344- AJ r
197 . "IIEE @wrow f or Seismic Qua li fication f or Class J : - -- -'r ' 1- d Equipment for Nuclear Power Generation Stations." JE Testing will be the principal qualification method. Analysis witheat testing will be acceptable only in those cases where structural integrity alone can assure the intended function; l
%./ where electrical equipment must function, testing will be I, performed. When testing alone is impracticable, a combination 8 of test and analysis will be used.
3.10.1 SEISMIC.__ DESIGN CRITEFIA 3.10.1.1 Seismic Cat eoory _ I Equi onent Identification Defer to Section 3.2.1 for a listing of all Seismic category I Instrumentation and Class IE Electrical Equipment requiring seismic qualification.
3.10.1.2 General feismic Desien criteriff All the Plant Seismic Category I Instrumentation and Electrical Equipment will be designed to resist and withstand the ef fects of the postulated earthquakes. For the safe Shutdown Earthquake (SSE) defined in Section 3.7.1, Seismic Category I
~*
Instrumentation and Electrical Equipment will be designed to withstand the ef fects of the earthquake without functional impai rment.
3.10-1
.. Ill.l
__a*- __
. _ ~ . .
3.10.1.3 comp 2fance with Seis-je Ouelificatier. Pe uire e-t3 _
see 251 NSSS GESSAR.
k r
3.10.1.3.1 Equipment Supplied other than by GI Qualificatien and doeurentation procedures used for Seisrie
" Category I electrical and instrurentation equip.e..t, supplied w by other than GE, will meet the provisions of IEEE Standard 344-1975 - '
- :- r_ -
1'I - ,- ,.-
E 9
M Tm ;::r- C : ;~ 2; R M.Jsideut2. k.....VNRL A% wle h O u t l.let ' d 3.10.2 I L SIISMic ANALYSES, TESTING PPDCEDCFES, AND ris:FA1ur utxsvFis The following sections outline the seismic analyses, testing procedures, and restraint ressures f or the Sei sr.d e category I instrume ntation and electrical equipment.
3.10.2.1 Seis-ie cataeory I Equirrent Tne following procedures will be applicable to the analysis 2
of seisr.ic design adequacy of Seismic Category I instru.en- _
tation and electrical equipment, including supports such as .~
cable tray supports, battery racks, instrument and control
- consoles.
l l
Seis.ic specification will be provided to the vendor with appropriate response-spectrum curves at the related floor elevations and instructions on their use in qualifying the specified equipment and components.
l I
The general approach employed in the dynamic analysis of seismic Category I equipment and component design will be based on the response-spectrum technique, where applicatie. The time-history I analysis of seismic Category I wtructures generates in-structure i
' response-spectrum curves and time histories at various support elevations for use in the analysis of systers and equipment.
l At each level of the structure where vital items are located, horizontal spectra for each of the two major axes of the structure and a vertical response spectrue will be developed.
Simplified analytical models will be used for analysis of syster.s and equipment; however, where one or two degrees-of-freedom
. . . models do not provide a suitable representation of the systers or equipment under consideration, multi-mass models will be used in
. 10-3
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- l. bicfwties not$ the M..eng Dr$13 ewppi.ev this informe>on $K4 CIT NUCLE Am POWE R PROJE CT l System aw the b yven in Fyre 7.12 of PHELiMINA AY SAF ETY Flameceb.f ty Controf ANALYSIS REPORT l
system fin t Ww N153 251 GES$AR. f I
of ww wi aurtion ' erstem sho. n A r m. = ,,",
hi F,re 7.12 et gg, a e Au.. Sadt HVAC. lv CODES AND STANDARDS sWEIS 251 GE ESA M1. POST Loc < owerion.
- APPLsC A81LITY MATRIX
. y ,
'w See St M . C.: Tre.t a Mort a, Comp, aree see w Also see se .tsoa 6 " 2.3 l
w sectio.. 8 3 1.3.1 .
FIGURE 7.12 for e.coptioem. "
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2 11.\
3:::'r
,,, 7.3.2.3.4 Qualifying class 2 Electric Equiprent IEEE-308 (1971), *1EEE standard criteria for cla ss I) Electric geysterr.s for Nuclear Power Generating stations." Ref er to Subs ection 8. 3.1. 2.1.
IEEE-323 (1971), 'Irrt standard for Qualifying class I E3e:-
'tric Equipment for Noelear Power Generating Stations." Refer -
to Section 8.3.1.3. gg tj gh E i
corpliance with IEEE-3sa/(197h , *2EEE3 &rede for selsr.ie l Qualification of class I rte ri- Equipment for Welear Power i]k Generating stations = is described in section 3.10. L i
1 i
7.3.2.3.5 conformance to 10 CTR 50 Appendix A
- a. Criterion 13 - cooling water to essential corponents during reactor norr.a1 shutdown isolation modes and following a locA is assured by monitoring appropriate signals which start the SSWS when required.
I
- b. Criterion 20 - The SSWS control System will automatically initiate appropriate action with no operation action required.
- c. Criterion 21 - The high funettonal reliability, redundancy, and inservice testability of the two separate trip systers assures that the SSWS control system will function as required,
- d. Criterion 22 - The two redundant separate trip systems are physically and electrically separated so that no single failure can prevent an initiation of the asWS.
- e. Criterion 23 - The system logic and actuator signals
~are designed to ensure that the SSWS control system will fail to a safe state. Slotor operated valves will fail as-is on loss of power.
- f. Criterion 2e - The pr' cess control system and ssus I
control system will be physleally and electrically separated so that failure in the process system will cot cause failure in the ssWS control system.
.g. Criterion 29 - separation, redundancy, functional reliability, and physical and electrical independency ensure that no anticipated operational occurrence will prevent initiation of the SSWS.
S.
7.3-17
_: 2t. l m-e _
gyster Repair f2EEE-279 Par. 4. 21) . The Cor.bustible Ga s Co . trol System is designed to permit repair or replacement of corpor ents.
(
All devices are designed for a 40-year lifetime, su$ ject to 9 replacement of limited identified parts, under the irposed duty cycles. Since this duty cycle is corposed painly of periode
. testing rather than operation, lifetite is more a matter of shelf lif e than active lif e. Ecwever, all ec&ponents are selected f or continuous duty plus thousands of cycles of operations, far beyond that anticipated in actual service.
Recognition and location of a f ailed component will be accomplished during periodic testing. The simplicity of the logic will make the detection and location relatively easy, and components are pounted in such a way that they can be conveniently replaced in a short time.
Identification (IEEE-279 Par. s.22) . A nameplate identified each control panel and instrument panel that is part of the Corbustible Gas Control Syster. The nameplate shows the division
- to which each panel is assigned, and also identifies the f unctier.
in the systen. of each item on the control panel. The syster. to which each relay belongs is identified on the relay panels.
Cor.forr.ance to_IEEE-338. The syster. will be testable during reactor operation. The test will corpletely test each logic through to the final actuators and demonstrate independence of
, [~ channels and bare any credible failures while not neglecting its l
safety function.
i Dualifying Class IE Electrical Equipeent. IEEE-308 (1971) , *IEEE Standard Criteria for Class II Electric Systems for Nuclear Power Generating Stations". Refer to subsection 8. 3.1. 2.1 (5) .
IEEE-323 (1974), IEEE Standard for Qualifying Class I Electric ,
e Equipment for Nuclear Power Generating StationsO . , M +(Jee As pection 8.3.1.3.
j Corpliance with IEEE-34 (197 , "IEEE ecide for Seismic M -J l Qualification of Class II:9 & Equipr c+ Q Nuclear Power 1 Generating stations" is described in;95n~ Fj ,
1s&sen 810 l
l l l
conformance to 10 CFP 50 Appenoix A 1
- a. Criterion 13. Any concentration of hydrogen within the l
' Arywell and ContainPent following a loss of Coolant accident will be monitored and can be recirculated by initation of the Bydrogen Mixing Control system.
- b. criterion 21 The high functional reliability, i
redundancy, and in-service testability of the two 5.s-2,
.. l
'. 11.1 i
l l
All devices in the system are designed for a long lifetime under the imposed duty cycles, since this duty cycle is compcsed mainly of Periodic testing rather than operation, lifetime is '
more a matter of shelf life than active life. Ecwever all ewponents are selected for continuous duty plus thous os of i j
cAles of operation far beyond that anticipated in actual service.
po' cognition and location of a failed component will be -
accomplished during periodic testing. The simplicity of the l logic will make the detection and location relatively eary, and corponents are mounted in such a way that they can be conveniently replaced in a short time.
1 The design diagnosis andof repair.
the SGTs Isolation control system facilitates rapid i Provisions have been made to facilitate :
repair of the radiation monitors and replacement of the gamma '
detectors during reactor operation.
Identi fi ca tion ffEEE-279 Par. 4.22). A nameplate identifies each control and instrument panel that is part of the SGTS. The nameplate shows the division to which each panel is assigned, and also identifies the function in the system of each iter on the panel. The system to which each relay belongs is identified on l 1
the relay panels.
Conformance to_IEEE-338. The system will he testable during reactor operation. The test will completely check each sensor through to the final actuators and demonstrate independence of channels and bare any credible failures while not neglecting its saf ety function.
Dualifying Class _7 Electrie_ rouirment. IEEE-308 (1971), *IEEE standard Criterla for Class IE Electric Systems for Nuclear Power Generating stations". Refer to section 8.3.
IEEE-323 (April 1971), "IEEE Trial-Use standard: General Guide for Gualifying class I Electric Equipment for Nuclear Power Generation stations". R fer to utDO-10698. 1>
f S- ka 7
Corpliance with IEEE-3an)(197fJ ,
- EEE4 ewide for seissde 3 Qualification of Class 1 :!;;tr' Equi - at Nuclear Power Generation stations' is described i ' ~ ~ ^^^^
% .* %.o.
Conformance to 10 CrR 50 Arrendix A -
- a. Criterion 13. The release of radioactive materials to the environment will be prevented by the monitoring of appropriate plant variables and, upon detection of abnormal conditions, close the appropriate dampers and activate the SGTS.
\
7.3-36
S/HNP-PSAR 03/04/8?
QUESTION 403.1 ( 8. 2.1)
Section 8.2.1 of the PSAR states that the proposed Ashe-Hanford No. 2 line will be looped through the S/HNP Sub-station to serve Unit 2 and that the Ashe-Hanford No. 2 line will be constructed as the generating base in Hanford area increases. Describe the impact on the offsite power system to S/HNP if the Ashe-Hanford No. 2 line is not in place by completion of Unit 2, and the Ashe-Hanford No. 1 line must service both Units. How will the stability analysis and line loading be affected?
RESPONSE
Puget Sound Power & Light Company (PSPL) plans, for economic reasons, to loop the Ashe-Hanford No. 2 500-kV line into the S/HNP Substation for the startup of Unit 2.
l l If, for reasons beyond PSPL's control, the Ashe-Hanford No. 2 500-kV line is not looped through the S/HNP Substation at the completion of Unit 2, there will be no reduction in the reliability of the 500-kV buses that serve the Plant Substation Transformers (PSXs) from that of Unit 1. The S/HNP Substation will contain two i physically independent 500-kV buses and two (PSXs) for Unit 1 and Unit 2 as shown on PSAR Figure 8.2-6. The Amendment 24
S/HNP-PSAR 03/04/82 500-kV portion of the substation will have a breaker-and-a-half configuration for each circuit together with breaker failure backup protection. This system will maximize the reliability of the 500-kV buses that serve the PSXs.
With the assumption that the Ashe-Hanford No. 2 500-kV line would not be completed at the startup of Unit 2, the line loading of 3278 megawatts on the S/HNP - Hanford No. 1 line would be less than the summer rating of 4000 megawatts for this circuit as shown on Figure 403.1-1.
The system is dynamically stable when the Ashe-Hanford No. 1 500-kV line is looped into the S/HNP Substation at startup as shown on Figure 403.1-1. The generator rotor swing curves for a 3-phase 4 cycle fault are shown on Figures 403.1-2 and 403.1-3 and are within safe limits.
A summary of the dynamic stability cases analyzed are shown in Table 403.1-1.
Amendment 24
e TABLE 403.1-1 DYNAMIC STABILITY
SUMMARY
TABLE Dynamic Stability Number of Transmission Line Powe r fl ow Swing Curve Generators Configuration Figure Numbers Figure Numbers at the Project Three-Phase Fault on Comment Result 403.2-1 403.2-2 2 Project Line-End of WPPSS Unita 1 6 4 Stable Project-to-Hanford Disconnected il 500 kV Line 403.2-a 403.2-3 2 Project Line-End of WPPSS Unit 1& 4 Stable Project-to-Ashe Disconnected il 500 kV Line 403.1-1 403.1-2 2 Project Line-End of Pr oj ect-to-H anf ord Stable Project-to-Hanford 92 and Project-to-el 500 kV Line Ashe 02 500 kV Line Out of Service 403.1-1 403.1-3 2 Project Line-End of Project-to-Hanford Stable Project-to-Ashe $2 anJ Project-to-il 500 kV Line Ashe 02 500 kV Line out of Service 403.3-1 40?.3-2 2 Project Line-End of None Stable Project-to-Ashe 9 2 500 kV Line. Drop Ashe-Project-Hanford 42 500 kV Lines (Simulating Backup Breaker Clearing of A Fault) 403.3-1 403.3-3 2 Project Line-End of hone Stable Project-to-Ashe il 500 kV Line. Drop Proj ec t-t o- Ashe 500 kV Line (Simu-lating Backup Breaker Clearing of A Fault) 403.3-1 403.3-4 2 Project Line-End of None Stable Project-t o-Hanf ord el 500 kV Line. Drop P r oj ect-t o-H anf or d 500 kV Line and Project Unit 62 (Simulating Backup Breaker Clearing of A Fault)
l i
MONROE CHIEF JOSEPH MAPLE VALLEY e 1442 MW e 118 MW -+
?
3 512 kV RAVER 4 516 kV 1028 MW 4 4
KITTITAS SICKLER 1278 MW 4- 638 MW I 4.0_ _ HANFORD I
531 kV 924 V e 1082 M W VANTAGE
/
JOHN DAY g
536 kV UNIT 1 y 01334*MW ISKAGIT/
HANFORD NUCLEAR OSTRANDER PROJECT UNIT 2 511 kV g MW. 1 546 kV t
SLATT I
YY 1 542 kV BADC
( 1
! H
a l E ( .
I
' I COULEE DYNAMIC f DESCRIPTION "366 M# SWING CURVE ( OF DYNAMIC RESULTSl j
CASE NUMBER STABILITY TEST l
CO,%C t? hase fault on Stable
<d 1369 MW '
ject's 500 kV bus, d _1369 MW yo3,) 2, ar Project to-Hanford 500 kV line in 548 kV 4c ycles.
? Stable M Shx M ,3-Phase fault on vo3 *1 - 3 , Project's 500 kV bus, telear Project to Ashe 500 kV line in 4
{ cycles.
M t 1
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Y ASHE MENTAL l a
'l 656 54%kV WPPSS 1 & 4 s
- 873 MW 4-- 2300 (
MW m t
WPPSS 2 L PUGET SOUND POWER & UGHT COMPANY
.1117 MW 1100 SKAGIT I HANFORD MUCLEAR PROJECT 6
.iMW PREUMINARY SAFETY ANALYSIS REPORT 549 kV POWER FLOW DIAGRAM .
AND !
iER DYNAMIC STABILITY TEST.
:1- = :c
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1 DYNAMIC STABILITY PLOTS
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SKAGIT / HANFORD NUCLEAR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT DYNAMIC STABILITY PLOTS 7 903.)-3 FIGU AE N4
S/HNP-PSAR 03/04/82 QUESTION 403.2 ( 8. 2. 2) ;
l l
1 The stability analysis discussed in Section 8.2.2.2 of i the PSAR was based on the grid configuration shown in Figure 8.2-7. Since Washington Public Power Supply System Unit 4 has been cancelled, describe the effects on the stability analysis and power flow without Unit 4.
RESPONSE
Power flow drawing Figure 403.2-1 shows the transmission grid without Washington Public Power Supply System (WPPSS) Unit 1 and Unit 4. With the assumption that Unit 4 is not built and Unit 1 is out of service, the line loadings are reduced between Ashe-S/HNP and S/HNP-Hanford. The system is dynamically stable for a 3-phase 4 cycle fault as shown on generator rotor swing curves, Figures 403.2-2 and 403.2-3. The rotor swings are about l 4 degrees greater with Units 1 and 4 out of service which is within safe limits.
l l Because the difference in dynamic stability with and without WNP-4 are relatively small, the analysis provided in PSAR Section 8.2.2.2 is stfficiently applicable to cover dynamic stability witheat WNP-4.
Amendment 24
S/HNP-PSAR 03/04/82 A summary of the dynamic stability cases analyzed is shown in Table 403.1-1.
Amendment 24
i I
I MAPLE VALLEY MONROE CHIEF JOSEPH 90 MW +
.t 8
y 513 kV /
974 MW ->
2
/
1 KITTITAS SICKLER /
<_1220 MW < 807 MW l / HANFOR[
l e 1033 MW 528 kV 413 VANTAGE e 1457 MW ~ 1146 MW -
$19.7 kV
+ 233 MW 526 kV 536 MW 976_ MW 1) k JOHN DAY i
kV UNIT 1 01334A MW SKAGIT/
HANFORD -
1 OSTRANDER NUCLEAR PROJECT UNIT 2 -
$18 kV C1334'->
MW 1 547!kV F
?
4
.t MM 1 P
545 kV w
BA@
l
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I 4
COULEE RESULTS DYNAMIC STABILITY DESCRIPTION c._. 524 MW '
SWING CURVE OF DYNAMIC !
CASE NUMBER STABILITY TEST i
V6 3,1 - 7. Project's 500 kV bus.
<J-.1464 MW ,
Clear Project-to- .
r ;
Hanford #1500 kV line i 540 kV ! . In 4 cycles. i Ar24f5)ceH9 SPhase fault on Stable Project's 500 kV bus.
N3.1"k' Clear Project to Ashe
- 1500 kV line in 4 cycles. l MW 2
, h if l
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i 8
N I 238 MW 2 [ t l LOWER g MONU-
- ASHE MENTAL 4* 5 -
g[ 549.'kV 440MW ,
WPPSS 2 PUGET SOUND POWER & UGHT COMPANYi
- . 863 MW 4 --- 1100-sKAGIT / HANFORD NUCLEAR PROJECT I MW PREUMINARY SAFETY 548 kV 4 ANALYSIS REPORT i POWER FLOW DIAGRAM AND 3ER [ DYNAMIC STABILITY TEST i'
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( 403.2 3 FIGURE CG e _ -9T
S/HNP-PSAR 03/04/82 QUESTION 403.3 (8.2.2)
The cases analyzed in the stability analysis for faulted lines which are identified in Figure 8.2-7 assume the faults to be cleared in four cycles. Will the system still remain stable for longer fault clearing times such as would occur given a circuit breaker failure.
RESPONSE
Power flow drawing Figure 403.3-1 shows the transmission grid when the Ashe-Hanford No. 1 and No. 2 lines are looped into the S/HNP Substation. With the assumption that a short circuit has occurred on the line to either Ashe or Hanford, the system is dynamically stable for a 3-phase 8 cycle fault. However, in normal operation, the 500 kV transmission system would not experience a 3-phase 8 cycle fault but would generally experience a 3-phase fault for a duration of 4 cycles which would then continue on as a single-phase fault for a duration of 8 cycles. At or near the end of 12 cycles, the backup protective relays on the system would operate and clear the circuit breaker that failed to open. The stress on the system of a 3-phase 8 cycle fault is nearly equal to a three-to-one phase 12 cycle fault. A 3-phase 8 cycle fault as shovn on generator rotor swing curves, Amendment 24
j S/HNP-PSAR 03/04/82 1
l i
i~
Figures 403.3-2, 403.3-3 and 403.3-4 are stable. There-i fore, the three-to-one phase 12 cycle fault will be stable.
a The results of all of the dynamic stability cases analyzed are summarized in Table 403.1-1. ,
1 i
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1 Amendment 24 e
I l l
1 M APLE VALLEY MONROE CHIEF JOSEPH 118 MW ; -*-"'- 1410 M W I f
.h 512 kV ,
RAVER 4 543 kV !
y 516!kV
- \
1029 MW ->
/ \
__/
KITTITAS SICKLER
-*-- 1291 MW + 631 MWJ 10- HANFORD,,
I \
' "I
- - 1093 MW 533 kV VANTAGE
522 kV '
i
- -. -334 MW. :
627 kV 535 kV 1048 MW 2 ! 545 kV !
~
JOHN DAY -
1 538 kV UNIT 1 {
1334! --.-
+ MWISKAGIT/ I 18.
0- HANFORD OSTRANDER NUCLEAR R PROJECT LUNIT 2.
515 kV '
1334 MW 1
-1548 kU r
'SLATT E /
gfS :1 544 kV BADGl l
i
S/HNP-PSAR 12/21/81 1
DYNAMIC STABILITY ! DESCRIPTION SWING CURVE i OF DYNAMIC RESULTS ,
CASE NUMBER '
STABILITY TEST COULEE ' f
- 370 MW 8 Phase raua on Ine staDie
\.2 8 Sheet 3 oject's 500 kV bus, k
j fo 1 kV line
- ! 1374 MW i Di 4 cycles, continue
." / 1374 MW ) ,
kn for 10 seconds. l 549 kV j 8.2-8 Sheet 4 i FPhase fault on tha / S*: 1 l
!seyectT500 kv bus,
{
ar Project-to-As l 00 kV line in "n fn j _
i sbw,e Is , nds. c;n'
} I j S.2-8, Sheet 9 W:- . . ontne 4an d 506 V bus, Stable y
f i' d r Hanford- John
( [ *Y,_500 p ,Hnei { ,n, 65econdA.
6.2-8, S 3 N .:,e rault on the table bhe 500 kV bus, clear p
l Whe-to-Slatt #1500 kV ,
- Ene in 4 cyue ,
i kntinue run for 10 J;
- . w onds ~
i .
JMttrl$W4 > Phase fault on the Stable '
Noject's 500 kV bus, ug*3 T A blear Project to.
Hanford #2 and Project to-Ashe 500 kV lines in 8 cycles, con-
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\
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, WWW SPhase f ault on the Noject's 500 kV bus, Stable g g*3,3 blear Project to Ashe 4
F1500 kV line and drop 2 LOWER : P. roiect's unit #2 in 8 MONU.' cycles, continue run
'~1 MENTA for 10 seconds.
- w%.3 2"r Nase fault on the Stable koASHE v* 349 kV
'3 Project's 500 kV bus, clear Project to.
l t 418- Hantord #1500 kV line PSS 1 Q and drop unit #2 in 8 759 MW --
2300 O tycles, continue run t
,MW d for 10 seconds.
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( V 03,3 - 4 FIGURE T ' " ' N-
S/HNP-PSAR 03/04/82 QUESTION 403.4 ( 8. 2)
Inconsistencies which exist in the following amendment 23 figure should be corrected:
Figure 8.3 This drawing does not agree with existing PSAR Figure 8.3-2 in the Class lE power feed to the non Class lE load center.
Figure 8.2-7 (Sheets 1 and 2A) -
Figure 8.2-8 sheet 1 is referenced on this drawing for two separate stability cases.
Figure 8.2-7 (Sheet 3) -
The generating unit labeled as 'WPSS 1 and 2" appears mislabeled.
RESPONSE
PSAR Figure 8.3-2 has been revised to be consistent with Figure 8.3-1.
1 Amendment 24
S/HNP-PSAR 03/04/82 See revised Figure 8.2-7 Sheet (2A of 8). The reference to Fjgure 8.2-8 Sheet 1 on Figure 8.2-7 Sheet (2A of 8) has been corrected to Figure 8.2-8 Sheet 3.
The generating u. sit labeled as 'WPPSS 1 & 2' on Figure 8.2-7 Sheet (3 of 8) has been corrected to 'WPPSS 1 & 4'.
) See revised Figure 8.2-7 (Sheet 3 of 8).
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, ., ,~ - _ . . , _ _ _ . _ - - - - , - _, . , _ , - , __ .
S/II?:P-PSAR ,,16%.2 WCT~
370 MW DYNAMIC STABILITY DESCRIPTION SWING CURVE OF DYNAMIC RESULTS CASE NUMBIR STABILITY TEST
)
- [ 1374 MW 8.2 8 Sheet \ 3-Phase fault on the Stable g e/ 1374 DAW Project's 500 kV bus, g i
3 clear Project to-549 kV Hanford #1500 kV line in 4 cycles, continue run for 10 seconds.
8.2 8. Sheet 4 3-Phase fault on the Stable Project's 500 kV bus, clear Project-to-Ashe
- 1500 kV line in 4
/ cycles, continue run for 10 seconds.
8.2-8, Sheet 9 3-Phase fauft on the Stable Hanford 500 kV bus, clear Hanford to-John Day 500 kV line in 4 cycles, continue run for 10 seconds.
8.2-8 Sheet 10 3-Phase fault on the Stable Ashe 500 kV bus, clear Ashe-to-Statt #1500 kV line in 4 cycles, continue run for 10 seconds.
\ \
MW 3 2
V ce t LOWER R MONU-13 MENTAL l u ASHE
$k
- 418* 5'S kV WPPSS 1 & 4
'9MW _ 2300 MW
. 1 00 PUGET SOUND POWER & LIGHT COVPANY g pg SKAGIT / HANFORD NUCLE AR PROJECT 549 kV PRELIMINARY SAFETY ANALYSIS REPORT POWER FLOW DIAGRAM AND DYNAMIC STABILITY TEST FIGURE 8 2-7 SHEET (2A OF 8)
A:nendrnen t At 7'i
S/Ht;P-PSAR s%r COULEE DYfiAMIC STABILITY DESCRIPTION SWING CURVE OF DYNAMIC RESULTS
- 366 MW CASE NUMBER STABILITY TEST 8.2-8. Sheet 5 Project-to Hanford 82 Stable 4- / 1369 MW 500 kV line out of l
- / 1369 MW semice.3 phase fault on the Project's 500 k I 548 kV bus, clear Project to- 1 Hanford #1 line in 4 l cycles, continue run for l 10 seconds.
1 2
LOWER MONU-ENTAL ASHE
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'MW 638 549kV 1 WPPSS4+2 WPPSh i$4 MW 4--- 2300 1MW Mw WPPSS 2 4- 1100 PUGET SOUND POWER & LIGHT COUSANY '
j MW SKAGIT 1 H ANFORD NUCLEAR PAOJECT 549 kV PRELIMINARY SAFETY ANALYSIS REPORT POWER FLOW DIAGRAM AND e DYNAMIC STABILITY TEST FIGURE 8 2 7 SMEET (3 OF 8)
Amendment 2+
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S/HNP-PSAR 03/04/82 QUESTION 410.1 (3.4.1 & 3.4.5) .
In Amendment 23 to the PSAR you state that flooding of safety-related equipment in the auxiliary building from the turbine building due a rupture of a circulating water system expansion joint will be prevented by operator action following an alarm in the control room which is annunciated by safety-grade detectors located 3 inches above the auxiliary building floor within the airlock between the turbine and auxiliary buildings. In order for us to determine that sufficient time is available for operator action, provide the following information:
- a. Provide the elevation of the airlock between the two buildings and revise your plant arrangement drawings to show the airlock.
- b. Verify that the airlock is normally closed and discuss any indications and alarms that are provided in the control room.
- c. Provide the resul_ts of an analysis to show that sufficient time is available for operator action following receipt of a water level detector alarm in the airlock considering the circulating water system flooding rate.
Amendment 24
S/HNP-PSAR 03/04/82
RESPONSE
(a) The airlock between the Turbine and Auxiliary Buildings is located elevation 527'. Figure 1.2-2 has been updated to show the location of the airlock.
(b) The Tirlock is a secondary containment boundary and will consist of two normally closed water resistant doors in series which are interlocked such that one door cannot be opened unless the second door is closed. Interlocked operation will be capable of being bypassed under appropriate administrative controls to allow both doors to be open at the same time. An alarm or warning light will be provided to alert control room personnel that the interlock has been bypassed. Loss of integrity of the airlock will be annunciated in the control room.
(c) A time dependent analysis of Turbine Building flooding due to breaks in the Circulating Water (CW) System has been performed. Two separate cases were analyzed:
Amendment 24
S/HNP-PSAR 03/04/82 o A large CW break resulting in a flood rate equal to runout flow of all operating CW pumps o A small CW break.
For the large break case, no operator action is relied upon. The water level in the Turbine Building rises rapidly (approximately 2 to 3 ft/ min) until it reaches elevation 531' (4' above the ground floor) at which point the pressure relief panels in the railroad access door open.
Approximately 9 minutes into the accident the cooling tower basin is empty and the CW pumps will lose suction. Because of the short duration of this accident and the presence of water resistant doors at the Auxiliary Building airlock and Control Building lobby, water levels in the Auxi-liary and Control Buildings will not reach more than a few inches above elevation 527'.
The small break case can be more limiting in that the time for the water to reach elevation 531' in the Turbine Building and open the relief panels can be much longer, thus allowing more time for leakage past the water resistant doors. Because Amendment 24
S/HNP-PSAR 03/04/82 of the broad spectrum of leak rates possible, it was conservatively assumed that the flood reaches elevation 531' undetected, which time is defined as t = 0, and then remains at that level without opening the relief panels.
Because the door sill leading to the Control Building is at elevation 531', no flooding of the Control Building will occur for the small break case. For the Auxiliary and Fuel Buildings the door sills for safety-related rooms are at elevation 528'. The water level is calculated to reach elevation 527' 3" due to leakage past the airlock door at approximately t = 15 min. and elevation 528' at approximately t = 45 min. Thus, 30 minutes will be available for operator action between the time of a flood level alarm (3 inches) and the time when water could enter a safety-related equipment room.
l This analysis is considered very conservative because:
l o No credit was taken for floor drain water removal or leakage out of the building through exterior doors.
Amendment 24
S/HNP-PSAR 03/04/82 o The condensate pump motors, located at elevation 517', would be damaged and go out of service early in the accident causing a loss of feedwater accident leading to reactor scram and a prompt investigation of the cause of the feedwater loss.
o Flood level was assumed to be at elevation 531' at t = 0 which assumes a long term water level rise from the Turbine Building basement floor elevation 517' to elevation 531' undetected which then suggests that all of the auxiliary equipment located between those elevations in the Turbine Building have gone under water, are out of service and their failure remains unknown to the operators.
Such an operating circumstance is, of course, not credible as the turbine generator will have earlier tripped off for multiple reasons.
Amendment 24
S/H:!P-PS AR 12/21/81 s- ,
i ,
y/o .1 I
t Becaose flooding will not iepact any cf the structures housing safety-related components er systems, no special flood protection requirenents will be required.
The pamphouse, which provides makeup water for the Plant I will be located on the west bar.h of the Columbia River at
! approximately River Mile 361.5. This pumphouse is not l j safety-related and, thus, if it were impacted by a PMF cr a'
- dam-break flood on the Colunbia River, the saf ety of the Plant would not be af f ected. 23 i
! 3.4.1.2 ' Permanent Dewatering Systen t .
! Because the ground water table of the site is 125 f eet below i ground surf ace (see Section 2.5.4.6) groundwater will not
! affect the operation of any of the safety related systens or components. No permanent dewatering systems will be
- required.
3.4.2 INTER"AL FLOODING 23 The maxirom postulated internal flood would result from a
- f ailure cd the Circulating Water System in conjunction with a failure of the circulating water pumps to trip.
The ECCS pump rooms will be separate watertight compartments
. with their entrances located well above the level that the
- CirculatLng Water System flood could reach, and the water systems inside these roans will be Seismic Category I 020.21 systems. 020.22 042.14 i There will be no connection between the Diesel Generator Building and other Plant buildings. The large water systems
- in the Diesel Generator Building will be Seismic Category I i and the three generator roons will be separated from each other.
The UHS complex will be separated from the other Plant
. buildings, and each of the safety-related pumps will be located bn separate rooms containing only Seismic Category I piping. i The control Building, which is accessed from the Turbine i suilding at Elevation 527' through a corridor and lobby,
', will be protected against the maximum postulated internal 23
- flood by use of pressure relief panels in the railroad W access door at Elevation 527' in the _.1 Tur,bine _ I-2w ev. Building .m m
- Kd - Ath E.REddef4o Ucekw g4%
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- M M A Twh 544% hl..As to 9mtwMg 3.4-2 Anendment 23
B/HNF-PSAR 12/21/81 .
5pF yulltad deor
- Elevation sit', water pressure will cause thegpanels to open, releasing the flood water to the outside A p . .n s !.. , f m ;;.. . sw. . it.e - te r le se 2 a- i;,4 L . ; ..s j iMi .;_ "2diti
- 1 0 c,;;;! L i; ding faced p c;w i;m. l.
- ;i; in
- d by ; c,eting :t.: :::;es Jsar te-i!.e G..; e:
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23 4 e% ,
jp-pFlooding the Circulating in the Water Auxiliary Building System caused will be by abyf ailure detected use ofof I
! flood level detectors which trigger Control Room alarms.
Redundant Class IE detectors will be located 3 inches above
! the .'. ilir r; "r'I : floor i eid: the rir10 7 ::- 2 *! ; 6
, Operator action is thW h.-bin: nd Auxiliary Buildings.
j 3
the elied upon to limit the effects of AuxiliarygBuilding
! flo ing. o,.wa Amt FM The condensate and refueling water storage tanks will be surrounded by a curb about 12 inches in height to provide 33 collection and a drainage path for tank leakage. Other
- tanke located outside the main Plant buildings as well as 320.21 tanks in the. Turbine, Auxiliary and Fuel Buildings will not 320.22
- cause a flood of suf ficient depth to endanger any safety- e42.14 i related equipment. The tanks in the Control Building will he small enough that any water lost from them will be within the capacity of the floor drain system.
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S/HNP-PSAR 03/04/82 QUESTION 410.2 (9.2.1)
In Amendment 23 to the PSAR, you revised the design of the Service Water System (SWS) as a result of the site move to Hanford. The proposed new design (Section 9.2.1 of the PSAR) deleted the cross-connection to the Standby Service Water System (SSWS) as reflected on the SWS drawing, PSAR Figure 9.2.1. However, Section 9.2.11 (SSWS) of the PSAR and Figure 9.2.17 still indicate that the cross-connection between the two systems has not changed. From PSAR Section 10.4.5 (circulating water system) it appears that the SSWS will now be supplied water from the circulating water booster pumps during normal plant operations. Revise Section 9.2.11 and Figure 9.2-17 to reflect the cross-connection between the safety-related SSWS and the nonsafety-related circulating water system. Also revise Sections 9.2.2 (reactor component cooling water system), 9.2.9 (essential chilled water system) and 9.1.3 (containment and fuel pools cooling and cleanup system) to indicate the correct normal source of cooling water.
Amendment 24
I l
S/HNP-PSAR 03/04/82
RESPONSE
See revised Sections 9.2.2, 9.2.9 and 9.2.11, and revised-Figures 9.1-1 and 9.2-17. Section 9.1.3 does not require changes.
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- m. Recirculation Pump 'A' Upper Bearing Cooler
- n. Decirculation Pump 'A' Seals Cooler
- o. Recirculation Pump 'B' Winding Cooler y P. Recirculation Pump *B" Iower Bearing Coolb
- q. Decirculation Pump "B" Upper Bearing Cooler
~
- r. Recirculation Pump *B' Seals Cooler During normal operation one RCCW pump and one heat exchanger are in service. The other pump and heat exchanger are on standby. The tube side flow to the heat exchanger is supplied by the h M :?;n System during normal operation <
and by the Standby Service W er System during loss of off- ""
Site power.
ce,,gqw,,w g,gg, The RCCW System is a balanced closed loop system. The water is circulated throughout the closed loop by the pump and the capacity required by each individual component is set by a manual control valve located on the discharge side of each component.
The water is discharged from the RCCW heat exchangers to the low inlet temperature components at a temperature of 77'F and discharged from these components at approximately 85'F. 7 This tempered water is then circulated through other compo-nents having an acceptable inlet temperature in the range of 95'F to 105'F. The water discharged from these components, o at a temperature of approximately ll7*F, is then returned to the RCCW heat exchanger. In the heat exchangers the heat is transferred to the service water flow on the tube side.
The RCCW System pumps, heat exchangers, and chemical addition tank will be located in the Fuel Building. The head tank will be the highest point in the loop.
9.2.2.3 Safety Evaluation The RCCW System has no safety-ralated function. Failure of I
the system will not compromise any safety-related system or component or prevent a safe reactor shutdown. Piping for the RCCW System is routed so that a pipe break will not flood or damage any safety-related equipment. The System will remain operable during all modes of normal operation including startup and shutdown of the Plant.
O 9.2-4 Amendment 23 L
S/ENP-PSAR 12/21/81 ggy A
- a. Two, 100 percent capacity water chillers
- b. Two, 100 percent capacity chilled water pumps
- c. Two, open type expansion tanks. (
=v hkiews '
CNs'qA /'ooling water to the water chiller condensers "*vvAs\will be,provided w Section 9.2.11, Standby Service Water System N t wA%4 M' m na.
Normal wthwater makeup eMh ehmof mc.nelad to each the two chilled water loops yk4t_
will be supplied with (SC-II) demineralized makeup water, discussed in Section 9.2.3, Demineralized Water Makeup, Standby service water will be Storage and Transfer System.
provided as backup makeup water, when demineralized water is not available.
Chemical treatment ,will be provided in the ECW TheSystem type of to prevent pipe corrosion and scale buildup. chemicals and the methods us chemical treatment will be provided in the FSAR.
9.2.9.2.1 Cemponent Description Design parameters for the major components of the ECW Syste=
are presented in Table 9.2-7. Major i;omponents are described as follows:
a.
Water Chiller Package - Each water chiller will be the f actory assembled and tested centrifugal hermetic type, complete with evaporator, water cooled shell and tube condenser, and complete with safety and refrigeration controls, standard low accessories such as high-low pressure cutouts, water temperature switch, and piping connections.
the Air Conditioning Each water chiller will meet and Refrigeration Institute (ARI) Standard 550-72,
" Standards for Application and Ratings of Centrifugal Water Chilling Packages" and the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section III, Class 3, " Nuclear Power Plant Components.'
- b. Chilled Water Pumps - Each chilled rater pump will be the centrifugal type with direct-coupled motor and base.
I 9.2-40 Amendment 23
V S/ENP-PS AR 12/21/81 gj. g 7,
- h. Puel Pool Heat Exchanger B
- 1. Essential Chilled Water Condenser j.
Reactor Component Cooling Water Heat Exchanqpr B.
bIVISIONIII, THE BPCS SERVICE WATER SYSTEM
- a. HPCS Diesel Generator Heat Exchangers:
(1) Turbocharger Aftercooler (2) Lube Oil Heat Exchanger (3) Jacket Water Heat Exchanger
- b. HPCS Pump Room Cooler
- c. HPCS Pump Bearing Cooler.
The SSW System operates only during reactor shutdown.
During normal operation the following SSW System components are to the SSW System headers:
cooled by the Cc rvicc Litcr (Ch'; ySystem via an intertie g, y %g
- a. Reactor Component Cooling Water System
- b. Essential Chilled Water Condenser
- c. Fuel Pool Heat Exchangers.
cw w The interties between the SSW endg9W Systems supply and discharge headers, shown in Figure 9.2-17, are provided with aswL's i . s J . .d. c. t ) motor operated isolation valves. The valves required to be open during normal Plant operation automatically close during SSW System operation initiated by a LOCA and/or loss of off-Site power.
The SW pumps draw water from the Ultimate Heat Sink basin during normal Plant operation. Drawdown of the water level in the basin is precluded by operator action in response to 020.54 safety-related level indicators and in response to alarms actuated by safety-related level switches.
- A conservative time in excess of 30 minutes has been allowed between reaching the alarm for basin low water level and the 30 day water Ic vel assuming failure of RWS makeup water and continucus los of basin water at a rate equal to the flow of both SW pumps. This time will be suf ficient for the operator to take action to preclude depletion of the level below that required for 30 days of shutdown heat removal.
9.2-47 Amendment 23
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- PRELIMIN ARY SAF ETY 4----.etI2 ANALYSIS REPORT
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S/HNP-PSAR 03/04/82 QUESTION 420.1 Summarize changes to the instrumentation and control systems which will be made as a result of the site relocation and confirm that these changes do not involve changes to the safety-related design bases or criteria from that previously submitted by the applicant in the Skagit/Hanford PSAR through Amendment 22. Also confirm that the instrumentation and control system changes resulting from the site relocation do not depend upon '
advancements in technology beyond the state of the art used for the instrumentation and control systems previously submitted in the PSAR through Amendment 22.
If there are changes to the safety-related design bases or criteria or advancements in technology beyond the state of the art used for the instrumentation and control systems discussed up through Amendment 22 of the PSAR,
! the changes should be itemized and a discussion provided for each to justify that, with these changes, the General Design Criteria contained in Table 7-1 of the Standard Review Plan (NUREG-0800) and IEEE-279 can be met.
I l
Amendment 24
S/HNP-PSAR 03/04/82
RESPONSE
PSAR Amendment 23 change to the instrumentatior, and control systems which will be made as a result of the site relocation is limited to addition of the following:
Anhydrous ammonia detection in the control room air intake, and complete and automatic isolation of the control room upon detection of anhydrous ammonia. The ammonia detection and isolation is not safety-related.
Failure will. not prevent any safety-related function.
This change in the instrumentation and control systems does not depend upon advancements in technology beyond the state of the art.
Amendment 24
l S/HNP-PSAR 03/31/82 l l
QUESTION 421.2 (13.7) (Supplement 19, Response (2) (a) )
Confirm that guidance in Regulatory Guide 5.44 will be used in selecting and installing a perimeter intrusion detection system.
RESPONSE
The guidance contained in Regulatory Guide 5.44 will be incorporated in the design, and used in the installation of the S/HNP perimeter intrusion detection system.
Supplement 19a will be provided under separate cover.
This supplement will provide details of the proposed security plan and Figure 13.7-1 and 13.7-2. Supplement 19a is to be withheld from public disclosure pursuant to 10 CFR 2.790.
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Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 460.1 Figure 11.3-2 of the amended FSAR indicates that the Turbine Building has 3 ventilation exhaust roof vents; Figure 1.2-8 (page 6, 7, and 8 of 9), however, shows eight vents identified as " smoke vents". Are the " smoke vents" in addition to the three ventilation exhaust vents or is this an error?
RESPONSE
Figure 11.3-2 shows only one vent port over the t'rbine u
building roof for HVAC exhaust during normal plant operation. The words "Typ of 3" on Figure 11.3-2 mean one vent port with horizontal air discharge is provided for each of the following non-seismic Category I buildings shown on the drawing:
o Turbine Building o Radwaste Building o Service Building The 8 vent ports shown on Figure 1.2-8 are the smoke and heat vents which will be automatically opened to release the smoke and heat caused by a fire. They are a part of the Fire Protection System and not part of the heating Amendment 24 l
S/HNP-PSAR 03/04/82 and ventilating system. General arrangement drawings do not normally show H&V intakes, ductwork, or exhaust ports. Since the effluent release points are an extension of the H&V system, they are not shown on Figure 1.2-8.
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S/HNP-PSAR 03/04/82 QUESTION 460.2 Provide the elevation, above grade or mean sea level, of the radioactive gaseous release points shown in Figure 11.3-3. Also indicate the shape and inside dimensions (not area) of each vent.
RESPONSE
The potential radioactive gaseous effluent release points during normal plant operations are detailed below.
Approx. El. Exhaust Release Release Release Above Duct Point Vent Point Grade Size Dimen. Config.
Main Plant 235' 42" # 36 " 5 circular Vent Fuel Bldg. 95' 30"x30" 4"x10"x52" square Roof Vent annulus Aux. Bldg. 90' 46"x46" 4"x22"x84" square Roof Vent annulus Turbine Bldg. 125' 72"x?2" 4"x48"x108" rect.
Roof Vent louvers Radwaste Bldg. 50' 56"x56" 4"x24"x78" rect.
Roof Vent louvers Service Bldg. 55' 40"x40" 2"x28"x106" rect.
, (Admin. Bldg.) 40"x18" 2"x28"x52" louvers i Roof Vent 40"x44" l
- See Figure 11.3-2 for orientation of airflow, i
I Amendment 24 l
S/HNP-PSAR 03/04/82 QUESTION 460.3 The response to NUREG-0737, Item II.F.1, Attachment 1, is inadequate in the following respects:
(1) Our examination of the amended PSAR indicates that ,
the Skagit Plant, Unit No. 1, has five gaseous radioactive release vents which could be sources of accident releases of radioactive material.
These are the multipurpose vent (vent or reactor building), the fuel building vent, and the radwaste building vent, the turbine building vent, the auxiliary building vent, the turbine building vents, and the radwaste building vent. Your response on page 1B-94 (Amendment 22) indicates you plan to monitor only the reactor building vent. While this vent would be the release point most likely to contain high concentrations of noble gases after an accident, the other release points cannot be overlooked.
(2) Your response on page 1B-94, Subpart (1), Items A and B, seems to indicate you are providing two accident monitors with ranges differing by a factor of ten but monitoring the same release point. Please clarify why you are using two Amendment 24 1
S/HNP-PSAR 03/04/82 monitors for this release point when one would apparently suffice.
RESPONSE
(1) The S/HNP design utilizes two common vents for airborne radioactive materials that may be released from the Plant during and following an accident. These are the Main Plant vent and the Fuel Building vent. The Turbine Building is not connected to the containment and consequently does not contain accident associated airborne radio-active materials subject to release through the Turbine Building vent. The Auxiliary Building vent is isolated under accident conditions and is vented by means of the Standby Gas Treatment System through the Fuel Building vent. Accident level releases, both liquid and gaseous, are
- isolated and not released to radwaste under accident conditions. " Consequently, the radwaste vent is not subject to accident releases of air-borne radioactive materials.
(2) The response on S/HNP PSAR page 1B-94 Subpart (1) indicates the two general categories of noble gas monitors that could be utilized. The Main Plant i
Amendment 24 i
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, S/HNP-PSAR 03/04/82 vent, which includes the drywell purge, and the Fuel Building vent, which includes the SGTS purge, 1
are both Type B release points.
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Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 460.4 The response to NUREG-0737, Item II.F.1, Attachment 2, is inadequate in the following respect:
Your discussion on Page 1B-96 indicates that you plan to sample only the multipurpose vent for radioiodine and particulates from accident releases of radioactive materials. NUREG-0737 requires monitoring for all potential accident release paths, which would include vents from the turbine building, fuel building, auxiliary building, and radwaste buildi
RESPONSE
A post-accident particulate and iodine sampling capability will be installed on the Fuel Building vent and the Main Plant-vent. As noted in the response to Question 460.3, the potential accident release paths are only through the Main Plant vent and the Fuel Building vent. The design was found acceptable in S/HNP SER NUREG-0309, Supplement 2, page II-20.
}
Amendment 24
S/HNP-PSAR 03/04/82 QUESTION 471.1 Section 12.3.1 of Regulatory Guide 1.70 specifies that layout drawings should show shield thickness. In Section 12.1.2.4.7 of th'e PSAR, the control room shielding is specified to be 2'-0" thick concrete walls, but the roof i shield thickness is not specified. The shield thickness above the control room should be specified, either in Subsection 12.1.2.4.7 or in Figure 12.1-16.
RESPONSE
See revised Section 12.1.6.1.
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i Amendment 24 J
1
5/H!!P-FSAR 12/21/81
- Q 11.f i .
i 12 1 5 0FERATIf!3 PROCEDURES
(
The health physics progra. and access control describro in kept tenance.on low as practicable during Plant operation t
- Operatirig evalcu ed experience of other SWR Plants will be continually to deterrine radiation levela present.
Ievels in areas not previously conside red will be noted.Any high Domes received by Plant personnel will also be noted.
radiation cuide 4 0. Ais traintained ALAP in accordance with 6
provided in the FSAR. description of these procedures will be 12.1.s 2stmarts or EXPOSURT J 2 1. 6.1 Anticipated Doses 331.'
be considered as the maxir.un dose for which the are noned (Section 12 1 2 1). These doses are not expected to
(. . occur during normal operation because the Plant shielding is based on maximum coolant activities while the average isotopic concentrations will be considerably less than the maximum.
auch as l'nside The highest dose rates will occur in Zone V areas the drywell, in the turbine-condenser area, and in rooms radioactive fluids. containing equipnent and piping handling highly The than direct the radiation doses to the control room will be less 2'-0*
thick concrete wall surrounding the control roon. The toned
- areas of other buildings adjacent to the control room are designated exceed 2.$ ar/hr.
as Ione II, so the dose rate in them will not due solely to N-15 garr.a radiation, theConservatively assuming this 2'-0" thick concrete 23 dos wall willthan to l'ess reduce 0 05the radiation level inside the control roon er/hr.
within the design basis of 0.5 mr/hr.Hence, the dose will be well The annual dose to the construction workers employed in Unit 2 while Unit 1 la in operation bas been entirated for varicas points in the Unit 2 construction areas the l
locations of these dose points are shown on Figure 12 1-17. 23 Stadiation dose to construction workers will be due rnostly 333*8
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. #3.endment 23
S/HNP-PSAR 03/04/82 QUESTION 471.2 Section 12.2.1 of Regulatory Guide 1.70 specifies that applicants should describe the source terms used for shield design for the turbine system. Section 12.1.3.9 states that the source strengths given by Table 12.1-15 are adjusted to reflect the self-absorption in the components and that the equivalent inventory was found to be 117 Ci of N-16, including exposed piping associated with the H.P. Turbine.
471.2(b)
In addition, you should explain why only " exposed piping associated with the H.P. Turbine" was used as the N-16 source. It appears that the dose contribution should include a minimum:
H.P. Turbine, (25.37 Ci),
Moisture separators / reheaters (118.25 Ci),
Crossover piping (17.13 C1), and L.P. Turbines (19,74 Ci), etc.
Based on Table 12.1-15 the total N-16 source should be at least 180 Ci of N-16. If the above quoted 117 Ci were Amendment 24
S/HNP-PSAR 03/04/82 '
l
^
QUESTION 471.2 Section 12.2.1 of Regulatory Guide 1.70 specifies that applicants should describe the source terms used for shield design for the turbine system. Section 12.1.3.9 states that the source strengths given by Table 12.1-15 are adjusted to reflect the self-absorption in the components and that the equivalent inventory was found to be 117 Ci of N-16, including exposed piping associated with the H.P. Turbir.e.
471.2(a)
You should specify if the N-16 inventories quoted in Table 12.1-15 are adjusted for component-self absorption, or if they are actual estimated inventories.
RESPONSE
4 See revisions to Section 12.1.3.9.
Amendment 24
S/HNP-PSAR 03/04/82 obtained by adjusting the 180 Ci for the source self-absorption effect, then it should be so stated.
i
RESPONSE
See revisions to Section 12.1.3.9.
Amendment 24
^
S/HNP-PSAR 03/04/82 QUESTION 471.2 '
Section 12.2.1 of Regulatory Guide 1.70 specifies that applicants should describe the source terms used for shield design for the turbine system. Section 12.1.3.9 7 states that the source strengths given by Table 12.1-15 are adjusted to reflect the self-absorption in the components and that the equivalent inventory was found to be 117 Ci of N-16, including exposed piping associated with the H.P. Turbine.
471.2(c)
In Table 12.1-15, "N-16 inventories in equipment in the turbine building", reference is made to notes (1), (2),
and (3). Note (2) refers to main steam piping.
References for Notes (1) and (3) are not specified. You should provide references for Notes (1) and (3).
RESPONSE
See revisions to Table 12.1-15. '
Amendment 24 l
l
S/HNP-PSAR 03/04/82 QUESTION 471.2 Section 12.2.1 of Regulatory Guide 1.70 specifies that applicants should describe the source terms used for shield design for the turbine system. Section 12.1.3.9 states that the source strengths given by Table 12.1-15 are adjusted to reflect the self-absorption in the components and that the equivalent inventory was found to be 117 Ci of N-16, including exposed piping associated with the H.P. Turbine.
471.2(d)
In subsection 12.1.3.9, " Turbine Shine Dose," reference is made to the " Exclusion Area Boundry", which is approximately 1.9 miles from the turbine building. At this distance you calculate the dose to be 0.5 mrem / year.
Is the exclusion area boundary the closest unrestricted area from radiation control standpoint? If not, you should provide the dose at the closest unrestricted area.
RESPONSE
See revisions to Section 12.1.3.9.
Amendment 24 L
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- lines, turbines, and roisture separators can coy ribute to y the Exclusion Area Brundary dese as a result of 'the high energy gamas which it emits as it decays. 23 i Turbine shine doses are calculated using the SKYSHINC I corpcter program described in Table 121-3. Point sources are caed to re I he w ee c'eFresent the co..ponents on the turbine deck. 331 3
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equivalent-imcnt-ey m f a u- 4 en > = 3 1 4 -C + ;,f :: }g ThIr h3 I
7 Joe 4Vdes suposad # pi.3 .. 3ciatod-M The sources are surrounded by 24'-6" high walls on the north, ,
'23 south, and east and a 31'-0* bigh wall on the west. The
. center and of the 50'-0* mid-LP from turbine the north wall.is 6C'-10* from the east wall The area enclosed by the
- wally is 200'-0* in the north-south direction and 204' in I the east-west direction.
EAO The e.xpectegdu(rbine shine dcse at the txclusion Area 4 BoundaryFW ach is aFyroxiriately 1.9 r fles from the Turbine auilding, is conservatively estinated to be less than 0.5
[ 23 with an availability factor of 80 percent. mrem /yr. ,this estir. ate is based The dose rates due to radiation scattered by air and the walls surrounding the co.yonents en the turbine deck were
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evaluated at several locations within the Turbine Building I but outside the shield walls. The raximun curulative 331.32 contribution from these sources of expesure was found to be I lese than 1 0 strem/hr. The shield walls will be designed to maintain the sone II maxirnun of 2 5 rirem/hr.
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j _12.1 3 10 Field Run Pipe Routing
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The procedurer ici uuting of field run p1 ping are discussed in Section 12 1.2 3 2 I-12.1 4 AREA FADI ATION MONITORING See Astendia IA, Section 14 1 and FSAR Table 12.1-4 16 1
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! FJ -Ib . ; I
, i l Irnsert R> & .p 12. t-28
. i i 3
e l b4M O.OO.#4rV lW6 CC04/.58,. 1 1$
&- ( M .j l
, closee! 4,i 4hc Put 4kan 4he news + unces+p'ded
. Z m ch.
ce4.qs -s t (.'W'fE Barrtcode. , cpEy[Eaie y
i 1
- - I
_ ,.m orerws ow ame -m<*s.=;ian c, l '
I
~
093075 i.
' f7/.0 b
~
TABLE 12.1-15 6 W-15 2:
! RVENTOA2E8 IN EQUIPMENT IN THE TUABINE BUILDI =
y
' Estimated Transit Component
- Estimated Volume Time to Component K-16 Inven
, ( ft s ) (see)
_ _ _ (C1) 1 Main Steam Piping (211' 4448 0 201.6 pigh Pressaie fu bine - 1023 2.04 25.37
, Crossunder Piping l' 7005 2.357 61.e4 N moisture separators - 18.400 3.273 118.25 j Crosso'ver Piping -
3719 5.291 17.13
. gew Pressure Turbinea " 44,441 5.872 19.*74 Gland Steam Evaporator 862 2.357 2.3 Nigh Presante sea,ters 3000 2.357 a, . 36.6 i '.,
( e- '
l l1 Based on 251 NSSS GESSAR Table 11.1.4 N-16 activities, Figure 10.1-2, j (. ) and the, asamnption that the N-16 la uniformly partitioned.
(2) Includu piping in the Containment and the Ausiliary Building.
t(3) Saturation vplue.
Represents total N-36 Inventory present beyond
- the estraction point. -
e
.l .. ..
t ,
. t t .
I -
i .
3 l .
i : .
- l
. e