ML20246L326
| ML20246L326 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/31/1989 |
| From: | William Cahill, Walker R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89623, NUDOCS 8909060305 | |
| Download: ML20246L326 (9) | |
Text
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EE Log # TXX-89623 C
File # 10010
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903.8 1UELECTRIC William J. Cahill, Jr.
Esecutin Vice Presulera U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 ADVANCE FSAR SUBMITTAL APPLICABILITY OF CODE STAMPING FOR PENETRATIONS Gentlemen:
This letter provides an advance copy of a change to be included in a future FSAR amendment. This change adds clarification that the equipment hatch and the sleeves for the fuel transfer tube, electrical penetrations, and process piping penetrations are not code stamped and are tested as part of the-containment structure.
In order to facilitate NRC Staff review of this change, the attachment is organized as follows:
1.
Draft revised FSAR pages, with change portions indicated by a bar in the margin, as they are to appear in a future amendment.
2.
A detailed description / justification for the change.
3.
A copy of a related SER section.
4.
A pnge containing the title of a " bullet" which consolidates and categorizes similar individual changes by subject and related SER section.
I I l
i 8909060305 890831 PDR ADOCK 05000445 A
PDC 400 North Olive Street LB 81 Dallas, Texas 75201
TXX-89623 August 31, 1989 Page 2 of 2 5.
The bold / overstrike version of the revised FSAR pages referenced by the detailed description / jus +4'ication for the changes identified-above. The bold / overstrike vers). i facilitates review of the of new text in bold type font revision by highlighting each addh
'o and overstriking with a slash (/) th sortion of the text that is deleted.
Sincerely, y
)
William J. Cahill, Jr.
b By:
s Roge# D. Walker Manager, Nuclear Licensing CBC/dje Attachment c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3)
- A'ttachment to TXX-89623 Page 1 of 7 Attachment to TXX-89623 August 31, 1989 Advance FSAR Submittal Concerning Applicability of Code Stamping for Penetrations 11gm Subiect Pages 1
Draft Revised FSAR Pages 2
2 Detailed Descripti :,/ Justification for 3
(
Changes 3
Related SER Section 4 thru 5 4
Bullet Title 6
5 Bold /0verstrike Version of Revised FSAR 7
Pages l
l I
b Attachment to TXX-89623 CPSES/FSAR.
7f Pag 2 2 of 7 including cattrials, design, fabrication, examination,. testing, and so DRAFT L
. forth, except the equipment hatch and the sleeves for the electrical penetrations, fuel transfer tube, and process piping penetrations are.
not code stamped and tre pressure tested as part of the reinforced concrete containment structure as discussed in Section 3.8.1.7.
The 1971. edition of the code, through and including the summer 1973 68 addenda, is used for the electrical. penetration sleeves, fuel transfer tube' penetration sleeve, emergency and personnel airlocks, and equipment hatch. The 1974 edition through and including the Summer 1976 Addenda is used for the process piping penetrations.
3.8.2.2.2 Other Applicable Codes, Specifications, and Standards 1.
See Section 3.8.1.2.3 and 3.8.2.6.6 of this FSAR.
68 3.8.2.2.3 Aoolicable NRC Reaulatory Guide 68-NRC Regulatory Guide 1.57 Design Limits and loading Combinations 68 for Metal Primary Reactor Containment System Components (6-73) (Applicable only to appropriate containment components constructed in accordance with Subsection NE of the ASME B&PV Code,Section III, Class MC.)
3.8.2.3 Loads and load Combinations The applicable loads stated in Subsection 3.8.1.3 are considered in the design of the ASME B&PV Code,Section III, Class MC steel components. For the various load combinations used in the design of k
Class M components and the related allowable stresses and strains, see Subsection 3.8.2.5.
Draft Version 3.8-74
CPSES FSAR AMENDMENT 77 Attachment to TXX-89623 DETAILED DESCRIPTION Pag] 1
. Page.3 of.7 oc FSAR Page.
IAs amended)
Group Description l
3.8-74 2
Adds clarification that equipment hatch and sleeves for J
fuel transfer tube, electrical and process piping pene-trations are not code stamped and are pressure tested as part of the containment structure.
Addition The equipment hatch and the sleeves for the electrical penetrations, fuel transfer tube, and the process pip-ing are provided as part of the reinforced concrete containment structure. The basic code for the mater-ials, examination, testing, and surveillance of the Containment is the appropriate portions (specified in Section 3.8.1.2 of the FSAR) of the Proposed Standard Code for Concrete Reactor-Vessels and Containments
( April 1973), which was issued for trial use and com-ments.
Specific references to the articles in Subsec-tion CA, General Requirements, which.are of legal nat-ure rather than a technical nature have been omitted.
These legal requirements are not applicable to CPSES since the Code edition in force for this project is the trial use and comments issue. Thus the components of the containment structure, which are not part of a code stamped system, are subject to the'same requirements as the containment structure and are not code stamped.
~
The equipment hatch and those po-tions of the sleeves described above that are not backed by concrete are constructed to the rules and requirements of the ASME B&PV Code Section III, Division 1. Subsection NE (Class MC), per section 3.8.2 of the FSAR. These com-ponents are enclosed at one end only: thus they were not pressure tested in accordance with the requirements of NE-6000 but are tested as part of the containment structure during the structural acceptance test.
FSAR Change Request Number: 89-497 Related SER Section: 3.8.1 SER/SSER Impact: No
Attachsent to TXX-89623
. Page.4 of 7 i
(2) an equivalent static load method (3) the simplified design method The applicant has indicated the analysis method used for each piping system and has provided technical justification for use of the equivalent static load and sisplified desiipi methods. Both of these methods are based on static seismic analysis.
"he applied seismic loads correspond to accelerations equal to at least the zero period accelerations of the appropriate floor response spectra.
The staff has reviewed the appifcant's precedures and concludes that the seismic ev_aluation methods and precedures described by the appifcant for nuclear steam supply system and nonnuclear steam supply system Seismic Category I piping systems and equipment are acceptable.
3.7.4 Seismic Instrumentation Program The type, number, location, and uttifration of strong-motten accelerographs to record seismic events and to provide data en the freguancy, amplitude, and phase relationship of the seismic resplase of the containment structure comply with Pegulatory Guide 1.12.
Supporting instrumastation is being installed on Category I structures, systems, and components 'te provide data for the verifice-tion of the seismic reopenses determined analytically for such Category I items.
The installation of the specified seismic instrumentation in the reactor con-tainment structure and at other Category I, structures, systems, and components constitutes an acceptable program to record data en seismic ground motion as well as data en the frequency and sup11 tads relationship of the response of major structures and systems. A prompt readout'of pertinent data.at the control room can be espected to yield sufficient information to guide ths operator on a timely basis for the purpose of. evaluating the setemic response in the event of an earthquake. Data obtained from such installed seismic instrumentation will be, sufficient to deterwise that ths seismic analysis asseptions and the analytical model' used fer'the design of the plant are adequate and that alleuable stresses-are not eneseded under conditions where continuity of aparation is intended.
prowleien of such seismic instrumentation comp 11es with Regulatory Guide 1.12.,
- g..
3.8 Desian of tainmig r am I 5turtures.
3.8.1 Concate testefament
+
The reacter coelemt system is enclosed in a steel-lined, reinforced concrete i
containment strutturer-This structure consists of a vertical cylinder and a hemispherical does, and is agported on an essentially flat foundation met with a reactor cavity pit projection. The. containment structure uns designed in accordance with American Concrete Institute (ACI)/ASE Code (ACI-369) and Regulatory Guides 1.10, 1.15, 1.18, 1.19, and 1.55. Various combinations of dead loads, live loads, environmental loads (including those caused by wind, tornadoes, 00E, and SSE), and leads generated by the design-basis accident (inciteting pressure, temperature, and associated pipe rupture effects) were consiGred. The load combinations used and presented in the ESAR are mese conservative than those specified in SRp Section 3.8.1.
3-17
Attachaent to TXX-89623-
. Page,5 of 7
$tetic anel is of the containment shall and base is fe e ded on methods pre-
)
vieusly fed.
Likewise, the liner desip for the containment employs J
sothods s lar to these previously accepted by the staff.
1he choice of asterials,imilar to of anchers, the desip criteria, and 1
the arr the desip methods are s evaluated fer previously licensed nuclear plants. Materials, construction methods, and quality assurance and quality control measures are covered in the FSAR.
In general, they are steiler to these used for previously licensed facilities.
j Before the plant begins operation the containment will be sub4ected to as acceptancetestinaccordancewitE 1 story Guide 1.18. Durlng this test, the internal pressure will be 1.15 times containment desip pressure.
The criteria used in the analysis, dest and construction of the concrete containmentstructuretoaccountforambipotedleadingsandpostulated conditions that any be taposed on the structure during its service lifeties are in confermance with established criteria, codes, standards, guides, and specifications acceptable to the staff.
The use of these criteria (as defined by applicable codes standards, guides, i and specifications); the leads and leedl cosepfnetiens; Ihe desip and analys procelhsres; the structural acceptance cri a; the meterials, quality centrol programs, and special construction techniques; and the testing and faservice surveillance requiremeM.s ide reasonable assurance that In the event of winds, tornadoes, eart.
, and verlous postulated accidents occuring within and eastside the containeest, the structure will withstand the specified design conditions without tapetrennt of structural futegrity or safety functions Conformance with those criteria constitutes an acceptable basis for satisfying, in part, the regoirements of B C 1, 4, 38, and 50.
3.8.2 Caecrete and Structurel Steel Isternal Structures The contalement internal structures are constructed primarily of reinforced concate and coasiet of the fle11ering elements: primary shield wel1*
sediate fleers;;reassable slabs and wells; polar crane; sissile sh eperatias floor refleeltag scritytti hose slab L
seppertins elements
- md sqsports for reacter pressere vessel steengenerators,reactercoolank pimps, pressurizer, and leap ofpi TEcssconcreteandsteelinternal structures are dost to reslot.
eus combinettees of deed and live leads, and accident-i leads,inel ng pressure jet leeds, and seismic leads.
The appl $sent has vertffed that the internal sfruttures meet the desty require-seats stated in S W 5ecties 3.4.3.
}
The criteria used is the desise analysis, and construction of the containment internal structures to account for anticipated leading and postuisted conditions that any be imposed epse the structures derfag their service lifeties, are in full confermance with established criteria, and with codes, standards, and specifications acceptable to the staff.
The use of these criteria (as defieed by aeolicable codes, standards, and specifications); the Icede and leading cessinettens; the desip and analysis precedures; the structural acceptance criteria; the esterials, gus11ty control 3-18
,; y,
I i
Attachment to TXX-89623
,e fa 9 3 c ge.6 of 7-d 3.8.
Desian of Cateaory I Structures-3.8.1 Concrete Containment l
?'
ESGB 21.
The FSAR has been revised to state that'the equipment hatch and
. sleeves for the fuel transfer tube, ' electrical. and process piping penetrations are not code stamped and are pressure tested as part.of the-containment structure.
9 N-----
Attachment to TXX-89623 CPSES/FSAR
< Page.7 of 7
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including caterials, design, fabrication, exaainaticn, testing, and so forth, except the equipment hatch and the sleeves for the electrical
-penetrations, fuel transfer tube, and process piping penetrations are not code stamped and are pressure tested as part of the reinforced 68 concrete containment structure as discussed in Section 3.8.1.7.
The 1971 edition of the code, through and including the summer 1973 addenda, is used for the electrical penetration sleeves, fuel transfer tube penetration sleeve, emergency and personnel airlocks, and equipment hatch. The 1974 edition through and including the Summer 1976 Addenda is used for the process piping penetrations.
3.8.2.2.2 Other Applicable Codes, Specifications, and Standards 68 See Section 3.8.1.2.3 and 3.8.2.6.6 of this FSAR.
68 3.8.2.2.3 ADolicable NRC Reaulatory Guide 68 NRC Regulatory Guide 1.57 Design Limits and loading Combinations for Metal Primary Reactor Containment System Components (6-73) (Applicable only to appropriate containment components constructed in accordance with Subsection NE of the ASME B&PV Code,Section III, Class MC.)
3.8.2.3 Loads and Load Combinations The applicable loads stated in Subsection 3.8.1.3 are considered in the design of the ASME B&PV Code,Section III, Class MC steel components. For the various load combinations used in the design of Class M components and the related allowable stresses and strains, see Subsection 3.8.2.5.
Bold /0verstrike 3.8-74 Version
_ _.