ML20054J371
| ML20054J371 | |
| Person / Time | |
|---|---|
| Site: | 05000000, San Onofre |
| Issue date: | 02/10/1981 |
| From: | Bessette D Advisory Committee on Reactor Safeguards |
| To: | Bender M, Mathis W, Ward D Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20033C447 | List: |
| References | |
| FOIA-82-176 NUDOCS 8206280548 | |
| Download: ML20054J371 (9) | |
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's Februa ry 10, 1981 M. 9 ender W. M. Mathis D. A. Ward J. J. Ray I. Catton W. C. Lipinski PROJECT STATdS REPORT: ACRS SAN ONOFRE 233 SUBCOMMITTEE v.EETING AS PART OF THE ACRS REVIEW FOR AM OPERATING LICENSE, WASHINGTON, DC, FEBRUARY 18, 1981 The ACRS San Onofre Units 2A3 Subcommittee will iact in Washington, D.C. on February 18, 1931. The purpose is to conduct a review of the Units as part of the ACR3 review in preparation for issuance of an operating license. A proposed agenda is attached. The Subcommittee met previously on January 31, 1931 to revie9 seismology and geology of the site.
The ACRS Electrical Systeas Subcommittee will meet February 24, 1981 in Washington, DC, to review the San Onofre computerized reactor protection system.
The design is similar to Arkansas Nuclear One-Unit 2 (ANO-2). At the tim
- of AM0-2 licensing about 2 years ago, the c&aputerized protection system was reviewed in some detail. The purpose of the forthcoming review will be to compare the San Onofre 213 system to that of ANO-2, and to review operating experience.
The. Staff's Safety Evaluation report has been recently distributed to you.
It did not include Til-related items. The Staff expects to complete this portion of the SER in the beginning of Xarch 1981, at which time the SER will be fully complete.
An additional Subconmittee meeting aay be held on March 11,1931 to revieu TMI related itens. The Committec :aay write a letter at the March full Connittee meeting, March 12-14, 1931.
Guarantee.1 reservations have been made as follows.
If for soac reason you cannot attend, the reservations must be cancelled.
Mane Nights of Hotel M. Bender Feb 17 Lombardy (202-828-2600)
- 9. Mathis Feb 17-13 Anny-Navy (202-628-8400)
D. Ward J. Ray Feb 17 Lombardy (202-828-2600)
I. Catton Feb 17-18 Park Central (202-393-4700)
U. Lipinski Feb 17 Park Central (202-393-4700)
I can be reached at 202/634-3267 if you need further infonnation.
D. E. Gessette Nuclear Engineer 8206280548 820512 l
MCMURR A82-176 PDR i
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February 10, 1981 San Onofre 2&3 The San Onofre site is located on the coast of southern California in San Diego County, approximately 62 miles southeast of Los Angeles and 51 miles northwest of San Diego. The site is located entirely within the boundaries of the United States Marine Corps Base, Camp Pendleton, California, near the northwest end of the 18-mile shoreline.
A permanent concrete seawall is provided along the seaward side of the site to protect the site against sea erosion. This seawall functions as a retaining wall.
It is designed to assure that it will withstand, with-out loss of functional capability, the design basis earthquake (DBE) fol-lowed by a tsunami, with coincident storm wave action.
The population of Camp Pendleton, which surrounds the San Onofre site, is extremely variable.
It is expected, however, that the pcpulation will not exceed 51,000.
No personnel will be quartered closer than 1-1/2 miles from the station site.
The principal administration and main personnel housing areas are located 12 to 15 miles to the southeast. The nearest sizable community is San Clemente (population approximately 18,200) located about 4 miles to the northwest.
Oceanside and San Diego are 17 miles and 51 miles to the southeast of the site, respectively.
The San Onofre Nuclear Generating Station Units 2 and 3 is comprised of two CE nuclear steam supply systems (NSSS) that produce a nominal net output of 1,100 MWe per unit.
Of the operating CE plants, San Onofre 283 are most similar to Arkansas Nuclear One-Unit 2.
San Onofre 2&3 features separate containments, safety equipment buildings, turbine buildings, diesel generator buildings, and fuel handling buildings, and a shared auxiliary building and intake structure.
The ultimate heat sink for all Seismic Category I cooling water systems is saltwater from the Pacific Ocean, supplied to the component cooling water heat exchangers by saltwater cooling pumps, located within separate intake conduits for each unit.
Seawater pumped from the intake conduits by the circulating water pumps serves as the heat sink for heat rejected by the main condensers and the turbine plant cooling water system. The 220-kV switchyard is located directly northsast of the power block.
The NSSS generates approximately 3410 MWt. Each of the two NSSS units 1
contains two independent primary coolant loops, each of which has two reactor coolant pumps, a steam generator, a 42-inch ID outlet (hot) pipe and two 30-inch ID inlet (cold) pipes.
The core consists of 217 fuel assemblies which will be initially loaded with three different U-235 enrichments. The NSSS full-thermal output is 3410 MWt with a core thermal output of 3390 MWt. No provisions for stretch capability are included in the design of the NSSS.
j Ooen Items The Staff lists a number of items as being coen.
In some cases, it is simply a matter of the Staff completing their review.
l The open items, with appropriate references to subsections of the SER, have been excerpted from the SER and are listed below.
(1) Emplosion hazarcs. Section 2.2.2, page 2-13 (2) Tomic gas hazarcs. Section 2.2.2, page 2-14 (3) Systems Interaction. Section 3.8.6. page 3-22 (4) Seismic avalification of equipment. Sections 3.10, page 3-28 (5) Reactor internals analysis. Section 3.9.2.3, page 3 23 (6) Incopendent mining analysis. Section 3.9.3.1, page 3 25 (7) Environmental cualification of ecufgeent. Section 3.11.2, page 3-29 (8) Seismic plus LOCA loads on FEA. Section 4.2.2.10, sage 4-7 (9) Core protection calculator. Section 4.4, page 4-21; Section 7.2.2, page 7-3; Section 15.1.1, page 15-3; Section 15.2.3, page 15-S (10) DNet testing of revised FEA. Section 4.4, page a-16 (11) Containment Pressure Bouncery Fracture Tougnness. Section 6.2.1.4 page 6-8.
(12) Emergency planning. Section 13.3., page 13-2 (13) Incustrial security. Section 13.6. page 13-15 (14) Review of CENPD-133. Section 15.1.2, page 15-6 I
(15) seview of Q-list. Section 17.3, sage 17-4 Each of these issues is summarized metow.
(1) Exclosion Mazares Tne applicants nave not coesnstrates that the emolosion risks associated with transportation of natsrecus materials past the site are sufficiently low to to acceptacle. Tney nave agreed to revise their procamility analysis, and to evaluate the amtlity of plant structures to withstand overpressures greater than the tornaco loacing. W will recuire snat any cortions of the plant founo to te vulnersole to significant clast aamage te socified suen that there will te reasonsole assurance tnat they will retain their functional cacao 111ty in the event of overgressure cue to emotosions.
(2) 'orie Oas warsees we are unaole to verify the motor carrier accicent
- ate =nien is presentea in Section 6.4 of the F5AR. Tne value usec in Section 6.4 is acout four orcers of magnituce less than the truct accicent rate cased on nationally everagea statistt used my the acclicants in Section 2.2 analyses. Our position is that the acofi-cants sust suostantiate tne truck accicent este usec in their toxic gas analysis or revise it accorcingly, and orotect tne coetrol room from any accitional
- ca1C gases that are a 9423rc to tne olant ocerators.
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s (3) Systees interaction We have requested, and the acclicants have proviced, aeditional information Concerning the eDjectives and SCoos of the applicants' systees interaction program, the methodoloQr and criteria used to postulate the interactions, and the organization esta011shed to inclement the prog?am. We are evaluating the applicants' response to our request and plan to conduct en onsite audit of the applicants' program.
(4) Seismic Qualification of feuiement Our review of the information presented in the F5AR is in progress. Our findings will be based on our review and on the information ootained during the Septeamer 1980 site visit my our Seismic Qualification Review Team. Our review is not yet conclete.
(5) Reactor Internals analvsis we have informed the applicants that the dynamic systees analysis described in F5AR Section 3.7.3.14 require further amplification and clarification. The applicants have agreed to provice the additional information, f
(6) Inceoenrent Dipino analysis We are perfoming an independent confimatory analysis of the shutdown cooling line. This analysis will not only verify that this piping system meets the acclicacle ASME Coce requirements, but will also provide a check on the
~appfidnd'~acility to correctly accel and analyze piping systems. We have contracted with the Energ Technology Engineering Center (ETEC) to p3rfom the confirmatory analysis, and it is in progress.
(7) Envirensertal cualification of Ecuiement We recuested that the applicants reassess their cualification docueentation for eouisment installed at the facility, to snow that the qualification methocs usec and results cetained conform to the staff positions in NUREG-0588. We nelieve that this additional review will confirm our steviously-reached conclusions that the San Onofre 2 anc 3 cualification cocumentation is aceouste. Nevertheless, we reQure that the accitional review te comoletec prior to issuance of a full power operating license.
(8) Seismic slus CCA '.caos sn Cuel fie= eat Asse-olv (Fia)
The analicants 9 ave referercec the tcotcal recort *ENPD-178, " Structural Analysis of fuel Asseeclies for S moinec Seismic anc Loss-of-Coolant Accicent Loacing,"
.nich accress,es this sattaa. As a esult of our oreliminary review,.e conclucec that *ENPO-178 310 not c:ntain an aceouate accel for analyzing lateral loacs on 1-9
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the fuel assemoly nor did it present sufficient information on spacer grid tests. The applicants have stated that they will previos additional inforestion on analytical setnocs and test results as an amendment to the Final Safety Analysis Report.
(9) Core erotection Calculatee he have reouired the San Onofre 2 and 3 acclicants to sucait a sweary of any socifications for their core protection calculator as compared to the Arkansas Nuclear One Unit 2 core protection calculator. Decause of our significant review effort on the ANO-2 computer.
The sp;nlicants noted sodifications in the following areas and for the following reasons:
(1) Core protection calculator / control element assemoly protection algorithms -
these changes are a result of the enange in the numoer of control element assemolies and control eleoent assemoly suegroups for San Onofre 2 and 3.
(2) Core protection calculator / control element asseemly data base constants -
I these changes are due to the specific core and coolant system characteristic:
l (3) Software changes related to thermal-nyoraulic metaccs - the enanges incor-porate current Coacustion Engineering metnods.
Our review of these modifications is still in progress.
(10) ON89 Testino of Revised FEA The departure free nucleate boiling correlation used for the design of the San Onofre 2 and 3 core is the Comoustion Engineering CE-1 correlation. However, the San Onofre 2 and 3 reactors will use fuel assemolies with succort grics i
l which are thicker and wicer than concaracle grids for the 16x16 fuel design in Artansas Nuclear One Unit 2 (ANO-2). Also, tne grid spacing has been increased relative to the grid scacing for AND-2 by using one less grid for the Duncle. The effect of these changes in grid cesign say be to reduce tre critical heat flux for San Onofre fuel relative to that for ANO-2 and other plants which use the same grid casign as ANO-2.
Therefore, we recuested that the applicants provice cata to justify the use of the OE-1 CHF correlation.
This data had Deen submittec Dy the applicants, Dut our review of it is not yet complete.
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l (11) Contairment pressure Soundary Fracture Tovanness l
The San Onofre 2 and 3 contairment pressure councary is comprised of ASME Code Class 1, 2 and MC components. In late 1979, we generically reviewed the fracture tougnness requirements of the ferritic materials of Class MC, Class 1 and Class 2 components which typically constitute the containment pressure boundary. Based on this review, we detereined that the fracture toughness requirements containoa in ASME Code Editions and Aasenca, typical of those used in the oesign of the San Onofre 2 and 3 primary containment, may not ensure compliance with GDC 51 for all areas of the containment pressure councery. We initiated a program to review fracture toughness requirements for contairment pressure bouncary meterials
- for the purpose of cefining those fracture toughness criteria that most appropri-ately accress the requirements of GCC 51. Prior to completion of this generic stucy, we elected to apply in our licensing reviews the criteria identified in the Summer 1977 Addenca of Section III of the ASME Code for Class Z components.
These criteria were selected to ensure uniform fracture toughness requirements, consistent with the containment safety function, are applied to all components in the containment pressure boundary. Accordingly, we have reviewed the Class 1, 2 and MC components in the San Onofre 2 and 3 can'ainment pressure bouncary according to the fracture toughness recuirements of the Summer 1977 Acconca of Section !!! for Class 2 components. However, in orcer to cosclete our review we recuire additional information, Decause the San Onofre 2 ano 3 FSAR does not
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provice the information necessary to character 1:e the fracture toughness of the reactor containment pressure pouncary within the context of GOC 51. We have recuested that the applicants provice the necessary information, and we will review it when it becomes availacle.
(12) Emerceaev 81annine We have reviewed the San Onofre site emergency plan against the criteria in NUREG-0654 Revision 1. ' Criteria for Preparation and Evaluation of Raciological Emergency Response Plans anc Prepareeness in Suppert of Nuclear Power Plants,"
Novencer 1980. Based on our review, we conclude that the San Onofre site emer-gency plan, wnen revisec in accordance with the acclicants' commitments, will provide an adequate planning basis for an accentable state of emergency preparecness, and meets the recuirements of 10 CFR 50 anc Appendix E thereto.
However, the San Onofre site emergency plan must te revised to accress the final criteria anc inclementation schecule for the emergency centers and thei-functions, emergency manoower levels, and meteorological systems.
The acclicants nave been recuesten to explicitly accress protective action determination and isoleeentation after an eartacuase in the revisec site plan.
In accition. FEMA nas been ecuested as part of their review of Feceral, Stats, anc local emergency plans to *eview the planning efforts for the areas arounc the site to assure that Drotective actions to te recomenced 3y the a:011 cants af ter earthouates could te inclemented anc are adecuate.
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After receiving the findings and detareinations ease by FElu on Feoeral. State, and local energency response plans, and after reviewing the revised site plan j
from the soplicants, we will provice our overall conclusions on the status of emergency prepareeness for San Onofre and relatea eeergency planning zones.
Our final approval of the state of energency prepareaness at San Onofre will l-be sace following implementation of the emergency' plans to include development of procedures, training and Qualifying of personnel, installation of equipment and facilities, and a joint exercise of all the plans (site, Federal, State, and local).
(13) Industria1 Security The applicants submitted a Mocified Amenced Security Plan as required by 10 CFR 8srt 73.55 encompassing protection of the San Onofre Nuclear Generating Station Units 1, 2, and 3.
The isolementation of this plan at Units 2 and 3 is currently undergoing a review prior to the issuance of operating licenses for these units and will ce reviewed throughout the plant's operating life to assure continuing comoliance with the requirements of Part 73.55 of 10 CFR 73.
The identification of vital areas and esasures used to control access to these areas, as coscribed in tre plan, eay be sucject to future amenments based upon a confirmatory evaluation of Units 2 and 3 to cetermine those areas where acts of sanotage signt cause a release of radionuclices in sufficient cuantities to result in dose ratas equal to or exceecing 10 CFR Part 100 limits.
(14) Review of CENDS-193 The analysis method usec for less-of-flow transients is described in CENPD-183, this report originally was dependent on the approval of CENPD-177, $ut CENPD-177 was witharawn free review at the rwuest of Concustion Engineering (Scherer, 1980a). Therefore, the staff review of CEMPD-183 was deferrec. Subsecuently, Comoustion Engineering amenceo CENPD-183 and removea the cepencence on CENPD-177 (Scherer,19800). we are cur *ently in the process of rescheculing our review of CENPD-183.
(15) teview of 0-tist we nave coroletea our review of the list of structures, systems, and components to unich the cuality assurance program applies (the Q-List), and Mave identifiec a nummer of systems which.e 3elieve should te acaec to the list. We have acvisec the acclicants of our position on these items, and they plan to respond in the met future.
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Tentative Presentation Schedule i
ACRS Subcommittee on San Onofre 2&3 Washington, D.C.
February 18, 1981 Organizational Presentation Approximate Speaker Time Time 1.
Opening Remarks by the Subcommittee 5 min 8:30 am Chairman 2.
Staff Presentation 2.1 Introduction NRR/H. Rood 5 min 8:35 am 2.2 Scope of NRC Review NRR/
15 min 9:40 am OSeismic Audit OSystems Interactions Studies 0Detailed Systems Reviews 2.3 Discussion of Open Items NRR/
10 min 9:05 am 0Differing Professional Opinions 3.
Applicant Presentation l
l 3.1 Introduction SCE/W. Moody 5 min 9:20 am 3.2 Applicant Organization SCE/
25 min 9:25 am l
I 0!nteraction with INP0 0Personnel Qualifications Break 10 min 10:05 am 3.3 Emergency Planning SCE/
10 min 10:15 am 3.4 Preoperational Testing SCE/
15 min 10:30 am 3.5 Quality Control Experience SCE/
20 min 10:55 am l
and Problems l
l 3.6 Security Industrial (Closed)
SCE/
20 min 11:25 am 0Separation of Units 283 Following Startup of Unit 2 l
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Tentative Presentation Schedule ACRS Subcommittee on San Onofre 2&3 Washington, D.C.
February 18, 1981 Organizational Presentation Approximate Speaker Time Time 3.7 Secondary System Chemistry SCE/
15 min 11:55 pm and Materials: Condensor, feedwater heaters, steam generator, water chemistry, plant chemist Break for Lunch 60 min 12:20 pm 3.8 TMI Related Issues 0Control Room Engineering, SCE/
15 min 1:20 pm including safety parameter display system and mimicking 0Hydrogen Control Provisions 10 min 1:45 pm 0Regulatory Guide 1.97, includ-15 min 2:00 pm ing vessel level instrumentation 0Natural Circulation 10 min 2:25 pm UAuxillary Feedwater System 10 min 2:40 pm Improvements 0Station Black-Out Analyses 10 min 2:55 pm 3.9 Summary and Conclusions 10 min 3:10 pm 4.0 Executive Session 10 min 3:20 pm Adjournment 3:30 pm e
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