ML20040C493
| ML20040C493 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/31/1981 |
| From: | Schewencer A Office of Nuclear Reactor Regulation |
| To: | William Cahill GULF STATES UTILITIES CO. |
| References | |
| NUDOCS 8201270731 | |
| Download: ML20040C493 (44) | |
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December 31, 1 981 o
DISTRIBUTION:
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. Attorne, OELD Mr. William J. Cahill, Jr.
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OME (3 Senior Vice President Ii c.
River Bend Nuclear Group
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bec' TERA Gulf States Utilities Company q
NSIC Post Office Box 2951 a, :-
C' NRC PDR Beatsnont, Texas 77704
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Dear Mr. Cahill:
SUBJECT:
RIVER BEND STATION UNIT NOS,1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION As a result of our review of your application for operatbg licenses for the River Bend Station, we find that we need additiomd trifomation in several areas of review. The specific information required is listed in the enclosure.
In order to maintain the current schedule for your licensfrg review, responses to the requested additional infomation should be available for our review within 45 days from the date of this letter. Please contact the NRC project manager for your facility if you desire any discussion or clarification of the infomation requested.
Sf cerely, origina.\\ signed b78 A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
As stated cc: See next page
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Mr. William J. Cahill, Jr.
Senior Vice President River Bend Nuclear Group Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 cc:
Troy B. Conner, Jr., Esquire Conner and Wetterhahn 1747 Pennsylvania Avenue, N. W.
Washington, D. C. 20006 Mr. J. E. Booker Manager -Technical Programs Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 Stanley Plettman, Esquire Orgain, Bell and Tucker Beaumont Savings Building Beaumont, Texas 77701 Karin P. Sheldon, Esquire Sheldon, Harmon & Weiss 1725 I Street, N. W.
Washington, D. C. 20006 William J. Guste, Jr., Esquire Attorney General -
State of Louisiana Post Office Box 44005 State Capitol Baton Rouge, Louisiana'70804 Richard fl. Troy, Jr., Esquire Assistant Attorney General in Charge State of Louisiana Department of Justice 234 Loyola Avenue flew Orleans, Louisiana 70112 e
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" REQUEST FOR ADBITIONAL GEOLOGIC AND SEISMOLOGICAL INFORMATION RIVER BEND STATION 230.3 As. discussed in Section 2.5.2.7.1 the probability of exceeding the (2.5.2)
OBE has been evaluated in the River Bend FSAR. Prohideabrief discussion on how uncertainty was accounted for iq this analysis.
Also provide the staff with the recurrence statistics assumed for each seismic source zone.
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230. 4 As discussed in Section 2.5.2.3.1 some of the seismic events in (2.5.2) southern Louisiana are considered to be due to growth or slump faults.
As shown in figure 2.5-26 the Baton Rou,ge growth fault,izone is located from about 11 to 35 kilometers south-southeast from the River Bend site.
Additionally the staff has recently become aware of mic' rose'ismicity associated with dehelopment of Gulf Coast geopressured-geothermal energy wells in regions of growth faulting (Mauk et al. presented at the Fifth Geopressured / Geothermal Energy Conference at Baton Rouge, LA Oct.1981).
Prohide a discussion of the potential for gecpressure-geothermal well actihitywithin35kilometersofthesite. Alsoprohideadiscussionof currentobservationsofseismicactihityinareasofgrowthfaultingfrom both natural and man-inducsd sources including the. potential for strong ground motion from these sources.
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231.3 Present 5 summary, with conclusions and date of 1$st rehision, of GulfStatesUtilitiesCompany'spost-CPSafetyEh51uEtionReport geologic'andseismologiceffortsrelatihetoupdatingtheRiverBend Station FSAR through December, 1981. This sunraary is to include information derihed/ produced by both.the. applicant as well as'by others.
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PleaserehiseappropriatesectionsoftheFSARaccordingly.
231.4 In order to conform with the Standard Rehiew Plan (NUREG-08bO), Section 2.5.3,
- and to confirm the absence of the surficial expression of geologic
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structure expressed by lineaments, please conduct A remote sensing lineament analysisofthedreawithinatleastE5-mileradiusoftheRiverBend Station.
AsindicEtedintheStEndardReviewPl$r.,prohidethest$ff with a duplicate set of the photographs and imagery used in your analysis."
231.S The Cretaceous Tuscaloosa formation, a prolific gas producing horizon
( 2.5,,1 )
in areas south of the Riher Bend Station, underlies the site Et depths greater than 15,000 ft.
Extensive exploration (drilling, doUnhale geophysicallogging,perhapsseismicreflection)hasbeenconductedinthe near-site area since discovery of gas in the Tuscaloosa at the False River field (12 miles southwest of the site) in 1975. On this basis, prepare a new FSAR figure (or revise an existing. figure, if appropriate) showing theseismicreflectionsurheycoheragewithin5milesoftheRiher Bend. Station.
Include a plot of the petroleum exploration test holes
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(either completed, being drilled or planned).on this figure..
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231.6 In addition to the north-south seismic reflection survey profile shot' (2.5.1) in 1970 which is located 1.5 mi east of the site, the FSAR (p. 2.5-35) indicatesthatmorerecentr. ear-sitepropriet$ryseismicprofileswere examined and were determimd to be fault-fre'e. Design $te the locations of these additional seismic profiles (FSAR References 87 and 88) *on the new figure (or other appropriate FSAR figure) suggested in imC RAI 231.3 and provide new figures depicting the non-proprietary portions of these seismicreflectionsurheyprofiles. These new figures should be similar to Figure 2.5-35.
Furnish, for staff reYiew, $n $ctu$1-size copy of each of the processed profiles (197b $nd post-197b)'.,
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241.1 The applicant has stated that surface and foundati.on deformations (Section 2.5.4.13) will be monitored until such time as the movements have essentially (RSP) ceased.
It is the staff's opinion that regular monitoring of foundation movements th'roughout the. life of the plant is necessary to provide assurance that the foundations are performing their functions as expected and that external influences (such as groundwater level changas) are, in fact, having no adverse effects on the plant structures.
We request that the applicant describe the monito' ring program that will be adopted to monitor structure movements during the life of the plant including the frequency of monitoring and the evaluation of results.
We also require that t'he monitoring' program be made a Technical Specification.
Our basis for requiring the monitoring program to be part of the Technica'l Specifications is that unexpected structural movements can occur and if they are not monitrcad and evaluated regularly, potentially damaging mov.ements will not be recognized and corrected in a timely manner.
The Technical Specification will assure that'the safety of the plant will not be compromised by excessive structure movements.
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- 260.0 Quality Assurance-: a3 260.1 Provide an organization chart which clearly identifies all the major "onsite"
' (17.2) and "offsite" organizational elements which function under the cognizance of the QA program (such as design, engineering, procurement, manufacturing, con--
struction, inspection, test, instrumentation and control.-nuclear engineering, etc.), and describe their responsibilities if not already identified in the SAR.
260.2 Provide a more complete description of the authority and responsibilities of (17.2) each of the ~ quality assurance and quality control positions identified in Figure 17.2-1, "GSU QA Department Organization.".
L(260.3Describe the. criteria used for determining-the size of the QA and QC organiza--
17,2) tional dbpartments.-
(260.4 Describe the prom sions which assure that designated QA individuals are involved 17.2) in day-to-day plant activities imprtant to safety (i.e., the QA organization routinely attends and participates in daily plant work schedule and status meet-
-ings'.to assure they are kept abreast of day-to-day work assignments throughout.
the plant and that there is adequate QA coverage relative to procedural. and inspectica ' controls, acceptance criteria, and QA staffing and qualification of personnel to carry out QA assignments).
250.5 Clarify that the indoctrination and training programs (Ref. Section 17.2.2.4)
(17.2) are established'such that:
f (a)
Proficiency tests are given te +. nose' personnel per. forming the. veri-fying activities affecting quality, and acceptance. criteria are de-
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veloped to determine if individuals are properly trained and qualified.
(b)
Proficiency of personnel p'erfcming and verifying activities affect-ing quality 'is maintained by retraining, reexamining, and/or recerti-tying as determined by management or program commitment.
260.6 Clarify that procedures are established requiring a documented check to verify (17.2) the dimensional accuracy and completeness of design drawing _ and s~pecifications.
260.7 Clarify that procedures are: established requiring that design drawings and,
(17.2) specifications be reviewed by the QA organization to assure that the documents are. prepared. reviewed, and approved in accordance with company procedures and that the docu.ments contain the necessary quality assurance requirements such as inspection and test requirements, acceptance requirements, and the extent of documenting inspection and test results.
260.8 Clarify that guidelines or criteria are established;and described for determin-(17.2.)
ing the method of. design verification ^ (design review, alternate calculations,
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260.9 Clarify that measurcs are provided to assure that responsible plant peisonnel (17.2) are made aware' of design changes / modifications which may affect the performance of their duties.
260.10 Clarify tnat procedures are established for design verification activities which (17.2) assure the following:
(a)
Design verification, if other than by qualification testing of a pro-totype or lead production unit, is completed prior :to release for pro-curement, modification, or to ancther organization for use in other
. design activities.
In those cases where this timing cannot be met, the design verification may be deferred, providing that the justifica-tion for this action is documented and the unverified. portion of the design output document and all design output docume'nts, based on the unverified data, are appropriately identified and controlled. Activities associated with a design or design change should not proceed without verification past the point where the installation would become irrever-sible (i.e., require extensive demolition and rework).
In all cases, the design verification should be complete prior to relying upon the component, system, or structure to perform its function.
(b)
Procedural control is established for design documents that reflect the commitments of the SAR; this control differentiates between docu-
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ments that receive formal design verification by interdisciplinary or multi-organizational teams and those which can be reviewed by a single individual (a signature and date is acceptable documentation for per-sonnel certification).
Design documents subject to procedural control include, but are not limited to, specifications, ca1culations, computer progrems, system descriptions, SAR when used as a design document, and drawings including flow diagrams, piping and instrument diagrams, struc-tural systems for major facilities, site arrangements, and equipment locations.
Specialized reviews shculd be used when uniqueness or spe-cial design considerations warrant.
(c)
The responsibilities of the verifier, the areas and features to be veri-fied, the pertinent considerations to be verified, and the extent of documentation are identified.
260.11 If the verification method is only by test, describe the provisio'ns which (17.2) assure that:
(a)
Procedures provide criteria that specify when verification should be by test.
(b)
Prototype, component or feature testing is performed as early as pos-sible prior to installation of plant equipment, or prior to the point when the installation would become irreversible.
260.12 Describe in more detail.your controls for the preparation of as-built drawings (17.2) and related documentati'on in a timely manner to accurately reflect the actual plant design and the specific responsibilities of the QA organization in this area.
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260,13 Describe the provisions which assure that procurement of spare and replacement (17.2) parts for structures, systems, and components important to safety is subject to present QA program controls, to codes and standards, and. to technical re-quirements equal to or better than the original technical requirements, or as required to preclude repetition of defects.
260.14 Describe the provisions which assure that for ' commercial "off-the-shelf" items (17.2) where specific quality assurance controls appropriate for nuclear applications cannot be imposed in a practicable manner, special quality verification require-ments shall be established and described to provide the necessary assurance of an acceptable item by the purchaser.
260.15 Describe the provisions for assuring that correct identification of material, (17.2) parts, and components is verified and docraented prior t'o release for fabrica-tion, assembling, shipping, and installation.
260.16 Describe the QA organizational responsibilities for qualification of special (17.2) processes, including procedures, equipment, and personnel and in the inspec-
' tion and verification of these activities to assure they are properly com-plied with.
260.17 Clarify that criteria are established to determine the accuracy requirem'ents (17.2) of inspection and test equipment and to determine when a test is required or how and when testing activities are performed.
Describe the responsibilities of the QA organization in this area.
250.18
~ Describe the QA organizational responsibilities for establishing, implementing, (17.2) and assuring effectiveness of the calibration program.
260.19 Identify the organization (s) responsible for the review and concurrence of
' (17.2) those procedures relating to calibration.
260.20 Describe in more detail the responsibilities and involvement of the QA organi-(17.2) zation in the review and documented concurrence of nonconformances including the close outs of nonconformances and in determining and assuring adequate cause and corrective actions to nonconformances.
The response should address the extent the description of 'the nonconformance and the implementation of the disposition is verified and the guidance used in determining when to initiate corrective actions.
Also describe the extent nonconformance control and cor-rective actions apply to activities which do not conform to requirements.
260.21 Describe the QA and other organizations' responsibilities for managing and (17.2) maintaining QA records.
260.22 Describe those provisions to assure that activities under 10 CFR Part 50, (17.2) 50.55(e) will be conducted in accordance with the applicable controls of the River Bend QA program.
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CHEMICAL. ENGINEERING '
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281',2 Verify that the inital total capacity of new demineralizer. resins (condensate and primary coolant) will be measured. Describe the (5.4.8) method to be used for this measurement (Regulatory Position C 3 (10.4.6).
of Regulatory Guide 1.56, revision 1).
281.3 Describe the method of determining the condition of the demineralizer (5.4.8) units so that the ion exchange resin can be replaced before an (10.4.6)
. unacceptable level of depletion is reached (Regulatory Position C.4 of Regulatory Guide 1.56, revision 1).
Dsecribe the method by which (a) the conductivity meter re'adings for the condensate cleanup system will be calibrated, (b) the flow rates through each demineral,
izer will be measured, (c) the quantity of the principal ions likely tocausedemineralizerbreakthroughwillbecalculated,and(d)the accuracy of-the calculation of resin capacity will be checked.
281.4 Indicate the control room alarm set points of the conductivity meters
.(5.4.8) at the. inlet and outlet demineralizers in the condensate and reactor (10.4.6) water cleanup systems when eith'er(Regulatory Position C.5 of Regulatory Guide 1.56, revision 1):
a.
The conductivity indicates marginal performance of the demineralizer system; b.
1he conductivity indicates noticeablebreakthrough of one or more demineralizers.
281.5 Indicate the reactor coolant limits and corrective action to be (5.4.8) taken if the conductivity, pH, or chloride content is exceeded these (10.4.6) established in the Technical Specifications. Describe the chemical analysis methods to be used for their determination (Regulatory Position C.6 of Regulatory Guide 1.56, revision 1).
281.6 In accordance with Regulatory Position C.1 of Regulatory Guide 1.56, (10.4.6)
, revision 1, describe the sampling frequency, chemical analyses, and established limits for purified condensate dissolved and. suspended solids that will be performed and the basis for these limits 281J Establish and state the sequential resin repiccement frequency in (10.4.6) order to maintain adecuate capacity margin in the condensate treatment system (Pec u tory Posit.iu6 C.2 of Regulatory Guide 1.56, revision 1).
Include the basis for the resin replacement; frequency.
281. 8 Describe the water chemistry control program to assure maintenance (10.4.6) of condensate demineralizer influent and effluent conductivity within the limits of Table 2cf Regulatory Guide 1.56, revision 1.
Include conductivity meter alarm set points and the corrective action to be taken if the ' limits of Table 2are exceeded.
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281.9 Indicate the chemical limits, such as pN, chloride, and conductivity, (6.1.1) of the suppression pool water and the condensate storage tank water, and describe how they are maintained to ensure that the -
chemical limits stated in Section II.B.1.6 of SRP 6.1.1 are met a.
under normal operating conditions.
281.10 a.
Indicate the total amount of paint or protective coatings (6.1.2)
(area and film thickness) used inside containment that do not meet the requirements of ANSI N101.2 (1972) and Regulatory Guide 1.54.
We will use the above information"to estimate the rate of combustible gas generation vs. time and the amount (volume) of
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solid debris that can be formed from these unqualified organic materials under DBA conditions and that can potentially reach the containment sump. A G value of 5 will be used unless a lower G value is justified technically.
'b.
In order for the staff to estimate the rate of combustible gas generation vs. time due to exposure of organic cable insulation to DBA conditions inside containment, provide the following information:
- 1) The approximate total quantity (weight and volume) of organic cable insulation material used inside containment, including uncovered cable and cable in closed metal conduit or closed
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cable trays. 'We will give credit for beta radiation shielding for cable in closed conduit or trays if information is provided as to the respective quantities of cable in closed conduits or trays vs. uncovered cable.
- 2) The approximate breakdown of cable diameters and conductor or cross section associated or an equivalent cable diameter and conductor cross section which is represer)tative of the total cable surface area consistent with the quantity of cable surface area identified in 1) above.
^3) The major crganic polymer or plastic material associated with the cable in 1) above.
If this information is not provided, we will assume the cable insulation to be polyethylene and a G value of 3.
281. 11 Describe the samples and instrument readings and the frequency of (9.1.3) measurement that will be performed to monitor the water purity and need for spent fuel pit cleanup system demineralizer resin and filter replacement. State the chemical and radiochemical limits to be used in monitor.ing the spent fuel pool water and initiating corrective actions. Provide basis for establishing these limits.
Your respcase should consider variables such as:
gross gama and iodine activity, demineralizer or filter differential pressure, demineralizer decontamination factor, pH, and crud level.
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281.12
. Verify that sample line purge flows and duration times are
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(9.3.2) sufficient to fluih' out stagnant lines to assure that a representative sample is obtained.
(The NSSS vendor recannends a flush of 6-10 sample line volume, at a purge flush rate of about 2-3 times the sample flow rate).
281. 13 Acceptance Criterion 3f in Standard Review Plan Section 9.3.2 (9.3.2) states that passive flow restrictions or environmentally qualified, remotely operated isolation-Vilves.t.o limit reactor coolant loss from a rupture of the sample line should be pro-vided. This has not been addressed in your FSAR. Describe how the requirement of maintaining radiati6n exposures to as low as is reasonably achievable will be met in the event of a rupture of the sample line containing contaminated primary coolant, in accordance with Regulatory Position C.2 i(6):6f Regulatory Guide 8.8, revision 3 (June 1978).
281. 14.
(9.3.2)
Acceptance criterion 1.b in Standard Review Plan Section 9.3.2 indicates that sumps inside containment and the standy liquid control storage tank shculd be sampled.. Describe prov,isions to sample sump water inside the containment in accordance with the requirements of General Design Criterion 64 in Appendix A to 10 CFR Part 50.
281. 15 Provide information that satisfies the attached proposed (9.3.2) license conditions for post-accident sampling.
281. 16 Describe provisions for monitoring filter /demineralizer (9.3.4 )
differential. pressure to assure that pressure differential limits are not exceeded (Section II.8.b of the Standard Review Plan 9.3.4).
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311.5 It is noted in Tables 2.1-5 thru 19 that the population data has (2.1 )
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been given in metric measu..rement. Please provide this data in the english system of miles to correspond with the distances listed a ::.
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in Regulatory Guide 1.70, Section 2.1.3 Population Distribution.
311.6 Figure 2.1-1 of the River Bend FSAR does not show an airport near (2.1 )
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St. Francisville while the 1965 USGS topographic map of the area.
(St.Francisvillequadrangle)indicatesthattheDipple-Enette
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airport with.2>ranways of approximately b mile in length is
. located approximately 4h miles northwest of the reactor site.
Please indicate the current use, if any, of this airport.
311. 7 Section' 2.2.2.1 and Appendix 2A of the FSAR indicates that in" -
( 2. 2. 2.1 )
1979.the Louisiana Department of Natural Resources.was initiating the maintenance of records of highway shipments of hazardous materials in 1975.
Please provide a copy of the shipments of.
these materials on Route 61 adjacent to the River Bend facility.
311 8 It is noted from FSAR section 2.2.3.2 " Effects of Design Basis
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(2.2.3.2)
Events" that chlorine is the only chemical judged to have a p'otential impact on the ' main control room habitability.. It has recently come to our attention that a= tonia compounds are currently beingshippedbybargepasttheGrandGulfsitenorthdfthe River Bend site.
Please indicate the frequency and maximum quantity, if any, of the shipment of these materials past the River Bend site'.
If the number of shipments exceeds the Regulatory Guide 1.78 value of 50 per. year', please provide an analysis of a gaseous release to the environment near the nuclear facility.
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311.9 Page 2.2-8 of the River Bend F5AR indicates that an
( 2. 2.3.1 )
accidental release from an ammonia truck on,US Highway' 61 would not pose a habitability problss (for the control room) because of gas buoyancy.
Volume 22 " Ammonia Plant Safety - a tec6nical manual published by the American Institute of Chemical Engineers - 1980 indicates that in addition to the buoyant fraction of the gas released to the environ-ment, there is also a fraction of the release in which the air will be adiabatically saturated with ammonia.
The latter becomes denser than the ambient air and fog.
is also created because of water vapor.
In view of this manual the potential for non-buoyant ammonia vapor clouds and the accident histories of events as published in the ilational Tr'ansportation Safety Board reports on anhydrous ammonia releases, please provide an analysis which censiders the effect of the denser-than-air fraction of an s=monia release on the control room from an accidental spill on Route 61, approximately 1 mile from the River Bend Station.
os 410.14 Table 3.4-1 and Section 2.4.2.2 show'the design basis flood level (3.4.1 )
(DBFL) to be 95 feet-one inch. The Table also shows that pipe penetrations in the reactor building, at the 99 feet level, are subject to the DBFL.
Explain what appears to be an inconsistency 2nd make appropriate corrections where necessary.
.410.15 The staff finds that your rationale _ is insufficient justification
( 3. 5.1.1 for not viewing components such as valve bonnets, thermowells, 3.5.1.2) nuts, bolts, studs and valve stems as possible missiles. There fore, you must provide satisfactory assurance that these components will damage neither safety-related. structures, systems or components (SSC) nor SSC which could cause safety-related.SSC to fail to per-form their safety functions either by showing that such component missiles will not affect the safety-related o' appropriate non-r safety-related SSC or that suitable missile barriers have been or will be provided for the protection of the appropriate SSC.
410.16 Provide assurance that in your review you have considered missiles
( 3. 5.1.1 due to gravitational effects (of such components as electrical 3.5.1.2) hoists or any unrestrained equipment and non-safety-related items
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such as piping, non-Class 1E conduit, instrument tubing trays, structures, HVAC ducting, and non-Class 1E cable trays) during maintenance times, reactor operation and following any abnormal event.
410.17 Verify that the Scram Discharge System satisfies the criteria (4.6) enumerated in the Generic Safety Evaluation Report, BWR Scram.
Discharge System, dated December 1,1980, which was transmitted to you by NRC letter. dated December 22, 1930.
410.18 Verify that the River Bend CRDS design is in full compliance with (4 6) the applicable criteria enumerated in NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking."
410.19 The following errors are seen on the P and ID's' for the CRDS (4.6)
(Figures 4.6-Sa and 4.6-5b):
a.
Figure SA, Area A-2, arrow refers to Sheet 2. H-11 Figure 5A, Area 0-2, arrow refers to Sheet 2,:B-11 Figure SA, Area H-2, arrow refers to Sheet 2, B-11 There is no coordinate 11 on either figure.
b.
Figure 5b, Area E This appears to be the scram accumulator but is not titled to indicate what this is.
Correct these figures to eliminate all incorrect locations such,
as H-11, B-11.
Indicate what you mean when you refer to Sheet 2 on Figure 4.6-Sa; is this /fgure 4.6-5b?
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410.20 Verify that the particulate channel of the fission products (5.2.5) monitoring subsystem receives its power from a Class 1E source.
410.21 The drywell equipment drain sump receives hot and cold reactor (5.2.5) coolant leakage. Leakage from " hot" sources such as the reactor vessel head flange vent drain and valve packings may flash into steam which must be condensed to reach the sump.
Provide assurance
.that such steam will be condensed for leak detection monitoring purposes.
410.22 The floor drain system, which is used to detect liquid leakage, (5.2.5) should be tested periodically for blocked lines.
Discuss any surveillance program planned to minimize the potential for drain system blockage.
410.23 In tne FSAR (Section 5.2.5.1.4), you discuss the detection of (5.2.5) intersystem leakage but do not show how you detect intersystem leakage from the RCPB into the RHR (LPCI), LPCS, and RCIC (both water and steam turbine sides)~ systems and into the secondary side of the recirculation pump heat exchangers as required by the Standard Review Plan (Section 5.2.5, " Reactor Coolant Pressure Boundary Leak Detection").
Provide for detection of
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such intersystem leakage or provide suitable justification for
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any such omission.
410.24 In Section 9.1.1.3.1 of the FSAR you indicate that the new fuel
( 9.1.1 )
storage arrangement will not exceed a k of 0.95 assuming the f
new fuel storage area was dry or floode8 with unborated water.
Verify that a k equal to or less than 0.98 will be maintained with new fuel oi (he highest anticipated reactivity assuming f
optimum moderation, for example, foam, spray, small droplets' or mist.
410.25 Show either that an SSE will not cause the fa[1ure of the st'ain-(9.1.2) less steel plates which are used to line the containment and spent fuel pools or that such failure will not result in any of the following:
(1 ) Significant release of radioactive material due to mechanical damage to the spent fuel; (2) Significant loss of water from the pool wh'ich could uncover the fuel, cause cladding faage and lead to release of radioactivity due to heat-up; (3) Loss of ability to cool the fuel due to flow blockage caused by one complete section of the liner plate (or portion thereof) falling on top of the fuel racks.
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410.26 Document the seismic category of the gates used to separate the (9.1.2) various sub-pools inside containment and the fuel building.
If the design does not meet Category I requirements, discuss how a failure of the gates as a result of an SSE will not result in similar conditions to those stated' for the pool liner in Question 410.25 above.
410.27 In Section 9.1.3.2.1 you state that the RHR system may be used to (9.1.3) cool the containment fuel storage water only during the refueling stage. Does this imply that the RHR system must be used to cool the primary system in order for it also to be used to cool the containment fuel pool?
410.28 The P&ID (Figures 9.1-23 a and b) shows one return line and one (9.1.3) feedline from the RHR system to the spent fuel cooling system while the RHR systm P& ids (5.4-12 and 5.4-13) show redundant feeds and returns.
Explain these apparent discrepancies and correct the P& ids accordingly.
410.29 Section 9.1.3.2.1 further states that one fuel pool cooling pump (9.1.3) and heat exchanger set will be in operation. cooling the fuel building pool while the other set will be in standby or used for containment fuel pool cooling. Assuming both sets in operation cooling the two separate pools, demonstrate that any single failure in one of the sets or the piping will not prevent successful short-term or long-term cooling of the spent ~fdeT~in ettheFpooT.
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410.30 Will there be a technical specification or other provision pro-(9.1.3) hibiting storage of spent fuel inside containment while the reactor is in operation?
41 0.31 Show that the loss of one RHR train (either "A" or "B" train) will (9.1.3) not prevent satisfactory cooling of the reactor core, containment fuel pool or spent fuel pool during the refueling operation.
In your consideration, select that single failure to make the RHR train inoperative which would produce the most severe consequences and explain your selection.
410.32 Table 9.1.5 in.iicates that a normal fuel pool cooling load is 11.87 (9.1.3) x 106 BTU /HR and the abnormal load is 24.73 x 106 BTU /HR.
Indicate how these loads were arrived at.
410.33 It is anticipated that there may be no spent fuel to store at River (9.1.3)
Bend urtfl-at least after completion of $q f.f rst dperative cycle; under these conditions the ' fuel pool cooling subsystem'and'the fuel pool purification subsystem may not be'needed or'used until the start of the first refueling cycle.
Demonstrate that operability of either or both subsystems (fuel pool cooling and fuel pool purification) will be assured sufficiently in advance of anticipated need to permit any required repair or adjustment.
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410.34 You cited Section 5.2.1 in the narrative in Section 9.1.3.2 which
( 9.1. 3) states that water from the containment fuel ;torage pool may be circulated through the RHR system. Section 5.2.1, however, deals with ASME code compliance.
Correct the FSAR to cite the section of the RHR description which deals with containment fuel pool water.
410.35 On December 22. 1980, a letter was sent to you concerning the con -
(9.1.4) trol of heavy loads. of the letter identified interim criteria.
Provide the staff with your commitment to implement the interim criteria prior to the final implementation of NUREG-0612 guidelines and prior to the receipt of an operating license.
410.36 Discuss how you intend to control the handling of objects over the (9.1.4) spent fuels in both the containment and spent fuel buildings so that the maximum kinetic energy of any such object, if dropped from the height at which it is normally handled above a storage rack, will not exceed the kinetic energy of one fuel assenbly and its associated handling tool.
Include objects of less weight than a spent fuel assembly in your consideration.
410.37 The Standby Service Water (SSW) System contains a number of motor
( 9. 2.1 )
operated valves required to function, in an automatic or manual mode, following an emergency condition concurrent with loss of offsite power.
Verify that these valves are connected to Class 1E buses, and that the loss of any diesel generator will not result in loss of proper operation of the system.
410.38 Table 9.2-13 indicates that only SSV pump A operates following a
( 9. 2.1 )
DBA.
Section 9.2.7.3 says that two pumps must operate when the RHR heat exchangers are in use. Explain SSW systems operation in de'tatl, 'following"a DBA',' including ithe effect of a single failure in the SSW system.
410.39 Table 9.2-15 indicates that there are two HPCS diesel Jacket water
( 9. 2.1 )
coolers.
Figure 9.2-la shows only one cooler.
Correct this dis-crepancy.
410.40 Table 9.2-15 indicates that each standby diesel generator jacket
( 9. 2.1 )
water cooler requires 700 gpm of cooling water.
The table also indicates that the HPCS diesel has two jacket water coolers each requiring 800 gpm of cooling water. Why does the HPCS diesel, which is considerably smaller than the standby diesels require so much more cooling water? -
410.41 The HPCS room unit cooler and the auxiliary building unit cooler
( 9. 2.1 )
1 HVR-UCl(Z-1) are not shown on the P& ids. Correct this discrepancy.
410.42 Section 9.2.7.2 states that the redundant SSW headers are cross
( 9. 2.1 )
connected with redundant normally closed valves.
The P&I3 show:
the valves, 505 A and B, to be normally open.
State the correct position of the valves and correct any discrepancy.
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410.43 Provide a drawing which shows the routing of the piping in the yard (9.2.1 )
between the ultimate heat sink and the reactor complex together with infomation on non-safety related piping in the yard that may cross the path of the safety related piping.
410.44 Figure 9.2-21 is a rudimentary schematic of the C'ondensate Storage (9.2.6)
Facil i ties.
Provide a P&ID of the system that delineates the seismic category and quality group of the various parts of the system and includes the isolation provisions at the interfaces with sa fety-related systems.
The lines from the RCIC system,not shown, should be included in the P&ID.
410.45 In Section 9.2.6.3 you state that the water supply to the high pres-( 9. 2. 6) sure core spray pump (and RCIC pump) is automatically transferred to the suppression pool on low condensate storage tank level.
(1) What is the seismic category and quality group of the HPCS actuation line?
(2) Are the level instruments together with their power supplies, transmitters, readout equipment, etc., safety related?
(3) Are the level instruments connected to the condensate storage tank or to the pipe (s) leading to the suction of the HPCS.and RCICpump(s)?
410.46 Discuss what would happen if there were a crack in the piping leading (9.2.6) to the HPCS and/or RCIC pump (s) when either or both pumps were in operation, taking water from the condensate storage tank.
Consider that crack which would cause the most severe consequences.
The crack should be consistent with the provisions of Branch Technical Position MEB 3-1, entitled " Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment."
410.47 The P&ID for the instrument air system (Figure 9.3.1) does not show
( 9. 3.1 )
a supply to the ultimate heat sink or emergency service water pump house.
Confirm that these safety-related systems do not require instrument air or indicate how they are supplied with cc: pressed air of instrument air quality.
410.48 Section 9.3.6.2.3 of the FSAR -(Penetration Valve Leakage Control (9.3.1)
System) states that the air compressors must operate for at least 30
~
days following a LOCA.
The P&ID (Fig. 9.3-13) shows 'the use of service water as the source of cooling for the compressors and a fter-coolers.
How are the compressors and after-coolers supplied with cooling water when the non-safety-related service water system is not available (i.e., following an SSE or loss of offsite power)?
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410.49 Table 3.2-1, Item XLII, indicates that the compressors are seismic
( 9. 3.1 )
Category I, Safety Class 2, but does not list the after-coolers, moisture separators, accumulators, piping and valves to the main steam safety and relief valve system, and containment isolation valves and piping.
Expand the table to include these items.
Verify the seismic and safety classification of the compressors.
410.50 We cannot find a P&ID for the ma'in steam safety and relief valve (9.3.1)
-system.
Provide such a PSIC.
41 0.51 The Penetration Valve Leakage Control System (FSAR Section 9.3.6),
(9.3.1) which supplies compressed air to the safety-related compressed air system (SRCAS), does not contain air dryers and after-filters.
Discuss how you will meet the requirements of ANSI MC 11.1-1976 for air quality as stated in Section 9.3.1 of the Standard Review Plan (Paragraph III.2) for the compressed air supplied to the SRCAS.
41 0.52 Provide the capacities of the sumps, drain tanks, and sump pumps in (9.3.3) the Equipment and Floor Drain System, 410.53 In Section 9.3.3.3 you state that the sump pump discharge lines from (9.3.3) the ECC cubicles contain two spring loaded check valves located on each side of the cubicle wall penetration.
Show this arrang(ment on the P&ID along with the safety class information.
410.54 On Page 9.3-17 (Section 9.3.3.3, " Safety Evaluation") of the FSAR, (9.3.3) you note that the walls of the LPCS, RHR(3), RCIC and HPCS cubicles in the auxiliary building are waterproofed to an elevation of 95'0."
Provide assurance that waterproofing to this level is correct in view of the question of pipe penetrations at the 99 foot level being subject to the DBFL.
410.55 Provide a key to the symbols used in the heating and ventilating
( 9. 4.1 )
P& ids.
410.56 Item XXXI of Table 3.2-1, " Auxiliary Building Ventilation System,"
( 9. 4.1 )
says "all components with safety functions" are seismic Category I, Safety Class 3, etc.
Expand this item of the table to list all the components as is done in other parts of the table.
Provide a similar listing for the control building ventilation system.
~410. 57 The chilled water system described in Section 9.2.10 of the FSAR
( 9. 4.1 )
providcs the cooling for the Control Room Area Ventilation System and will be reviewed as part of this system.
Table 3.2-1 of the FSAR makes no mention of this safety-related chilled water system.
Correct this deficiency.
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410.58 Which subsystem provides the heating and ventilating for the access
( 9. 4.1 )
control area and the computer room?
410.59 Figure 9.4,la shows redundant radiation detectors in the remote
( 9. 4.1 )
outside air intake and the local outside air. intake but does not show any chlorine detectors as stated in Section 9.4.1.3.
Verify that redundant chlorine detectors are located in both outside air intakes and correct the drawing to show them.
410.60 Figure 9.4.2a indicated (Note 2) that the exhaust ductwork and (9.4.2) dampers shown on this figure and on Figure 9.4-2b is SC-3.
Note 2 on Figure 9.4-2b indicates that the ductwork and dampers are NNS.
Confirm the seismic and quality classification of the exhaust duct-work and dampers and correct the drawings to reflect this.
41 0.61 Item XXXIX of Table 3.L,1 says that only the charcoal filter system (9.4.2) exhaust fans are seismic Category I, Safety Class 3, while Section 9.4.2.2.1 mentions that the outside air intake louver.. tornado dam-per, fire dampers and ductwork to filter plenum are also designed to Safety Class 3, seismic Category I standards.
Expand Item XXXIX to show the seismic and safety classification of all of the components of the Fuel Building Ventilation System.
410.62 Describe the means provided to assure that the temperature'in the (9.4.2) rooms housing the spent fuel pool cooling pumps and the charcoal filter exhaust trains.and exhaust fans can-be maintained at accept-able levels for operation under accident and emergency conditions when the normal fuel building HVAC system is not operating.
410.63 Table 9.4-1 indicates a design temperature range in the auxiliary (9.4.3) building supply and exhaust system of 122'F maximum and 40*F minimum.
Describe the means provided for meeting the minimum temperature in the winter when the plant is shut down.
41 0.64 Table 9.4-1 indicates that a minimum temperature of 40*F will be (9.4.5) maintained in the diesel generator buildings and standby service-water pumphouse.
Describe the means pr'ovidedfor maintaining this temperature in the winter when the diesels or standby service water pumps are not operating.
410.65 Describe the means providedto assure that the tenperature in the (9.4.5) rooms hcusing the safety-related hydrogen recombiners, the safety-related penetration valve leakage control system air compressors, and the various safety-related unit coolers can be maintained at acceptable levels for operation under accident and emergency con-ditions when the normal building HVAC system is not operating.
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410.66 Neither Table 3.2-1 nor the P& ids indicate the seismic category or (9.4.5) safety class of the components in the auxiliary building unit cooler systems.
Identify the coolers that serve safety-related equipment and the coolers (if any) that serve only non-safety-related equipment.
Verify that the coolers and their associated ductwork, dampers, etc.
that serve safety related equipment are classified seismic Category I, Safety Class 3, and amend Table 3.2-1 and the P& ids to show this in forma tion.
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450. 3 Provide the. locations of the control room remote air intakes.
(6.4) 450.4 For the protection of control room personnel against the release of
-(6.4) chlorine, Regulatory Guide 1.95 recommends (for the type of control' room at River Bend Station).a. control room leak rate equal to or less than 0.06 air change per hou'r.
In FSAR section 6.4.2.3, however, h leakage testing program with a maximum leak rate of one' air change per hour - a value that is more than 15 times the reconnended value - is proposed.
Justify your departure from the guidance of Regulatory Guide-1.95.
450.5 The description given in FSAR Section 6.4.2.6 relative to portable self-(6.4) contained air breathing units falls short of the staff and requirement that 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of onsite bottled air supply be.available for at.least 5 men with unlimited offsite replenishment capability from nearby location (s).
Justify your departure from the staff position.
450.6 In the evaluation of toxic gas concentrations, justify the use of 0.5 m/
(6.4) see (1.1 mph) wind speed. Our experience hus been that lower' wind speeds may not be conservative in the calculation of toxic gas concentrations because higher wind speeds decrease the time available for the operator to take protective actions following a toxic gas release and will increase the infiltration of toxic. gases into the control room.
450.7 (15.6) In Section 15.6.1 it is stated that the " plant design does not have instrument lines containing primary coolant outside the containment;'!
Figure 6.2-64, however, shows " typical" instrument lines which do penetrate containment.
In section 13.6, the main steam lines are the only lines connected to the reactor for which radiological consequences have been analyzed, on the stated grounds (15.6.6.5.1) that they envelope-the consequences of any other piping failure outside containment.
Justify this contention, assuming a coincident " iodine spike".
Do all instr'ument lines, sample lines, and other small. diameter pipes which penetrate the containment terminate within volumes served by engineered safety feature ventilation system?
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460.3 Table 6.5.1 of the FSAR discusses compliance of ESF Filter Systems (6. 5.1')
with Regulatory Guide 1.52 requirements. Address the following areas of non-compliance for the SGTS, Fuel Building Charcoal Filtration System and Main Control Room Air-Conditioning Subsystem:
- 1. Item C-2.g Either describe alternate mechanism provided to assure adequate flow or justify why facilities to record pressure and flow readings are not provided.
- 2. Item C-2.3 Compliance with Regulatory Guide 8.8 must be addressed.
- 3. Item C-3.k An air bleed cooling mechanism is not provide'd because analysis has shown that the charcoal will remain well' bel ow ignition temperature. These provisions are require $ not only for' pre-vention against self ignition, but also because at elevated temperatures, charcoal may have its adsorptive capacity for iodine reduced.
In light of this fact, an air cooling mechanism l
should be provided, or justification for'not providing it should
~
be presented.
- 4. Item C-4.d l
ESF filter systems should be run a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per l
month.
460.4 Provide a simplified flow diagram of the. liquid waste management (11.2) system which' clearly shows inputs to the system, processing equip-ment, and flowpaths (both nomal and alternate).
460.5 Page 11.2-5 of the FSAR states that the condensate storage tanks (11.2) are not provided with dikes because an analysis has shown that the dose consequences to surrounding or down river population are negligible.
If the tank fails, what will be the flow path of the leaking water? Is the tank on a concrete pad' to prevent the water from going into the ground? If not, what prevents the water from leaking into the ground?
460.6 Page 11.4-2 of the FSAR commits to supplying detailed information (11.4) on the solid radwaste system. This information must be supplied in order for the staff. to complete the reivew of the River Bend Station.
Please furnish the approximate date(s) when this information dill be supplied.
460.7 Table 1A-1 of the FSAR provides River Bend Station positions on (0737 II.F.1)
Post-TMI requirements (NUREG-0737). The response for Item II.F.1,
~
Items 1 and 2, is unacceptable. Provide inf6rmation in sufficient
- detail to show a commitment to how you will meet these requirements.
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Containment Systems Branch l
480.05
- General Electric has provided, on the GESSAR-II docket, th'e latest load criteria for both SRV and LOCA related pool dyn r.ic loads. The staff is currently reviewing these load specifications and will prepare a technical evaluation report by early 1982. In order for the staff to complete its review of the pool dynamic load definitions for your facility, we need the following:
a) A statement of your intention to completely utilize the load definitions presented in GESSAR-II as modi-fied, if necessary, by the staff's evaluation report; or b) A detailed list of'the exceptions to the generic load definition criteria (as modified by the staff's evaluation report) that will be used in the design of your facility; c) provide detailed plan and section drawings of the TIP statiori, : equipment hatch, personnel hat'ch, and any other structure 1ccated within 20 feet of the suppression pool surface whose width is greater than 20 inches.
Show on these drawings your plan to extend these structures into the suppression pool, thus eliminating iinpact loads due to pool swell; and d) Provide an analysis for suppression pool temperature response to various SRV events to demonstrate that the bulk temperature of the pool meets the prescribed limit.
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480.0 6
- Provide a comparison between the predicted mass and energy (6.2) released following a LOCA using both the LOCTVS computer pro-C gram (that is used for the River Bend analysis) and the General' Electric methodology described in NEDO-20533.
480.0 7 Describe and justify the nodalization study performed to cal-(6.2) culate the transient forces end moments acting on the RPV and tajor components.
Also, provide the following information:
a) Discuss the manner in which movable obstructions to vent flow (s'uch as insulation, ducting, plugs, seals and doors) were treated.
Provide analytical and experimental justi-fication that vent areas will not be partially or completely plugged by displaced objects.
Discuss how insulation'for piping and coraponents-was-considered-in-detemining-volumes and vent areas.
b) Provide the projected area used to calculate these loads and identify the location of the area projections on plan and section dtawings in the selected coordinate system.
This information should be presented in such a manner that con-firmatory evaluations on the loads and moments can be made.
c) Provide the peak and transient loading on the major components used to establish the ade;gacy of the design.
This should in-clude the idad forefog
- nctions [e.g., f (t), f (t), F (t)]
x y
and transient ne. ? r.zt, <. 9., M (t), Mj(t)] as resolved about x
t a specific, identified coordinate system.
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o O-2 480.08 Provide the mass and energy release data (in tabular form) used (6.2) in the short term response calculations for both the main steam line break analysis and recirculation line break analysis.-
480.09 In the subcompartment pressure analyses due to a break in the (6.2) reactor water clean up system (RWCU), credit for closing of the containment isolation valve was taken ( Table 6.2-12, 6.2-28 and 6.2.31). Therefore, provide the folloWing information:
a) The signal used to~. isolate. these valves' and the time these signals are gen ~erated including instnzmentatica delay times.
b) if credit is taken for reduction in f 3cw due to the closing
~
of valves (prior to full closure), prcride justification for such an assumpt_fon.
c)
Identify the vent area used in the surcompartment analyses that were provided by blowout panels.
For these vent arcas, discuss how the flow area and flow res'istance var'ies With time.
Provide the experimental data that support these assumptions or propose a testing program that will demonstrate this capabi-lity.
Also, provide an analysis that shows there will be no missiles generated.
~
480,10 Provide the mode of heat transfer assumed, i.e., condensing or (6.2) convective, and the correlation used to compute the heat transfer coefficient throughout the containment response calculation for the main-steam line break and recirculation line break analysis.
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480.11 Section 6.2.1.1.3.3.2 presents the containment external pressure (6.2) differential calculation. Provide the following information:
a) the results of all potential single failure analyses to arrive at the worst single failure assumption used in the
- analysis, b) the number of the pressure switches that isolate the chilled water control valves, and if they> are powered from different sources.
~
480.12-Explain why in the containment externa'l pressure differential (6.2) analysis, operation of all three containment unit toolers was not considered.
480.13 River Bend station design does not utilize vacuum relief devices. -
(6.2) to minimize the external pressurt. ciffe'rential. Provide the fol-lowing information to justify the c'ons'ervhti'sm.of the design.
' a) the experiment'al data used to determine that thb dynamic loads associated with reverse vent clearing are negligible, snd b) the potential flooding of' equipment and components in the drywell due.to reverst vent flow.
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O' 480.14 FSAR Table 6.2-40 shows the High Pressure Core Spray Pump Suction
~
(6.2) for Suppression Pool valve as GDC 55 FSA'R Figure 6.2-64 the valve as GDC 56. Please show the correct designations.
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480.15 FSAR Table 6.2-40 shows tbt the Reactor Water Clean Up Pump (6.2)
Discharge system is classified as an ESF system, FSAR Figure 6.2-65 indicates that the outboard and inboard piping to be Safety Class 3 (SC-3).
All ESF Systems must be classified as either Safety Class 1 or Safety Class 2 (SRP 6.2.4).
Provide
~
evidence that the requirements of SRP 6.2.4 are satisfied.
480.16 With regard to suppression pool bypass lea.kage provide the (6.2)
(9.4) following infor: nation:
a) the containment pressure following a small steamline break assuming.
2 bypass leakge path of AdK = 1.0 ft.and the containment cooler actuated no sooner than 30 minutes from-the time the contain-
~
ment pressure ' reaches the high containment pressure set point, b) if the analysis stated above, shows that containment design pressure will be reached, discuss your plan for including.
automated actuation of the containment coolers, and c) with regard to the drywell low ' pressure leclage test, it is our position that this test should be perfo'rmed at each refueling outage instead of' your proposed testing frequency of 3 per 10 it
t years. Therefore, provide your plan to comply with our position.
d) The FSAR Section 9.4.6.2.1,statcs that the containment unit i
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4
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coolers are designed to Safety Class 3 and Seismic Category I requirements. Since W.:oolers are' needed to mitigate the consequence of LOCA typa accidents, it is our position t' hat the reactor containmerit unit coolers must be designed, fabrica-
- i ted, ecected and tested to Quality Group B standards, as re-comended by Regulatory Guide 1.26.
Provide.information on how you will comply w.ith this position.
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U-480.17 From the discussion in the FSAR it is not clear whether or not (6.2) debris screens are included in the design of the containment purge system.
It is our position that Section B.l.g of BTP CSB 6-4 should be met. Guidance is provided below which.if followed, would represent an acceptable debris screen design.
a) The debris screen should be seismic Category 1 and installed typically about one pipe diameter away from the inner side of the inboard isolation ' valve.
b) The piping between the debris screen and the valve should also be seismic Category I design.
c) The debris screen should be designed to withstand the LOCA differential pressure; d) The debris screen opening tagcally should be about 2 inches by 1 3/16 inches.
State your intention to comply with our position and provide.a j
description of the debris screen design.
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480.18 Section 9.4.6 of the FSAR states that the.c'ontainment/drywell at-l (9.4) mosphere is purged during normal operation. However, we believe that purging / venting should be minimized during reactor operation because the plant is inherently safer with closed purge valves than with open lines requiring valve action to provide containment isola -
j l
tion.
In fact, serious consider.ation should be given to a plant design'such that purging / venting is not required during operation.
Therefore, provide a detakled discussion of the reasons why the
L.
River Bend Station Plant need to purge, and an estimate of the number of bours per year that purging is expected th'ough r
each particular valve.
Provide evidence that all containment purge valves are quali-fied for operation under the worst accident conditions including temperature and pressure.
(Reference BTP 6-4 position B.l.a) 480.19 SRP 6.2.4 Item II.5 establishes the design requirenents for closed (6.2) systems.
Provide evidence that the isolation systems shown in FSAR figure 6'.2-65 and classified as GDC 57 systems in 'the River Bend Station meet these r6quirements.
If these systems do not meet General Design Criterion 57 requirements, these systems must be' classified as GDC 55 or 56.
Revise the River Bend Station's FSAR accordingly.
Provide a complete drawing for each closed system outside contain-ment for which credit is claimed as an isolation barrier.
Show all piping connecting to the closed system up to a second isolation barri er.
Identify all lines connected to the closed system that leave the secondary containment, 480. 20 Describe the. provisions to detect possible leakage outside contain-
..,(6.2) ment from lines in engineered safety features (ESF) oi ESF-related systems, or in systems needed for safe shutdown of the plant, that contain remote-manual valves in accordance with SRP Section 6.2.4.
l 480.21.
FSAR.Section 3.2.2 sta.tes that. the designation NNS is comparable to (3.2.2)
Regulatory Guide 1.26' Quality Group D classification.
Quality Group Et
m D components are not acceptable for components performing a containment isolation function, including the isolation b:r-riers and the piping between them, or the piping between the containment and the outermost ' isolation ' barrier.
Provide in-formation demonstrating that the design provisions for penetra-tions meets the requirements of Standard Review Plan 6.2.4.
480.22 The RCIC Turbine Exhaust Yacuum Breaker from above the sur-(6.2) pression pool, listed in Table 6.2-40 and shown in Figure 6.2-64 does not meet the explicit requirements of General Design Criterion 56.
No isolation valve is provided inside containment.
Provide justification for non-compliance to this requirement.
480.23 The RCIC steam supply isolation valves are listed as ESF auto-(6.2) matically controlled valves.
The RCIC turbine exhaust isolation valves are listed as non-ESF, manually operated valves.
Verify that these valves.are listed correctly in FSAR Table 6.2-40.and provide the rationale for classifying the RCIC system as a non-engineered safety feature system
- 480.24 SRP 6.2.4 states that if a fluid system does not have a post ac-(6.2) cident function, the isolation valves in the lines :hould be rutomatically closed.
For engineered safety feature or engine-ered safety feature-related systems, isolation valves in the lines may remain open or be opened.
Do the following non-ESF valves, listed in Table 6.2-40, have a post accident function and if not, justify why they should b'e open under post accident conditions.
32-
RCIC Turbine Exhaust to Suppression Pool lE51 MOV F068 Containment and Drywell Purge Supply to Drywell lHVR A0V Fire Protection Header 1FPW MOV 121 & 122 Service Air Supply to Containm.ent and Drywell 15AS MOV 103 Instrument Air Supply to Cont, and Drywell lIAS V 79 Service Water Supply ISWP MOV 507A & 507B
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Service Water Supply to Drywell ISWP V 205 Air Supply for MS Safety & Relief
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Valve System.
ISWP MOV 1 A & 18 Air Supply for MS Safety & Relief.
Valve System ISVV V'E0 & V 53 Condensate Makeup Supply ICNS MOV 588, 598, 60B & 61B Vent Line 1B21 MOV F005 Instrument Air Supply IIAS V 237 & V 238 480.25 SRP 6.2.4 section II states that in gen'eral, two iselation barriers (6.2) in series are required to assure tnat the isolation function is l
satisfied assuming any single active failure in the containment isolation provisions.
The Service Water supply and retura systems t
isolation valves are not powered from diverse power sources, there-l fore subject to a single active failure.
Provide justification for powering the series isolation valves from a single power source.
480. 26 SRP Section 6.2.4 and NUREG 0737 Item II.E.4.2 Position 3 states (6.2) that if a fluid system.does not have a post eccid.ent function, the-35
O' isolation valves in the lines should be automatically cl'osed.
Some of the ' valves listed in FSAR Table 6.2-40 and show'n to be remote manual operated valves and closed post accident are listed bel ow.
Justify why these valves are not automaticakly clo~ sed.
Turbine Plant Miscellaneous Drains 1B21 MOV F085 Reactor Water Cleanup Backwash Discharge IWCS MOV:lli, lWCS MOV 173 HPCS Test Return IE22 MOV F023 lE22 MOV F012 RCIC Pump Minimum Flow Bypass lE51 MOV F019 RHR Heat Exchanger Vent lE12 MOV. F073A, B RHR and LPCS Test. Returns lE12 MOV F064A, B RHR Return "C" Minimum Bypass 1E12 MOV F064C Containment and Drywell Purge Supply ICPP SOV 140, lHVR A0V 123, i
lHVR A0V 125, t
ICPP MOV 104, ICPP MOV 105, 1,HVR A0V 148, and
~
t lHVR A0V 126 Instrument Air S'upply lI AS MOV 107.
Containment Atmosphere Monitoring Probe 1 CMS SOV 34F 480.27 The Same (EFF) A0 valves to reactor vessel shown on Figure 6.2-63 l
(6.2) are not li:ted in Table 6.2-40.
In addition, the following listed i
valves shown on Figure 6.2'-64 are not listed in Table 6.2-40:
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Containment and Drywell H2 Sample "A" and "B" solenoid operated valdes and located in the containment and drywell Drywell Sampic solenoid operated valves located in containment Provide-information normally listed in Table 6.2-40 for these va1ves In addition, provide the isolation provis,ica for the following systems; reactor building equipment drain instrumentation penetra-tion through the drywell and the TIP system 480.28 Table 6.2-40 of the FSAR does not list all containment isolation valves (e.g., MSLLCS 'is shown'to'have one valve in Table 6.2-4 an'd two valves in Figure 6.2-63).
In addition information such as GDC,
-containment isolation signsls...etc are missing from the table.
Revise Table 6.2-40 to include all pertinent information.
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480.29 Figures 6.2-63, 64 and 65 show the arrangement of the various iso-(6.2) lation valves listed in Table 6.2.40.
Many of the lines penetrating the containment have test lines between',the isolation valves.
Pro-vide justification why these test lines shculd not be treated as branch lires and ' included in the containment isolation valve tables and tested in accordance wi:5 Appendix J..
480.30 List the systems which penetrate the containment and are not vented (5.2) 4 and drained for. the Type A containment leak rate test.
These systems that are not vented and' drained for the Type A test must meet the following requirements:
1)
The system is protected against missiles and pipe whip; 2)
The system is de'signated seismic Category I;
- 3) The system is classified Safety Class 2;
- 4) The system pressure is greater than the containment pressure at all times during the course of the accident;
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- 5) The system will remain full of water for 30 days; and
- 6) Both items 4 and 5 will be maintained when a single active failure is assumed in the system.
State whether or not these systems meet th'e above requirements.
%"te :ystems have lines that are sealed free the containment ats.
'480.31 machere because their lines terminate below the water level of the suppression pool. Therefore, the3+ systems are not vented and drained for the Type A containment leak rate test.' However,
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to be considered a sealed system, the piping between the suppressior pool and isolation valve: should meet the following requirements:
- 1) The piping is protected against missile and pipe whip;
- 2) The piping is designated seismic' Category 1; and
- 3) The piping is classified Safety Class 2.
State whetner or not the piping between the suppression pool and
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isolation valves meet the above requirements for th'e penetration mentioned above. Also, specify the fluid that is used to pressuriz the valves to perform the Type C test.
Closed systems outside containment having a post accit'ent function
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480. 3j (6.2 become extensions of the containment tioundary follow'ing a LOCA. Cc tain of these systems may also be identified as one of the redundai containment isolation barriers. Since these systems may circulate 36
a contaminated water or the containment atmosphere, system com-ponents which may leak are relied on to provide containment integrity.
Therefore, discuss your plans for specifying a leakage limit for each system that becomes an extension of the containment boundary following 'a LOCA, and leak tactina the
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systems either hydrostatically or in accordance with appendix J testing requirements. Also, disc 0ss how the leakage will be in-cluded in the radiological' assessement of the site.
480.33 Table 6.2-40 identifies certain containment isolation valves that (6.2) vill not be Type C tested. Therefore, justify that they do not constitute potential containment atmosphere leak paths following-a LOCA.
In this regard, a water seal may be shown to exist that will preclude containment atmosphere leakage. 7.f this approach -
is taken, discuss how a water seal can be established and maintained using safety grade pipes and components, and considering single failure o~f active components.
System drawings showing the routing and elevation of piping should be used to show the existence of a.
water secl.
When operation of a system is needed to maintain a water seal in the system, the ECCS for examcle, show that the system will keep its water seal for a sufficient period of time if the system is removed from operation.
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480.34 Note 20 to Table 6 2-40 of the FSAR discusses locked-open manual (6.2)'
operated instrument valves such as reactor plant ventilation pressure cell that are required to remain open during a (OCA.
' What are the provisions to maintain contairunent leaktight
. integrity in the event of a rupture of any component in the instrument line outside primary containment 7 It is our posi-tion that instrument lines should be designed to the requirements of Regulatory Guide 1.11. Provide information to demonstrate com-pliance with the Regulatory Guide.
480.35 Verify that the pressure relief dampers, referenced in FSAR Sectinn (9.4) 9.4.6.2.1, provided to protect the containment ventilation system unit cooler fan, motors and ductwork against possible pressure transient following a LOCA will function for both positive and negative presw re transients...
480. 36 FSAR Section 1.8, Appendix 1 A, RIVER BEto STATIQ*iS POSITIONS
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(6.2)
ON THE NUCLEAR REGULATORY COMMISSION'S POST-TMI REQUIREMENTS, NUREG-0737, does no.t provide the requ' ired information for certain Items. Regarding -II.E.4.2, describe specifically how each. para-graph of this item in NUREG-07,37 is satisfied.
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48a 37 Regarding NUREG-0737, Item II.F.1, At'tachment' 4, "Contai,nment (6.2)
Pressure Monitor"; Attachment 5. " Containment Water Level Mo'nitor". -
Attac.hment 6 " Containment Hydrogen Monitor", provide information to indicate that the re'quir' ements of NUREG-0737. It'em II'.F.1 Attach-ment 4, 5, and 6.will be met.
480.38 Sealed closed barriers used in place of automatic isolation valves
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(6.2) should be under administrative control to assure that they cannot be inadvertently opened. Administrative control includes mechanical devices to seal or lock the valve closed,'or to prevent power from being supplied to the valve operator. Verify that all leakage monitoring connections, normally closed manual valves in test, vent drain instrument, and oth'er similar types of branch lines which serve as containment isolation barriers will be sealed closed as defined in SRP Section 6.2.4.
480.39 Verify that the design of the containment isolation system allows (6.2) the operato'r in the main control room to.know when to isolate fluid systems that at ; equipped with remote manual isolation valves, l
as defined in SRP Section 6.2.4.
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480. 40 The accident at Three Mile Island, Unit 2 involved a large amount (6.2) of metal-water re. action in the core with rrsulting hydrogen gene-ration well in excess of the amounts considered in 10 CFR Section 50.44 of the Commission's regulations. During the past year the staff has been studying the potential of excess hydrogen gene-ration, the effects such concentrations of hydrogen would have on -
the various types of plants, and the effectiveness of various mitigation systems in protecting the plant against such situations.
The results of our studies to date are presi..;ted in the SECY 80-107 series of documents.
In t'hese reports, we recommend that all BWR l
l Mark I and II containment plants be inerted and that owners of all.
other plants be reonired to provide an analysis 'and/or, proposed.'
design (or designs)~ to mitigate the consequences' of 1.arge amounts.
- of hydrogen in containment. The associat6d proposed inte'rfm rsle was published in The Federal Lagist6c on October 2,1980.
l-i Subsequent to the issuance of SECY 80-107, a substantial amount of additional w'ork has been performed on this issue with emphasis on l
ice condensers.
With respect to ice cor$dansers, and specifically-l Sequoyah, the Cor.caission has decided that the matter of hydrogen control for degraded core accidents in ' plants with small contain-ments needs to be resolved in the near term, i.e., the resolution l
should not be deferred to rulemaking.
In SECY 80-107, the staff showed that Mark III containments are similar to ice condenser containments in regard to their ability.
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to accomodate large amounts of metal-water reaction without jeopardizing contairunent integrity.
We, therefore request a description of the program to improve the -
~e hydrogen control capability at the River Bend Station, Units 1 and 2.
In addition provide the analysis.of hydrogen generation based on 75% metal-water reaction.
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ReactorPhisics'Section
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491. 4,
'What is the relationship of the one-dimensional transient (4.3.2.5) thermal-hydraulic model of Reference 11 to that used in the CDYN code? Are the procedures,used'to obEain cross l
sections (i.e., to collapse from three dimensions to one-dimension) the same as those'use.d in ODYN? Do the con-clusions quoted in the ODYN verification report, NEDE-24154P, with respect to comparison of, scram reactivity
-with three-dimensional calculations hold for the model
. described.in Reference 11?
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