ML19263C116
| ML19263C116 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 11/09/1978 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | Caffey L ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| References | |
| NUDOCS 7902070067 | |
| Download: ML19263C116 (33) | |
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gt NUCLEAR REGULATORY CCMMISSION M
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NOVs go THIS DOCUMEtiT CONTA!NS Docket No. 50-537 POOR QUl\\UlY PAGES Mr. Lochlin W. Caffey, Director Clinch River Breeder Reactor Plant Project THIS DOCUMENT CONTAINS P. O. Box U Oak Ridge, Tennessee 37830 POOR QUAUTY PAGES
Dear Mr. Caffey:
As you are aware, following President Carter's energy policy message of April 22, 1977, the Energy Research and Development Administration requested an indefinite suspension of the public hearing schedule associ-ated with the licensing of the Clinch River Breeder Reactor Project (CRBRP). As a result, the Nuclear Regulatory Commission (NRC) deter-mined that the NRC staff should not continue its safety review in regard to the construction permit application for the CRBRP on an indefinite schedul e.
Thus, the staff redirected its review activities to bring the safety review to a point where the effort to date would be adequately documented and any nearly complete efforts would be completed.
In light of these considerations, we have compiled a status report which sumarizes the staff's positions regarding the major outstanding items existing at the time of the redirection of the safety review. A copy of this status report is enclosed with this letter. This report is not to be considered a Safety Evaluation Report, should not be con-strued as addressing necessarily all unresolved issues or outstanding items regarding the CRBRP safety review, and should not be interpreted as commenting on the satisfactory resolution of any review area. The purpose of this status report is to provide general overview documenta-tion of the status of the review in major areas at the time of the suspension of our review.
I hope that you will find the enclosed document helpful.
Sincerely, Wh William P Gami 1, Assistan irector for Standardization and Advanced Reactors Division of Project Management Office of Nuclear Reactor Regulation
Enclosure:
Summary of Outstanding Items cc:
See next page 790207oc67
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cc:
George L. Edgar D. C. Gibbs Morgan, Lewis & Sockius General Manager, Project 1140 Connecticut Avenue, N. W.
Management Corporation Suite 1100 P. O. Box U Washington, D. C.
20036 Oak Ridge, Tennessee 37830 J. E. Nolan, Project Manager William F. Hubbard LMFBR Demonstration Plant Assistant Attorney General
_ _,., Westinghouse Electrig_ Corp.
State of Tennessee CRBR Project Office Supreme Court Building, Rm. 421 P. O. Box U Nashville, Tennessee 37219 Oak Ridge, Tennessee 37830 S. Wallace Brewer
_ __ _.. Luther M. Reed County Judge and Attorney Attorney for the City of Roane County Courthouse Oak Ridge Kingston, Tennessee 37763 E53 Main Street, East Oak Ridge, Tennessee 37830 Godwin Williams, Jr.
Manager of Power Tennessee Valley Authority 819 Power Buildina Chattanooga, Tennessee 37401 M. M. Hoyle Project Manager, CRBRP Burns & Roe, Inc.
700 Kinderkamack Road Oradell, New Jersey 07649 Anthony Roisman Roisman, Kessler & Cashdan 1712 N Street, N. W.
Washington, D. C. 20036 T. Cochran Natural Resources Defense Council, Inc.
917 - 15th Street, N. W.
8th Floor Washington, D. C.
20005
Enc 1:sare NRC STAFF REVIEW OF CRERP - SO'"ARY CF CUT 3TA'C:NG ITEMS I
This enclosure highlights the status of major outstancing issues associated with the radiological safety review of the Clinch River Breeder Reactor
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Project (CRERP).
As such, this listing deals only with 'open items or unresolved issues.
No attempt is made herein to document the satisfactory resolution of any item or the satisfactory completion of any review area.
I.
GENERT.L A.
Control Room Design Conformance with CRBRP Criterion 17 1.
The adequacy of the control room design to. limit radiation doses received by operating personnel is an open item since the Preliminary Safety Analysis Report (PSAR) does not yet provide a radiological analysis for the control room assuming the staff's specified site suitability source term.
The adequacy of the control room design for events associated with core melt and the adequacy of fresh air intake locations (Q312.66 and Q312.67) have not been resolved.
2.
The capability for remote control of the reactor has been addressed in PSAR Amendment 32.
Staff review is necessary.
B.
Large Sodium Releases in Steam Generator (S/G) Building 1.
The r$AR indicates that postulated Intermediate Heat Transport System (IHTS) pipe breaks and ensuing sodium fires result in overpressures in the S/G building.
T'ie staff must understand and evaluate the models and assumptions used in the applicant's analyses.
(It should be noted that sodium-air reactions are not fully treated in the.PSAR.) The applicant relies essentially on instantaneous vent'ing of the S/G building (0.5 seconds) to prevent overpressures (refer to PSAR Section 15.6.1.5).
The adequacy of S/G building design pressure and temperature has not been established.
The staff has requestec additional information (see Q001.703) regarding the analyses of sodium fires in the S/G building.
2.
The staff must assess the adequacy of proposed fire suppression and nitrogen flooding systems for sodium fires in the S/G building.
C.
Implementation of Design Criteria The staff issued design criteria in January 1976, to which the applicant responded in January 1977.
As the safety review proceeds, the staff will need to evaluate the degree of design conformance with these criteria.
The staff also issued a position on safety systems classification in January 1976 to which applicant responded in January 1977 stating acceptance of staff position.
The staff review to confirm implementation of this position has not been comp!etec.
6 M
4 D.
Operator Licensing The applicant must commit to comply with the guidance of Regula-tory Guide 1.33 or must develop an acceptable alternative.
E.
Safeguards 7-New regulations (10 CFR 73.55) were issued in February 1977 for operating plants.
The applicant must revise the PSAR to reflect spirit and intent of the new requirements.
Previous staff requests regarding this issue (refer to 421 series of questions) were not adequately addressed by the applicant since the interpre-tation of the requirements as proposed by the applicant differed from that stated by the staff.
10 CFR 73.55 resolves this issue and places the burden on the applicant to revise the PSAR.
The staff must assess the susceptibility of principal design features to sabotage.
Such an assessment was initiated for the Reactor Shutdown System but was terminated before completion.
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F.
Emergency Planning The staff requested time-distance-dose relationships during the
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acceptance review (June 1975).
The applicant provided a response in Amendment 40 (July 1977) which has not been reviewed.
Depend-ing on the results of staff review, the staff may need to review further the acceptability of the low populaticn zone and the scope of specific emergency planning measures, in particular in relation to facilities at the Holifield National Laboratory.
G.
Quality Assurance and Operations s
The staff has reviewed the administrative controls to be used in the development and conduct of the initial plant test programs as described in the PSAR through Amendment 37.
The description of these controls has ~een found to be generally acceptable.
D Amendment 38 has been submitted for our review and includes a response to staff concerns regarding test summaries for first-of-a-kind plant design features.
The staff has not reviewed that portion of Amendment 38.
The staff has reviewed the Quality Assurance (QA) program described
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in the PSAR through Amendment 38 and found it acceptable subject to (1) receipt of an adequate QA program description for the fuel and blanket assemblies, and (2) a list of structures, systems and components under control of the QA program. Amendment 40 to the PSAR addressed these items but has not been reviewed.
In a
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general, the QA program to be used during the design and construction phases was found to be acceptable.
Amendment 40 included about 200 pages as revisions to PSAR Chapter 17.
The applicant states these changes augment the previously submitted description and provide more detailed procedures.
Amendment 41 provided a new Appendix 17I.
Based on a cursory review of Amendments 40 and 41, the staff finds that although many of the changes will not affect its previous conclusion of accepta-bility, scme of the changes may be significant.
Further staff review is necessary.
H.
Foundation Engineering T'e staff and its consultants, the Corps of Engineers, have not completed their reviews of matters relating to foundation engi-neering.
Recent applicant responses to staff questions have not been reviewed.
Topics for which our reviews remain to be com-pleted include subsurface solution cavities, quality control and assurance programs for Class "A" fill and backfill, dynamic lateral earth pressures on subsurface walls, and evaluation of foundation conditions encountered during excavation.
I.
Geology-Seismology On August 29, 1977 the applicant submitted for staff review the results of geological investigations addressing the possible existence of unidentified linears in the vicinity of the pro-posed CRBRP site.
Initial staff review.of these results revealed nothing to modify the staff's conclusion presented in Section IV.C of the Site Suitabilitj Report (SSR).
However, further staff review is required.
J.
Meteorology The onsite meteorological measurements program and atmospheric diffusion estimates have encountered a history of problems related to data collection.
The applicant's program has been upgraded several times.
Prior to Amendment 38, all annual low-level wind data were based on measurements at the 75-foot level (in conjunction with high-level data at the 200-foot level) rather than at the recommended 33-foot level.
(The staff's final atmospheric diffusion estimate will use wind data measured at the 200-foot and 33-foot levels).
In Amendment 38 the applicant provided one year of onsite data using 33-foot winds and 200-75 foot temperature differences.
The applicant has not pr0vided a joint correlation which relates both wind
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_4 data and temperature difference data.;easured at the 75-foot level to those measured at the 33-foot level.
The data provided in Amendment 38 appear acceptable except for a very low fre-quency of calm wind occurrences, as compared to other data sets at the proposed CRBRP site and at other sites.
The use of this data without resolution of this apparent abnormality could result in the estimation of a much better diffusion condition than actually may exist at the site.
We will require that the applicant resolve this issue.
II.
CONTAINMENT SYSTEM DESIGN A.
Large Sodium Releases and Design Basis Accidents In addition to Engineered Safety Features (ESFs) provided for containment safety, the applicant has proposed containment systems to cope with accidents beyond the design basis.
These Third Level Thermal Margin (TLTM) systems include a reactor containment building annulus cooling system, a containment vent and purge system, and a reactor cavity liner and vent system.
These systems are provided to maintain containment pressure and temperature within acceptable limits following hypothetical core melt accidents which could result in significant sodium-fuel-concrete reactions.
It is proposed that the production and buildup of hydrogen will be controlled by autocatalytic recombi-nation in the main containment volume.
The TLDi systems a.re provided to assure that containment integrity will be maintained for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a core melt accident and at the same time provide confine.7.ent of the radioactive fission products.
s The design basis accident for which the containment system has been evaluated is the rupture of a sodium storage tank during reactor maintenance.
This relatively benign accident results in essentially no challenge to the capability of the containment system to maintain a barrier for the protection of the public health and safety.
The staff is evaluating postulated accident scenarios which may be selected as more appropriate for a design basis accident. We have been able to perform a preliminary review of the containment isolation system, the applicant's approach to containment / confinement bypass leakage, and the conformance of the proposed containment leakage testing program to Appendix J of 10 CFR 50.
Further staff effort is necessary to determine accept ;ility in these areas.
The requirements for containment functional design, heat removal systems, combustible gas control systems and the containment purge system will be dependent upon the identification and evaluation of an acceptable design basis accident.
With regard to TLTM scenarios there is a
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particular concern regarding the capability to control the accumulation of hydrogen within the containment.
Autocatalytic recombination has been proposed as the means whereby hydrogen
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would be maintained at acceptable levels without the need for
.d hydrogen recombiners.
Significant review of test results and their applicability to the CRBRP is needed. Additional tests and analyses may be required.
The capability of TLTM systems and components to operate effectively in the hostile environment
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associated with accidents beyond the design basis i_s also a
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matter which will require additional review effort.
Additional staff concerns requiring resolution are:
1.
The staff requer ted an analysis of a spectrum of pipe breaks within c;ntainment (see Q001.581 issued October 7, 1976).
The applicant responded in Amendment 40 (July 1977) which included a 73 page re e rt addressing this issue.
The staff has performed a cursay review of this response.
The applicant performed analyses ts je,tify the proposed design by using time-dependent sodium leak rates.
The design basis sodium leak continues to be 8 gallons per minute (gpm) and recent responses to Q040.4 do not include leaks beyond 8 gpm.
Addition.al staff effort is needed to ascer-tain need for further detailed review of the applicant's analyses and the codes used.
2.
Staff analyses indicate sensitivity of accident calculations to cell liner integrity.
Recent submittals by the appli-cant (see response to Q001.581) include the effects of failed liners for so:ite cases. A decision by the applicant is needed regarding the status of cell liners as an Engi-neered Safety Feature (ESF).
A proposal is required from the applicant addressing cell liner design requirements and acceptance criteria.
Further staff review of the response to Q001.581 will be needed.
Responses to staff questions on cell liner design have been provided in the PSAR but have not been reviewed.
3.
PSAR Chapter 15.6 addresses sodium spills but does not include a verification program for the SPRAY, SOFIRE and CACECO analytical models.
The staff met with the applicant.
in Spring 1976 to discuss this issue and documented its concerns in the meeting summary.
Sodium fire questions were included in a March 30, 1977 letter requesting addi-tional information.
The adequacy of the CACECO code is still in question since numerous code deficiencies and errors exist (refer to the March 30, 1977 letter). The
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method of analysis and the assumptions for sodium-air reactions are not fully treated in the PSAR.
Acditional information is to be submitted in the future.
7, 4.
The proposed PSAR design basis accident for the Reactor Containment Building (RCB) is the rupture of the primary sodium storage tank during maintenance.
The adequacy of
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the RC8 design pressure is an open issue pending resolution of the spectrum of sodium spills and behavior of cell liners.
The spectrum of postulated in-containment fires analyzed in the PSAR (e.g., Section 6.2) is neitner complete nor adequate.
The applicant originally proposed a RCS design pressure of ten psig, a PHTS cell design pressure of ten psig and a reactor cavity design pressure of 30 psig.
Recent revisions to the cell design pressures by the appli-cant are not yet fully reflected in the PSAR.
The applicant must update the PSAR with revised analyses to reflect the current design.
Additional analyses of sodium spills may result in higher design basis pr;ssures if sodium-concrete and attendant sodium-water reactions are considered.
The staff has been conducting independent parametric analyses in these areas.
An updated listing of containment pene-trations has been provided in PSAR Table 6.2-5 but staff review is required since many isolation valves normally open during operation require remote manual closure in the event of an accident.
B.
Accommodation of Energetics 1.
The staff issued its-position regarding Core Disruptive Accident (CCA) energetics on May 6, 1976 stating that an accommodation of 1200 Megajoules (MJ) is required.
Documen-tation of the basis for this position was provided in a curarehensive report, NUREG-0122, issued February 1977.
On Septei.5er 17, 1976 the applicant appeal'd the staff position.
In accoruance with the staff letter of December 6, 1976 outlining the scope and structure of the process to resolve the appeal, three general areas were to be discussed in detail.
These areas were:
- 1) Loss-of-Flow (LOF) accident analysis; 2) Meltaown Recriticality accident analysis; and
- 3) Transient Overpower accident analysis.
Of these general areas, only the initial phase of the LOF has been reviewed by the staff.
On January 26 and 27,1977 a meeting was held at Argonne National Laboratory where the applicant presented scme new information and data related to the LCF.
A meeting summary was issued on July 18, 1977.
The meeting was structured to consider fuel dispersai mechanisms and related
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topics imoortant in assessing the accident behavior during the initial phase.
The staff ice-d insufficient evidence to support invoking any potential fL?1 dispersal mechanisms to the extant that they would cause acequate fuel dispersion.
Numerous unkr. awns concerning the magnittde, timing and duration of the;e effects still exist.
halating to these analyses the applicant introduced a new version of the SAS code, SAS3D, which (1) increased the number of subassembly simulation channels from ten to 33, (2) introduced certain modeling changes, (3) altered inputr and (4) changed the end-of-equilibrium cycle loading pat 3rn.
To facilitate the staff's assessment of the new code, a meeting was held March 3,1977 for which a c.eeting summary was issued on May 20, 1977. Of the above listed code changes, the only one which has a maior impact on the energetics evaluation is under the heading "model changes."
In its subsequent review, the staff has determined that the use of the 33 channel SAS3D acpears to proacce results siniiar to those cbtained earlier by the staff using the ten channel SAS3A for the cases analyzed.
The general conclusion of the staff, based on its review including information pre-sented in appeal meetings, is that no new information has been develcped which changes the basic staff position that CDAs should be considered energetic for evaluating contain-ment design features.
Therefore, the basic position of the applicant that CDAs are nonenergetic cannot be justified at this time.
The applicant at this time has the option either to develcp additional data in support of its position or to withdraw the appeal and implement the requirements of the May 6, 1976 lette,r.
2.
The applicant's design approach to protect the containment frca CDA energetics has been to rely on the reactor vessel head to accommodate the energetics.
No fallback options have been suggested by the applicant.
The applicant's analyses, presented in Not ember 1976, show an inadequate design margin due to the buckling of risers.
Most recent staff analyses also have shown the applicant's present design not adequate for either 1200 MJ or 661 MJ but for reasons other than those determined by the applicant.
Staff evaluation of shear ring capability has indicated the need for additional analyses ty the applicant to demonstrate design adequacy.
The applicant has provided additional PSAR infr~
on regarding the analysis of the shear rings so that une staff may assess the cacability of the large rc*.ating plug to accommodate 561 MJ.
Present indication (see response to QC01.509) is that the applicant is not m
um
performing any analysis for energetics exceeding 661 MJ.
Staff review is required to assess the adequacy of re:ent.
submittals (e.g., WARD D-0178 entitled " Closure Head Capa-bility for Third Level Structural Margin Leading," June 1977).
3.
Recent information on the Stanford Research Institute (SRI) test program was submitted in Amendments 39 and 40 in response to our request for additional information, dated March 22, 1977.
This information has not been reviewed for completeness or technical adecuacy.
Basee on the differing points of view between the applicant and the staff regarding the analytical aspects of this outstanding issue, the SRI test program may be crucial in assessing the accommodation of CDA energetics within the reactor vessel head structural ccmplex.
C.
Acco.mmedation of Melt-Down On May 6, 1976 the staff issuec its position that containment integrity be provided for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> folicwing a postulated core disruptive accident.
By letter dated Novem::er 5,1976.ne applicant responded, stating that it did not agree that the
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24-hour criterion is necessary or appropriate.
The staff's request for additional information dated March 30, 1977 high-lighted specific deficiencies of the applicant's analyses and design modifications which were proposed as a basis for resolv-ing the 24-hour issue.
These include, but are not limited to:
1.
Substantiation of the assumptions regarding materials interactions and core debris - The Department of Energy (previously Energy Re' search and Development Administration or ERDA) research and development effort appears minimal, not sufficiently defined, and not timely.
The assumption of a fragmented, level, coolable debris bed is critical to the hypothesized melt-down behavior and requires substantiation; 2.
Evolution and control of hydrogen - No apparent attempt has been made to limit production of hydrogen and reliance will be on " controlled" ignition and purging to control hydrogen concentrations.
Recent claims by the applicant that hydrogen concentration control will be accomplished by auto-ignition require further substantiation; 3.
Concrete structural analyses - No detailed analysis of residual structural capacility was provided for staff
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review.
This must be provided; I
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4.
Dose miti;sti.g features - Eropcsed details were recently submitted in pSAR Section 9.6.2 but have not been reviewed.
Furthermore, the intent to cool the annulus defeats the purpose of the " dual" containment sooner than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with possibly marginal benefits in limiting containment vessel temperatures; 5.
Interaction of Third Level Thermal Margin (TLTM) features with Engineered Safety Features (ESFs) - The applicant proposes to initiate TLTM features manually.
The available information does not sufficiently describe any interlocks or controls between TLTM features and ESFs.
The staff is concerned that reliance solely on manual initiation for TLTM features can potentially degrade the ESFs if inacvertent or erroneous decisions are made regarding the initiation of TLTM features.
Conversely, inadvertant actuation of ESFs could degrade the cperational capability of TLTM features.
6.
Response to staff letter of March 30, 1977 - Based on discussions with the applicant, a revisec TLTM-type report was to be submitted in the fall of 1977.
This report has not yet been submitted. Moreover, the applicant does not intend to respond to the staff's questions issued March 30, 1977 and suggested in a letter dated May 9, 1977 that these questions be withdrawn.
The staff letter of May 27, 1977 provided a response to the applicant's suggestion.
Based on informal discussions with the applicant it appears that the applicant will not provide complete responses to our questions and will not submit a cross reference list to identify where the response to each question can be found.
It appears that the applicant will simply submit another updated report. We find this unacceptable.
A lengthy and burdensome staff review is foreseen.
D.
Acccmmodation of Site Suitability Source Term The applicant has agreed to comply with staff's position of May 6, 1977 regarding site suitability source term accommodation.
PSAR Chapter 15A was revised in Amendment 40 to incorporate the results of the applicant's analysis in this respect.
Based on a cursory review, the results appear to be in general agreement with those presented by the staff in the Site Suitability Report.
1.
Filtration and recirculation systems - The apolicant met with staff in October 1976 to describe these systems concep-tually.
Amendment 36 provided acditional docunentation.
The staff review is not complete.
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2.
Bypass leakage - The PSAR has been revised to include a listing of containment penetrations.
Mcwever, staff review is not complete.
Filtration of the Reactor Service Building has been prcposed but is not yet described in the PSAR.
III. THERMAL-HYDRAULIC DESIGN A.
Natural Circulation and Low Sodium Ficws Natural circulation in the reactor vessel, the Primary Heat Transport System (PHTS), and the Intermediate Heat Transport System (IHTS) was originally proposed to provide emergency c0 cling for the core in the event of the loss of forced ficw.
The applicant submitted a natural circulation verification p'an which outlined analyses and tas.ts to be performed.
The staff concluded that insufficient analyses and data exist at this time to accept a design which relies solely on natural circulation (refer te staff position C001.530).
In response to the staff pcsitica, the a:plicant has preposed a disarse electrical pcwer suoply for the PHTS and IHTS pony motors and the PSAR has recently been revised to incorporate this change.
The staff review of this revision is not complete.
Areas of staff concern include:
1.
Core flow distribution - Core and blanket flow distribution during normal operation appears acceptable.
However, core flow distribution at low sodium flow rates is based solely on the predictions of unverified analytical models. Although tests are planned at the Fast Flux. Test Facility (FFTF) it is not clear that the results can be extrapolated to CRERP due to theeffectsoftheblanketregion.
Variation of power distri-bution with time, especially at end of life (E0L), requires additional staff review to assure adequacy of assumed flow through core, blanket and bypass ports.
Bypass flow is used to keep the reactor ressel temperature below 750 F.
Recent changes in decay heat loads (see Amendment 37) result in higher temperatures in blankat assemblies than before.
2.
Verification of ccmputer models
'The applicant has recently provided a new code FORE-2M for analysis of core thermal-hydraulic behavior.
Verification of this code and a number of others is lacking and the staff review is not complete.
B.
Hot Channel Factors The applicant has proposed an approach to hot channel factors similar to that used for FFTF.
CRERP Report (WARD-CC50) was reviewed by staff, and a meeting was held with the a:piicant.
a.
..mo mm o
11 The staff has recommended that WARD-0050 be rewritten and bases provided.
The applicant has statea thct WARD-0-CC50 and PSAR Section 4.4 will be revised, incorporating changes in the statis-tical format including the use of tolerance intervals and, for certain random variables, nonparametric statistical methods if r
appropriate.
This information is still outstanding.
Furthermore, WARD-D-0050 in most instances contains the same information as that provided for FFTF and is therefore amenable to consolidation as a generic topical report.
Such an approacn sould facilitate the review of this engineering approach.
C.
Acceptability of Limits The core thermal limits for steady-state operation, for transients and for accidents have not been completely described by the appli-cant.
In lieu of specific limits, the applicant has proposed an approach which does not define strict adherency to numerical limits althcugh in some instances " guidelines" have been proposed.
It is the staff cpinion that fixed limits should be established to the maximum extent possible for' steady-state and transient cor.ditions.
" Guidelines" are too indefinite, can be revised at will, and are not an adequate substitute for design specifications.
The amount of conservatism in any proposed limits should be investigated and documented and the staff must determi-adequate margins. The applicant's reliance on no-sodium-boii;ng as a faulted condition limit is acceptable.
However, the degree of conservatism in the calculations requires review and evaluation since, in certain instances, calculations indicate that socium boiling is approached within about 20 F.
For example, the applicant relies on dynamic head as one means to increase the allowable temperature limits.
Therefore, the review of those plant characteristics affecting pressure drops throughout the system must be more vigorous.
D.
Loose Parts Monitoring The applicant has not proposed any loose parts monitoring system similar to those currently required for LWRs.
The staff has developed a position which requires such a system (refer to page II-23 of the Site Suitability' Report). This is regarded as an R&D item requiring developmental effort on items such as sensors, cables and associated system components.
IV.
MECHANICAL / STRUCTURAL DESIGN A.
Seismic Design On March 9, 1977 the staff met with the applicant to consider the seismic design aspects of CRERP.
On April 27, 1977 the applicant 9
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submitted a report entitled " Reserve Seismic Cacability Recort."
This document has not been reviewed by the staff.
The following topics are of particular concern to the staff:
1.
Additional information must be submitted concerning the
" Hybrid" foundation analyses involving static finite elecent calculations to identify spring constants and related properties, followed by lumped mass model analyses of the overall system.
The information must include a ccmparative
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discussion of this approach with a fixed base analysis or with an analysis based on the half-space theory.
2.
The applicant has used Tsai's weighted damping method in an
" altered form" to accommodate the three-dimensional analysis.
Documentation of the changes is required in order to evaluate the adequacy of this approach.
3.
Additional information must be submitted concerning the time increments and duration used in the analysis and the developcent of foundation time histories in order to evaluate the accuracy of the response calculations which are based on modal analysis with modal frequencies up to 50 Hz.
4.
The applicant should provide details of the containment /
confinement structure (drawings), including its effect on component design response spectra.
In the current design, the confinement structure extends down to the basemat and is keyed to the containment structure near the ope' rating floor resulting in interaction.
5.
The applicant should' provide specific delineation of the systems for which nonlinear analysis was used, and addi-tional details of this approach.
6.
Additional information is needed to clarify the method used in determining response spectra at various elevations and its compliance with Section 3.7 of the Standard Review Plan.
7.
The method used for analysis of soil-supported tanks needs further clarification.
It is not clear how strain-dependent properties of soil-and fluid-str Icture coupling arising from seismic excitations are considered in the analysis.
8.
Additional information concerning material degradation is required.
The applicant should clarify how material degraca-tion is accounted for and how the material properties are m
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selected (code cases 1592 through 1596) for seismic analyses of high tempecature components.
9.
The me* lod of accounting for the hydrodynamic and fluid coupling effects employed in the seismic analysis for the reactor vessel should be properly documented.
B.
Dynamic and Static Analysis for Seismic Category I Ccmponents
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Of principal concern in the dynamic and static analysis for seismic Category I components is the assurance that the analytical models will adequately represent the physical situations of concern.
~his includes proper failure modes (static and creep fatigue rupture, vibrational distortions, stability and deforma-tions), adequate constitutive relationships (elastic, inelastic, and degradation with time), compatible boundary conditions and realistic component material prcperites.
Also of concern is verification of the adequacy of the ccmputer programs to prcduce sound quantitative results.
Scme rec:gnized ccmputer programs are capable of producing very good results for one class of ccmponent models while possibly not providing ac urate resalts for others.
Convincing evidence must be presented to demon;trate that the ccmputer programs used are adequate for their particular applications.
Supporting tests are generally required to directly verify the structural capabilities of components and to confirm the analytical methods used.
Such tests can be either component model or prototype tests.
The staff must be assured that such tests accomplish their purpose.
The staff has issued numerous questions relating to the methods of analysis and tests to be used for the design of safety-related equipment (see Q110.78 issued December 1,1976).
The applicant indicated that the information will be provided in the FSAR during the operating license stage of review.
The staff must understand the methodology and criteria to be employed prior to Construction Permit (CP) issuance.
There is no ccmmitment by the applicant to resolve this issue at the CP review stage.
In e
a meeting on January 16, 1976 (for which meeting minutes were issued February 9,1976) the staff proposed an acceptable pro-cedure for the applicant to submit the requested information for staff review prior to start of fabrication of each ccmponent in question.
The manufacture of these components would be contingent on staff approval of this information.
No response has been provided by the applicant to this proposal.
C.
Centrol Roc Systems It is not clear that the PSAR reflects the current design of the secondary centrol rod drive mechanisms (CRCMs).
Further
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documentation and staff review are recuired for the sucperting systems associated with the functisning of both prii.;ary and secondary CRCMs, since at eacn refueling botn systems must be temporarily discennected.
The precedures for this disconnection
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and subsequent reconnection muet be pursued to assure that the reliability and performance are not affected by either intentional or randem failures.
Common mode failures of the primary and secondary CRDMs must be further examined to determine the degree of diversity.
D.
Flow Induced Vibrations The applicant has stated in the PSAR that the structural integrity of the reactor internals due to flow induced vibrations will be assured through a comprehensive program of analysis, ccmponent testing, reactor system model testing and in-clant measurements.
It is statad that a cescriction of the analysis methcds and mathematical redels to be used for the dynamic system analysis will be described in the FSAR.
Few details of this rethod are summarized in the PSAR.
Cc.Tpenent tests evaluating the effects of the interaction between the upper internals structure and core assemblics are being run at Argonne National Laboratory (ANL) in a 1/3 scale model.
A reactor systems model test of the reactor internals is planned in the Integral Reactor Flcw Model (IRFM) test at the Handford Engineering Develcpment Labcratory (HEDL).
This test program includes a thorough vibration evaluation of components in a 1/4 scale water test.
To verify similarity of these various models with the CRBRP, the applicant prcposed to locate two triaxial accelerometers on the uppe,r internals structure to measure the vibrational behavior in th CRSRP during preeperational testing.
Lee intent is to compare the vibrations cenitored from these accelerometers with those anticipated in the model tests and thereby judge the adequacy of these models and tests to model the CRSRP reactor internals and predict their vibrations.
It is the staff's position that, due to the complicated structure of the reactor, more instrumentation than that prcposed by the applicant will be required to prove the adequacy of the internals and demonstrate a reliable interpretation of the internal vibration from the model tests.
The staff is not convinced that the IRFM, as parametrically described in the PSAR, will adequately represent the CRSRP in tests.
For example, several of the T.odel-to pretotyce scale parameters ratios are far from unity.
Thus, it remains to be demcnstrated that the IRFM, as well as the ANL 1/3 scale com;cnent, m
h
test results can be used to predict the dynamic behavior of the CRERP reactor interna',s.
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Since the reactor internals of the CRERP are similar to the internals of the Fast Flux Test Facility (FFTF), the applicant intends to use the operational results of the FFTF as a further aid in verifying the adequacy of the CRSRP internals.
- Hewever, it must be demonstrated that the actual dynamic behavior of the FFTF reactor internals will be similar to that prcjected for the CRBRP internals.
The proposed program concept is reasonable and can be used to assure the adequacy of the CRSRP reactor internals after the above mentioned problem areas are resolved.
A meeting with the applicant may be necessary to obtain additional details and clarify the staff positic E.
Load Ccabinations
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The originally orocosed lead comoinations were rejected by the staff since they lacked clear definition 2nd since the methods for combination for plant and system upset, emergency and faulted conditions were unclear.
Amendment 36 provided a response to the staff position.
Based on a preliminary review of this response, the applicant essentially has deferred the specific and detailed explanations of the manner in which load ccmbina-tions will be affected.
Of principal concern to the staff is the definition of loads resulting from plant and system emergency conditions for which PSAR service limit "C" is appropriate.
The staff believes that the issue of load combination must be resolved during the Construction Pe'rmit review.
F.
Pipe Whip Analysis The staff requested the applicant to specify the pipe wh'p design criteria and to formulate the dynamic ocel for the pipe and its constraints.
System conditions, forcing functions, and bounding and pipe rebound conditions are also required.
A recent response has been provided by the applicant but has not been reviewed.
G.
Snubbers Snubbers are to be used as component supports.
The basis for the location, required load capacity, and structural and mechan-ical performance parameters of safety-relatad snubbers must be provided in the PSAR.
A description of the treatment of locked J
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snubbers in the analysis of the PHTS and IHT5 piping is recuired as weil as a commitment to incluoe in tne FSAR a ce: ailed cescrip-
~~
tion of the snubber systems, design specifications, qualification tests, and system structural analyses.
H.
Active Pump and '.'alve Operability The applicant was requested to provide (1) a. ore detailed description of the types of analyses which will be performed, (2) a commitment to perform static tests to simulate faulted condition loads on representative active components, (3) a commitment to include faulted condition nozzle loads in the aforementioned analyses and tests, (4) a commitment to seismically qualify all appurtenances which are required for operation of the active component, and (5) a commitment to demonstrate by test and/or analysis that all active Class 1, 2 and 3 pumps and valves will operate when subjer'.ed to appropriate stress limits.
The applicant has not responded to this request.
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I.
Structural Cesign Of particular staff concern are the structural design considera-tions of the ceil liner and anchor system.
The applicant must submit a mathematical model for the analysis of the whole system, including the concrete in which the anchors are embedded.
The models and assumptions should be such that the results of the analysis will indicate stresses and/or strains in the liner, the anchors and the concrete.
V.
PIPING INTEGRITY s
s..
A.
Cold leg The staff stated its position in a letter dated October 6, 1976 that the double-ended rupture of the PHT5 cold leg need not be considered a design basis event subject to resciution of the three concerns discussed below.
For purposes of leak detection and assessing the safety margin, a conservative design basis failure area or leak has not been proposed yet by the applicant.
1.
Leak detection system - In response to staff questions, numerous letters have been submitted by the applicant.
PSAR Amendment 36 provided additional information.
The staff reiiew indicates that the applicant's responses have not provided sufficient informaticn to cenclude that the o
reliability and adequacy of the sodium-t: gas leak detec-tion system to be incorporated in CRSRP has been verified.
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Verification of the performance of these components through long term environmental verification tests prototypical of conditions expected in the plant is the primary concern.
2.
Pre-and in-service inspection - The staff position was issued August 17, 1976 (see Q120.66).
The PSAR was revised February 18, 1977 and a new Appendix G was incorporated as a response.
The staff finds that this new information is inadequate and not fully responsive to the staff position.
Additional staff positions and questions must be resolved and Appendix G must be revised before a satisfactory conclusion can be o'de.
3.
Material surveillance program - The staff position was issued August 17, 1976 (Q120.68) to which the applicant responded in Amendment 32.
The staff review of this response is not complete.
S.
Hot Leg The applicant submitted a draft reoort in January 1977 to adoress hot leg piping integrity.
The staff met with the applicant and stated that many problems common to cold leg integrity are still unanswered regarding hot leg integrity.
Additional information will be necessary to conduct a review.
No estimate was given by the applicant for submittal of a final report.
Also, the applicant has not developed or proposed a conservative design basis f ailure area or leak rate.
The applicant's final report will be prepared and submitted in the near future.
Prelimin'ary anali cs of information contained in the draft report indicate in certain instances that the dtsign is inadequate and that additional pipe supports will be required.
C.
Moderate Energy Pipe Break Criteria for Sodium Piping The applicant has proposed moderate energy pipe break criteria for treatment of sodium and NaK piping as a moderate energy system.
The staff review is not complete.
D.
Intermediate Heat Transport System Piping 1.
IHTS piping within containment is located in inerted cells.
Th' applicant proposes to treat this piping in a manner similar to that of rHTS piping, i.e., couble-ended rupture is not treated as a Design Basis Accident (CEA).
Occumenta-tion has not yet been provided by the applicant in scoport
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of this position although it is expected to be similar to that cf the PHTS hot leg.
2.
IHTS piping outside containment is located in an air environ-ment.
The PSAR indicates that double-ended rupture of this piping is a CBA.
E.
Mechanical Properties
-- In response to staff position Q120.62, the applicant agreed to submit a report addressing,long-term thermal aging ef fects upon the mechanical properties, fracture toughness, and crack propa-gation rates in welds using materials and procedures specified for the CRERP. The report has not been received.
VI.
ELECTRICAL A.
Reactor Shutdewn System (RSS)
The staff has stated its position in Eoth the FES and SSR that
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the reactor shutdown system can potentially satisfy tne follow-ing necessary requirements:
1.
Redundancy - The PSAR states that the two subsystems are neither redundant nor intended to be redundant for all design basis events.
To assess the applicant's approach, additional information is required in order to review those events for which the applicant claims redundancy is not needed on the basis of low probability.
2.
Diversity - The desig'n appears.to potentially satisfy this requirement although the staff review is not complete.
Additional detailed information is required.
3.
Reliability - As an additional design aid, the applicant has proposed using reliability analysis to demonstrate design adequacy and has submitted in Amendment 36 a revised Appendix C containing no numerical analyses or results. To assess the reliability of the reactor shutdown system, additional information is required in the following areas:
(a) the interfaces between the RSS and other portions of the plant protection system and between the RSS and the nonsafety portions of the plant.
(b) the physical separation of the channels within the two subsystems and the separation between the ao subsystems.
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(c) the testing of both subsystems, including the adecuacy of the overlapping testing.
B.
Environmental and Seismic Qualification The applicant has committed to the appropriate standards and related Regulatory Guides regarding environmental and seismic qualification.
However, the determination of the parameters for the environmental qualification testing has not been resolved since the applicant intends to cetermine these requirements on basis of individual accidents within each separate plant area where safety grade equipment is to be located.
C.
Electrical Power System Recent changes (Amendment 37) included revisions to the electrical pcwer system.
The staff review of these revisions is not ccmplete.
Questions related to grid stability and ethar 2: pacta cf the electrical pcwer system need to be resolved.
D.
Engineered Safety Features (E5Fs)
Certain ESF systems, such as the containment isolation system and the shutdown heat removal system, have undergone variou:
design changes.
The staff review of the electrical, instrumenta-tion and control systems associated with these systems is not ccmplete pending satisfactory determination of the adequacy of these ESF systems.
E.
Piping and Equipment Electrical Heating System It is not clear whether the piping and equipment electrical heating system for safety-related systems is connected to a standby AC power supply.
The applicant should provide additional information to demonstrate that a Safe Shutdcwn Earthcuake or loss of offsite pcwer will not ccepremise the capability for safe plant shutdown as a result of excessive plant temperature differentials or sodium solidification in the auxiliary systems and stagnant areas.
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VII. FUEL DESIGN A.
Cumulative Damage Function The fundamental CRSRP fuel rod design recuirement is the assurance of cladding integrity during steady state cperation assuming a history of a specified number of anticipated upset events and a single emergency event.
The design lifetime therefore coincides
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with that point in time at which an emergency event would just cause failure.
All aspects of tne prior mechanical and environ-mental nistories as well as all analytical and eperational uncertainties must be considered.
A time-dependent function of temperature and stress, called the Cumulative Damage Function (CDF), was developed.
A more generalized approach involving the definition of an envelope (transient limit curve) of failure conditions for a general class of emergency events was pr: posed.
A transient limit curve (TLC) provides a graphic representation of the likelihood of a fuel rod failure during a given-type of emergency event.
The following staff concerns exist regarding the proposed cumulative damage function approach:
1.
Verification - Applicability of the CDF concept is predicated on the existence of a damage function such that failure occurs at a predictable time.
The principal assumptions in this approach are (1) each type of.cansient can be analyzed separately by superimposing the transient parameters u;cn those existing during s,teady sta e operation and (2) all the CCF ircrements are independent of one another.
Therefore, critical evaluation of the data base is required.
The staff has evaluated the fuel cladding transient tester (FCTT) test data and finds that:
a.
with some excentions, the CDF predictions appear to be in reasonable agreement with FCTT data on unfueled plenum region tubing; b.
the CDF method has not been shown to predict time to failure or margi,n to failure for cladding in tta fuel column region at'any neutron fl'uence.
There is no current evidence to show that the CDF method is able to predict margin to failure at low fluence for cladding irradiated at temperatures exceeding 1000 F since no data are available for these conditions.
Therefore, no verifica-tion of CDF predictive capability exists for CRBRP fuel undergoing an emergency event.
It has not been demonstrated also that the method can be used to envelop transient events and that it can be applied toward a TLC for the irradiation exposures and temperatures associated with the specific CRSRP design burnup'(80,000 megawatt days per tonne).
2.
Fallback options - The staff has not been provided the detailed plans for further tests on the CR3RP fuel and cladding.
The LMFSR fuel development test planning has been uncer the ccgnizance of ERDA (now OCE).
In the event
thi-R&D program for LMFERs does not verify the conservatism of the CDF methodology, the staff believes that the available options include:
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the operational burnup limit can be reduced to less a.
than 80,000 megawatt days per tonne; b.
power / flow ratio can be reduced; c.
FFTF operational experience can be used to demonstrate whether steady state burnup limits can be achieved; d.
experimental verification can be provided to determine whether transients terminated by the plant protective system will cause cladding limits to be exceeded.
B.
Design Recipe In a parallel but separate development to the CDF, the applicant presented an alternative procedure which has ccme to be kncwn as the FCF-213 or design recipe.
In the FCF-213 methodology, designated plastic strain limits (< 0.2% for steady-state and
<0.1% for transient conditions) are transformed into calculated lifetimes (burnup or transient time to failure) and into a design guideline of 1600*F for an emergency cvent.
The basic premise in the FCF-213 methodology < that it is conservative to substitute the creep rate and tensi'se stress / strain relationships of unirradiated, solution annealed 316 55 for those of highly irradiated 20% cold worked 316 SS.
This assumption is unsubstantiated.
1 To der.vostrate the conservatism of this approach, it must be demonstrated that the Plant Protection System (PPS) system will initiate scram and that sufficient negative reactivity will be inserted so that either the calculated 0.3% total cladding strain limit will not be violated or a significant margin to failure exists for an emergency event.
This has not been done.
1.
Conclusion - The staff concludes that the conservatism of the FCF-213 methodology coupled with specified strain limits has not been demonstrated for the CRBRP fuel rods with regard to (1) the models in the methodology, (2) the experimental verification of the models, and (3) the speed of response of the PPS during an emergency event.
2.
Fallback Options - If the R&D program does not verify the conservatism claimed, the staff believes the folicwing options may need to be pursued:
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a.
rei:ce Operational power; D.
reduce cperational burnup; c.
add subassembly (local) transient detection instrumentation; or
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d.
redesign the fuel to satisfy the current?y proposed limits C.
Fuel and Blanket Design Limits Fuel and blanket steady state and transient design limits are stated in the PSAR.
However, no temperature limit for the cladding has been developed for either steady state or transient conditions.
Instead, it has been proposed as a guideline that the maximum cladding midsall temperature during an emergency event should be less than 1600 F for an extended pericd of time (minutes).
For faulted conditions, it has been proposed that the sodium temperature be maintained below the saturation temperature 'anich varies aith pra:3ure.
The idaquacy of the emergency and faulted event limits has not been accepted oy the staff and additional revisu is ne::::ary.
The app!icant's approacn requires a detailed understanding of all the scenarios associated with these events.
D.
Fuel Failure The staff position regarding fuel failure propagation is stated in the staff May 6, 1976 letter and in the Site Suitability Report.
The applicant disagrees with the staff (refer to letter of September 20, 1976).
The applicant believes that fuel failure mechanisms are suf ficiently understood that design provisions are not necessary.
The applicant relies on future R&D results to confirm this.
The staff is unable to concur with the applicant's projections since the experimental evidence is not available.
Further resolution is required in the following areas:
1-Surveillance - The applicant proposes to cperate the plant with failed fuel.
PSAR Section 4.2.1.1.3.8 states tnat if only fission gas releases are detected, fuel will not be removed from the core (it is claimed that the cleanup system will limit extent of fuel failures).
It is proposed to allow fuel exposure to sodium up to a limit which has not yet been defined.
Development of this limit is an R&D effort (refer to Q241.75 and response).
The apolicant proposes to remove defective fuel rods experiencing fuel-sodium contact only at each refueling (see 0241.76 and response).
The staff position expressed in the SIR notes that operation with failec fuel sill be restrictad anc
23 -
increased surveillance will be reouired.
Further review of this issue is required.
2.
Instrumentation requirements - The applicant proposes to
~~
rely on gas monitoring and delayed neutron detectors for failed fuel monitoring.
It is not clear how the applicant will differentiate failed fuel with a resultant limited fuel-sodium contact from failed fuel with gross fuel losses or exposures ia excess of a " limit." The basic approach presumes that future R&D will assure the benign nature of operation with such failed fuel.
No specific R&D effort has been noted for CRSRP for developing additional sensors although other sensors are being develcped as part of the base LMFBR program for ccamerical plants.
If the R&D results indicate a need for additional sensors, the CRERP design should be modi fied to incorporate them.
E.
Post-CP !ssuas By letter dated October 6, l'975 the siaf f recuested several reports for identified topics.
In the fuel design area many of these reports were intended to be reviewed in the post-CP phase.
Maray reports have recently been submitted and the staff has not initiated a review program for them.
The PSAR section entitled
" Responses to 10/6/75 NRC Letter" summarizes the status of each topic.
The topics which have been judged by the staff to be amenable to a post-CP review are presented belcw and were intended to be reviewed as topical reports:
1.
fuel densification; -
2.
fuel rod bowing; 3.
fuel rod vibration; 4.
fuel rod wire wrap interaction; 5.
fuel assembly structural evaluation; 6.
fuel thermal performance calculations; 7.
fuel rod seismic analysis; 8.
internal / external cladding degradation; 9.
fuel restructuring; 10.
fuel rod failure criteria; and 11.
exposure dependent cladding deformation.
The majority of the topical reports were submitted in late 1976 and early 1977.
Some of the reports are expected to be submitted in 1978.
Ncne have been reviewed.
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VIII. SYSTEMS A.
Fire Protection Systems On August 17, 1976 the staff issued its position regarding fire protection stating that the applicant ihould specifically address the guidelines in Appendix A to Branch Technical Position APCSE 9.5-1.
The staff concern relates to non-sodium fires (refer to Question 020.47).
The applicant's response of October 1976 committed to provide the results of an evaluation by February 15, 1977.
No results have been provided to date.
Originally, the applicant proposed to use DOWTHERM as an intermediate coolant in certain systems.
Some of these systems have been replaced by chilled water systems.
The staff will require the folicwing information to ccmplete its review:
1.
a ccmplete response to QO20.47, with particular emphasis on Appendix A to BTP APCSS 9.5-1; 2.
a fire hazards analysis.;
3.
clarification of all the intermediate coolant systems with specif1. emphasis on tnose which are flammable or ccmbustible; 4.
ratings of the fire barriers listed in PSAR Table 9.13-1; 5.
piping and instrumentation diagrams for the Halon and carbon dioxide fire suppression systems and the nitrogen flooding distribution system; and, 6.
an assessment of the suitability of sodium carbonate fire extinguishers in liquid metal areas.
B.
Shutdown Heat Removal System On August 17, 1976 the sta'ff issued a position regarding shutdcwn heat removal capability reinforcing its position of May 6,1976 which required redundant and diverse shutdown heat removal systems.
The applicant has provided several submittals in response, the latest of which was received in June 1977 (Amendment 39).
Several portions'of the response are argumenta-tive and not acceptable since chey do not address the technical issue.
For example, the staff has stated that the Direct Heat Removal Service (DHRS) must be capable of functioning assuming the existence of the design-specified minimum sodium level in the reactor vessel.
The applicant claims that its system is adequate for all current design basis events and does not address this issue.
However, the DHRS canr.ot function unless the sodium level is at its normal operating level.
This issue remains open.
For two-loop operation, the staff has stated that the current design is not adequate.
The applicant has deferred this issue to the FSAR.
This is not acceptable since design charges will be necessary if twc-loop cperatien is contemplated.
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two-loop operation appears to be a requirement if the demonstra-tion cDjectives for CRERP are to be met.
This re ains an open item.
The staff also required that experimental evidence, based on sodium testing, be provided to support the adequacy of the locations for the DHRS intake and discharge nozzles. The applicant continues to rely on water tests for this purpose.
The redundant Primary Heat Transport System (PHTS) flow paths are not diverse in principle since they require proper functioning of the extensive piping, the steam generators and the Steam Generator Auxiliary Heat Removal System (SGAHRS).
The applicant believes that DHR5 provices a diverse system.
However, it is not redundant to the other heat removal systems since it requires PHTS piping and component integrity and forced convection through the PHTS loops. Water portions of the decay heat removal systems continue to depend on natural convection instead of redundant motive and control power supplies.
The ability cf the apolicant's decay heat removal simulations to predict full scale natural convection in the water portions of the systems must be de=cnstrated.
Heat transfer coupling between the water and socium portions must also ce accounted for.
C.
Reliability The applicant's preliminary reliability studies were documented in a GE report, NECM-14082, submitted January 1976.
This report has been reviewed and found to be unacceptable.
The optimistic assumptions and engineering estimates were not supported by a technical basis.
The staff and its consultants have been evaluating the design.
Of principal concern is the capability of the systems to function, with a loss of AC power for a limited period of time.
In PSAR Amendment 39 the applicant has described a DC motive power source for a train of PHTS and IHTS pony motors.
Conceptually, this approach is acceptable but the staff has not reviewed the implementation of it.
Furthermore, there are various combinations of SGAHRS and DHRS operating modes (for example, the initiation of CHRS at various time intervals after reactor shutdown follcwed by the initiation of SGAHRS at various a
time intervals folle'
'RS initiation).
The applicant must evaluate the reliabili,
these systems interacting in a manner other than they a.e currently described in the PSAR.
re staff and its consultants cannot complete their review until such an evaluation is conducted.
System success or failure as presented in the PSAR is a very limited and narrow interpretation since it ignores potential system interactions.
Resolution of this item is of potential benefit to the applicant in its assessment of overall system reliability for removing residual heat and could form the basis for revising the staff position on decay heat removal.
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D.
Effluent Treatment Systems The staff has issued questior.s regarding preposed TLIM features.
The applicant has not responded to these questions.
E.
Steam Generator Tube Rupture For the steam generator tube rupture the applicant has proposed a design basis accident of one guillotine tube rupture followed by six sequential tube ruptures.
The staff has net yet accepted this approach.
Items of concern include:
1.
Adequecy of TRANSWRAP - The TRANS'nRAP code analyzes pressure, temperature, and flow transients resulting from the exothermic water / sodium reaction in a steam generator following rupture of a steam generator tube.
The PSAR and supporting documenta-tion refer to an older version of the code while the ccmputer tape supplied to the staff by the apolicant is apparently a newer version.
The coce has been in a state of continuing chan;a and it appears that the PSAR analyses may have been performed with an inter eciate version. These inconsistan-cies hamper the staf f's ability to review the applicant's analyses.
The use and documentation of TRANSWRAP are outstanding items due to these incensi,stencies.
Also the staff has not been able to obtain adequate documentation of the code inputs, options and logic.
Further documentation and verification are required.
2.
Adequacy of SWkPRS - The sodium / water reaction pressure relief subsystem (SW3 PRS) is briefly described in the PSAR, hcwever insufficient information exists to evaluate it.
Additional information is need-d for the staff to independently verify the capabilities of the system to limit IHTS overpressures.
Of particular concern is the lack of information in the PSAR regarding the rupture discs and the analyses related to the n]rmal and ab.;rmal per-formance of these discs.
Staff questions were prepared, addressing both these items but were not issued to the applicant.
F.
Steam Generator Auxiliary Heat Removal System (SGAHRS)
The applicant states in the PSAR that the volume of the protected water storage tank (PWST) is based on cperator action within 10 minutes in the event of a feedwater pipe break.
It is the staff's opinion that, for an accident of that magnitude in which several actions are required by the cperator under emergency
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conditions, 30 minutes would be more appropriate and that the Pk3T sncuid ce sized in accordance with a time delay of 30 7.inutes before an operator action.
Assuming a feedwater pipe break and a concurrent most limiting single active failure, the applicant also should demonstrate that sufficient water will be supplied to the intact steam drums to assure safe shutdown during the time period defined above prior to operator action.
Acditionally the applicant should describe the low steam drum nressure autcmatic isolation feature and the auxiliary feedwater flow controller menticned in the PSAR.
G.
Emergency Plant Service Water System (EPSWS)
The following additional information must be submitted in order for the staff to complete its review of the EPSWS design:
1.
system P& ids; 2.
the solume of the E?SWS expansion tank, the adequacy of the available water storage volume based on maximum ex:ected system leakage, and the capability for water replenishment in the event of plant emergencies; 3.
the heat duty and adequacy of the Airblast Heat Exchangers and their degree of protection from tcrnado missiles.
H.
Compressed Air System The following additional information must be submitted in order for the staff to complete Qts review of the compressed air system:
l.
a list of safety-related air operated valves to which seismic Category I, Quality Group C air receivers and connecting piping will be provided.
2.
the design criteria and bases for establishing the stored air volume assuming that the air operated valves should be operable from the control room for cold shutdown of the plant.
I.
Equipment and Floor Drainage Jystem It is proposed that the equipment and floor drainage systams are not safety-related except for piping and valves required for containment isolation.
It is the staff's position that (1) the equipr.ent and' floor drainage system is considered safety-related where backflow can cause flooding of safety-related areas, and
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(2) these portions of the system should maintain their function ohile sustair ing loss of any active co ponent and shcuic be desigr.ed to seismic Category I and Quality Group C criteria.
The CRSRP design should be revised to demonstrate that the above positions are met.
Also we will require a P&ID of the equipment and floor drainage system.
J.
Turbine Missiles The turbine proposed for use in CRERP is different from those currently used in light water reactors (e.g., higher rotation rate).
The proposed design and plant layout afford considerable protection against potentially damaging tur';ne failures.
- However, in order for the staff to complete its re iew of this matter, additional information regarding the turaine design is required.
IX.
ACC CENT ANALYSIS (RACICLOGICAL)
A.
Continuous Contair ent Purge, The acplicant proposes to maintain tne envirorment insice containment by continuous purge.
Since docketing, the purge valve size has been reduced from 36 inches to 24 inches and atmospheric temperature control within inerted cells has been changed frcm 00WTHERM to a chilled water system.
In light of the applicant's decision to use chilled water inside containment and apparent reliance on active operation of valves to institute containment isolation, the staff believes that alternate schemes for controlling the RCB environment should be pursued.
The Site Suitability Report summarizes the staff concern.
Based on (1) the applicant's design for'open hatch refueling, (2) the analysis of RAPS surge vessel rupture, and (3) the analysis of cold trap fires, the concept of a passive contair. ment system is not achieved.
The staff has expressed concern regarding the release of radioactive materials in the event of accidents inside containment, such as a cold trap fire or a rupture of the RAPS (refer to Question 310.45).
The initiation of containment isolation is depencent on the w
adequacy and location of selected sensors inside containment.
Assuming a thorough analysis and understanding of the various accident sequences and scenarios, such an approach may be acceptable.
However, the staff does not agree with the applicant's characterization of these events.- Adequately conservative assessments of the spectrum of design basis accidents must be developed before the staff could consider this purge concept to be acceptable.
At this time, the continuous purge concept is not considered to be a conservative or appropriate means for assuring a lcw risk to public health and safety.
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Open Hatch Refueling The applicant has prcposed to conduct open hatch refueling operations with the containment in free communication with the refueling building. This latter structure is not leak-tight.
The staff has not agreed with the applicant on a design basis fuel handling accident and specifically has taker, exception to (1) the proposal to rely on elastomeric seals as the primary barrier to a fission product release for EVTM-related accidents, and (2) the apparent lack of positive means to control and mitigate other postulated accidents during refueling.
The current reliance on the ex-vessel handling machine as a containment recuires detailed staff review of the device and an assessment of the potential for radiological releases due to failures of cooling systems associated with the device.
To assess the adequacy of croposed filter systems, the staff must concur with the applicant regarding the fuel handling accident radiological
- ource term and must completa its evaluation of the ventilation systems.
C.
Steam Generator Tube Failures The staff has not completed its evaluation of the consequences of primary sodium leakage through the Intermediate Heat Exchanger
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(IHX) and subsequent release to the environment through the sodium / water reaction pressure relief subsystem (SWRPRS). As a result of tube failures, there is potential for driving primary sodium through the IHX, IHTS, and ruptured steam generator tubes to the atmosphere, thus breaching containment.
There are no isolation valves in either the PHTS or IHTS and reliance is placed on the integrity of' the IHX' to prevent such atmospheric releases.
Additional information is required.
D.
Sodium Fires Rupture of IHTS piping outside containment results in significant overpressures in the steam generator building.
High capacity venting to the atmosphere is required in the building to prevent overpressurization.
The sodium toxicity levels resulting from an IHTS rupture outside containment and the radiological consequences resulting from IHX leakage have not been fully addressed.
The applicant's analyses in PSAR Section 15.6.1.5 regarding the radiological consequences assumes 150 gallons of primary coolant has leaked undetected into the IHTS prior to the accident and 1.4 pounds leaked during the accident.
An analysis supporting these values has not ;.et been submitted for our review and, thus, the conservatisms of the radiological and system analyses have not yet been evaluated.
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E.
RAPS Surge Vessel Rupture A rupture of the RAPS Surge Vessel is analyzed in PSAR Chapter 15.7.2.4.
The applicant had preposed to locate this vessel in a centrolled leakage environment such that the controlled
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leakage would be used to limit offsite deses.
The staff position in this regard was that the propcsed design criteria provided insufficient control of the consequences of a Surge Vessel ructure.
The applicant is modifying his appr:ach by relocating the vessel inside the RC3.
Documentation was to be provided by May 1,1977 (see PSAR response to Q 310.48 in Amendment 36).
The adequacy of the analyses associated with this design change has not been reviewed.
F.
Cold Trap Fires The pplicant originally proposed a site suitability source term
..-hich wcs based en quasi-machanistic secuences (refer to DSAR Chapter 15A).
Due to the radioactive. invent:ry of cold trapc, a cold trap fire is a severe event.
The staff has completed a sufficiently cetailed review to determine that the assumptions regarding this event are not conservative.
In this regard, the staff requested add'tional information on August 19, 1976 (refer to Question 310.49).
No response has been provided. Of particular staff concern is the applicant's analysis of aerosol depletion effects.
G.
Instrumentation to Follow the Course of an Accident
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Regulatory Guide 1.97 was-issued for comment in December 1975.
The spirit and intent of this guide is to' provide sufficient information to plant operators during an unlikely plant event so that proper actions can be taken.
For the CRERP these events can be characterized:
]
1.
Cesign basis events - The applicant has described in the PSAR those instruments to be provided botn inside and outside the RCS.
The staff has not completed its review of this subject.
2.
Events beyond the design basis - The staff has requested additional information (refer to staff letter of March 30, 1977) regarding the applicant's TLTM report, which addresses core melt down.
At a minimum, the staff will impose Position 3 of Regulatory Guide 1.97 (Rev.1) which in essence requires kncwledge of the radiation level inside containment, the containment pressure, the ? HTS and/or cell 1
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pressures, and release T.cnitor indications.
It is not apparent that auch features are provided for in the CRSRP design.
H.
Risks Associated with Nearly Industrial Activities
,7 The CRSRP is located adjacent to the Holifield National Laboratory.
The applicant was requested to provide an analysis of the pctential impacts to the CRERP of accidents at nearby facilities.
The applicant's response was incceplete -ith regard to the prepose Exxon facility and information remains to be provi ad.
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