ML19339A182
| ML19339A182 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/16/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Gary R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| References | |
| NUDOCS 8011030208 | |
| Download: ML19339A182 (12) | |
Text
.
75cA 0
f **%gg
?,
UNITED STATES
[ 7 ).-o i
NUCLEAR REGULATORY COMMISSION 3-I WASHINGTON, O. C. 20555 s
%..v y OCT 161980 Docket Nos. 50-445 and 50-446 Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Towers Dallas, Texas 75201
Dear Mr. Gary:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR COMANCHE PEAK STEN 4 ELECTRIC STATION, UNITS 1 AND 2 Enclosed is a request for additional infonnation which we require to complete our evaluation of your application for operating licenses for Comanche Peak Steam Electric Station, Units 1 and 2.
This request for additional information is the result of our continuing review by the Core Performance Branch, Accident Evaluation Branch, Radiological Assessment Branch, Meteorology Section and Quality Assurance Branch. Please amend your FSAR to include the information requested in the Enclosure.
Your response to the enclosed request for additional information should be submitted within six (6) weeks. Should you have questions concerning this request for additional information, please contact us.
Sincerely, o
,gRobertL.Te sco, Assistant Director M
for Licensing
% Division of Licensing
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page 801 j 03 OM p
s l
007 18 1380 Mr. R. J. Gary Executive Vice President and 7
General Manager Texas Utilities Generating Company 2001 Bryan Towers Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq.
Mr. Richard L. Fouke Debevoise & Liberman Citizens for Fair Utility Regulation 1200 Seventeenth Street 1668-B Carter Drive Washington, D. C.
20036 Arlington, Texas 76010 Spencer C. Relyea, Esq.
Resident Inspector / Comanche Peak Worsham, Forsythe & Sampels Nuclear Power Station 2001 Bryan Tower c/o U. S. Nuclear Regulatory Connission Dallas, Texas 75201 P. O. Box 38 11en Rose, Texas 76043 Mr. Homer C. Schmidt Manager - Nuclear Services Texas Utilities Services, Inc.
2001 Bryan Tower Dallas, Texas 75201 Mr. H. R. Rock Gibbs and Hill, Inc.
393 Seventh Avenue New York, New York 10001 Mr. A. T. Parker Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 David J. Preister Assistant Attorney General Environmental Protection Division P. O. Box 12548, Capitol Station Austin, Texas 78711 Mrs. Juanita Ellis, President Cicizens Association for Sound Energy 1426 South Polk Dallas, Texas 75224 Geoffrey M. Gay, Esq.
West Texas Legal Services 406 W. T. Waggoner Building i
810 Houston Street Fort Worth, Texas 76102
T i
1 1
J ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 & 2 DOCKET NOS: 50-445 AND 50-446 i
9 i
I 4
i 1
i l
\\
~.
O"D TED Y%@'
IAs S
, S..\\b 230 CORE PERFORMANCE BRANCH 237.14, The NRC staff has been generically evaluating :nree ma:erials medels :na:
are used in ECCS evaluations. Those models predi:: :ladding ructure tem erature, claccing burs strain, and fuel assem:ly now clocxage. We have (a) discussed Our evaluation wi n venders and c:ner incustry recre-sen atives (Reference 1), (b) suciisned NUREG-0630, "Claccing Swelling and Rupture Models for LOCA Analysis" (Reference 2), and (c) recuired licensees to c:nfinn :na: their Ocerating reacter, would con-inue :: be in : nformance wi n 10 CFR 50.46 if ne NUREG-0620 meceis were su:stitu ec for ne cresent materials moceis in :neir ECCS evalut. icns anc :er ain 0:ner ccm:ensat:ry mocei :nanges were allcwed (References 3 inc
).
Until we have ccm letec cur generic review anc imclemented new ac:e :ance cri: aria for claccing medels, we will recuire :na: :ne ECOS analyses in your FSAR Oe ac: mpanied by supplemen:ai calculations :: be performed wi n :ne materials mccels of MUREG-0630.
For these succlemen al calculations only, we will ac:e:: c:ner c:meensa::ry mecel enanges :na: may nc: ye: be a: roved by -he NRC, :u are : nsisten wi n :ne enanges alicwed for :ne confirma Ory coerating reac:Or calculations men:icnec at:ve.
?!?ERINCII 1.
Mem:randum fr:m R. P. Denise, NRC, c R. J. Mat:scn, "Su=r.ary Minu es Of Mae:ing en Claccing Rupture Tem:erature, Ciac ':g Strain, anc Assemolv Ficw 310ckage," Novem er 20, 1979.
2.
D. A. Pcwers anc R. O. Meyer, 'Cladcing Swelling and Rupture Mccels fer LOCA*
Analysis," NRC Reper NURE2-0520, A:ril 1980.
1.
- s er fr
- = 0.1. Eisenna:, NRC, :: ali Coera-ing Lign: Wa:ar React:rs,
- a ac Novem er 9,1979.
4 Mem:rincum from M. R. Dent:n, NRC, :: C mmissioners, "PO:antiai Jeficiencies in 2005 Evaluation Moceis," Ncvem:er 26, 1979.
31 0 ACCIDENT EVALUATION BRANCH 312.22 Provide the locations of all safety-related equipment not contained (3.5.1.4) within reinforced concrete buildings or structures.
Provide the structural composition of all walls and roofs of buildings housing safety-related equipment, as well as the building locations. Discuss the sizes and directional orientations of any openings in these buildings.
312.23 Describe the protection of the control room air intake (s) and diesel (3.5.1.4) generator exhaust pipes from tornado-generated missiles.
312.24 Provide description and drawing showing the locations of control room (6.4) outside air inlet (s) relative to potential radiation releases.
312.25 Staff position is to use steam flash fraction instead of iodine partition (15.6.2) coefficient in estimating the dose from the failure of 3-inch CVCS istdown line. Provide your basis for using iodine partition coefficient.
312.26 In evaluating the radiological consequences of fuel handling accidents, (15.7.4) provide the reason Regulatory Guide 1.25 gap fraction is not directly used.
Section 15.7.4.2 states incorrectly that the reason is discussed in Appendix 1A.
330 RADIOLOGICAL ASSESSMENT BRANCH 331.12 Section 12.3.4.1.1 addresses design criteria of fixed systems of airborne radioactivity monitoring. Design Objective 5 on page 12.3-44 indicates that continuous surveillance is made within enclosures by monitoring exhaust duct radiation levels which alarms when abnormal activity levels are reached so that inadvertent entry will not be made into enclosures where the levels may exceed Part 20 limits. With respect to this objective, please answer the following questions:
(a)
In specifying " exhaust duct radiation levels" and " alarming at abnormal activity levels" there are two units of measure-ment inferred; mr/hr and uci/cc respectively.
Please indicate the unit of measurement for the detector. What is its lower limit of sensitivity for particulates and iodines in MPC-hrs based on this measurement unit (i.e., how long would it take to detect multiple MPC's eW1ved in a single enclosure assuming particulate and/or iodine activity). What is its sensitivity for noble gases.
(b) Describe how one would identify the specific enclosure having the airborne radioactivity when an exhaust duct monitor, located in the plant area specified in Section 12.3.4.1.2, starts to alarm.
331.13 Section 12.5.1 specifies that the H.P. program will be implemented in accordance with hgulatory Guide 8.8, Rev.1.
Since this Guide has been superse' with Revision 2 (with Revision 3 in draft out for connents). Puse indicate your plans for correcting this section to indicate implementation to Revision 2 or address your alternatives.
331.14 Section 12.5.2, pg.12.5.5 refers to a " hot instrument shop or other appropriate areas" for calibration and maintenance of portable health-physics instruments. Please describe the function of the " hot instru-ment shop" and where "other appropriate areas" may be located for calibration and maintenance of the aforementioned instruments.
331.15 Table 12.5.2 " Portable Health Physics Equipment" shows quantities of instrumentation not adequate to meet the anticipated needs of a two unit plant.
The staff position is that sufficient numbers of instru-mentation be available in operating condition to accomodate the need to monitor such large numbers of operations that may be required in radiation creas and high radiation areas throughout the plant during major maintenance and refueling outages and/or accidents.
In arriving at a total number, consideration should also be given to the survey instruments that may be in a calibration, maintenance or inoperative-on-the-shelf s+atus during the outage and/or accident. Additionally the inventory should include the requirements for selected ranges, sensitivities, types of radiation to be monitored, accuracy required, and types of monitoring to be performed. Therefore the table should be revised to reflect these needs to operate a two unit plant. Also
ment of fast and thennal neutron flux densities ik counter for the neutron counter using a moderated and bare Bf an unacceptable method for neutron dosimetry based on state-of-the-art techniques.
Rem-meter monitor such as the Snoopy or PNR-4 should be used to measure the neutron dose equivalent rate directly. Please state your intentions regarding replacement of or addition to the neutron monitor listed in Table 12.5.2.
331.16 3ase:: on information contained in tne draft document " Criteria for i
utility Management and Technical Competence" it is our position that your station organization chain (Figure 13.1-4) should show that the radiation protection group is a separate organization from the Chemistry group and that the radiation protection manager report directly to the General Superintendent. Regulatory Guide 8.8 states that the RPM should be independent of the technical support division and should have direct recourse to the plant manager to resolve questions relating to the conduct of the radiation protection program.
Your FSAR and proposed Technical Specification should therefore be revised accordingly.
331.17 Concurrent to the change request in 331.16 above, Figure 13.1-4 should also show that Health Physics technicians and Chemistry technicians become separate groups and each report directly to their respective Radiation Protection and Chemistry group managers. This change request is also in accordance with the aforementioned draft document.
331.18 Please describe your plan to provide backup coverage in the event of the absence of the RPM and outline the qualifications of the individual who will act as the backup. The December 1979 revision of ANSI 3.1 specifies that the temporary replacement for an RPM should have a BS degree in science or engineering, 2 years experience in radiation protection,1 year of which should be nuclear power plant experience, 6 months of which should be on-site.
331.19 Section 13.1.2.3 specifying shift crew composition, does not state that an H.P. technician will be onsite at all times (e.g., including back-shift and weekends). NUREG-0654 " Criteria for Pre;;aration and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants" requires that a radiation protection technician, whose qualifications are described in ANSI 18.1, shall be onsite at all times. Section 13.1.2.3, as written, would allow a designated member of the shift crew (e.g., reactor operator) to act as a health physics technician if he is qualified to implement radiation protection procedures.
It should be noted that this qualification is no longer acceptable to the staff after the reactor is at power. Only an assigned health physics tehnician will be acceptable based on new staff require-ments. Therefore, Section 13.1.2.3 should be revised accordingly.
331.20 Figure 13.1-4 shows that the number of Chemistry and Health Physics technicians for one unit is 11 and for both units 12. Assuming that the change addressed in question 331.17 is consummated and that the number of health physics technicians is 5 for one unit and 6 for both units, justify that these numbers of technicians are sufficient to accomodate the operation of the station during normal operation of the station anticipated oper'.it"nal occurrences aad accidents.
State your pleas for staffing of H.P. technicians during major maintenance and refueling outages.
Include the nunber of additional H.P. technicians that may be needed and their requirement to be qualified in accordance with ANSI 18.1.
331.21Property "ANSI code" (as page type) with input value "ANSI 18.1.</br></br>331.21" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. Your response to Q331.10 requires sr.me additional information. Please state the dose rate levels at contact with the 8" lead shieiding shown in Figure 12.3-33.
Qu'.* position is that the shielding shall be such that the resultant contact radiation levels during fuel transfer shall be no greater than 100 Rads /hr. Also explain the gap in the lead shield shown in Section 3-3 of Figure 12.3-33 (sheet 3 of 3).
- Finally, what is the maximum expected dose rate in the area where the lockable wire mesh doors will be installed at elevation 302', (Figure 12.3-33),
during fuel transfer
370 METEOROLOGY SECTION 372.31 Provide estimates of the weight of the 100-year return period (2.3.1) snowpack and the weight of the 48-hour Probable Maximum Winter Precipitation for the site. Using the above estimates, provide the weight of snow and ice on the roof of each safety-related structure. (Regulatory Guide 1.70, " Standard Fomat and Content of Safety Analysis Reports for Nuclear Power Plants," Section 2.3.1.2.)
3T2.32 The meteorological data used in evaluating the performance of the (2.3.1) ultimate heat sink (UHS) contains 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the temperature was greater than 100 F.
Present the criteria used in selecting 1974 as "the most severe year with regard to ultimate heat sink thermal perfomance." Provide a comparison of the offsite data used to determine the most severe year and the onsite data used in the UHS evaluation. The summer of 1980 has had a long period of persistent high temperatures (greater than 100 F). Provide a comparison of these data with the most severe year previously selected.
372.33 Explain why most of the onsite precipitation data was invalid for the (2.3.2) four-year data period. Outline the corrective measures to ensure an acceptable data recovery (as defined in Regulatory Guide 1.23) of the precipitation measurements for the operational program.
372.34 Section 2.3.2.1.9 compares Forth Worth data with Love Field data. The (2.3.2) conclusion that the onsite data is representative of the longer term climatological conditions is not supported by the comparison. Pro-vide a direct comparison of the Fort Worth data and the onsite data in order to support the conclusion.
372.35 Cold air drainage down Squaw Creek (less than 700 ft. MSL) cannot be (2.3.2) adequately determined by comparing the 10m (-850 ft. MSL) and 60m (~1000 ft. MSL) wind directions. Variable terrain can influence other local diffusion characteristics. Other effects include mechan-ically-and themally-induced flow and flow blocking. Discuss terrain effects that influence the dispersion and transport characteristics at the Comanche Peak site.
372.36 An operational meteorological measurements program is needed to make (2.3.3) the required assessments to satisfy the requirements of Appendix E and Appendix I to 10 CFR Part 50. Therefore, the program should be installed and operating by the fuel load date. Regulatory Guide 1.23 provides guidance on system acceptability (Revision 1 of Regulatory Guide 1.23 has been released for consnent and may impact your meteor-ological measurements program). Provide a detailed description of and the expected date for initiation of the operational meteorological measurements program.
Include descriptions of any instrumentation which will not be the same as in the preoperational program (including system accuracies). Provide assurance that the area around the tower will be free of obstructions (buildings, storage piles, etc.) by the beginning of the operational program.
If not, provide locations and
dimensions of those items within 600 meters (ten times the measuring height) of the tower and justify that they will not influence the measurements.
Provide the dimensions of and the distance to the fill dirt pile at the end of the peninsula. Justify that it will not affect the meteorological measurements or previde the date for removal of the pile.
372.37 Data display in the control room should meet the position of Regula-(2.3.3) tory Guide 1.23.
Computer terminal display of data in the control room would be satisfactory if a continuous time-history of data were displayed such that no significant transient response csn occur inside the recording and display interval. (Fifteen mi..utes is considered the maximum acceptable averaging interval.)
Provide a description of a program to display the meteorological data in the control room, the Technical Support Center and the Emergency Offsite Facility (as required in NUREG-0696, " Functional Criteria for Emergency Response Facilities," July 1980).
372.38 A description of an upgraded mer.eorological program in compliance (2.3.3) with NUREG-0654, Appendix 2 " Criteria for Preparation and Evaluation of Radiological Emergency Restonse Plans and Pr2paredness in Support of Nuclear Power Plants," is required by Appendix E to 10.CFR 20 at the time at application for.. full power license.
The basic elements of a prceram which address the meteorological aspects of this requirement are:
l 1.
A primary meteorological measurements program with redundant power sources for the primary system.
2.
A backup meteorological measurenents system with redundant power sources.
3.
A system for making real-time predictions of the atmospheric effluent transport and diffusion.
4 A capability for remote interrogation, on-demand, of the atmos-pheric measurements and prediction systems by the licensee, emer-gency response organizations, and the NRC staff with primary and backup communications systems.
. The meteorological program of the Emergency Response Plan should be implemented by April 1,1981 or by the date of issuance of the full power license, whf chever f s later, Item i should be addressed in response to question 372.36.
Provide a commitment for the remainder of your upgraded meteorological program along with a completion schedule to meet these criteria.
372.39 Provide more details on the model used to estimate the atmospheric (2.2.3) dispersion in the gas pipeline accident.
Include: (1) the rationale for using design basis accident meteorological assumptions determined for a radioactive release iri the analysis of a flammable and/or explosive source; (2) the procedures and numeric values used for the virtual source distance correction to account for initial finite source size; (3) a quantitative description of how Briggs plume rise equations were applicd; (4) definition of how a wind speed of 1 m/s and stable atmospheric conditions were determined to be
" conservative" for this type of event.
o 420 QUALITY ASSURANCE BRANCH 421.81 Paragraph C on page 9.5.246 references Section 17.3 for those systems, structures, and compcnents covered by the operation's QA program and page 17.2-10, Section 17.2.2 references Table 17A-1 as the listing of the systems, structures, and corponents covered by tre ocerational CA program. We are unable to locate Section 17.3 in the FSAR and believe this is a typographical error.
Please correct this error or orovide us with Section 17.3 of the FSAR.
421.32 7he response to Q421.44 is not totally acceptable as your description in Section 17.1.8, page 17.1-27 of the FSAR does not exp'ain how receipt inspection controls the status of acceptable matarial and equipment. The deleted portion of your previous description clearly stated that a status tag, sticker, or stanp was placed on the equipment to identify the stacus of an item whereby the present descriptio:. implies only " hold" or " reject" status tags are applied.
It is our position that you clearly describe in more detail, how the acceptance status of material or equipmant is con-trolled following receipt inspection.
421.83 Amendment 11, page 17.2-26, paragraph 17.2-10 of the FSAA deletes the previously accepted portion whereby QC inspectors' certifications were to be maintained current through the TUGC0 training program. The pre-sent revision in Amendment 11 implies inspectors need only be qualified.
This is contrary to Section 2.2 ',f AfiSI N45.2.6-1973 which delineates certification requirements for inspectors. Please explain your rationale for this deletion and provide your alternatives in sufficient detail for our review and evaluation.
-