ML19224D184

Forwards Request for Addl Info Re Application for OL

ML19224D184
Person / Time
Site:	Byron, Braidwood
Issue date:	06/01/1979
From:	Varga S Office of Nuclear Reactor Regulation
To:	Reed C COMMONWEALTH EDISON CO.
References
NUDOCS 7907110051
Download: ML19224D184 (16)

Text

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NUCLE AR REGULATORY COMMISSION y 3 or g W ASHING T O N, D. C. 20555 C S o %, *.../ JUN 1 1979 Docket Nos.- STN 50-454/455 STN 50-456/457 Mr. Cordell Reed Assistant ' lice President Coninonwealth Edison Company P. O. Box 767 ~ Chicago, Illinois 60690

Dear Mr. Reed:

FIRST ROUND QUESTIONS ON THE BYRON AND BRAIDWOOD OL APPLICATI

SUBJECT:

In our review of your application for operating licenses for the Byron Station, Units 1 and 2, and the Braidwond Station. Units 1 and 2, we have identified a need for additional infonnation which wt. require to complete our review. The specific requests contained in the enclosure to this letter are the second set of our round one questions and cover those areas of our review performed by the following: (1) Accident Analysis Branch and (2) Quality Assurance (Conduct of Operations). In order to maintain our present schedule as stated in our letter of February 22, 1979, we need a completely adequaie response to all questions in the enclosure by July 21, 1979. Please contact us if you desire any discussion or clarification of the enclosed requests. Sincerely, l / tev n A. h rg,(Cf f Light Water Reactors Branch No. 4 Division of Project Management

Enclosure:

( Request for Additional Information n., cc: See next page ))0 L. \\ U { '[.i 8 700711005/,

Commonwealth Edison Company cc: Mr. William Kortier Atomic Powr Distribution Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 John W. Rowe, Esq. Isham, Lincoln & Beale One First fiational Plaza ~ 42nd Floor Chicago, Illinois 60690 Mrs. Phillip B. Johnson 1907 Stratf ord Lano Rockford, Illinois 61107 Ms. Marilyn J. Shineflug 1816 Judy Lane DeKalb, Illinois 60115 P Ms. Beth L. Galbreath 734 Parkview Rockford, Illinois 61107 C. Allen Bock, Esq. P. O. Box 342 Urbana, Il1inois 61801 Thomas J. Gordon, Esq. F Waaler, Evans & Gordon 2503 S. tieil A-{t Champaign, Illinois 61820 n' A~y q. .m,. E.' ~. f. ts. w I ,.i L. l '7 L M 0

311-4 SEC110N A, ACCIDENT ANALYS. BRANCH 311.0 For the urban centers listed in Table 2.1-10 for Braidwood, 311.14 (Table 2.1-10) previde the estimated population for the year 2020 and the basis for such estimates. For the Byron site, please clarify the intentions of Mr. Yost 311.15 with respect to his restricted use airport. In particular, will (2.2.2) the FAA designation of this facility be changed and will it be removed from future editions of the Chicago FAA Sectional Aero-nautical Chart? Some of the missile velocities presented in Table 3.5-4 appear 311.16 (3.5.1.4) to be lower than the Revision 0 spectrum of Standard Review Plan Section (SRP) 3.5.1.4 (Revision 1). As noted in SRP 3.5.1.4, mixing of velocities between spectra is not considered acceptable. demonstration that the plant is protected against We require one of the missile spectra of SRP 3.5.1.4 Rev. 1, Pr ovide a table which lists all systems or system components for 311.1 7 (3.5.1.4) which tornado missile protection is pro /ided in accordance with Regulation Guide 1.117. Include in the table the location of the system or component and the means and degree of protection (e.g. roof and wall thicknesses, concrete strength, etc.). Section 6.4.4 and Table 6.4-1 indicate that your calculated 311.18 (6.4) occupational doses to control room operators after a DBA are within the levels required by Criterion 19 of the General Design Criteria. Clarify the doses listed in the referenced table to indicate the applicable plant (Byron or Braidwood) and the type In addition, of dose (i.e., beta, gamma, whole body, thyroid). outline the analysis that was performed, identifying all assump-tions for each plant used in the analysis including, as appropriate: '* \\ b a 7 L.

311-5 Assumed credit for engineered safety features such as a. containment cleanup systems and the control room makeup air f ilters. b. Assumed rate of unfiltered air inleakage af ter the DP.A, including such leak paths as control room doors, ducts, penetrations, outside air isolation dampers and contaminated air from rooms adjacent to those served by the control room HVAC. Assumed atnospheric dispersion (X/Q) factors for the control c. room air intake vents, the data source (e.g., meteorological records, literary references) for thesc X/Q values, and other assumptions nade in rcaching the X/Q values used in your analysis (release height, distance and direction to receptor-control room air intake vents, building wake factor, projected containnent area, wind direction changes, control room occu-pancy factor). These data need to be supplied, as. ypropriate, for both the Byron and Braidwood stations. d. The volume of the control room envelope. For your reference in this dose analysis, see the following: U.S. NRC Standard Review Plan Section 6.4, "Hab.tability a. Systems," (t"JUEG-75/087, Section 6.4). b. Ilurphy, K.G. and K.H. Campe, "i;uclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion !9," Proceedings of the Thirteenth AEC Air Clean ing Conference, August 1974. U.S. I;RC Standard RJview Plan Section 15.6.5, Appendix A, c. 330 nJ'q

311-6 " Radiological Consequences of a Design Basis loss-of-Coolant Accident: Containment Leakage Contribution," and Appendix B, "... Leakage from Engineered Safety Features Components out-side Containment." 311.19 It is not clear from your description of the spray system what (6.5.2) type of spray nozzle you propose to use in the containment. Clarify your description by including the model number and nozzle manufacturer for the containment spray nozzles in your proposed design. Your description of the spray ring header placement is not suffi-311.20 (6.5.2) cient to permit a determination of the frat. tion of the containment which will not be sprayed directly. Provide the volumes for the following regions of the containment: region which is sprayed directly ; region which is not sprayed and with no comunication with the sprayed region; region which 's not sprayed and in poor comunication with the sprayed region, and the region which is not sprayed and in good communication with the sprayed region. Provide all the assump ions used to calculate these volumes in-cluding appropriate figures and analyses. 311.21 Provide additional information regarding the negative pressure (6.5.1) (9.4.5) that will be established in the fuel handling builoing following a postulated fuel handling accident. Your response should include the specific negative pressure required, the basis for this pres-sure, the time required to establish the design negative pretsure, the areas or buildings that are maintained at the negative pres-go u"O V)

311-7 sure, the necessary supply and exhaust flow rates, the methods of detecting and controlling changes in the negative pressure, and the fur.1 handling building isolation procedures. Conforn ance to the criteria of SRP 15.7.4 should be demonstrated. 311.22 The fuel handling building exhaust system is designed to Safety ( 6. 5.1. ) (9.4.5) Class I requirements. To meet the guidelines of Regulatory Guide (3.2.1) Appendix A 1.29, this system should be designed to Seirnic Category I require-A1.29 ments. It is not apparent from the discussian in FSAR Section 3.2.1 if your Safety Class I requiremente neet the staff require-ments as given in Regulatory Guide 1.29 and Appendix A to 10 CFR Part 100. Provide the necessary clarification. 311.23 According to the FSAR, the Fuel Handling Building Exhaust System (9.4.5) (6.5.1) (FHBES) operates continuously during all normal plant operating conditions, discharging the exhaust from the plant via the char-coal bypass line. Additional information and clarification is required with respect to the following:

1) Two dampars are installed in series in the bypass line, one labeled normally open, the other normally closed (Fig. 9.4-5, sheet 12).

Explain or correct this apparent functional inconsistency. 2) As stated in the FSAR, the bypass line is closed automati-cally en a high radiation signal in the exhaust dec+. However, on FSAR Fig. 9.4-5, sheet 12, the damper actuators are nanual hand switches (HS). Explain this inconsistency. O ') ) ))(} L. L. '

311-8 3) The high radiation signal for iscelation of the charcoal bypass line and opening of the dampers in the charcoal booster fan lines is generated by a single radiation monitor in the exhaust duct of t8:e FHBES ', Fig. 9.4-5, sheet 3). Since the exhaust system is an ESF system, this arrange-ment does net appear tu satisfy the requirement for the delivery of the isolation and switching signal in the event of a single failura. Clarify this inconsistency. 4) Since the FHBES docs not include a means for humidity control of the exhaust prior to entering the charcoal fil-ters, provide a discussion of the operation of the FHBES for the cr.se of high radioactivity and high humidity in the exhaust flow upstrear.n of the exhaust filter plenums (Figure 9.4-5, sheet 12). Provide a discussion of the po-tcntial prooleais which could arise from operating the FHBES u:1 der these conditions. 5) In FSAR Section 6.5.1.2.3, the FHBES is referred to as the Auxiliary Building Exhust System. Correct this inconsis-tency. Table 15.1-3 contains an esitry for the long term steam release 311.24 (15.1.5) from a dafective steam generator and quotes a value of 1,000 pounds for the 0-8 hour period. Powever, the initial steam re-lease in the detective steam generator over a 0-2 hour period is given as 163,000 pounds. Provide sr.c discussion which clear-ly defines the differences between these two values for the affected steam generator over the period 0-8 hours. O')9 L U

311 -9 Provide either in FSAR Section 15.3.3.2 or in Table 15.3-3 the 311.2 5 (15.3.3) amount of feel. hat would be expected to experience DNS and there-fore would be assumed to experience clad perforation and/or fuel melt for the Reactor Coolant Pump Shaft Seizure accident. There is an apparent discrepancy between the steam release 311.2 6 (15.4.8) values and other inportant parameters of Table 15.4-4 and those For example, provided in the text of FSAR Section 15.4.8.3. the text states that for the case of loss-of-offsite power, 83,000 pounds of steam will be released by a steam dump through the relief values for a period of only 350 seconds. Table 15.4-4, on the other hand, states that this same steam dump would release 113,000 pounds of steam over a period of 500 In light of this information, revise FSAR Section seconds. 16.4.8.3 and Table 15.4 4 as appropriate such that the assump-aiven in the text and listed in the table are consistent tiorr with e_ a other. Also Table 15.4-4 indicates that even though the condenser is available in the realistic case, no steam dump to the condenser Clarify why no steam dump to the condenser occurs would occur. in the rea.istic analysis for the consequences of this accident. Since Acceptance Review Question 311.10 incorrectly identified 311.2 7 (15.5.2) the assumed location of the letdown line break, a response to staff question 311.10 is not required. However, perform an analysis of the radiological consequences of a CVCS letdown line rupture outside containment and downsteam of the outboard isolation L. L.

311-10 salve. The analysis should be performed using the most re'tric-tive single failure, include in your analysis t he t iw rein t red to generate the isolation signal, isclation valve closure time, I single failure of one isolation valve, mass of reactor coolant released prior to isolation and increased iodine reactor coolant concentrations because of the transient conditions (iodine spike). Provide all the assumptions and parameters used in the radio-logical consequence analysis. 311.28 Describe in more detail the design and operation of the proposed (15.6.5) (9.4) containment minipurge system. Include in your discussion the pro-jected number of hours of cperation per year and a radiological consequence analysis of a LOCA while the minipurge is in operation. 311.29 Table 15.6-15 is incomplete. Provide all possible sources of (15.6.5) ECCS recirculation leakage and the leakage rates. If these are referenced from the FSAR, provide sufficient references for these values. 311.30 Provide an analysis idiological consequences from (15.6.5) expected ECCS leakage for the first 24 hours af ter a LOCA and from the assumed failure described in SRP Section 15.6.5 Appendix B for the course of the accident. Also provide a descripcion of the opera' ion of all systems which will be used to mitigate If filters are proposed for iodine dose any such consequences. reduction, provide sufficient informatic, for the staff to deter-mine that the proposed filters meet the recommendations of Reg-ulatory Guide 1.52 and to cetermine that the ECCS pump rooms or any rooms where radioactive leakage might occu can be maintained at a negalisa pressure with respect to the ambient atmosphere to insure that any leakage will be processed prior to releaseV) liT L v-to the envirenment.

311-11 in Table 15.6-9 the applicant uses iodine spray removal coe-311.51 (15.6.5) fficient of 29.9 hr-1 in their conservative analyses. The branch positionstatesthatamaximumiodinesprayrembvalcoefficient of 10 hr-1 can be used when 50?.' instantaneous plateout of radio-active iodine is assumed. Provide a revised LOCA radiological consequence analysis using the spray removal coefficient of 10 hr-1 o25 330

311-12 311.32 In an appendix to the Generating Station Emergency Plan (GSEP) specific (13.3) to the Byron ar.d the Braidwood sites, provide the information cited at the follow 9 places in App 1 dix A to Regulatory Guide 1.101, (Revision 1, March 1977): 3. 4.1.3, 4.1.4, 4.1.5, 4.2, 5.1, 5.2.1, 5.3.2, 5.4 (Local u,. e:. ) L.4, 6.4.1, 6.4.2, 6.4.3.2, 6.5, 7., 7.1, 7.3, 7.4 (re: m is ably e.a for site evt.

es, including construction workers, and e+fs' r e a gency operation centers), 7.5, 7.6, 8.1.2, 8.2 (last sentence' i.1 and 10.

in lieu of submitting 311.33 With regard to section 5.4 of Regulatory Guide 1.101: (13.3) State and locul agencies' radiological response plans as evidence of reason-able assurance that appropriate and timely response measures can and will be taken in behaif of the population-at-risk in the plant environ, you may address tne applicable elements of the list below for each State ana local agency having a re;ponse role in support of the Byron and the Braidwood If you choose to submit any State or local plans in lieu Emergency P.an. of the above, ensure that the plc..s are reviewed for completeness with respect to the applicable elements listed below. If necessary, request the State or local agencies to include such information in their plans, or you may supplement their plans with the necessary information in Section 13.3 Note that in the absence of a State or local agency plan, the of the FSAR. written agreement with that agency should reflect their concurrence with ? your docketed description of the applicable elements from the following list as related to tt.at ajency's role in support of the emergency resperise pians developed for the Byron and the Braidwood plant. 330 226

311-13 l I 1. The identity of the agency. A description of the authority and responsibility for 2. emergency response functions. A description of the concept of operations including the 3. operational ir terrelationships of all organizations having emergency response roles. The designation and location of the Emergency Operations Center 4. for the direction and/or coordination of emergency support activities. The established relationship and interface with State and/or 5. local government emergency response plans. i The provisions established with the Department of Energy 6. nal Coordinating Office for radiological assistance under Reg the RAP and IRAP programs. A description of the communication plan for emergencies 7. including titles and alternates for both ends of the communi-cation links, and primary and backup means of communication. g Where consistent with the agency function, include the following: Provision for 24-hour / day manning of communication link. a. Provision for administrative control methods for ensuring b. the effective coordination and controi of the emergency support activities. Provisioa for conmunications arrangements with contiguous c. local governments where applicable, Provision for communications arrangements with Federal d. emergency response organizations. Provision for communications with the nuclear facility, e. State and/or local emergency operations centers, and field assessment teams. A description of the communications methods for issuing 8. emergency instructions to the public in the potentialTy affected environs of the nuclear facility. A description of the methods and equipment to be employed -3 9. in determining the magnitude and locations of any radio-logical hazards following liquid or gaseous radioactivity rd eases. n-)] , f, [) j J

c. L '

~ 4 311-11

10. The designation of protective action guides and/or other criteria to be used for implementing specific protective actions and the information needs (e.g., dose rates, projected dose levels, contamination levels, airborne or.

waterborne activity levels-) for implementing such actions. ~

11. A description of the methods for protecting the public from consumption of contaminated foodstuffs.

12. A description of the evacuation plans for the Low Population Zone (LPZ) including survey maps for the facility environs showing evacuation routes as well as relocation and shelter The plans may extend to areas beyond the LFZ and areas.

should include the following: Population and their distribution around the nuclear a. facility. Means for notification of the potentially affected I b. population. Disabilities, institutional confinement, or other factors which may impair mobility of parts of the population. c. d. Means of effecting relocation. Potential egress routes and their traffic capacities, e. Potential impediments to use of egress routes. f.

13. The provisions for maintaining dose records of all potentially uposed energency wnrkers involved in response activities.

14. The provisions for emergency drills and exercises to test and evaluate the response role of the agency, including provisions for critique by qualified observers.

15. A description of the training prcgram established for those individuals having an emergency response assignment.

16. The provisions for periodic review and. updating of the emergency response plans of the agency.

f l We note that section 13.3 references the GSEP filed February 18, 1,975. We 311.34 (13,3) have on file a document dated December 1977 (Revision 0) with page changes dated January 1978 (Revision 1). Please icientify the appropriate document ? to be used for the Byron and Braidwood plants. 330 220

422-1 422.0 QUALITY ASSURAMCE BRANCH - CONDUCT OF OPERATIONS 422.3 Descrit e the nu:'ber of persons reporting to the Project (13.1.1.3) Engineers, P'.-lR section of the SNED, in terms of total numbers by discipline (Electrical Engineers, Mechanical Engineers, Nuclear Engineers) of those assigned specifically to the Byron /Braidwood Station project and those assigned to other projects. 422.4 Describe the number of perso'is reporting to each of the (13.1.1.3) supervisors (tech s taff supervisor, supervisor of radiation ~ pro tec ti on.., supervisor for nuclear waste processi.g...) reporting to the Assis tant Superint ?ndent - Rad / Chem Systems for Production Sys te.n Analysis. Include the proportion of time they will be available to work on the Byron /Braidwood Stations project. In addition, provide the resumes of the principal supervisory personnel in the areas of health physics and rad-chemistry. 422.5 Describe the nunber of professional persons on the staff of (13.1.1.3) the Nuclear Fuel Services Department. In addition, provide the resume of the person filling the position of Director, Nuclear Fuel Services. 422.6 The organization chart provided in subsection 16.6.1 does not (13.1.2.1) indica te the number of persons assigned to common or duplicate positions. Additionally, these numbers should not be included in the proposed Techi.ical Specifications. Therefore, provide in a new plant staff figure, or some other manner, the number of persons a% igned to the positions of Operating Engineers, Shif t Engineers, Shif t Foreman, Nuclear Station Operator, Fquipmen t Opera tor, Equipment Attendant, Instrument Maintenance Foreman, Mechanical Maintenance Foreman, Electrical Maintenance Foreman, Fuel Handling Foreman, Rad / Chem Fe eman and Engineers, Group l eader, Rad / Chem Technicians, Control System Technicians and Mechanics for both single-unit and two-unit operation. 422.7 Expand this section to describe the responsibilities and (13.1.2.2) authori ty of your Shi f t Foreman, Startup Foreman, Nuclear Station Opera tor, Equipment Opera tor, Equipment Attendant, Station Manager, Rad Foreman and Engineers, and Group Leaders. In addi tion, expand your station organization chart to show these organizational units. n JG -jj()r LLs

422-2 Expand your pl int staff organizational chart to show the 422.8 (13.1.2.2) functional units responsible for nuclear engineering and plant quali ty control. For these functional units, show any further functional units reporting to those foranen and the number of persons to be assigned to these units for both single-unit and two-unit operation. Clarify Table 13.1-2 to show the type of NRC license to be 422.9 (13.1.2.3) held by those persons filling the positions for which you state an NRC license will be required. Please clarify the specific line of succession of authority 422.10 (13.1.2.2) and responsibility for station operation when the Station Superintendent is unavailable by describing the line of succession by position titie. 422.ll(RSP) Your proposed shi f t crew complement for two-unit opera tion is not acceptable. It is the staff's position that, in (13.1.2.3) regard to licensed operators, that your shif t crew composition for two-unit nperation should include at least two senior licensed opera tors and thr ee licensed operators. 422.12 Expand Table 13.1-2 to show the qualification requirements (13.1.3.1) for the positions of Assistant Superintendent, Administrative Assistant, as ter Instrument Mechanic, Instrument Mechanical Foreman, Electrical Foreman, Mechanical Foreman, Operating Fngineers, Equipment Operators, Equipment Attendents, Rad Foreman and Fngineers, Quality Control Engineer, and Training Supervisor. This may be done by reference to a specific ANSI N18.i position title or by describing your proposed qualification requirements for the position. 422.13(RSP) Your qualification requirements for the position of Rad / Chem (13.1.3.1) Supervisor is unacceptable. It is the staff's position, in regard to the health physics qualifications, that the qualifi-Cd tion require.nents for this position should meet those described in Revision 1 to Regulatory Guide 1.8 for the position of Radia tion Pro tection Manager. 422.14(RSP) The qualification requirements for the position of Radiation-( 13.1. 3.1 ) Chemistry Technicians are not satisfactory. It is our position that the qualifications for this pesition should meet that described in Section 4.5.2 of ANSI N18.1-1971. 330 230

422-3 Identify the upper level of fsite management position (s) 422.15 thich has overall responsibility for the fonnulation, ( 9. 5.1 ) implementation, and assessment of the ef fectiveness of the station fire protection program. While the station sunc. intendent is generally responsible 422.16 for all activities at the facility, describe any further (9.5.1) delegation of these responsibilities for the fire protection program such as training, maintenance of fire protection systens, testing of fire protection equipment, fire safety inspections, fire fighting procedures, and fire drills. Describe the tm, position of your shif t fire brigade in 422.17 (9.5.1) terms of numbers and job titles. Describe the authority of your shif t fire brigade leader 422.18 (9.5.1) relative to that of your shif t engineers. 971 , 7) n ) U LJ'}}