ML20038B312

Forwards Request for Addl Info Prepared by Containment Sys Branch & Lll After Completing Review of Appropriate Sections of FSAR

ML20038B312
Person / Time
Site:	Byron, Braidwood
Issue date:	11/17/1981
From:	Youngblood B Office of Nuclear Reactor Regulation
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
NUDOCS 8112070586
Download: ML20038B312 (27)

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l NOV 171981 LB#1 Rdg.

TERA JYoungblood PDR MRushbrook LPDR KKiper NSIC Docket flos.: STN 50-454, STil 50-455

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and STN 50-456, STri 50-457 "'

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e VA Mr. Louis 0. De1 George RVollmer gv

' * [,",]RMattson gpj TMurley

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Director of Nuclear Licensing "

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Comonwealth Edison Company u

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RHartfield, MPA 0

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Post Office Box 767

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Subject:

Request for Additional.. Information,,, Byron /Braidwood Stations

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The enclosed request for additional information.for the Byron /Braidwood Stations has been prepared by the Containment Systems Pranch, through its contractor, the Lawrence Livermore. National Laboratnry, aftar completing a review of the appro-priate sections of'tha FSAR.

In the courw of the reyj-w, the need fo.r additional information in the following areas has been identi,fied:

1.

Containment Functional Design (SRP. Section,6.2.1.1);

2.

Subcompartment Analysis (SRP Section 6.2.1.2);,

3.

Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (SRP Section 6.2.1.3);

4.

Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures (SRP Section 6.2.1.4);

5.

Minimum Containment Pressure Analysis for. Emergency Core Cooling Systen Performance Capability Studies (SRP Section' 6.2.1.5);

6.

Containment Heat Removal Systems (SRP Section 6.2.2);

7.

Containment Isolation Systen (SRP Section 6.2.4):

8.

Combustible Gas Control in Containment (SRP Section 6.2.5); and 9.

Containment Leakage Testing (SRP.Section.6 2,6),.

The enclosed request for additional information identifies the open items in our review. Your responses to the enclosed request should be provided not later than December 4,1981.

8112070506 011117 PDR ADOCK 05000454 A

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OFFIClAL RECORD COPY usam mi--mmo nac rosa ais oo-so.i nacu eaa

Mr. Louis 0. DelGeorge

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. NOV 171981 If you require any clarification of this. request please contact the staff's assigned project manager.

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Sincerely, t

Original eigned bg4 54' GordenE.Edi g j

B. J. Youngblood, Chief, Licensing Branch No. 1 Division of Licensing

Enclosure:

As stated cc w/ encl. : See next page e

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i Mr. Louis 0. De1 George Director.of Nuclear Licensing Commonwealth-Edison Company Post Office Box 767 Chicago, Illinois 60590 cc: Mr. William Kortier Atomic Power Distribution Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230.

Paul M. Murphy, Esq.

Isham, Lincoln & Beale One First National Plaza 42nd Floor 1

Chicago, Illinois 60603 i

C. Allen Bock, Esq.

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Post Office Box 242 Urbana, Illinois 61801 Thomas J. Gordon, Esq.

Waaler, Evans & Gordon 2503 S. Neil Champaign, Illinois 61820 Ms. Bridget Little Rorem Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935 Mr. Edward R. Crass Nucl'ar Safeguards and Licensing Division e

Sargent & Lundy Engineers 55 East Monroe Street Chicago, Illinois 60603 i

':cclear Reculatory Commission, Recion III 4

Office of Inspection and Enforcement 799~R60sevelt Road Glen Ellyn, Illinois 60137 1

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4 Mr. Louis 0. DelGeorge s:

Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 i

Chicago, Illinois 60690 ccs:

Mr. William' Kortier U. S. Nuclear Regulatory Comission Atomic Power. Distribution Resident Inspectors Office' Westinghouse Electric Corpcration 4448 German Church Road P. O. Box 355 Byron, Illinois 61010 Pittsburgh, Pennsylvania 15230 1~

Ms. Diane Chavez.

Paul M.- Murphy, Esq.-

602 Oak Street Isham, Lincoln & Beale Rockford, Illinois 61104 One First National Plaza 42nd Floor

. Chicago, Illinois 60603 Mrs. Phillip B. Johnson 3

1907 Stratford Lane l

Rockford,-Illinois 61107 Ms. Bridget Little Rorem

, Appleseed Coordinator i

117 North Linden Street j

Essex, Illinois 60935 Dr. Bruce von Zellin Department of Biological Sciences Northern Illinois University.

- DeKalb, Illinois 61107

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2 Mr. Edward R. Crass 4

Nuclear Safeguards and Licensing Division Sargent & Lundy Engineers i

55 East Monroe Street Chi cago,- Illinois 60603 4

Nuclear Regulatory Commission Region III Office of Inspection and Enforcement 799 Roosevelt Road Glen Ellyn, Illinois 60137 i

Myron Cherry, Esq.

j Cherry, Flynn and Kanter -

i 1 IBM Plaza, Suite 4501 Chicago, Illinois 60611 nr.n--

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REQUEST FOR ADDITIONAL !NFORMATION FOR CONTAINMENT SYSTEMS REVIEW OF BYRON /BRAIDWOOD FSAR 5

t 022.8 Provide information on the required instrumentation to monitor (6.2.1.0 containment atmosphere temperature and sump water temperature in the post-accident environment. Include the instrument range, occuracy, and response time.

022.9 The containment pressere and temperature occident environment (6.2.1.1) conditions listed in FSAR Table 3.ll-2 do not envelop the containment pressures and temperatures calculated fur the containment functional analysis and presented in FSAR Figures 6.2-1 to -14. Provide justification for why this is acceptable or modify the environmental qualifiection specifications to envelop the results of the onolyses.

022.10 Revise the analysis of inadvertent spray cetuation using a spray water (6.2.1.1) temperature of 35 F (Technical Specification 16.3/4.5-9) and a relative humidity of 100% and provide the results.

022.11 Describe the conservatisms in the passive heat sink data provided in Table (6.2.1.1) 6.2-4 which tend to maximize the co!culoted containment temperature and pressure in the containment functional onelysis and in Table 6.2-55 which tend to minimize heat transfer for the minimum containment pressure onalysis for performance capability studies of ECCS.

022.12 Provide on analysis demonstrating that the assumed times for full operation (6.2.1.1) of the RCFC system and containment spray system in the containment furictional onelyses are conservative, i.e.,

40.0 and 45.0 seconds, respectively, for the LOCA cases (FSAR Tcbles 6.2-6, 7, and 8) and 40.0 on'd 88.0 seconds, respectively, for the MSLB (FSAR Table 6.2-9).

022.13 Justify the containment spray system heat removal assumpflen (duration

( * 'I*I) and flow) used in the containment functipnol analyses in light of Section 022.23 Provide on analysis of the effect of the miniflow purge system, i.e., on (6.2.1.5) open purge line, on the minimum containment pressure analysis for performance capability studies on the ECCS (Reference Bronch Technical Position CSB 6-4 B.S.c).

022.24 Provide the assumed. essential cooling water temperature used in FSAR (6.2.1.5)

Table 6.2-25 to verify that the minimum essential service water temperature has been used to maximize the heat removal capacity of the reactor containment fan coolers used in the minimum containment pressure analysis for ECCS performance evoluotion (Reference Branch Technical Position CSB 6-l).

022.25 FSAR Section 6.5.2.2 states " Containment spray injection and coustic (6.2.2) eduction... will continue until... the low-low level clarm of the RWST is annunciated. Containment spray injection and coustic addition may then be terminated, and the operating personnel may transfer the containment spray pumps from the injection to the recirculation mode by first closing the motor-operated volves in the suction line from the RWST, the water and coustic lines to the eductor, and then opening the motor-operated volves in the suction lines from the containment sumps." State clearly whether transferring the containment spray pumps from the injection to the recirculation mode involves stopping and restarting the containment spray pumps os implied in the above statement b-couse the valves in the su'etion line from the RWST ore closed before the valves in the suction line; from the containment sumps are opened.

l 022.26 In Appendix A of the FSAR, it is stated that the applicant complies with (6.2.2)

Regulatory Guide 1.82 with comments and clarifications keyed only to paragraphs 2, 4, and 7 in the Position. Using engineering drawings os oppropriate, describe specifically how each paragraph of the Regulatory Guide 1.82 Position hos been satisfied, and expand the already provided comments and clarif:cotions es follows:

l 2.

Describe the measures taken to preclude damage to'the containment recirculation sump intake filters by whipping pipes or high-velocity

jets of water or steam resulting from high-energy piping breaks outside the primary coolont pressure boundary.

4.

Describe the design measures taken to preclude heavy pieces of debris

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from accumulating near the containment recirculation sump.

7.

Provide the design data and calculations used to determine the design coolant velocity of the vertical inner screen.

Verify that the available surfoce crea used in determining the ' esign coolant velocity d

l is based on one-half of the free surface creo of the inner screen to conservatively account for partial blockage by slowly settling debris.

Since your reported coolont velocity at the vertical inner screen (approximately 0.5 ft/sec) is greater than the recommended value of 0.2 ft/sec, provide test results or on analysis demonstrating that your design coolant velocity at the vertical inner screen will allow debris with a specific gravity of 1.05 or more to settle behind the baffle wa!!s versus on the vertical screen surfoce.

022.27 Verify that the containment recirculation sumps (including trash rocks, (6.2.2) screens, and pump suction inlets) are designated Seismic Category I, as recommended oy Regulatory Guide 1.29 (SRP Section 6.2.2 Il.7).

022.28 Provide the results of the performance testing of the ESW cooling coil f

(6.2.2) assembly and the qualification tests of the back draft dampers referenced in FSAR Section 6.2.2.2.1 (i).

022.29 Describe the containment isolation provisions for the test connection (6.2.4) penetration (P-4)' and the spare penetrations listed in FSAR Table 3.8-l.

Specifically include the design criterio that were applied to the isolation barriers and interconnecting piping.

022.30 Penetrations P-2, 3, 7, 9,14,15, S, 6, 8,10, 22, 2S, and 48 have been listed (6.2.4) in FSAR Table 6.2-58 as having met the requirements of GDC 57.

However, these penetrations serve systems inside containtnent that include components which cre not Seismic Category I and/or not Safety Class 2.

Therefore, in occordance with the provisions of SRP.Section 6.2.4 ll.9.c 1.

4 cnd d, the systems inside containment associated with these penetrations are not acceptable os closed systems and cannot be considered as one of the isolation barriers. Provide information demonstrating that the design provisions for these penetrations meet the requirement for two acceptable isolation barriers in series.

022.31 Provide the relief valve setpoints for ICC9428A and B (Penetration P-22)

(6.2.4) to verify that they are greater than I.S times the containment design pressure in occordance with SRP Section 6.2.4 11.3.g.

022.32 Verify that all normally closed manual volves in test, vent, drain, (6.2.4) instrument, and other similar types of branch lines which, serve os containment irclotion barriers will be sealed closed as defined in SRP Section 6.2.4 1 1.3.f.

022.33 Screwed caps are not occeptable os sealed closed barriers (see SRP (6.2.4)'

Section 6.2.4 II.3.f.). Therefore, describe for the following test and vent connection lines how acceptable containment isolation provisions will be provided.

Penetration Line Number P-25 ICC83A 3/4 P-26 IS184A 3/4 P-27 IRY40A 3/4 P-33 ICVE6AC 3/4 P-33 ICVE6AD 3/4 P-44 IRY43A 3/4 P-48 ICC84AA 3/4 P-48 ICC89AB 3/4 P-53 ICVE6AA 3/4 P-53 ICVE6AB 3/4 P-59 ISl89A 3/4 P-66 ISl8SA 3/4 P-7I ICVF4A 3/4 022.34 Containment isolation valves ISD002A-H in the steam generator blowdown (6.2.4) line penetrations (P-80, 81, 82, 83, 88, 89, 90, and 91) are listed as remote-monvolly octuated in FSAR Table 6.2-58. They are shown on FSAR Figure 11.2-6 as having hand control stations, but the location of the hand control stations is not given. Verify that these valves are remotely operable from s.

the control rooni and have position indicators in the control room in accordance.with ANSI N271-1976, paragraphs 4.2.2 and 4.2.3.

022.35 Provide the volve or volve operator power source for all containment (6.2.0 isolation valves in order to verify electrical redundoney and that all power operated containment isolation volves receive Class IE emergency power,in accordance with ANSI N271-1976, Porograph 4.4.7.

With respect to the requirement for Class IE emergency power, we have already noticed that volves ICV 8355C, D (Penetration P-33) and ICV 8355A, B (Penetration P-53) in the reactor coolant pump seal water injection lines do not receive Class IE emergency power (see FSAR Figure 9.3-4 Sheets 2 and 3 of 9), which is unacceptable. Provide information on how this design deficiency will be corrected.

022.36 The outside automatic containment isolation valve (ICV 8105) on the (6.2.4) charging line, Penetration P-71, is designated essential, while the-inside containment isolation check valve (ICV 8381) is designated nonessential.

Provide information justifying the essential designation of volve ICV 8105 particularly since it is automatically isolated upon a safety injection signal

. and is shown as closed post-accident in FSAR Table 6.2-58.

Verify that the following normally closed.cc.itainment isolation valves will 022.37 (6.2.4) be sealed closed as defined in SRP Section 6.2.4 Il.3.f:

Penetration Number Valve Number System P-37 ICV 8346 RCS Fill Line P-57 IFC8787A,B Spent. Fuel Pool Cleaning Line P-32 IFC8767A,B Spent Fuel Pool Cleaning Line P-50,51 IS18890A,B Safety Injection Test Lines P-59 1S18881 Safety Injection Test Line P-73 1S18824 Safety injection Test Line P-60,

1518823 Safety injection Test Line P-66 1518825 Safety injection Test Line P-26 1518843 Safety injection Test Line 022.38 The following lines are designated nonessential in FSAR Table 6.2-58, and (6.2.4) yet the containment isolation volves are remote-manually operated. It is the NRC position that the containment isolation volves for nonessential lines a n

must be automatically closed when containment isolation is required (SRP Section 6.2.4 11.5 and NUREG-0737 Item E.4.2 Position 3).

Provide justification for why these lines are essential or provide information on how.

this design deficiency will be corrected.

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Penetration Number System P-7, 9,14,15 Essential Service Water Lines P-34 Fire Protection P-56 Service Air System l

022.39 Provide in FSAR Table 6.2-58 the missing distances to the outside (6.2.4) containment isolation valves (Column 11). Additionally, provide evidence that all containment isolation valves located outside contain~ ment have been placed as close to the containment as practical, as required by GDC 55, 56, and 57, since some of the distances listed in FSAR Table 6.2-58 appear to be excessive.

022.40 State which signal automatically isolates the waste disposal line, (6.2.4)

Penetration P-47, and the instrument air line, Penetration P-39.

022.41 The occumulator fill line, ISl23B 3/4", on FSAR Figure 6.3-l, Sheets 3 and (6.2.4) 6, is not clearly designated Safety Category I and Quality Group B between the inner and outer containment isolation volves for Penetration P-55.

Verify that this line is Safety Category I and Quality Group B.

022.42 Provide information justifying why the nitrogen strpply line (Penetration (6.2.4) j P-55) is designated essential in FSAR Table 6.2-58, especially since it is l

shown closed after an accident ~ and automatically isolated on a Phase A containment isolation signal.

022.43 Verify that valves IRH8734A, B; 1RH003A, B; and IRH004C, D are sealed (6.2.4) closed as defined in SRP Section 6.2.4 ll.3.f, in order to ensure that the RHR system is closed outside containment as required for Penetrations P-68 and 75.

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3 022.44 Verify that the IRHOISA, B volve containment assemblies (Penetrations P-(6.2.4) 92, 93) are.leck tight or controlled leakoge housings, os specified in SRP Section 6.2.4 ll.3.e.

Also verify that the compartment vent and drain isolation volves (IS1015A, B; IS1018A, B) are sealed closed as defined in SRP Section 6.2.4 11.3.f.

022.45 FSAR Table 6.2-58 lists the main steam line containment isclation volves (6.2.4) iMS001 A, B, C, and D (Penetrations P-78, 85, 86, and 77) as open post-LOCA. Verify that these volves are automatically closed by a main steam isolation signol, and that they fait closed in the event of a power failure to the volve operators.

022.46 Volves iVG003,1OG053A, B,1OG058, and IRE 9iS7 have been omitted from (6.2.4)

FSAR Toble 6.2-58 (Penetrations P-94, 95, 97, and 65). Provide all the required information, normally provided in FSAR Table 6.2-58, on these containment isolation volves.

022.47 Provide closure times for all power operated containment isolation volves (6.2.4) listed in FSAR Table 6.2-58.

022.48 State whether the containment purge system isolation volves (Penetrations (6.2.4)

P-94, 95, 96, and 97) are normally closed, as listed in Table.6.2-58, or normally open, as shown in Figure 9.4-11. Additionally, verify that the volves foil closed in the event of a failure of power to the volve operator.

022.49 Branch Technical Position CSB 6-4 pertains to system lines which con (6.2.4) provide on open path from the containment to the environs during normo!

plant operation; e.g., miniflow purge system. Describe specifically how each paragraph of the Branch Technical Position is satisfied. Concerning Porograph B.I.g, provide engineering drawings showing the materials and dimensions of the purge and vent system debris screens, and demonstrate compliance with the following criterio:

a.

The debris screen should be Seismic Category I design and insto!!ed about one pipe diameter 'oway from the inner side of the inboard isolation volve.

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b.

The piping between the debris screen and the isolation volve should also be Seismic Category I design, c.

The debris screen should be designed to withstand the LO,CA differential pressure.

022.50 NUREG-0737 Item I I.E.6.2 pertains to containment isolation (6.2.4) dependability. Describe specifically how each paragraph of this NUREG-l 0737 item is satisfied.

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022.51 Describe the provisions to detect possible leakage outside containment (6.2.4) from lines in engineered safety features (ESF)' or ESF-related systems, or in systems needed for safe shutdown of the plant, that contain remote-monual valves in accordance with SRP Section 6.2.4 II.3.b, ll.3.c, and.II.ll.

022.52 Add the following automatic containment isolation volves to FSAR Chapter (6.2.4) 16 Table 3.6-1 (p. 3/4.6-15), or justify their exclusion.

CV8105 Charging line Waste disposal line RF026, RF027

' Reactor and containment radwoste drains RE9159A, B; RE9160A, B RE1003, RE9170, RE9157 Reactor and containment radwaste drains SA032, SA033 Service air line 1 065,1A066 Instrument air line 022.53 The chemical and volume control system seal water supply lines to the

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reactor coolant pumps (Penetrations P-33 and 53) are classifed as l

nonessential. Provide the justification for why remote manual isolation instead of automatic isolation is acceptable for these lines.

0.?2.S4 Verify that the normal containment purge system isolation volves (5.2.4)

(IVQODI A, B, and IVQ002A, B) and post-LOCA purge system isolation volve' l

(IVQ003) will be sealed closed (as defined in SRP Section 6.2.4 ll.3.f) during the operational modes of power operation, startup, hot standby, and hot shutdown.

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022.55 Provide information demonstrating how SRP Section 6.2.4 Il.7 will be (6.2.4) met. This criterion concerns how system lines which provide on open path.

from the containment to the environs should be equipped with radiation monitors that are capable of isolating these lines upon a high radiot' ion signol.

022.56 The information in FSAR Table 6.2-58 is not consistent with FSAR Figures (6.2.4) 10.4-1 (Sheet I of 2) and 10.4-2. Revise Table 6.2-58 or provide updated F5AR figures to correct oil inconsistencies. Specifically our review of the containment isolation design of the main and auxiliary feedwater system lines has raised the following concerns:

c.

FSAR Table 6.2-58 does not include Penetrations P-99,100,101, and 102 shown on FSAR Figure 10.4-1 (Sheet I of 2), and some of the volves listed in Table 6.2-58 as nssociated with Penetrations P-79,84,87, and 76 are shown on Figure 10.4-1 (Sheet I of 2) and 10.4-2 as associated with Penetrations P-100,101,102, and 99.

b.

The manual containment isolation volves (IFWOISA, b, C, and D) provided on the chemical feed connections to the steam generator feedwater lines (Penetrations P-100,101,102, and 99 on Figure i

10.4-1, Sheet I of 2, and Penetrations P-79, 84, 87, and 76 in Table 6.2-58) are unocceptable. The NRC position is that manual volves in these nonessential lines must be sealed closed as defined in SRP Section 6.2.4 II.3.f.

Provide information on how this design deficiency will be corrected.

c.

Add to FSAR Table 6.2-58 steam generator feedwater line isolation volves IFWO35A, B, C, and D and IFWO40A, B, C, and D l

shown.on FSAR Figure 10.4-1 (Sheet I of 2) for Penetrations P-l 100,101,102, and 99.

l d.

FSAR Table 6.2-58 indicates that the main feedwater line volves IFWOO9A, B, C, and D (Penetrabons P-79, 84, 87, and 76) are essential and are open post-occident, but the table and FSAR

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O Section 15.1 indicate that a safety injection signal will isolate

-these valves. Justify the essential designation of these volves and explain the apparent inconsistencies in the above information.

Add to the FSAR Table 6.2-58 valves IFWO43A, B, C, and D (see e.

FSAR Figure 10.4-1, Sheet I of 2, Penetrations P-79,84,87, and 76).

022.57 Verify that for the instrument penetrations (P-1,2,3,4, and 5) the lines (6.2.0 inside and outside containment meet all the requirements of closed systems inside and outside containment listed in SRP Section 6.2.4 II.3.e and i1.9, respectively.

022.58 Concerning the capability of the hydrogen recombiner system to provide (6.2.5) odequate hydrogen control in containment following a LOCA:

J a.

Provide the analysis results showing the containment hydrogen concentration as a function of time following the worst cose LOCA with operation of one and two hydrogen recombiners and ~

with operation of the post-LOCA purge system.

Provide all analysis assumptions including when operation of the hydrogen recombiners or post-LOCA purge system is initiated, i.e., upon reaching a designated hydrogen concentration in containment or offer a stated period of time following the LOCA.

b.

Describe the procedures and all required actions (e.g., installation and connection cf. equipment, construction ~of temporary shielding, leak testing, etc.) necesscry to place into operation following a

LOCA the portable skid-mounted hydrogen recombiners with separate remote portable skid-mounted control and power panels.

Estimate the maximum length of time to accomplish the above.

Include in the time estimate the ~ time immediately following o LOCA when access to the Auxiliary Building may be prohibited due to high radiation (see otso Question No. 6.2.5-2).

Compare this estimated time with the time when hydrogen recombiner operation is required (see Part (a) above).

o c.

It is the NRC staff position that within a time period equal to or less than one-half the time before the. hydrogen recombiner system is required to operate (see Por[(a) above), two hydrogen

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recombiner packages must be available for containment hydrogen control to satisfy the single failure criterion (i.e., in the~ event of failure of one hydrogen recombiner package on independent and -

redundant recombiner package must be immediately available for hydrogen control). Based on our review of the FSAR it is unclear whether this requirement has been met.

Provide additional information on the design, installation, and procedural provisions that demonstrate compliance with this staff position.

If the hydrogen recombiner package from Unit 2 will be required to meet this requirement for Unit I, describe the Unit 2 equipment and design features (e.g., emergency Class IE power supply, auxiliary building HVAC,. shielding, instrumentation and controls) that must be operational or completed for Unit I uperation while Unit 2 is still under construction.

d.

The Atomics International hydrogen recombiner systems used at Byron /Braidwood have maximum operating limits of 150 F and 10 evoluotion, which demonstrates that the psig.

Provide on containment temperature and pressure will be reduced below 150 F and 10 psig prior to the time before the hydrogen recombiners are required 'to operate (see Port (c) above for when hydrogen recombiner operation is required and Question No.

6.2.5-13 for concerns on which LOCA should be considered as the worst case).

022.59 Provide information demonstrating that adequate :hielding provisions are (6.2.5) provided to allow personnel access to activate, maintain, and operate the hydrogen recombiner system, the hydrogen monitoring system, and the post-LOCA purge system following a LOCA.. (Note:

Reference the response to NUREG-0737 Item II.B.2, " Plant Shielding Review".)

022.60 Provide additional information concerning the hydrogen recombiner (6.2.5) subsystem supplying 3,000 cfm of air to cool the reacted gases possing from

e the reaction chamber of the hydrogen recombiner. State whether the cooling fan is on integral part of the hydrogen recombiner package. 'If the cooling cir supply and exhaust points are " local" verify that this added heat load has been included in the Auxiliary Building HVAC system design

basis, if the cooling air is drown from and exhausted to the out de environs, provide the following:

a.

State whether the cooling air ducts, filters, and louvers that are not part of the hydrogen recombiner package are designed to Seismic Category I and Quality Group B requirements.

b.

State whether the cooling air inlet and outlet structures are protected against tornados, floods, missiles, etc.

022.61 The design of the hydrogen recombiner system lines to and from (6.2.5) containment do not satisfy the dedicated hydrogen penekotion requirements of NUREG-0737 Item i1.E.4.1. Provide information on how the design will be modified to comply with the NUREG-0737 requirement.

022.62 Besed on our review of the FSAR, the hydrogen monitoring system does not (6.2.5) provide the required capability to adequately monitor combustible gas concentration in containment. ' Provide information demonstrating how the hydrogen monitoring system will meet the following:

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a.

NUREG-0737 Item li.F.I Attachment 6, " Containment Hydrogen Monitor."

Specifically address o!! position requirements and pesints of clarification. (NOTE: Local actuation of'the hydrogen analyzers is not occeptable since octuation within 30 minutes following safety injection connot be assured with local control.)

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l b.

The capability to continuously monitor containment hydrogen concentrction independent of hydrogen recombiner operation and assuming a single failure (i.e., redundancy).

t Regulatory Guide 1.26 (i.e., Quality Group B).

I c.

d.

Regulatory Guide 1.29 (i.e., Seismic Category I).

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022.63 State the temperature limitations of the hydrogen analyzer equipment and (6.2.5)'

show that these are compatible with the temperature of the containment sample gases fo!!owing a LOCA (Reference SRP Section 6.2.5 l1.5).

022.64 Provide information on the instrumentation that is capable of determining (6.2.5) that the hydrogen recombiner system and the hydrogen monitoring system are properly performing er that a hydrogen recombiner or hydrogen analyzer is malfunctioning.

Specifically indicate where this instrumentation has read-outs and alarms (i.e, control room and/or locally). (Reference SRP Section 6.2.511.11) 022.65 Verify that the hydrogen recombiner syst'em is designed, fabricated, (6.2.5) erected, and tested to Group B quality standards, as recommended in l

Regulatory Guide 1.26 and is designated as Seismic Category 1, as recommended in Regulatory Guide. l.29. Confusion regarding the obeve is due to FSAR Table 3.2-1 (p. 3.2-8) where the hydrogen recombiner quality 6

group is listed as "N/A " and Footnote 6 is ambiguous stating only that

" Quality measures equivalent in intent to those of Quality Group B have been opplied". Also in Tcble 3.2-1 no electricci classification has been given for the hydrogen recombiners, although this should be Class IE.

1 022.66 The FSAR (p. 6.2-46) references Al-72-61, Zion Station FSAR, Appendix (6.2.5) 6B, Amendment 25, January 1973 for &monstration test results on the hydrogen recombiners to be used at Byron /Brcidwood.

Reference and summarize the results of all subsequent testing on the Al hydrogen recombiners since the date of this reference.

022.67 Verify that the assumption used in the hydrogen generation analysis of 1.5%

(6.2.5) zirconium-water reaction, which is five times the maximum calculated reaction under 10CFR 50.46 (i.e.,5 times 0.3%), is greater than the amount of hydrogen evolved from a core-wide average depth of reaction into the original cladding of 0.00023 inches (BTP CSB 6-2, Table 1).

022.68 FSAR Section 6.2.5.3.1.2 states that the maximum allowable coolant (6.2.5) hydrogen concentration is 35 cm3 (STP)/kg of coolant.

Modify the Technical Specifications (Section 16.3.4.7) to include this limiting condition of operation.

(

022.69 FSAR Table 6.2-62 which presents the containment post-LOCA time-(0*

temperature history used in the hydrogen generation analysis cppears to be based on the double-ended pump suction LOCA with maximum safety injection (FSAR Figure 6.2-7). However, the time-temperature history for other loss-of-coolant occidents presented in FSAR Section 6.2.1 show that containment temperatures remain higher following on occident over longer periods of time than in the double-end pump suction with maximum safety injection LOCA.

Since aluminum and zine corrosion rates are highly dependent on temperature, justify the use of the time-temperature history presented in Table 6.2-62 for the hydrogen generation analysis.

022.70 Provide information demonstrating the c'opability. of the input and (6.2.5) discharge lines to and from the hydrogen recombiners to withstand dynamic effects, such as transient differential pressures, that would occur early in the blowdown phase of a loss-of-coolant accident (Reference SRP Section 6.2.511.4).

022.71 The following volves are listed in FSAR Table 6.2-58 as failing to the "as (6.2.4) is" position upon loss of power. Justify why the volves fail "as is" rather than foil closed (reference SRP Section 6.2.4 !!.5).

Penetration Valve System 28 ICV 8100 RCP Seal Water Return Line 28 ICV 8112 RCP Seal Water Return Line 71 ICV 8105 Chcrging Line 5, 8 IWOO20A, B Chilled Water Lines 6,10 IWOOO6 A, B Chilled Water Lines 21 ICC9414 Component Cooling Water Line 21 ICC9416 Component Coofing Water Line ICC685 Component Cooling Water Line 24 24 ICC9428 Component Cooling Water Line 25 ICC9413B Component Cooiing Water Line'

.-sea.-

022.72 Concerning the containdient isolation design of the hydrogen recombiner (6.2.0 lines to and from containment:

a.

Verify that the following containment isolation volves have positive position indication in the control room and are remote manually operable from the control room in accordance with SRP Section 6.2.4 !!.S.c and ANSI N271-1976, Perographs 4.2.2 and 4.2.3:

00GOS9 00G063 00G061' 00G064 00G062 00G065 b

Describe the is'olction provisions for the hidrogen recombiner disefiorge lines (00Q4SB_3 and 00G43B,3). Although the normally open volves (00G060 and 00G066) in these lines are supplie.d with power from emergency buses, they must receive on outomatic containment isolation signal,ge remote manually operable from the control. room, and have positive position indication in the.

control room to be acceptchle cs containment isolation bcrri~ers. '

022.73 FSAR Table 3.2-1 indicates that " quality measures equivalent in intent to *

(6

• I those in Quality Group C will be applied" to the reactor containment fan coolers. It is our position that the ' reactor containment fan coolers must be designed, fabricated, erected, and tested to Quality Group B standards, as recommended by Regulatory Guide 1.26 (SRP Section 6. 2. 2 11. 6 ).

Provide information on how you will comply with this position.

6 e

ab m__

022.74 Section III.D of Appendix J to 10 CFR Part 50 requires that, after (6.2.6) the preoperational Type A tests (containment integrated leak rate tests), a set of three Type A tests shall be performed at approxi-mately equal intervals during each 10-year service period.

FSAR section 6.2.6.1 indicates conformance with this requirement, but section 6.2.6.1.7 specifies a different, nonconforming schedule.

Revise the FSAR (ar.d your Type -A test schedule, if necessary) to conform to the Appendix J requirement. Also, FSAR section 6.2.6.1 states that the periodic Type A tests wil'1 be conducted at a pres-sure of 25 psig (reduced pressure), while other sections suggest that the test pressure will be Pa l(peakpressure).

Correct this apparent discrepancy and state clearly which pressure will be used.

022.75 Section III.A.3 of Appendix J to 10 CFR Part 50 requires that all (6.2.6)

Type A tests shall be conducted in accordance with the provisions of the American National Standard (ANSI) N45.4-1972, Leakage Rate Testing of Containment Structures for Nuclear Reactors, March 16, 1972.

Paragraph 7.6 of the standard requires that the Type A test duration shall extend to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..The FSAR describes a procedure that would a' low a test duration as short as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Revise your

~

test procedures to meet the duration requirement of ANSI N45.4-1972.

022Property "ANSI code" (as page type) with input value "ANSI N45.4-1972.</br></br>022" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..76 FSAR Table 6.2-66 states that La (maximum allowable leakage rate (6.2.6) for Type A test at pressure Pais 0.267% per 24 hours. Section 6.2.6.1 states that La equals 0.16% per day. Provide the correct value and revise the FSAR accordingly. ~4

4 y. 022.77 Specify whic'h fluid systems, or portions of fluid systems, are.to (6.2.6) be vented and drained during Type A tests. Identify and justify those systems not vented and drained, but which may be open to the containment atmosphere under post-accident conditions and become an extension of the boundary of the containment. For vented and drained systems, illustrate, by reference to figures, diagrams, and drawings, that the isolatio' valves are exposed to the contain-n ment test air pressure on one side and atmospheric pressure on the other side; that is, the systems are vented and drained both up-stream and downstream of the containment isolation valves. The hydrogen recombiner systems ' located outside containment should be included as vented systems. The recombiner systems should be open to the containment atmosphere during the' performance ' f the Type A tests. Alternately, a local leak test of the systems o may be done at the time of the Type A test and the measured leak rate added to the Type A result. Verify that this will be done. 022.78 a. Provide the test schedules for all Type B and Type C tests. (6.2.6) Provide the-test pressures for these test.s; Pa is an acceptable test pressure. Specify the test media (for example, air or nitrogen) and test durations for each of these tests. b. Describe in detail the leakage testing program for the two personnel air locks (access hatches) and show that the program complies with all of the requirements of section III.D.2(b) of 19 _

~ Appendix J to 10 CFR Part 50. Include test pressure and dura-tion for leak tests performed by pressurizing between-the door seals. 022.79 Describe, in detail, 'with text and figures, the permanently instal-(6.2.6) led leakage surveillance system used for continuous pressurization between the closure flanges of electrical penetrations. Provide the system pressure and state whether or not the system will be em-ployed at all times during normal plant operations. State which electrical penetrations, if any, are not serviced by this. system; provide the testing provision's for these penetrations. State. whether the use of this system is to be in lieu of conventional Type B testing provisions. Describe tne way in which leakage nessured with this system is to be added to the total of Type B and Type C leakage. Describe the high leakage alarm and its redundancy and single failure characteristics. 022.80 Many containment isolation valves are shown on Table 6.2-58 as not (6.2.6) receiving Type C tests. For each such valve, justify this lack of testing and show that this is in conformance with the requirements of Appendix J to 10 CFR Part 50. 022.81 Identify any instrumentation lines that will be isolated during the (6.2.6) Type A test. If instrumentation lines are isolated, they should be locally (Type C) tested and the measured leakage added to the Type A j result. Discuss your plans for complying with this position.

^ ~ 3 Discuss che administrative controls to' assure that instrument & tion lines isolated during the Type A test are restored to their oper-able status following the test. 022.82 The rationale given in FSxR section 6.2.6.3 to justify Type C (6.2.6) testing.of certain isolation valves with the test pressure applied in a direction opposite to that wh.ich would occur under accident ~ conditions,is unacceptable. Section III.C.1 of Appendix J to 10 CFR Part 50 requires, for such " reverse-direction" testing, that it be shown that the results from such tests provide equivalent or more conservative results, when compared to testing in the "fotward" direction. Therefore, justify the acceptability of each reverse; direction test. 022.83 The test, vent, and drain (TVD) connections that are used to facili-l (6.2.6) tate local (Type B and C) leak rate testing and the performance of the containment integrated leak rate (Type A) test si.Juld be under administrative control, and should be subject to periodic surveillance to assure their integrity and vrvify the effectiveness of administra-tive controls. Describe the administrative controls and periodic surveillance to be applied to TVD connections. Item B.4 of Branch Technical Position CSB 6-4, " Containment Purging 022.84 (6.2.6) During Normal plant Operations," specifies that provisions should be made for leakage rate testing of the purge /ver,t system isolation-

valves during reactor operation. As a result of numerous reports on unsatisfactory performance of the resilient seats for the iso-lation valves in containment purge red vent lines, the following position has been developed as a basis #or a leak rate testing program for purge / vent containment isolatio,, valves: 1. Require that plants provide the capability to leak test these valves during plant operation in Modes 1 thrco.pl 4; 2. Require that a leakage int'egrity test be performed on the con-tainment' isolation valves with resilient material seals in (a) active purge / vent sys.tems (i.e., those which may be operated during plant operating Modes 1 through 4) at least once every three months; and (b) passive purge systems (i.e., those which must be administratively controlled closed during reactor oper-ating Modes 1 through 4) at least once every six months; and 3. Require that the maximum allowable leakage rate for purge / vent system containment isolation valves with resilient material seals - be determined on a case-by-case basis. This determination should give appropriate consideration t.o valve size, maximum allowable leakage rate of the containment (La as defincd in Appendix J to 10 CFR Part 50) and, where appropriate, the maximum allowable by-pass leakage fraction for dual containment. We also recommend that in the development of technical specifications for those plants which have passive purge systems that the inlet and outlet lines be tested at intervals no greater than six months on a ____ _

e staggered test basis (as an example, inlet line isolation valvas would be tested in the first and third quarters of the year and the outlet line isolation valves in the second and fourth quarters). This should improve the probability of detecting large butterfly valve leakage due to seasonal weather variations. Provide the details of a surveillance program that will conform to the position stated above. Om f e o e S 8 9 e 23 - _}}