ML19261D936
| ML19261D936 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 05/15/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Reed C COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 7906290156 | |
| Download: ML19261D936 (77) | |
Text
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UNITEO STATES NUCLEAR REGULATORY COMMisslON y'r
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MAY 151979 Docket Nos.:
STN 50-454/455 STN 50-456/457 Mr. Cordell Reed Assistant Vice President Conmonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690
Dear Mr. Reed:
SUBJECT:
FIRST ROUND QUESTIONS ON THE BYRON AND BRAIDWOOD 0L APPLICATION In our review of your application for operating licenses for the Byron Station, Units 1 and 2, and the Braidwood Station, Units 1 and 2, we have identified a need for additional information which we require to complete our rev'iew. The specific requests contained in the enclosure to this letter are the first set of our round one questions and cover those areas of our review performed by the following:
(1) Auxiliary SystemsBranch,(2)PowerSystemsBranch(electrical),(3) Mechanical Engineering Branch, (4) Structural Engineering Branch, (5) Effluent Treatment Systems Branch, (6) Radiological Assessment Branch, (7) Hy-drology-Meteorology Branch (hydrology) and (8) Quality Assurance Branch.
In order to maintain our present schedule as stated in our letter of February 22, 1979, we need a completely adequate response to all questions in the enclosure by July 21, 1979.
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Please contact us if you desire any discussion or clarification of the enclosed requests.
Sincerely,0
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3 D
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/ Steven A.,Varga, Ch f
Light Water Reactors ranch No. 4 Division of Project Management
Enclosure:
Request for Additional Information cc: See next page h
39062005 2157 161
r I
Comonwealth Edison Company cts:
Mr. William Kortier Atomic Power Distribution Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. John W. Rowe Isham, Lincoln & Beale One First National Plaza 3
42nd Floor
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Chicago, Illinois 60690 Mrs. Phillip B. Johnson 1907 Stratford Lane Rockford, Illinois 61107 Ms. Marilyn J. Shineflug 1816 Judy Lane DeKalb, Illinois 60115 Ms. Beth L. Galbreath 3
734 Parkview l
Rockford, Illinois 61107 I
Ms. Bridget Little Rorem f
Braidwood Area Co-ordinator Bailly Alliance-Illinois Braidwood, Illinoir. 60408 C. Allen Bock, Esq.
i P. O. Box 342 Urbana, Illinois 61801 Thomas J. Gordon, Esq.
WAAfEP., CVANS & GORDON 2503 S. Neil Champaign, Illinois 61820 2157 162
.m ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION BYRON STATION UNITS 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 BRAIDWOOD STATION UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457 These requests for additional information are numbered such that the three digits to the left of the decimal identify the technical review branch and the numbers to the right of the decimal are the sequential request numbers. The number in parenthesis indicates the relevant section in the Safety Analysis Report. The initials RSP indicate the request represents a regulatory staff position.
Branch Technical Positions referenced in these requests can be found in " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-75/087 dated September 1975.
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005-1 005.0 AUXILIARY SYSTEMS BRANCH - CODES AND STANDARDS l
005.1 In addition to Code Case 1528 identified in Section 5.2.1.2 of the FSAR, I
l identify all other ASME Code Cases (including those that are listed as j
acceptable in Regulatory Guides 1.84 and 1.85) that were used in the i
construction of each Quality Group A components within the reactor coolant I
pressure boundary. These code cases should be identified by code case l
number, revision, and title, for each components to which the code case 1
has been applied.
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005.2 Verify that all components within the reactor coolant pressure boundary as defined in 10 CiR Part 50.2(V) are classified Quality Group A and constructed to Section III, Class 1 of the ASME Boiler and Pressure Vessel Code in compliance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, or as a minimum, are classified Quality Grouc l
B and constructed to Section III, Class 2, of the code 'f the components l
meet the exclusion requirements of the rule.
l 005.3 As noted in Section 5.2.1.1 of the FSAR, the control valves for the Byron Station Units 1 and 2 are not in confomance with 10 CFR Part 50, Section 50.55a, Codes and Standards. These components are constructed l
to Section III, Class 1, of the ASME Boiler and Pressure Vessel Code, l
1971 Edition, through the Summer 1972 Addenda, whereas, the regulation requires the components to be constructed to the same code and edition through the Winter 1972 Addenda to the code.
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a 005-2 In order to assess the acceptability of these control valves, identify those portions in the 1972 Winter Addenda to the code with which the control valves are not in compliance.
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010-9 f
010.0 AUXILIARY SYSTEMS BRANCH l
010.15 You indicate that you have not nostulated potential internally generated (3.5.1.1) missile sources such as valve bonnets, instrument wells, and pump impellers outside of containment.
It is our position that you perfonn an analysis and evaluation for all areas of the plant housing safety rdated equipment assuming potentia' internal missiles generated from the above mentioned sources. This evaluation should verify that no damage to saf.ty 'related equipment will result which would prevent a
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010.16 You have not responded to question 010.2 concerning our request for a (3.5) j tabulation of all safety-related components located outdoors and their method of protection against tornado missiles. Provide a response.
I 010.17 Your response to question 010.3 concoming high and moderate energy (3.6) piping system failures is not complete. Provide results of analyses of the effects on safety related systems for all high and moderate energy piping system failures in accordance with our Branch Technical Position
{
ASB 3-1.
Include a table which identifies the method of protection pro-vided for a postulated break in each high energy system (separation, barriers, restraints) and postulated crack in each moderate energy sys-tem if sted in Table 3.6-2.
As an example, provide the results of the analysis that demonstrates how the residual heat removal system will be protected from the effects i
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010-10 of postulated piping failures.
Include piping layout drawings for the areas housing the residual heat removal system. Also, include the assump-tions used in your analysis such as flow rates through postulated cracks, pump room areas, s' ump capacities, and floor drainage systent capacities.
010.18 In Section 3.6 you do not provide adequate information to show protec-(3.6) tion of safety related equipment due to a failure of non-seismic Category I piping systems. A design bases earthquake may result in a failure of any non-seismic Category I piping system regardless of postulated break and crack locations as defined in BTP MEB 3-1.
Demonstrate that all safety related equipment will be protected from flooding, pipe whip, jet impingement, and environmental effects due to such an event in accordance with Regulatory Guide 1.29, Position C.2.
Place particular emphasis on flooding protection against a complete severance of non-seismic moderate energy systems. -
010.19 You state that the spent fuel storage pool will be lined on the inside (9.1.2) surface with a stainless steel liner plate and that leak tightness will be assured by means of a leak chase system. You do not, however, state that the plate will be designed to seismic Category I requirements. Show that a failure of the liner plate as a result of an SSE will not result in any of the following: significant release of radioactive material due to mechanical danage to the spent fuel; significant loss of water from the pool which could uncover the fuel and lead to release of radicactivity 2157 107
010-11 due to heat-up; loss of ability to cool the fuel due to flow blockage caused by a portion or one complete section of the liner plate falling on top of the fuel racks.
010.20 You have not provided' a response to question 010.6 which requests a (9.1.4)
(RSP) description of the containment polar crane. Previde a response. It is our position that you perform a load drop analysis for the polar crane and provide us with the details of this analysis and the resulting conclusions reached. This analysis should include all loads handled by this crane and an evaluation of the effects of cropping of these loads anywhere along the crane's path of travel where unacceotable damage to reactor coolant system components or fuel could occur.
If it cannot be demonstrated that adverse affects to primary system components or fuel will not occur as a result of dropping loads from the containment polar crane, then the crane 4must be designed in accordance with the guidelines of Branch Technical Position ASB 9-1 or Regulatory Guide 1.104.
010.21 Your response to question 010.8 concernin] the routing of the essential (9.2.1)
(RSP) service water system piping is unacceptable.
In the FSAR you indicate that the essential service water lines are routed through the turbine building basement. Provide piping arrangement drawings (plan and ele-vations) for the essential service water supply and return lines from the ultimate heat sink to the essential service water pcmps.
It is our position th.It the essential service water piping not be routed through 2157 108
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010-12 areas where a seismic event could prevent the system from performing its safety function. Provide further details to verify that your design will meet this position.
010.22 You indicate that a single essentiai service water pump is in operation (9.2.1) during normal operation. On occurrence of a LOCA or loss of offsite power, valves are automatically opened to pennit essential service water supply to those components reouired for shutdown but that are not nomally operating. However, the redundant essential service water pump is not started.
It is not clear that adequate capacity can be supplied to both essential loops by one pump. Verify that adeauate cooling water supply is provided to essential components considering a single failure.
Also, verify that the' redundant essential service water pump is auto--
matically started immediately upon failure of the other pump.
010.23 Your component cooling water system malfunction analysis indicates that (9.2.2) three of the five CCW pumps are required during emergency operation.
However, in Section 9.2.2.3 you indicate that four of five pumps are required during simultaneous LOCA in one unit and safe shutdown of the o ther. Explain this apparent discrepancy.
010.24 Expand Table 9.2-5, Component Cooling System Malfunction Analysis to (9.2.2) include a single failure analysis of valves in all Unit 1 to Unit 2 common supply and return headers. Also provide an analysis for passive failures (pipe ruptures) in these headers between isolation valves 2157 169 1
010-13 during long term ccoling. Any operator action required as a result of these failures cannot be assumed to occur for 30 minutes.
010.25 In Figure 9.2-2, Sheet 10 of 10 you show a locked open valve on the blow-(9.2.1)
(Byron down connection to each essential service water return header. The only) piping downstream of.this valve is non-seismic.
In the event _of an earthquake, this piping could fail creating an open flowpath resulting in loss of essential service water through the open valve. Verify that the portion of the essential service water ficw lost in such an event will not prevent the system from providing its safety function or modify 5 our design accordingly.
010.26 You indicate that the safety related essential service water maket? pumps (9.2.5)
(Byron in the river screen house may be lost during the probable maximum flood.
only)
Non-safety grade wells are provided as backup to these pumps for makeup to the essential service water cooling towers. Verify that these wells can supply adequate makeup water in the event of loss of offsite oower and assuming a single failure.
010.27 In order to permit an assessment of the Ultimate Heat Sink, provide the (9.2.5)
(Braidwood results of an analysis of the thirty-day period following a design basis only) accident in one unit and a normal shutdown and cooldown in the remaining unit, that determines the total heat rejected, the sensible heat rejected, the station auxiliary system heat rejected, and the decay heat release from the reactors.
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010-14 In submitting the results of the analysis requested, include the following infomation on a day by day basis in both tabular and graphical presenta-
'fons:
t 1.
The total integrated decay heat'.
2.
The heat rejection rate and integrated heat rejected by the station auxiliary systems, including all operating pumps, ventilation equip-ment, diesels, spent fuel pool makeup, and other heat sources for both units.
3.
The heat rejection rate and integrated heat rejected due to the sensible heat removed from containment and the primary ystem.
4.
The total integrated heat rejected due to the above.
5.
The maximum allowable inlet water temperature taking into account the rate at which the heat energy must be removed, cooling water flow rate, and the capabilities of the respective heat exchangers.
6.
The required and av,ailable NPSH to the essential service water pumps at the minimum Ultimate Heat Sink water level.
The above analysis, including pertinent backup infomation, is to demon-strate the capability to provide adequate water inventory and provide sufficient heat dissipation to limit essential cooling water operating temperatures within the design ranges of system components.
Use the methods set forth in Branch Technical Position ASB 9-2, "Resi-dual Decay Energy for Light Water Reactors for Long Term Cooling," to establish the input due to fission product decay. Assume an initial cooling water temperature based on the most adverse conditions for nomal coeration.
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010-15 010.28 You indicate that the non-safety related condensate storage tank is the (9.2.6)
(10.4.9) preferred source of auxiliary feedwater supply. You also indicate that the essential service water system supplies the auxiliary feedwater sys-tem in a seismic event.
It is our position that switchover from the condensate storage tank to the essential service water system be auto-matic in the event of loss of offsite power or LOCA. Verify that you will meet this position.
010.29 You indicate that the instrument air system is non-safety related with (9.3.1) the exception of the containment penetrations. Verify that all safety related system valves will fail in their safe position on loss of instru-ment air.
i 010.30 Your response to question 010.11 concerning the design of your equipment (9.3.3) and floor drainage system is not acceptable.
a.
You have not adequately described the equipment and floor drainage system for safety related plant areas, nor provided an analysis that demonstrates compartment and/or area drains serving safety related components or rystems have been sized for maximum flow conditions as a result of pipe breaks or cracks in these areas. Provide this informa tion.
b.
You indicate that certain safety related pump areas 'are protected by leak detection sumps which drain to the non-safety related auxiliary building floor drain sumps and annunciate alarms in the control room when full.
It is our position that you provide the following for all areas housir] both safety related equipment trains:
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010-16 1.
Leak detection sumps shall be equipped with redundant safety grade alams which annunciate in the control room. Verify that if operator action is required on receipt of the alarm that flooding of redundant safety grade equipment will not occur within 30 minutes. oz 2.
Provide separate water tight rooms and independent drainage paths with leak detection sumps for each redundant safety related component.
010.31 Verify that failure of the essential battery room exhaust fans is annun-(9.4.5.3) ciated in the control room.
010.32 You indicate that the river screen house ventilation system is non-safety (9.4.6)
(Byron related. H;nver, this building houses the safety related diesel only) engine driven essential service water makeup pumps. Verify that adequate air is available to these engines and that the oroper operating environ-ment can be maintained (for the batteries, etc.) assuming loss of this system under emergency conditions.
010.33 Your response to Question 010.12 concerning the potential for flooding (10.4.5) as a result of a failure of a circulating water system expansion joint is unacceptable.
In the FSAR you do not evaluate the effects of an expansion joint failure at the condenser. Expand the information pro-vided to include an evaluation regarding the effects of possible circu-lat:ng water system failure inside the turbine building.
Include the following:
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010-17 (1) The maximum flow rate through a completely failed expansion joint.
(2) The potantial for and the means provided to detect a failure in the circulating water transport system barrier such as the rubber exoan-sion joints.
Include the design and operating pressures of the various portions of the transport system barri.r and their relation to the pressures which could exist during malfunctions and failures in the system (rapid valve closure).
(3) The time required to stop the circulating water flow (time zero being the instant of failure) including all inherent delays such as operator reaction time, drop out tipas of the control circuitry and coastdown time.
(4) For each postulated failure in the circulating water transport system barrier give the rate of rise of water in the associated spaces and total height of the water when the circulating water flow has been stopped or overflows to site grade.
(5) For each flooded space provide a discussion, with the aid of drawings if necessary, of the protective barrier provided for all essential systems that could become affected as a result of flooding.
Include a discussion of the consideration given to passageways, pipe chases and/or the cableways joining the flooded space to the spaces containing safety related system components.
Discuss the effect of the flood water on all submerged essential electrical systems and components.
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010-18 010.34 Your response tc Question 010.14 concerning our request for a reliability (10.4.9) analysis for the auxiliary feedwater pump diesel driver is not acceptable.
As indicated in NUREG-75/023 Supplement 1 dated August,1975, we require a reliability analysis which demonstrates that the diesel engine driven auxiliary feedwater pump is at least as reliable as the onsite emergency diesel generators. This analysis should include comparisons of test data between the auxiliary feedwater pump driver and emergency diesel engine.
010.35 In Figure 10.4-2 (Auxiliary Feedwater System) you show a recirculation (10.4.9) line with a locked open valve from each auxiliary feedwater pump discharge to the non-safety related condensate storage tank. The piping downstream of the locked open valve. is also non-seismic.
In the event of loss of this line and/or tank, an open flowpath will be created resulting in loss of auxiliary feedwater. Verify that the portion of the auxiliary feedwater flow lost in such an event will not prevent the system from performing its safety function or modify your design accordingly.
010.36 In Figure 10.4-2 you show motor operated valves in the essential service (10.4.9) water supply line to and the discharge line from the diesel driven auxiliary feedwater pump. Discuss how you will meet the power diversity requirements of our Branch Technical Position ASB 10-1 for assuring auxiliary feedwater supply with this arrangement.
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040-22 040.0 POWER SYSTEMS BRANCH - ELECTRICAL AREA 040. 36 It is the staff position that motors outside as well as motors in-(3.11) side containment comply with the qualification testing requirements of Criterion III of Appendix B to 10 CFR Part 50. The procedures for conducting the qualification tests specified by IEEE standard 334-1974 describe an acceptable method of complying with this staff position.
Identify each continuous duty Class lE motor located outside as well l
as inside containment for NSSS supplied and for B0P supplied motors.
l For each representative type Class 1E motor provide detailed design and the qualification program infonnation.
040.37 Containment Fan Cooler Motor (3.11)
Section 6.2.2.2.1 of the FSAR indicates 1) that the motor for the con-tainment fan cooler is cooled by direct containment air, 2) that the motor insulation hot-spot temperature will not exceed 150 C under accident conditions, and 3) that the motor / fan assembly has been tested under simulated loss-of-coolant accident environment and in accordance with the requirements of IEEE Standards 323-1974 and 334-1971.
Provide the qualification test information that demonstrates thz.t the motor /
fan assembly aged to end of life is capable of performing its design function during design basis event conditions including post accident conditions, with motor insulation hot-spot temperatures of 150 C.
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040-23 040.38 Section 3.11.2 of the FSAR indicates that balance of plant Class lE (3.11)
(8.0) equipment will be in accordance with the requirements of IEEE Standard 323-1971.
It is the staff position that bc.th NSSS and BOP equipment meet IEEE Standard 323-1974.
Revise section 3.11.2 of the FSAR to indicate compliance with IEEE Standard 323-1974.
040.39 Identify each Class lE equipment that is not served by redundant (3.11)
Class lE environmental support systems. Define the limiting environ-mental conditions that are expected to occur, assuming a loss of the environmental support system.
040.40 The utility grid for Byron /Braidwood has not been described in Sec-(8.1) tion 8.1 of the FSAR. Provide a c'escription of the grid for both Byron and Braidwood stations. The description must clearly define the interconnections between the nuclear unit, the utility grid, and other grids. The descriptions s' mld state whether facilities are existing or planned; if planned, the respective completion dates should be provided.
(SRP Section 8.1, Part II, Item 1).
040.41 Section 8.1.6 of the FSAR implies that the Byron /Braidwood design has (8.1) two immediate access circuits from the transmission network. However, Section 8.3.1.1.1 of the FSAR indicates one immediate and one delayed access circuit from the transmission network.
Correct the discre-pancy between sections 8.1.6 and 8.3.1.1.1 of the FSAR.
Provide the results of an analysis to show that the delayed access circuit can be made available in sufficient time to assure that specified acceptable limits described by GDC 17 are not exceeded.
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040-24 040.42 The provisions for testing the DC power sources, relays, and switches (8.1) associated with the preferred power source protection system and the provisions for testing the onsite Class lE AC and DC power systems should satisfy the recomendations of Regulatory Guide 1.118 (SRP section 8.2, Part III, item 3).
Identify and justify each exception taken to Regulatory Guide 1.118.
040./3 To assure that the requirements of GDC 1 are met 1) in the preferred (8.1) offsite power system between the switchyard and the Class lE distribution system and 2) in the onsite Class lE AC and DC power systems, the quality assurance program must satisfy the requirements of IEEE Standard 336 as augmented by Regulatory Guide 1.30 (SRP Section 8.2 Part II item 2a, Section 8.3.1 Part II item 9b, and Section 8.3.2 Part II item 7b).
Provide a statement in Section 8.1.5 of the FSAR that the Byron /Braidwood design conforms with IEEE Standard 336-1971 and Regulatory Guide 1.30 for bcith the preferred offsite and Class lE onsite power systems, or identify and justify each exception taken to this standard and guide.
If exception is taken regarding the preferred offsite power system, provide a dercription of alternate quality assurance program measures for the installation, inspection and testing control for the offsite power system comensurate with its importance to safety. The descrip-tion should apply to all offsite power equipment in the switchyard, the plant, and the interccnnecting circuitry; and it should specifically address the two offsite power circuits required by GDC-17.
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040-25 040.u With regard to thE surveillance of the offsite preferred and onsite (8.1)
Class 1E AC and DC power system operability status, an acceptable design must satisfy the positions of Regulatory Guide 1.47, as aug-mented by Branch Technical Position (BTP) ICSB 21 (SRP Section 8.2, Part II, Item 2c).
Describe ar.d justify each exception taken to Regulator-Guide 1.47 and BTP ICSB 21.
040.45 Preoperational and initial startup test programs for the offsite pre-(8.1) ferred and onsit'e Class 1E AC and DC power systems should be in accordance with Regulatory Guide 1.68 as augmented by Regulatory Guide 1.41 (SRP Section 8.2, Part II, Item 2b). Describe and justify each exception taken to Regulatory Guides 1.68 and 1.41.
040.46 The type of electric power (ac or de) required by each safety load (8.1) i has not been identified in sec. tion 8.1 of the FSAR. Provide this information.
(SRP Section 0.1, Part III).
040.47 The interrelationship between the onsite and offsite power systems (8.l) has not been clearly defined in the FSAR. Provide, in Section 8.1 of the FSAR, an electrical one-line diagram that clearly describes the offsite power system switchyard, the onsite Class lE and non-Class lE power distribution system, and the interrelationship between the onsite and offsite systems.
(SRP Section 8.1, Part III).
Provide in Section 8.1 of the FSAR a positive statement with regard 040.48 (O' }
to conformance of the Byron /Braidwood design to each of the criteria listed in Table 8.1 of the Standard Review Plan that are applicable to Section 8.2, 8.3.1 and 8.3.2 of the FSAR.
Identify and justify each exception taken to the criteria listed in Tabie 8.1 of the Standard Review Plan, j]9
040-26 040.49 The electrical schematics and scaled physical arrangemer.t and layout (8.2) drawings for the offsite power systems have not been presented in the applicant's FSAR. Provide the subject drawings. The drawings should include locatior. of rights-of-way, transmission lines and towers, transformers, switchyard interconnection (breaker and bus arrangement),
switchyard control systems and power supplies, location of switchgear (in plant), interconnection between switchgear, cable routings, main generator disconnect, and the disconnect control system and power supply (SRP Section 8.2, Part 1. Items 1 and 2).
040.50 In regard to the separation and independence between the two preferred (8.2) power circuits from the 345 KV transmission terminal buses to each unit's Class lE distribution system, describe the protection provided these preferred power circuits that assures that no single ever.t (such as a fire or raceway failure) can simultaneously affect both circuits in such a way that neither can be returned to service in time to prevent fuel design limits or design conditions of the reactor coolant pressure boundary from being exceedeo s3RP 8.2, Part III, Items 2b, 7, and 8).
040.51 Provide the results of our analysis of the switchyard breaker control (8.2) system, power supply and breaker arrangement for the possibility of simultaneous failure of both preferred power circuits from single events such as a breaker not operating during fault conditions, loss of a control circuit power supply, etc.
(SRP Sections 8.2, Part III, items 2c and 2d).
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040-27 040.52 The reactor protection system must be designed to prevent any load (8.2) dispatch system actior, that would interfere with safety actions dur-ing periods when safety actions are required.
Provide the results of analyses of this sectior, which assure that no failure mode of the load dispatch system will cause an incident at the generating station or interfere with any protective action required.
(SRP Section 8.2, Part III, Item 11).
040.53 Describe the power grid frequency decay rate expected at Byron /Braidwood (8.2'. 2 )
stations as a result of disturbances occurring anywhere in the grid sys-tem. Describe the type of disturbintc considered in your study and define the maximum expected frequency decay rate at the Byron /Braidwood station.
(SRP 8.2, Part III, Iterr 2.f).
040.54 Provide the nominal value and the maximum and minimum limits of (8.2) voltage and frequency available to the station and to each safety related bus from the offsite power grid system.
Describe the procedures used to maintain the offsite power grid system within the maximum and minimum limits identified above.
(1) Describe how the nuclear plant operator is made aware of the maximum and minimum limits on the offsite power grid system.
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(2) Describe the effects on the offsite power grid system stability when the grid is operating at the maximum or minimum limits and when the grid system experiences a disturbar.ce such as,
- 1) loss of the largest single supply to the grid, 2) removal of the largest load from the grid, or 3) possibly loss of several large supplies or loads if they use a connon transmission tower, transformer, or a breaker in a remote switchyard or substation.
(3) Describe restriction placed on the offsite power grid system ooeration that requires specific spinning reserve /either real or reactive power) to be available within a dedicated distance from the nuclear station in order to maintain the grid system within the maximum and minimum limits during grid disturbarces.
040.55 On Table 8.3-1 of the FSAR it has been indicated that the following (S.3.1) loads 1) control room refrigeration units, 2) control room HVAC system supply fan, 3) control room HVAC system, and 4) F.H. Bldg.
Charcoal Booster fan are supplied power from only Unit l's power distribution system and will be supplied power through cross ties between Unit 1 and 2 during outage on Unit 1.
Describe how this design meets Position 2 of Revision 1 to Regulatory Guide 1.81.
Identify and justify each exception taken to this position.
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040-29 040.56 From the infcrmation in Section 8.1.2 and Table 8.3-1 of the FSAR, (8.3.1) it appears that the diesel generator set capacity for standby power supplies does not meet staff positions C2 and C4 of Regulatory Guide 1.9.
a.
Describe the diesel-generator specifications.
Provide a state-ment in the FSAR that the diesel generator design specifications are in conformance with positions C4 of Regulatory Guide 1.9 or justify the non-conformance.
b.
Provide the total KW load on each diesel generator for both loss of offsite power and for loss-of-coolant accident coincident with loss of offsite power, c.
Provide the basis for the hp or KW requirements for each load listed on table 8.3-1 of the FSAR for bott the Byron and Braidwood stations.
d.
The hp and KW loads listed on table 8.3-1 of the FSAR are in-consistent with those listed on drawings 20E-1-4001 and 6E-1-4001.
Justify or correct the inconsistencies.
e.
Drawings similar to 20E-1-4001 and 6E-1-4001 have not been provided for Unit 2 of Byron and for Unit 2 of Braidwood stations. Provide these drawings for the Byron /Braidwood Unit 2.
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040-30 l
f.
Table 8.3-1 does not include power requirements for actor operated valves or other informittent loads.
Provide this informatinn and describe how this type of loading is accounted fcr in sizing the diesel-generators.
I g.
Table 8.3-1 includes an item 18, "other loads". Provide a listing l
of these other loads with similar information to that provided for loads already listed on table 8.3-1.
l l
040.57 Table 8.3-1 does not describe the sequential loading of the diesel generator (8. 3.1 )
l with those loads required for safe shutdown of the reactor given a loss of offsite power and no design basis accident. Provide this description.
040.58 Section 7.1.2.1.10 of the FSAR implies that all Class lE motors are designed (8. 3.1 )
to accelerate their driven loads with 80 percent of motor-rated voltage at the motor terminal.
Voltage at the motor terminals will drop below 80 percent when the diesel generator breaker closes.
Provide justifica-tion for using motor designed to start their loads at a voltage below the minimum of 80 percent of rated voltage.
040.59 Sizing of the system auxiliary transformer - Preferred power to the two (8.3.1)
Class lE divisions for each unit is provided from the switchyard through two normally connected system auxiliary transformers. A second source of preferred power is provided for each unit by manual switching from the switcFyard through the other units auxiliary transformers.
- Thus, the syster auxiliary transformers must be capable of supplying the DBA loads of one division of one unit and the safe shutdown loads of one division 1
iur
040-31 of the othe-unit.
In addition the Byron /Braidwood design allows for automatic transfer of all non-safety loads to the system auxiliary transformer and the relocation of removable links. Thus, the system auxi-liary transformer must also be capable of supplying the DBA loads and all the non-safety loads.
Section 8.3.1.1 of the FSAR indicates that the system auxiliary transformer is only capable of furnishing startup and limited operating loads.
From this information it appears that the system auxiliary transformer does not have sufficient capacity for all modes of plant operation. Pro-vide a description of the maximum loading of the system auxiliary trans-former for all modes of plant operation. Justify the sizing of the sys-tem auxiliary transformer. (MVA capacity and voltage drop considerations) for worst case loading and grid voltage ccnditions.
040.60 Describe each mode of plant operation for which tie brer.kers 1411, 1412, (8. 3.1 )
2421, and 2711 between Class lE and non-Class lE buses are closed.
040.61 In section 8.3.1.2 of the FSAR it has t,een indicated thtt, for the (8. 3.1 )
diesel generator periodic load shedding test, the largest load versus full load will be used. This is an example and exception to the position 2a(4) of Revision 1 to Regulatory Guide 1.108.
The provisions for preoperational and periodic testing of diesel generator units should satisfy the recommendations of Regulatory Guide 1.108 (SRP Section 8.3, Part II, item 4e and 7).
Describe how the Byron /Braidwood design meets the recommendatici of Regulatory Guide 1.108.
Identi fy and justify each exception taken to the Regulatory Guide.
2157 185
040-32
~
040.62 Tabulate, for each Class lE system required to bring the plant to (8. 3.1 )
(9.5.1) a safe cold shutdown, the systems essential power, control, and in-i strumentation cable routing locations.
Identify the required systems essential cable and their location where thE separation from the cable redundant counterpart is less than a three hour rated fire barrier.
040.63 Class 1E cabling that leave one seismic Category I building and enters (8. 3.1 )
another seismic, Category I building must be routed through appropriately designed and classified raceways. Provide a description of your design criteria for Class lE cables routed between seismic Category I buildings.
l 040. 64 It is the staff position that the diesel generator qualifications (8.3.1) testing programs satisfy position 5 of Regulatory Guide 1.6, Regula-i tory Guide 1.9, and Branch Tea..nical Positions ICSB 2 (PSB).
(SRP f
Revision 1, Section 8.3.1, Part II, Item 4d).
i l
Describe how your design meets this position.
Identify and justify each exception.
i 040. 65 It is the staff position that cables installed in raceways shall (8.3.1) be marked (by color coding) at a sufficient number of points - at intervals not to exceed 5 feet throughout the entire cable length -
l to facilitate verification that the cable installations is in con-formance with separation criteria.
(Regulatory Guide 1.75 position C10 and IEEE Standard 384-1974 Section 5.1.2).
Section 8.3.1.3.4 of the FSAR implies non-compliance with this position. Justify the non-compliance.
2157 186
040-33 i
A 040.66 Figure 9.2-3 and Section 9.2.2.4.5 of the FSAR indicater a common (8. 3.1 )
(between Unit 1 and 2) or a fifth component cooling system pump i
for which the normal power supply is from the ESF buses. Section 9.2.2.4.5 also indicates this power supply is fully described in l
subsectier. 8.3.1.1 of the FSAR. Section 8.3.1.1 of the FSAR, however, I
l does not describe the power design requirements for this fifth com-ponent cooling system pump.
Provide the subject descriptions.
I The description must include but not be limited to 1) separation criteria for the power sources and cable (power, control and instrumentation) l to each of the five pumps and 2) the results of a analysis that i
demonstrates when the fifth pump is being used, that no single failure will reduce power requirements to the component cooling water syster.
below ac:eptable limits.
i 040.67 In order to meet Branch Technical Position RSB 5-1,it is anticipated (8.3.1) i that independent and redundant power supplies will be required to each of the four RHR valves shown on figure 5.4-4 of the FSAR (8701-A-1, 87018-2,8702A-1,87028-2).
If four power supplies are required, provide a description of the power supplies electrical independence and physical separation. Provide a description of the physical separatien provided the power and control cables to these valver. And, identify each location where the physical separation between any of the four cables (power and cor, trol) is less than a triree hour rated fire barrier.
2157 187
040-34 040. 68 A description of the DC power systems associated with the diesel (8.3.2) driven auxiliary feedwater pump has not been provided in section 8.3.2 of the FSAR with the description of the Class lE DC power sys-i tems. Provide the subject description. The description must include requirements for separation, capacity, charging, ventilation, loading, t
redundancy, testing, results of an analysis to demonstrate compli-ance with the Commission's general design criteria, and the extent to which recomendations of Regulatory Guides and other applicable i
I criteria are followed.
I l
040.69 Section 8.3.2.1.2 and Dmwings GE-0-4001 and 20E-0-4001 of the FSAR (8.3.2) indicates non-safety related as well as Class lE DC loads are connected l
to the redundant 125-Vdc Class lE battery buses.
[
a.
Provide a tabulation of all safety-related and non-safety related loads to be connected to each d-c supply.
I b.
For each load (safety ard non-safety) describe the length of time they would be operable in the event of loss of all a.c. power.
f Prcvide design information and analyses to demonstrate the suitability c.
of the batteries and battery chargers as d-c power supplies.
d.
Provide a description of the capability of the battery chargers i
{
to properly function and remain stable upon the disconnection of the battery.
I i
2157 188
?
i
040-35 040.70 Section 8.3.2.1.1 of the FSAR indicates that the Byron /Braidwood design (8.3.2) provides interconnections for possibly sharin9 d-c supplies between Unit 1 and Unit 2.
In this regard, it is the staff position that d-c systems in multi-unit nuclear power plants should not be shared (position 1 of Regulatory Guide 1.81). Justify the Byron /Brcidwood design in view of this position.
040. 71 It is the staff position that the reliability of the d-c supplies (8.3.2) should be assured by periodic discharge test of the batteries as required by section 5.3.6 of IEEE Standard 308-1971.
Regulatory Guide 1.129 or Branch Technical Positior. EICSB 6 (SRP dated 11/24/75, Appendix 7A) outline an acceptable approach in re-gard to this position.
Describe the extert to which the recommenda-l tions of Regulatory Guide 1.129 or Branch Technical Position EICSB l
6 are followed.
Identify and justify each exception taken to the i
Regulatory Guide or Branch Technical Position.
040. 72 Section 8.3.2.1 of the FSAR states that the battery charger is rated (8.3.2) to supply its associated d-c loads while fully recharging the battery within a 24-hour period.
It is our interpretation of this
{
statement that the Byron /Braidwood battery charger supply meets the recommendations of position C.l.b of Regulatory Guide 1.32.
Confirm that our interpreation is correct.
2157 189 m
040-36 040.73 To assure that the interconnections between nonsafety-related (8.3.1)
(8.3.2) loads and safety-related buses will not compromise the independence between redundant systems nor degrade either redundant system below an acceptable level, it is the staff position that the Byron /
Braidwood design cor. form to Regulatory Guide 1.75 with respect to the role isolation devices play (SRP Section 8.3.2, Part II, item 3c).
i For each non-Class lE load (ac and dc) connected to the Class (1) lE buses via isolation devices, provide qualification information to demonstrate that any failure in the non-Class lE circuit will not cause unacceptable influences in the Class lE circuits.
l (2) For each non-Class lE load (ac and de) that rely on overcurrent i
devices for isolation, provide the results of an analysis to demonstrate that any failure of non-Class lE circuits or failure of multiple non-Class lE circuits due to design basis events (such as a fire) will not degrade Class lE circuits below acceptable levels.
040. 74 From the information in Section 8.1.l(a) of the FSAR, it appears (8.3.1)
(8.3.2) that electrically powered safety loads (ac and dc) have not been separated into redundtnt load groups such that loss of any one group will not prevent the. minimum safety function from being per-formed.
Confirm that the Byron /Braidwood design is in conformance with position 0.1 of Regulatory Guide 1.6 for both ac and dc load groups or provide additional design information to demonstrate that an acceptable level of safety har been attained.
2157 190
I 040-37
)
i i
j 040.75 In regard to thermal overload protection for motors of motor-operated (8.3.1) l (8.3.2) safety related valves, Regulatory Guide 1.106 or Branch Technical Posi-l tion EICSB 27 (SRP dated 11/24/75, Appendix 7A) outline acceptable l
l design criteria.
Describe how the Byron /Braidwood design conforms to Regulatory Guide 1.106 or Branch Technical Position EICSB 27.
i t
Identify and justify each exception taken to the Regulatory Guide or Branch Technical Position.
j I
i l,
2157 19I i
l i
f
110-5 110 MECHANICAL ENGINEERING 110.22 Expand the definition of a " Leakage Crack" contained in l
( 3. 6.1.1 )
Item k of Section 3.6.1.1.1 (page 3.6-3) to include "an environment which wets all unprotected components within the compartment..." as described in Item c of Section i
3.6.2.1.2.3.3 ( page 3.6-13).
[
110.23 Expand Items b.2 and 3 ( page 3.6-9) and d (page 3.6-10) of
( 3.6.2.1 )
Section 3.6.2.1.2.1.1 to indicate that the stress will be calculated using the sum of equations (9) and (10) of i
l NC-3652.
110.24 Expand Section 3.6.2.1.2.1.1 ( pages 3.6-8 through 3.6-10)
( 3. 6. 2.1 )
to include the criteria used for postulating break i
locations in high energy piping systems which are not designed to seismic Category I standards.
110.25 Expand Section 3.6.2.1.2.2 (page 3.6-11) to include the
( 3.6.2.1 )
criteria for postulating the break locations in moderate energy piping systems, which are and are not designed to seismic Category I standards.7hese breaks are mentioned in Item (2) of the heading to Section 3.6.1.3 (page 3.6-5).
110.26 Expand Item a of Section 3.6.2.1.2.3.2 ( page 3.6-12) to
( 3.6.2.1 )
indicate how consideration of the maximum stress range is used to exempt longitudinal breaks when the postulated break location is due to a usage factor in excess of 0.1.
110.27 The heading to Section 3.6.2.2.2.1.3 ( page 3.6-16), Item a (3.6.2.2) of Section 3.6.2.2.2.1.4 ( page 3.6-17), and Attachment (Attach.
83.6 (page B3.6-1) cite a draf t ANSI Standard. As such B3.6) draft documents are subject to frequent changes during their develognent, they are not acceptable as a reference. Therefore, i
expand the appropriate sections to specifically identify the criteria used.
110.28 Expand Sections 3.6.2.3.1 (pages 3.6-25 through 27) and 3.6.2.3.2 (3.6.2.3)
(pages 3.6-27 through 30) to include a desription of tne l
procedures to be used to insure that, throughout the life of the plant, the restraints will not adversely affect the stresses in the pipes that have restraints installed.
i i
2157 192 i
i
110-6 110.29 The response to Question 110.2 is not totally acceptable.
(3.6.2.3)
The 5% increase in allowable stress due to strain hardening effects is not acceptdale as an across-the4)oard value. Ex pand Item a of Section 3.6.2.3.2.4 (page 3.6-28) or the response to Question 110.2 to indicate the criteria to be met before i
the 5% increase will be used.
110.30 The response to Question 110.3 is not clear.
.(3.6.2.3) i Expand Item a of Section 3.6.2.3.2.4 ( page 3.6-28) or the response to Question 110.3 to clearly indicate how the proposed criteria will insure that the honeyconb material will not experience a deflection in excess of that defined by tne horizontal portion of the load deflection curve.
i 110.31 While Section 3.6.2.8 indicates that guard pipes are not used.
(3.6.2.4) the penetration sleeves shown in Figures 3.8-40 through 42 are similar to guard pipes in that they may prevent the $ccess required for the augmented inservice inspectica committed to in response to Question 110.1.
To preclude such an occurrence, provide a connitment to provide accessibility to examine all welds located in the penetration region. Alternatively, a commitment to insure that process pipe welds are not located within the penetration sleeve would be acceptd)le.
110.32 Figures 3.8-41 and 42 show penetration sleeves which are (3.6.2.4) welded to the process pipe.
Expand Section 3.6.2.4 (page 3.6-35) to provide a commitment to the following criteria:
(1) An analysis is required sufficient to demonstrate that the limits of Items a.2 and 3 of Section 3.6.2.1.2.1.1 ( page 3.6-8) of the FSAR are met. The use of the special exemp-tions, relative to the omission of a fatigue analysis, contained in NB and NE-3222.4 is not permitted.
(2) A 100% volumetric examination of the penetration sleeve to process pipe weld during each inspection interval should be included in the augnented inservice inspection program committed to in the response to Question 110.1.
2157 193 1
9
't 8
110-7 110.33 Section 3.7.3.7 (pages 3.7-22 through 25) proposes two methods (3.7.3.7) for cod)ination of modal responses which are similar to those contained in Regulatory Guide 1.92.
Justify the following differences:
(1) Equation 3.7-32 of the FSAR sums the product of the pair of responses whereas equation 8 of the RG sums the d) solute value.
(2) Equation 3.7-34 of the FSAR uses the modified damping facter whereas equation 10 of the RG uses the uncorrected value.
(3) Equation 3.7-36 of the FSAR sums the product of the pair of responses whereas equation 5 of the RG sums the absolute value.
(4) Equation 2,7-36 of the FSAR uses a coupling factor whereas equation 5 of the RG,does not.
(5) Equation 3.7-37 of the FSAR uses the product of the pair of modified modal frequencies where as equation 9 uses the difference.
(6) Equation 3.7-38 of the FSAR uses the modified damping factor whereas equation 10 of the RG uses the uncorrected value.
110.34 Equations 3.7-31, 3.7-32 and 3.7-36 of Section 3.7.3.7 ( pages 3.7-22 (3.7.3.7) through 25) contain, what appear to be typographical, errors in the bounds of the summations. These errors make it difficult to follow the logic of the presentation and should be corrected.
110.35 Sections 3.9.1.4.3 ( pages ?,,0-23 through 3.9-25), 3.9.1.4.5 (3.9.1.4)
( page 3.9-27), and 3.9.1.4.7 (page 3.9-29) contain a number of typographical errors which should be corrected in order to eliminate any passible confusion and to permit a complete review.
(1) The mathematical symbol for tne displacement vector is missing from the third paragraph on page 3.9-23.
2157 194
- P
e 110-8 (2) Equations 3.9-1 through 3.9-5 are missing from page 3.9-24.
(3) The subscripts for T, M, and K are missing f rom pages 3.9-24 and 3.9-25.
(4) The subscripts for K are missing from page 3.9-27.
(5) The subscripts for S are missing from page 3.9-29.
110.36 The final paragraph of Section 3.9.1.4.7 ( page 3.9-30)
(3.9.1.4) indicates that, for ASME Class 1 components, the response to the LOCA and SSE are conbined using the SRSS method.
Expand this section to indicate:
(1) How, for ASME Class 1 components, the responses to other load combi:vtions are combined; and (2) How, for ASME Class 1 component supports, the responses are combined.
110.37 Expand Section 3.9.2.1 (page 3.9-31) to provide a commitment
( 3. 9. 2.1 )
to include the following systems in the preoperational vibration (RSP) and dynamic effects program:
(1) ASME Code Class 1, 2, and 3 systems; (2) Other high energy piping systems inside seismic Category I structures; (3) High energy portions of systems whose failure could reduce the functioning of any seismic Category I plant feature to an unacceptdale safety level; and (4) Seismic Category I portions of moderatc energy piping systems located outside containment.
Additionally, it is the Staff's position that an acceptable test program to confirm the adequacy of the designs should consist of tne following:
(1) A list of systems that will be monitored.
(2)
A listing of the different flow modes of operation and transients such as pump trips, valve closures, etc. to 2157 195
110-9 110.37 which the components will be subjected during the (Cont.)
test. For example, the transients associated with the reactor coolant system heatup tests should include, but not necessarily be limited to:
(a) Reactor coolant pump start.
(b ) Reactor coolant punp trip.
(c) Operation of primary and secondary pressure-relieving valves.
(d) Closure of a turbine stop valve.
(3) A list of selected locations in the piping system at which visual inspections and measurements (as needed) will be performed dJring the tests. For each of these selected locations, the deflection (peak-to-peak) or other appropriate criteria, to be used to show that the stress and fatigue limits are within the design levels, should be provided.
(4) A list of snubbers on systems which experience sufficient thermal movement to measure snubber travel from cold to hot position.
(3) A description of the thermal motion monitoring program, i.e.,
verification of snubber movement, adequate clearances and gaps, including acceptance criteria and how motion will be measured.
j (6)
If vibration is noted beyond the acceptance levels set by the criteria of (3), above, corrective restraints should be designed, incorporated in the piping system analysis, and installed.
If, during the test, piping system restraints are determined to be inadequate or are damaged, corrective restraints, should be installed and another test should be performed to determine that the vibrations have been reduced to an acceptable level.
If no snubber piston travel is measured at those stations indicated in (4), above, a description should be provided of the corrective action to be taken to assure that the snubber is operable.
2157 196 m
M
110-10 110.33 The footnotes to Section 3.9.3.1.1 (page 3.9-45) are not clear
( 3. 9. 3.1 )
as to whether or not active components include those wnich must function following the transient or event. Verify that such is, in fact, the case.
110.39 Expand Section 3.9.3.1 (pages 3.9-45 through 46) to provica
( 3.9.3.1 )
the stress limits for ASME Class 2 ana 3 piping.
110.40 Expand Section 3.9.3.1 (pages 3.9-44 through 46) to incluce
( 3.9. 3.1 )
the criteria used to assure the functional capability of essential systems when they are subjected to loads in excess of trose for which Service limit B limits are specified. By essential systems are meant those ASME Class 1, 2, and 3 piping systems which are necessary to shut down the piant following, or to mitigate the consequences of an accident.
110.41 The answer to Question 110.16 is not responsive to the
( 3.9.3.1 )
question asked.
Expand Section 3.9.3.1 (pages 3.9-44 through 46) to provide the load combinations and stress limits for all reactor internals.
110.42 Recent operating reactor experience indicates that vibratory
( 3.9.3.1 )
loads associated with the operation of positive displacement pumps have contributed to high cycle fatigue pipe failure.
Such failures are known to occur on both the suction and discharge sides of positivc displacement pumps in PWR charging systems.
Describe the measures that are proposed to be taken to absort these vibratory loads originating from the positive displacement charging pumps.
If pulsation dampers or other mechanical devices are to be used in the pumps vicinity, furnish a description of such devices, i.e., manufacturer, type, size, location, and effectiveness of the device.
In case pulsation dampers or other mechanical devices are not employed to damper vibratory loads:
(1) Descr'se the vibratory loao originating at the positive displacement pump and transmitted to the discharge and suction pipe and associated pipe supports.
2157 197 m
o
= -
e
110-11 110.42 (2) Describe in some detail how the maximum vibratory loads (Cont.!
were established for calculating the maximum alternating stress in the design of the pipe runs and associated su pports. Also describe the analytical procedure to determine the fatigue stresses in the affected piping system.
(3) Furnish an isometric sketch of the pump affected piping system showing the locatiod of the pipe supports and the peak alternating stresses. Also indicate the locations which till be monitored for vibration during the preoperational piping vibration and dynamic effects test program.
110.43 Expand Section 3.9.3.2.1.1 (pages 3.9-49 and 50) to indicate (3.9.3.2) how the operability of the balance of plant active pumps is assured when they are subjected to the nozzle loads associated with the safe shutdown earthquake.
110.44 Expand Section 3.9.3.2.1 (pages 3.9-47 through 50) to indicate (3.9.3.2) how the operability of the balance of plant active pumps is assured when they are subjected to loads, other than those from the safe shutdown earthquake, in excess of those for which Ser'tice Limit B limits are specified.
110.45 Expand Section 3.9.3.2.2 (pages 3.9-50 and 51) to indicate (3.9.3.2) how the operability of the balance of the plant active valves, with extended structure, is assured when tney are subjected to loads in excess of those for which Service Limit B limits are specified.
110.46 Expand the first paragraph on page 3.9-52 (Section 3.9.3.2.2)
(3.9.3.2) to indicte the criteria used in selecting which representative NSSS active valves will be tested.
110.47 Expand the second paragraph on page 3.9-52 (Section 3.9.3.2.2)
(3.9.3.2) to indicate how "... by analysis, the nozzle loads are sh7wn not to affect the operability of the valve.*
110.48 Sd) paragraph "a" on page 3.9-52 (Section 3.9.3.2.2) states (3.9.3.2) that "All the active valves are designed to have a first natural frecuency which is greater than 33 hertz." However, the first paragraph on page 3.9-53 states, "If the natural frequency of the valve is less than 33 hertz, a dynamic..."
Resolve this conflict.
2157 198
.?
110-12 110.49 The response to Question 110.9 is not acceptM)le.
(3.9.3.2)
( App. A1)
The justification for use of a DLF less than two shall be included in Section 3.9.3.3 (page 3.9-55) or the discussion of Regulatory Guide 1.67 in Appendix A1 (page A1.67-1).
Alternatively, a commitment to use DLF's not less than two should be included.
110.50 The responses to Questions 110.10 and 110.11 are not totally (3.9.3.4) acce ptable.
( App. A1)
Expand the response to clearly show how the two conservatisms incorporated in the analysis (namely, (1) using a response spectrum which is correct for the steam generator upper lateral. supports and (2) using the d) solute sum method of load combination) compensate for the lack of conservatism associated with the use of stresses 50* over the normal allowable limits for the faulted condition.
Similar statements are contained in the discussions of Regulatory Guides 1.124 and 1.130 (pages A1.124-1 and A1.130-1, respectively) in Appendix A1.
110.51 The response tc Question 110.13 is not totally acceptaole.
(3.9.3.4)
Provide the fol. lowing with respect to Table 0110.13-1:
(1) Define the term "Pr" used when discussing the criteria for Class I component supports at Service Level C.
(2)
Indicate why "Pm" (used when discussing the criteria for Class 1 ccmponent supports at Service Level C) is obtained from Table I-1.0 rather than from analysis.
(3) Expand the ta)le to include linear and component standard component supports. Alternatively, indicate that these other types of supports are not used.
(Note - for consistency, similar statements would be required in the response to Question 110.11 and the discussion of Regulatory Guide 1.124.)
(4) Expand the table to include the limits associated with the other methods of derign evaluation such as limit analysis, experimental stress analysis, and load rating. Alternatively, indicate tnat these other methods are not used.
2157 i99 a
110-13 110.52 Expand Section 3.9.3.4 (pages 3.9-55 through 62) to provide the (3.9.3.4) basis for selecting the location, required load capacity, structural and mechanical performance parameters of all safety-related snubbers (mechanical and hydraulic) and achieving a high level of operability assurance including:
(1) A description of the analytical and design methodology utilized to develop the required snubber locations and charateristics.
(2) A discussion of design specification requirements to assure that required structural and mechanical performance characteristics and product quality are achieved.
(3) Procedures, controls to assure correct installation of snd)bers and checking the hot and cold settings during plant startup tests.
(4) Provisions for accessibility for inspection, testing ar d repair or replacement of snubbers.
110.53 The responses to Question 110.8 and 110.13 are not totally (Tables acceptd>1e.
3.9-2, 3.9-5, The changes mentioned in the responses need to be made to 3.9-9 Tables 3.9-2, 3.9-5, 3.9-9, and 3.9-10.
3.9-10) 110.54 The response to Question 110.7 is not complete.
J (Table 3.9-3)
Although the response to the question (Amendment 18) indicated a revision to Table 3.9-3, none has been received through 1
Amendment 19.-
110.55 Expand Table 3.9-5 to indicate the method of con)ining the (Table res ponses.
3.9-5) 110.56 Expand Tables 3.9-8 and 3.9-10 to indicate the design criteria (Tables for ASME Class 1 active pumps and valves, respectively.
3.9-8 Alternatively, indicate that there are no ASME Class 1 active 3.9-10) pumps or valves, as appropriate.
110.57 Expand note 4 of Table 3.9-9 to indicate the criteria used to
( Tab l e insure that the valve disc will not fail should the valve be 3.9-9) subjected to "Pmax" while in the closed position.
57 200
=
110-14 110.58 A review of the design adequacy of your safety-related electrical (3.10) and mechanical equipment under seismic loadings will be performed by our Seismic Qualification Review Team (SQRT). A site visit at some future date will be necessary to inspect and otherwise evaluate selected equipment after our review of the following requested information. The SQRT effort will be primarily focused on the adequacy of the original single-axis, single-frequency tests or analyses in accordance with IEEE 344-1971.
Attached Enclosure 110-1 describes the SORT and its procedures.
Section V.2.A requires information which you should submit so that SQRT can perform its review.
110.59 The information presented in Sections 3.10 and Appendix Al
( 3.10) concerning seismic qualification of mechanical and electrical
( App. A1) equipment is not completely acceptable. For example, the discus-sion of Regulatory Guide 1.100 in Appendix Al references WCAP 8587 for the NSSS supplied equipnent. This topical report has not been approved by the Staff and it is considered highly unlikely that there will be an approved version prior to your fuel load date.
During an October 28, 1978 meeting, with the staff to discuss the qualification testing documented in WCAP-8587, Westinghouse presented an outline of additional information they can provide to supplement seismic test data and justify its adequacy for all plants for which CP aplications were docketed after October 27, 1972.
The staff reiterated its opinion that it would be extremely I
difficult to provide adequate justification for most equipment required to meet IEEE 323-1974 and Regulatory Guide 1.89, such as that for the Byron Braidwood Units, because they specify that environmental tests and aging be performed prior to the seismic tests. However, if there is some equipment for which availd)le seismic test data is believed to be adequate the staff stated that Westinghouse should provide a list of the equipment and a detailed justification for each item of equipment listed.
The staff further stated that it intends to review both seismic and environmental qualification for plants required to meet IEEE 323-1974 on a case-by-case basis rather than on a generic basis, however, full use will be made of previously approved information when appropriate.
To insure a timely review of the qualifications which have been
~
performed for the Byron /Braidwood NSSS Category IE equipment, it is requested that you submit the recuested information in an FSAR amendment.
2157 201
110-15 110.60 The response to Question 110.21 is not totally acceptable.
(5.4.2.5)
(16.4.4.5)
Expand Section 5.4.2.5.4 (page 5.4-16) and item a.6 of Section 16.4.4.5.4 ( page 16.3/4.4-13) to provide sufficient detail to permit a determination that the proposed tube plugging criteria is in accordance with Regulatory Guide 1.121.
110.61 Expand Appendix Al to include a commitment to comply with
( App. Al)
2i57 202
Appendix 110-1 SEISMIC QUALIFICATION REVIEW TEAM (SQRT)
Interim Precedures I.
SCOPE SQRT tasks include both generic and site specific reviews. Generic reviews cover equipment supplied by the NSSS and A/E common to more than one plant.
Specific plant reviews as delineated in the Standard Review Plan Sections 3.9.2 and 3.10 will be supplemented by SQRT site visits and evaluation.
II. OBJECTIVES SQRT is a group of NRC staff members established to conduct reviews of the design adequacy of safety related mechanical components, instrumentation and control equipment, and their supporting structures for various vibratory loads. SQRT is charged with accomplishing the following three tasks.
1.
Determine the design adequacy of mechanical and electrical ccaponents and their supports for the required vibratory loading conditions
~
which include:
(a) seismic (b) hydrodynamic (as applic?ble)
(c) offsite explosion (as applicable)
(d) other vibratory inputs frcm the operating environment (as applicable)
(e) appropriate comoinations of the above events.
2157 203
. 2.
Changes in seismic qualification criteria, such as the revision of IEEE Std. 344 and other IEEE Standards, and the issuance of Regula-tory Guides 1.100 and 1.89 require that the staff verify:
(a) For older plants having components qualified by previous criteria, that components have adequate margin to perform their intended design functions during and after a seismic event.
(b) For new plant applications; that there has been uniformity and consistency in implementing the current criteria.
3.
In the case of plants which have design basis seismic ground motion levels and/or other required vibratory loads increased, review to assure adequate design margin exists at the revised levels.
III. GENERAL CRITERIA The bases used by the staff to determine the acceptability of equipment qualification will be IEEE Std. 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, and Standard Review Plan Sections 3.9.2 and 3.10 IV. GENERAL PROCEDURES SQRT will conduct generic and plant specific reviews:
1.
Generic reviews will be conducted of all NSSS vendors and most architect engineers (major equipment vendors and testing laboratories may be included if necessary) to assure proper interpretation and implementation of the current equipment qualification criteria applied c i s) p 1
to plants applying for construction permits and operating licenses.
2.
A plant specific equipment qualification review will be conducted of each plant now undergoing licensing review having components qualified to criteria different from current requirements.
A.
For components having multi-plant application (such as those within tha scope of an NSSS vendor), an equipment qualification review at specific sites will provide generic qualifications.
B.
For components which have only specific plant application (mostly those within the scope of the BCP supply), an equipment qualification review at specific sites will provide site-specific qualifications.
3.
Equipment qualification review for plants with revised increased design basis seismic ground motion levels and/or other required vibratory loads will t'e conducted on a plant by plant basis.
V.
SPECIFIC PROCEDURES SQRT procedures provide for both generic discussion meetings and plant site visits.
1.
Generic Discussion Meeting:
To implement the generic review specified in IV.1 and IV.2.A, a generic discussion meeting will be held to discuss the following:
A.
Meeting Agenda Meeting Objectives by SCRT 2157 205
4 B.
NSSS or A/E personnel should be prepared to present the following information:
(1) A detailed description of current practice followed in equipment qualification, including acceptance criteria, methods, and procedures used in conducting testing and analysis. Present and discuss the equipment qualification program on certain specified items (i.e., pumps, valves, diesel generators, motors, bistable units, relays, electrical cabinets,etc.)
(2)
Information regarding administrative control of equipment qualification, especially the handling of interface problems, documentation, and internal review procedures.
(3)
Identifying the scope of their suppliers. A list of equip-ment should be made available if possible prior to the meeting.
C.
For the cases specified in IV.2.A, methods and procedures for conducting equipment qualification review are discussed, including selection of plants for site visits and setting up a tentative schedule for such visits.
O.
Discuss necessary documentation.
E.
Inspect testing facilities, if any. Testing capaoility, format of testing reports, wave forms of shaker table motions, and monitoring and centrol cevices are the major items for inscection.
2157 206
. F.
SQRT concludes the meeting and specifies the follow-up items.
5 2.
Plant Site Reviews:
To implement plant specific equipment qualification reviews specified in IV.2 above, on-site inspection of equipment and supporting structures in question is required.
Site visits generally follow the following procedures:
A.
Pre-visit information submission:
Steo 1 The applicant '
ant owner) receives initial information concerning the intended isit, and should subsequently submit the following:
(i) Two summary equipment lists (one for NSSS supplied equipment and one for BOP supplied equipment). These lists should include all safety related mechanical components, instrumentation, and control equipment, including valve actuators and other appurtenances of active pumps and valves.
In the lists, the following information should be specified for each item of equipment:
(1) Method of qualification used:
(a) Analysis or test (indicate the reference report number)
(b)
If by test, describe whether it was a single or multi-frequency test and whether input was single axis or bi-axial (c)
If by analysis, describe wnether static or dynamic, 2157 207
. single or multiple-axis analysis was used. Present natural frequency of equipment.
l (2)
Indicate whether the equipment is required for:
(a) hot stand-by (b ) cold shutdown (c) both (d) neither (3) Location of equipment, i.e., building, elevation.
(4) Availability for inspection (Is the equipment already installed at the plant site?)
(ii) An accceptd)le scenario of how to maintain hot stand 4)y and cold shutdown based on the following assumptions:
(1) SSE or OBE (2) Loss of offsite power
.(3) Any single failure (iii) A compilation of the required response spectra (RRS) for all applicable vibratory loads (individual and combined if required) for each floor of the nuclear station under consideration.
Step 2 SQRT screens the a)ove information and decides which items will be eva'uated during our forthcoming site visit. The applicant will be informed of these items and will be expected to sd)mit 2157 208
7-two weeks prior to the visit an equipment qualif cation summary as shown on pages 10-12 for each of the selected items.
B.
A brief meeting is held at the beginning of a site visit with the following agenda:
(1) SQRT explains the objectives of the site visit and procedures to conduct equipment inspection.
(2) Utility personnel or their designees present an overview of the seismic qualification program conducted.
(3) The seismic ' qualification of certain specified items may be discussed as necessary.
(4) SQRT specifies items that need to be inspected.
C.
SQRT conducts inspection of some specified items.
D.
SQRT reviews the qualification documents of the selected equipment.
E.
SQRT describes findings of the inspection and the review.
4 F.
General discussion.
G.
SQRT concludes the visit and specifies needed information and the follow-up actions.
3.
After each visit SQRT will issue a trip report, which 4.antifies findings, conclusions and follow-up items. Status reports may be issued as necessary. The site review will include the issuance of 2157 2139 I
Evaluation Report for the specific plant.. Generic evaluations an will be referenced to the NSSS vendor or A/E.
VI. RESPONSIBILITIES OF NRC PARTICIPANTS:
A.
The Seismic Qualification Review Team consists of memoers of t? u Mechanical Engineering Branch (MEB), the Instrumentation a..J Control Systems Branch (ICSB), and the Power Systems Branch (PSB). One additional member frem MEB will join the team when a review of a specific plant is going to be conducted.
This member will be the reviewer of the plant.
The Team Leader is responsible for scheduling actions, coordinating staff positions, and contacting appropriate authorities for work assignments to each member. He. reports to the MEB Branch Chief regarding the progress of SQRT performance. He will set up necessary contacts for generic reviews and will contact project management for specific plant site visits. He will specify the meeting objectives and concludes meetings.
The MEB members and Team Leader are responsible for reviewing assigned equipment qualifications in the area of responsibility of the Mechanical Engineering Branch, including the methods and procedures used in test and analysis.
Members representing the Power Systems Branch (PSB) and the Instrumentation
& Control Systems Branch (ICSB) are responsible for reviewing assigned equipment qualification in the area of responsibility of 2157 210
.g-their branch, including equipment signal interpretations for functional verification. They serve as a liaison cetween SQRT and ICSB and PSB.
All members shall present their opinion and professional judgement to the Team Leader in order to arrive at consistent and uniform SQRT positions.
B.
The MEB, PSB, and ICSB project reviewers will be advised of SQRT activities which relate to specific plants. The MEB project reviewer is responsible for evaluating the impact of SQRT activity on the specific plant review and for taking appropriate action to include pertinent information in the plant safety evaluation. The MEB project reviewer is expected to participate in the site visit and attend pertinent generic meetings as necessary.
The DPM project manager, after being informed of the intended plant visit, is expected to contact the applicant and arrange for the vi si t.
The project manager serves as a liaison between the SORT and the applicant.
C.
Generic meetings will be arranged by the SQRT or via the OPM generic project manager if one is assigned.
D.
Representatives from !&E Regional Offices and other interested organizational groups within NRC are welcome to attend either generic meetings or plant site visits as observers. The SCRT should be informed of expected attendance at such meetings or site visits.
2l57 211
~
Qualification Summary of Ecuioment I.
Plant Name:
Tyoe:
1.
Utility:
PWR -
2.
NSSS:
3.
A/E:
BWR II. Component Name 1.
Scope:
[ ] NSSS
[ ] BOP 2.
Model Number:
Quantity:
3.
Vendor:
4.
If the component is a cabinet or panel, name and model No. of the devices included:
5.
Physical Description a.
Appearance b.
Dimensions i
c.
Weight 6.
Location:
Building:
Elevation:
7.
Field Mounting Conditions [ ] Bolt (No.
, Size
)
[] Weld (Lengtn
)
[]
8.
Natural Frequencies in Each Direction (Side / Side, Front /Back, Vertical):
S/S:
F/B:
V:
9.
a.
Functional
Description:
b.
Is the equipment required for [] Hot Stancby [] Cold Shutdown
[] Both 10.
Pertinent Reference Design Specifications:
2157 2i2
~
_ 11 III.
Is Ecuioment Available for Insoection in the Plant:
[] Yes
[] No IV.
Ecuipment Cualification Method: Test:
Analysis:
j Combination of Test and Analysis:
Test and/or Analysis by (name of Company or LaDoratory & Report No.)
V.
Vibration Input:
1.
Loads considered:1.[ ] Seismic only 2.[ ] Hydrodynamic only 3.[ ] Explosive only 4.[ ] Other (Specify) 5.[ ] Combination of 6.
Method of combining RRS: [ ] Absolute Sum [ ] SRSS [ ]
(otner, specify) 2.
Required Response Spectra (attach the graphs):
l 3.
Required Acceleration in Each ' Direction:
S/S =
F/B =
V=
VI.
If Oualification by Test, then comolete:
[ ] random 1.
[ ] Single Frequency
[ ] Multi-Frequency:
[ ] sine beat
[]
2.
[ ] Single Axis
[ ] Multi-Axis 3.
No. of Qualification Tests: OBE SSE Other
( specify) 4.
Frequency Range:
5.
TRS enveloping RRS using Multi-Frequency Test [ ] Yes (Plot TRS on RRS graphs)
[ ] No i
6.
Input g-level Test at S/S =
F/B =
V=
7.
Laboratory Mounting:
1.
[ ] Bolt (No.
Si::e
)
[ ] Weld (Length
)
[]
8.
Functional operaDility veri fied:
[ ] Yes [ ] No [ ] Not ApplicaDie 9.
Test Results including ::todifications made:
- 10. Other tests perfor ned (such as fragility test, including results):
s r' N 11
~7 G)bl
&I) c
. VII.
If Oualification by Analysis er by the combination of Test and Analysis, then Comolete: "
l.
Description of Test including Results:
2.
Method of Analysis:
[ ] Static Analysis
[ ] Equivalent Static Analysis
[ ] Dynamic Analysis: [ ] Time-History
[ ] Response Spectrum 3.
Model Type:
[ ] 3D
[ ] 2D
[]10
[ ] Finite Element
[ ] Beam
[ ] Closed Form Solution 4.
[ 3 Computer Codes:
Frequency Range and No. of modes considered:
[ ] Hand Calculations 5.
Method of Combining Dynamic Responses:
[ ] Absolute Sum [ ] SRSS
[ ]Other:
(spec 1fy) 6.
Damping:
Basis for the damping used:
7.
Support Considerations. in the model:
8.
Critical Structural Elements:
Governing Load or Seismic Total Stress A.
Identi fication Location Resconse Conicination Stress Stress Allowable Effect Upon Functional B.
Max. Deflection Location Ocerability 2157 214
~
G 130-2 130.0 STRUCTURA;. ENGINEERING BRANCH I
130.06 The seismic analysis was performed by the response spectrum method.
(3.7.2.1) However, the response spectra at the foundation level generated by
.)
the synthetic time history have displayed a significant dip over i;
a large range of frequencies as compared with the design response
- j spectra-in R.G. l.60 (Figures 3.7-1 through 3.7-40). The use of such unconservative response spectra is unacceptable to the staff.
~
The deconvolution procedure as described in the FSAR is not appropriate for the Byron /Braidwood sites due to the shallow soil overburden (16ft to 38ft) on bedrock. Therefore, it is requested that the analysis shall be based on R.G. 1.60 free field surface design response spectra applied at the foundation level and the design time history shall generate response spectra envelope the R.G. 1.60 design response spectra at the foundation level.
130.07 Provide the specific mass and frequency ratios or other specific (3.7.2.3.
criteria used to decoupling subsystem from the primary system or 2) structure.
130.08 Describe the criteria used to ensure an adequate number of masses or (3.7.2.3) degrees of freedom were employed in your dynamic modeling to determine the response of all seismic category I and applicable non-sei_smic category I structures and plant equipment.
I f
130.09 The river screen house at the Byron station is founded on soil. Soil-2
^tructure interaction was performed by using the finite element (3.7.2.4) s method.
It is the staff's position that the methods for implementing the soil-structure interaction analysis should include both the half space lumped spring and mass representation and the finite element approaches. Category I structures, systems and components should be designed to responses obtained by any one of the following methods:
(a) Envelope of results of the two methoos, (b) Results of one method with conservative design consideration of impact frcm use of the other method.
(c) Combination of (a) and (b) with provision of adequate conservatism in_desi.gn.. _
Therefore, we request you to compare the responses obtained by the half space (lumped parameter) approach to those obtained by the finite element approach at a few typical locations.
Floor response spectra should be provided at least at the basemat, an intermediate elevation and an upper elevation. For the lumped parameter representation, the variation of soil properties should be considered.
2i57 215 r
1 130-3 1
130.10 Reference is made to Question 130.2 and your response as amended C
(3.7.4) in Anendment 18. The purpose of installing separate triaxial response spectrum recorder at the foundation of an independent seismic i
category I stacture where the response is different from that of g
the reactor containment structure is twofold:
(1) To check the validity of the modeling technique and design assumptions made for the structure and design input motion to the supported systems and components.
i (2) To provide the necessary infomation as the advisability of l
continuing the operation of the plant without a safety analysis following an earthquake.
Therefore, we request you to install a triaxial response spectrum recorder at the foundation of the Byron River Screen House regardless of whether the response at the foundation is higher or lower than that of the containment building.
130.11 Referring to Table 3.8-3 of the FSAR, some of the loading cases (3.8.1.3) are different from those of section CC-3000, ASME B & PV Code, Section l
III, Division 2.
Provide justification to this deviation.
i 130.12 Provide the rationality for using both the thin shell approach and (3.8.14.
the thick shell approach for the analysis of the containment structure.
2)
Compare the values of moments and forces obtained by the two approaches. If they are different, assess the significance of the difference and the validity and/or the accuracy of both methods.
130.13 Refering to section 3.8.1.4.3 of the FSAR, it is not clear whether (3.8.1.
you have adequately included containment wall in your basemat analytical 4.3) model. Because of the extremely large inplane stiffness of the containment wall, little or no relative displacements between the nodes (at the same elevation) af the containment wall should result.
Therefore your use of only the rotational spring at the top nodes of the wall may not be representative, and a justificaticn for such a modeling approach should be provided.
130.14 You have stated that, principal tension stress less than 3/Tc is assumed (3.8.1.
to be resisted by the concrete. The ASME B & PV Code.Section III, 5.2)
Division 2 states that concrete tensile strength shall not be relied upon to resist flexural and membrane tension. Provide justificaction for the exception that you have taken with the code.
130.15 There are scme editorial errors.
Item C should have been under (3.8.3.
item b and item e shculd have beer, under item d.
Also, provide 5.1) detailed justification for utilizi 5 shear friction to resist tangential shear force. Describe 1e procedures of how to ccmbine the stresses resulting from shear fri uien with the membrane and flexure stresses.
2l57 216
130- 4 I
l 130.16 You have listed in Table 3.8.2 that ACI-349-15 is one of the (3.8.3.
appropriate industry standardsto use in the resign and cons-1 2) truction of the containment internal structores. Through Reg-f ulatory Guide 1.142, the staff has endorsed the use of ACI-349-76 I
with a number of exceptions. Therefore, it is the staff's position that ACT-349-76 may be used only if the regulatory position delineated in Regulatory Guide 1.142 is complied with.
130.17 Refering to Table 3.8-10 the load combination for reinforced concrete (3.8.3.
ultimate strength design in the FSAR, the load factors for the live 3) load (L), the thermal effects and loads during normal operating or shutdown conditions lTo), and the pipe reactions during normal operating or shutdcwn conditions (Ro) should be nqual to 1.7 instead i
i of 1.3 as shown in the table in accordance with FCI-349-76.
s 2i57 217 9
321-3 321.0 EFFLUENT TREATMENT SYSTEMS BRANCH 321.6 A continuous'flo'w 3,200 cfm miniflow purge system is described on page 9.4-51 of the FSAR; however, on page 11.2-20 of the FSAR and en pages 3.5-12 and 3.5-28 of the Environmental Report, it is stated that there is no continuous purging of. containment. The discrepancies should be corrected.
If the gaseous effluent source terms shown in Section 11.1 of the FSAR were calculated on the basis of no continuous purging of containment, the source terms should be recalculated using appropriate parameters.
321.7 Your response.to question 321.3 is not satisfactory. For (11.3) gaseous waste systems which are not designed to withstand a hydrogen explosion, it is our position that dual gas analyzers with automatic control functions are required to preclude the formation or buildup of explosive hydrogen / oxygen mixtures.
Criteria for dual gas analyzers aad for automatic control features are given in Standard Review Plan 11.3, Revision 1.
321.8 Your response to question 321.2 is not satisfactory.
In accordance with Regulatory Guide 1.140, you should provide for the in-place testing of HEPA filters and charcoal adsorbers in non-safety-related ventilation exhaust systems in order to take credit for the decontamination factors assumed in the analyses performed to show compliance with 10 CFR Part 50, Appendix I.
321.9 Provide a table, or revise Tables 11.5-1 and 11.5-2, to in-(11.5) dicate how the process and effluent radiological monitors meet the criteria in Tables 1A and 1B of Standard Review Plan 11.5, Rev. 2.
Identify monitors with automatic control features interlocked with isolation valves or other devices to terminate or reduce radioactive discharges; for example, SRP 11.5 indicates automatic control features for containment purge effluent (including miniflow purge systems), gas decay tank effluents, and liquid radwaste effluent, none of which are identified with automatic control features in the system description in Section 11.5. All such automatic control features should have fail-safe provisions, e.g., isolation valves should close automatically upon instrumentation failure or loss of power.
321.10 Your description of the charcoal adsorber and HEPA filter system for the main condenser offgas system in Section 10.4.2 should be cross-referenced to the description of the offgas and miscellaneous tank vent filter system in Section 9.4.7 and should indicate that this system is designed to be bypassed in normal operation.
2157 218
t 8
321-4 321.11 The word " permeate" is used several times in figures and (11.2) tables describing the laundry waste system, e.g., Figure 11.2-1, Figure 11.2-9, Table 11.2-9.
This nomenclature is apparently a holdover from the PSAR, where a reverse osmosis unit was pro-posed for laundry waste treatment, but which was not included in the PSAR. Referenccs to " permeate" are misleading and should be deleted.
321.12 Your response to question 321.5 is generally satisfactory.
(11.4)
Your process control program for solidification of solid wastes calls for establishiog pretested formulas for each specific waste stream; however, you have not provided for periodic testing to assure that solidification takes place.
In accordance with Branch Technical. Position (BTP) ETSB 11-3 (Rev. 1) Item II.1.b, you should revise your response to provide for periodic testing for solidification and to provide for changing the solidification formulas as necessary.
321.13 Describe your provisions for the packaging and shipment of (11.4) radioactive or potentially radioactive " secondary side" ion exchange resins.
I
331-4 1
331.0 RADIOLOGICAL ASSESSMEriT BRANCH 331.12 Indicate whether, and if so how, the guidance followed (12.0) by the following Regulatory Guides has been followed; if not followed, describe the specific alternative methods used:
I R.G. 8.2, " Guide for Administrative Practices in l
l Radiation Monitoring".
R.G. 8.3, " Film Badje Performance Criteria".
R.G. 8.7, ' Occupational Radiation Exposure Records System".
}
R.G. 8.9, " Acceptable Concepts, Models, Equations, and l
Assumptions for a Bioassay Program".
I l
R.G. 8.10, " Operating Philosophy for Mcintaining Occupa-l tional Radiation Exposures as Low as is Reasonably Achievable".
j R.G. 8.12, " Criticality Accident Alarm Systems".
l R.G. 8.15 " Acceptable Programs for Respiratory Protection".
t j
R.G. 8.19, " Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants Design Stage i
Man-Rem Estimates".
331.13 The following design features are intended to complement (12.1.2) the ALARA program. Describe how your plant design reflects consideration of these features:
j 1.
Use of adequate and quick-service auxiliary lighting in high radiation areas.
2.
Clear identification of localized radiation sources.
l c,
clal 220 I
331
'5 331.14 Tables 12.2-46 through 12.2-48 give the calculated (Tables 12.2-46 l
thru 12.2-48) airborne iodine activity concentrations for the major I
plant buildings.
Provide similar tables listing cal-l culated airborne radioactivity concentrations in these l
t areas for krypton, xenon, tritium, and other expected airborne radionuclides.
I i
331.15 Provide an illustrative example of each of the following l
(12.3.1) comocnsnts describing how they are designed to minimize occupational dose:
demineralizers, recombiners, evaporators, and sampling stations.
Include diagrams of these examples showing the radiation protection features.
I l
331.16 Show that your change and locker room facilities are 1
(12.3.1) l sufficient to accommodate and provide services for an I
expected increased number of maintenance perscnnel : resent f
during major outages.
331.17 Provide the following information concerning the spent fuel (12.3.2) i transfer tubes:
a)
Describe the shielding or structural barriers used to l
prevent inadvertent access to potentially high radia-I tion areas near the fuel transfer tube and describe the shie.lJing provided to assure acceptable radiation levels in adjacent occupied areas.
i 7
cIJ/C' no 22]
i i
331-6 331.17
- cont'd b) Provide plan and elevation layout drawings of the areas through which the spent fuel transfer tube passes.
c) Discuss your procedure f,or positive access control and radiation monitoring for the areas where J.a spent fuel transfer tubes may be exposed.
331.18 Area and airborne radiation monitors located in high (12.3.4) noise areas should have visual as well as audible alarms.
State hcw you plan to comply.
331.19 Indicate the location of the following area monitors on the (Table 12.3-3) appropriate figures in Chapter 12:
ORE-AR001 through I
ORE-AR003, ORE-AR041 through ORE-AR050, IRE-AR001, IRE-AR002, 2RE-AR001, and 2RE-AR002.
331.20 Using the appropriate layout drawings as references, (Fig. 12.3-23 and 1.2-8) describe the path of the 55-gallen drum.".to container from the time it is removed from the nnpty drum storage area to the time it 'is loaded onto tne waste loading trailer. The section of the radwaste/ service building between grid H-L and 37-41 or level 401', depicted in Figure 1.2-8, appears to contain radwaste equipment.
This same area on the 401' level of figure 12.3-23 appears to contain radwaste storage areas.
Clarify this apparent discrepancy.
Also describe the functicn of the ficcr transfer system shown between grids i'-P qt E7
? L\\JI
331-7 331.20 - cont'd and 36-48 on the 401' level of the radwaste/ service building complex in Figure 1.2-8. Describe the relatic anip, if any, between this system and the drum transfer tunnel depicted on level 401 in Figure 12.3-23. 331.21 Figure 12.3-26 shows the laundry roca to be in a radia-(Fig. 12.3-26) tion zcne of f4 crem/hr., State how you plan to control access to this room so that plant personnel doses do not excead 100 mrem / week. 331.22 Using occupational exposure data from Zion 1&2 and other (12.4) ap?licable operating reactors for 1977 and 1978., update your operating exposure estimates for Byron /Braidwood 1&2 in Table Q-331.9-1. Incorporate the revised dose esti-mates from your response to 0-331.9 into the appropriate sections of Chapter 12.4. 331.23 Your response to Q-331.11 is inadequate. Supply the (12.5.3.3) specific frequencies to be used for plant and contractor bioassays and the criteria used for determining which personnel are to be tested. 2157 223 ~ e 9
371-5 i dYDR0 LOGY-METEOROLOGY BRANCH-HYDROLOGY ENGINEERING SECTION - 371.0 EiRON CTATION l i 371.7 You have not provided the information requested in questions (2.4.2.3) l 371.3 and 371.4. Revised figure 2.4-8, provided in Amendment f 18, is of very poor quality, and consequently, many of the lines and figures are only partially legible or not legible at all. Provide a detailed topographic map that shows the drainage feaf ores just outside the perimeter roads, This information is necessary to verify drainage patterns which e i occur outside the drainage areas identified on figure 2.4-8a. l It appears that there are additional drainage areas outside 1 the plant area that will convey water toward the plant area. If this is the case, these areas should be identified and the l l flows accounted for. Additionally, these surrounding drainage I patterns control the degree of submergence on culverts and road weirs and therefore are required for our review. The topographic I map should also contain sufficient detail to allow accurate f computation of water surface profiles between safety related buildings and road weirs. l You have not provided the requested peak discharges for all drainage areas identified. Provide the peak discharge or i hydrograph.resulting from PMP on each drainage area, including any new drainage areas identified as a result of this question. Provide the method used to develop peak discharges or hydrographs. l 5
371-6 Your discussion in section 2.4.3.2 indicates that you have apparently taken credit for culvert flow during the PMP event. Accordingly, provide for each culvert, the peak discharge, associated headwater and tailwater ~ elevations, type of flow, and flow parameters used (ie., discharge coefficients, entrance losses, roughness coefficients, etc.) A tabular presentation would assist in the review of this material. Also provide a discussion of your assumptions on debris accumulation and subsequent blockage for each culvert. 371.8 Provide the maximum allowable design return water temperature (9.2) from the Essential Service Water Cooling Tower Basins, as limited by safety related equipment or systems. Specify the equipment or system that is the basis for the limiting value. 371.9 Are there any ou'. side surface tanks containing radioactive (2.4.12) liquids that could spill and runoff directly to surface water? If there are such tanks, provide analyses to show dilution factors and travel times at the nearest downgradient potable water supply (either surface or groundwater, whichever is critical in tems of concentration). Provide discussions for both pathways. 2157 225
371-7 L 371.10 Are there any cewatering systems that will be used to j (2.4.13.5) I permanently lower groundwater levels under either safety i or non-safety related buildings during plant operation? I If there are, provide the information and analyses necessary to show compliance with the attached Branch I Technical Position HMB/GSB 1. I i f i i i i i 2157 226 i i i i -he
371-8 371.0 HYDROLOGY-METEOROLOGY BRANCH-HYDROLOGY ENGINEERING SECTION - BRAIDWOOD STA1 ION 371.4 You have not provided all the infomation requested in (2.4.2.3) question 371.2. Provide a detailed topographic map that shows the drainage features between plant safety related buildings and perimeter roads and railroads that act as weirs. Also show the topographic features, just outside the perimeter roads, sufficient to determine flow patterns and submergence effects. Provide the peak discharge or hydrograph resulting from PMP on each drainage sub-area within the perimeter roads and for any subareas outside the perimeter roads if they are pertinent to the analyses. Provide the method and parameters used to develop peak discharges or hydrographs. Your discussions in Section 2.4.2.3 indicate that you have apparently taken credit for culvert flow during the PMP event. Accordingly, provide for each culvert the peak discharge, associated headwater and tailwater elevations, and type of flow and flow parameters used (i.e., discharge coefficients, entrance losses, roughness coefficient:, etc.) A tabular presentation would assist in the review of this material. Also provide a discussion of your assumptions on debris accumulations and subsequent blockage for each culvert. t i 2157 227 k r
371-9 371.5 Are there any outside surface tanks containing radioactive (2.4.12) liquids that could spill and runoff directly to surface water? If there are such tanks, provide analyses to show dilution factors and travel times at the nearest downgradient potable water supply (either surface or groundwater, whichever is critical in terms of concentration). Provide discussions for both pathways. 371.6 Are there any dewatering system that will be used to (2.4.13.5) permanently lower groundwater levels under either safety or non-safety buildings during plant operation? If there i are, provide the information and analyses necessary to l l show compliance with the attached Branch Technical Position HMB/GSB 1. 2i57 228 (
BRANCH TECHNICAL POSITIONS HMB/GSB 1 SAFETY-RELATED PERMANENT DEWATERING SYSTEMS Surma ry This position has been formulated to minimize review probless ccman to pemanent dewater-ing systems that are depended upon to serve safety-related purposes by describing accept-able geotechnical and hydrologic engineering design bases and criteria. A safety-related designation for pemanent dewatering systems is provided since they protect other safety-related structures, systems and components from the effects of natural and man caused events such as groundwater. In addition, the level of documentation of data and studies which are considered necessary to support safety-related functions is defined. This position appliet to both active (e.g., uses pumps) and passive (e.g., uses gravity drains) dewatering systems. This position does not reflect structural, mechanical and electrical criterir. II. Backersund The staff has reviewed a neber of pemanent dewatering systems, including McGuire 1 & 2 Cherokee 1 & 2 Perkins 1 & 2. Perry 1 & 2 WPPSS 3 & 5, Douglas Point 1 & 2, and Catawba 1 & 2. Perry, beginning in 1975, was the first plant reviewed with suen systems, and was reviewed very late in the CP process. 'Only WPPSS 3 & 5 and Douglas Point use a passive systen (no p eps). Permanent dewatering systems lower groundwater levels to reduce subsurface water leads on plant structures. In addition, they can increase plant operational dependability and redJee costs. These effects are accomplished by providing added means of keeping seepage water out of lower butiding levels during the later stages of plant life when nomal water-proofing provisions may have deteriorated, and reducing radwaste system operating costs by minimizing the amount of drain water that must be treated. Benefits are, therefore, of two types, tangible (dollars) and intangible (" insurance"). We understand the construction costs of underdrains can vary widely depending on the design., Construction costs of between $125K to $1000K per unit have been suggested. The costs of coping with significant amounts of groundwater inleakage in safety-relats wilding areas, which underdrains are expected to minimize, is estimated to be in the raige of $100K to $200K per year per reactor. The construction costs of alternatives to underdrains for structural purposes alone (exclusive of inteakage treat:nent) is estimated to range upward frms $300K per unit and is highly dependent on site conditions. Structural alternatives to permanent underdrains include additional concrete and steel in the lower portions of buildings, and the use of anchor e systems to resist floatation. Dewatering systems are generally composed of three cmoonents; the collector systen, the drain system, and the discharge systen. Water is first collected in collector drains 2.4.13-9 Rev. 1 s.N ~'
a e adjacent to buildings or excavations. Interceptor drains or piping are then used to convey this water to a fiaal discharge system. The discharge system can be either gravity flow ~ or a pumping system. Most underdrain structures, systems and components are buried along-side and under structures, although some systems employ pumping systems within larger structures (such as reactor or auxiliary buildings) to discharge collected water. Finally, pemanent dewatering systeris are not a required feature at any plant, but may be proposed as a cost effective feature. Many permanent dewatering systems at nonnuclear facilities, such as dams and large build-ings, have functioned over the years. However, the likelihood of a portion of such a system becoming ineffective and, therefore, not performing its intended function may well be considerably greater than the prooability f occurrence of a nuclear power plant design basis event such as a Probable Maximum Hurr' ae, Probable Maximum Flood, or Safe Shutdown Earthquake. Losses of function in the past have generally been attributable to piping of fines, inadequate capacity, or clogging. We nave concluded that safety analyses of such systems should consider reliability and failures of features of the system itself, as well as potentially adverse effects of failures of nearby nonsafety-related features. Such systems need not be designed for design earthquakes if they are not intended to perform as underdrains fully during or insnediately following a severe earthquake, or if the system can be expected to perform an underdrain function in a degraded condition. Certain portions of such systems, however, may be required to regularly perfom other safety functions (e.g., po-ous concrete base mats) and should be casigned for severe earthquakes. Failure of a dewatering system could cause groundwater levels to ris,e above design levels, resulting in overloading concrete walls and mats not designed to withstand the resulting hydrostatic pressures. In addition to causing potential structural and equipment damage, groundwater could enter safety-related buildings and flood compenents necessary for plant safety. The basis for staff concerns over the use of such systems is whether they can be expected to perform their function, and prevent structural failures and interior flooding of safety-related structures. The degree of concern is directly related to the correspondir.g degree to which the safety of the structures and systems rely on the integrity of the dewatering systeri, particularly with a dewatering system in a degraded situation. For example, if structures can accormodate hydrostatic,, loads that would result with a total failure of a dewatering system, our concerns have been primarily limited to the capability of such systems to perform their functions under relatively infrequent earthquake situations. If, nowever, such systems must remain functional (e.g., keep water levels down), whether in a degraded situation or not to prevent structural failures and internal flooding under potentially frequent conditions, we have been very concerned with system reliability. Many applicants have indicated that their plants can withstand, or have been designed against, full hydrostatic loadings that would occur in the absence of the underdrain systems, but not if an earthquake were to occur. If the plant can withstand full hydrostatic loading, assuming degradation of the underdrafn system, many of the staff's concerns may be gliminated from further consideration because of the time available for remedial action after detection of system degracation. " ' ' ' I
- 2. a.13-10 1
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!!!. Situations identified During Previous Reviews Four general categories of situations have been iddntified during case reviews as follows: (a) Estimating and confirming permeab111ty ' Values It is necessary to estimate the amount of water that will be collected so that system components such as strip drains, blanket drains, collector pipes, and pumps are ade-quately designed and sized. One of the most important and most difficult parameters to evaluate is the penneability of the soil and rock existing at a site. A per-meability value could be affected significantly by conditions of concentrated flow along joints in fractured and weathered rocks, or within other aquifers affected tm foundation excavation. In addition, geological and foundation conditicns that were not detected in site txclorations may affect flow conditions and cause the estimated pemeability values and flow regimes to be substantially different from those assumed at the CP preliminary design stage.. These conditions are often first detected during construction dewatering. Therefore, we have required a consnitment to consider con-struction excavation and dewatering data in the final design of underdrain systems. (See situation (d) belcw.) (b) Operational Monitorino Requirements To guard against system malfunctions and to assure sufficient time is available for implementation of remedial measures before groundwater could rise to an unacceptable level, provisions must be made for early detection of system fa':ures, and contingency measures for these failures must be well defined prior to plant operation. Since drain systems are usually buried and concealed and there may be no direct way of inspecting them, reliance must be placed on piezameters, observation wells, manholes, [ and monitoring of collected water to detect problems or malfunctioning of the system. The details of an operational monitoring program are necessary prior to construction of the underdrain to assure that each of the following will be provided: (a) an early detection alam system during normal operating conditions; (b) regularly scheduled inspection and gonitoring; and (c) competent evaluation of observa. ions during both construction and operation. In addition, the bases for acceptabl4 contingency measures suitable for coping with various possible hazards mun - u oitshed at the CP stage. (c) Pipe Breaks A dewatering system might be overloaded by such conditions as leaks or brepks in either the circulating or service water systems. A leak through a pipe break may be a very small percentage of the total flow of the cooling water system, but large enough to exceed the hydraulic capacity of drains, pipes and pumps in the dewatering system. For example, a conglete failure of circulating water system piping has been required in the design of the dewatering systems reviewed to date. This requirement w s made to assure that such abnormal occurrences do not adversely affect the integ-rity of safety-related structures, systems, and components. (d) Sequence of Review Underdrain systems are usually one of tM first items constructed and, after back. filling and construction of subsurt.s facilities, are then no longer visible for 2.4.13-11 Rev. 1 s 1! r7 q,i Llal L.) l
a regular inspection. In most cases, these systems are initially designed based on t rather limited information from preconstruction field activities, and are tailored specifically for the site and facilities. By necessity then, final review and approval by the staff of the design must rely in some part on information gathered during construction. Therefore, the review and approval can be accompli:r.ad in two ways: (1) design details of the permanent underdrain system, the operational monitoring program and plans for construction dewatering can be submitted in the PSAR, with only con-firmation of the details required prior to actual construction; or (2) conceptual designs of the permanent underdrain system and the operational monitoring program and details of construction dewatering can be submitted in the PSAR with the more complete review and approval based on construction dewatering requiring review and approval prior to actual construction. Review and approval of unique designs as post-CP matters is based upon 10 CFR Part 50, Subsections 35(b) and 55(e)(1)(iii). To prevent extending the review schedule, the first procedure would be the most desirable, but the staff recognizes that the detail required may not always be avail-able at the time the PSAR is subnitted. IV. Proposed Staff Position l We have reviewed and approved the design of a limited number of permanent dewatering However, because of the importance of these systems to plant safety, we have f systems. always required that they be designed and used in a conservative manner. The following is a r list of required design provisions which are consistent with requirements in recent CP reviews-(a) if the dewatering system is relied upon for any safety-related function, the system mst meet the appropriate criteria of Appendix A and Appendix B to 10 CFR Part 50. f In addition, guidance for structural, mechanical and electrical design criteria is i provided in related sections of the Standard Review Plan for Category I structures. systems and components. However, all portions of the system need not be designed to accomodate all design basis events, such as earthquakes and tornados, provided that such events cannot either influence the system, or that the consequences of failure from such events is not imortant to safety; nevertheless, a clear demonstration of l the effectiveness of a backup system and the timeliness of its implementation must l be provided; (b) the potential for localized pressures developing in areas which are not in contact with the drainage system, or in areas where pipes enter or exit the structural walls or mat foundations, must be considered. (c) uncertainty in detecting operational problems and providing a suitable monitoring system mst be considered; (d) the potential for piping fines and clogging of filter and drainage layers must be considered; Rev. 1 '.4.13-12 2157 232 4 e
+ e assurance must be provided that the system as proposed can be expected to reliably (e) perform its function during the lifetime of the plant; and (f) where the system is safety-related, is not totally redundant or is not designed for all design basis events, provide the bases for a technical specification to assure that in the event of system failure, necessary remedial action can be implemented before design basis conditions are exceeded. $AR's (5td. Format & Content Information, Sections 2.4 & 2.b) for each of the plants with V. permanent dewatering systems should include the following information: (a) Provide a description of the proposed dewatering system, including drawings showing the proposed locations of affected structures, components and features of the system., Provide information related to the geotechnical and hydrologic design of all system components such as interceptors, drainage blankets, and pervious fills with descrip-tions of material source, gradation limits, material properties, special construc-tien features, and placement and quality control measures. (Note structural, mechanical and electrical information needs described elsewhere.) Where the dewater-ing system is important to safety, provide a discussion of its expected functional reliability. The discussion of the bases for reliability should include comparisons of proposed systems and components with the performance of existing and comparable systems and components for applications under site co'nditions similar to those proposed. Where such information is unavailable or unfavorable, or the application (design and/or site) is unique, the unusual features of the design should be supported by additional tests and analyses to demonstrate the conservative nature of the design. In such cases the staff will meet with the applicant, on request, to establish the bases for such additional tests and analyses, Provide estimates, and their bases, for soil and rock permeabilities, total porosity, so) effective porosity (specific yield), storage coefficient and other related parameters' used in the design of the dewatering system. In general, these site parameters should be determined utilizing field and, if necessary, laboratory tests of materials representative of the entire area of influence of the expected drawdown of the system. Unless it can be substantiated that aquifer materials are essentially homogeneous, or that obviously conservative estimates have been used as design bases, provide pre-construction punging tests and other in-situ tests performed to estimate the pertinent hydrologic parameters of the aquifer. Monitoring of puwping rates and flow patterns during dewatering for the construction excavation is also necessary to verify assumed In design bases relating to such factors as permeability arjd acuifer continuity. addition, the final design of the system should be based on construction dewatering data and related otservations to assure that the values estimated from site exploration data are conservative. Lastly, t' final design of the dewatering system and its hydrologic and geotechnical operational monitoring program should be confirmed by construction excavation and dewatering information. 2.4.13-13 Rev. 1 I 71 r, cis/ 2 ], ],
If such infomation fails to support the conservatism of design infomation previously reviewed by the staff, the changed information should be reviewed under 10 CFR Part 50, Sobsections 35(t,) and 55(e)(1)(111). (c) Provide analyses and their bases for estimates of groundwater flow rates in the various parts of the permanent dewatering system, the area of influence of drawdown, and the shapes of phreatic surfaces to be expected during operation of the system. The extent of influence of the drawdown may be especially important if a natural or man-made water Dody affects, or is affected by, the dewatering systems. (d) Provide analyses, including their bases, to establish conservative estimates of the time available to mitigate the consequences of system degradation
- that could cause groundwater levels to exceed design bases. Document the measures that will be taken to either repair the system, or provide an alternate dewatering system that would beco=a operetional before the design basis groundwater level is exceeded.
(e) Provide both the design basis and nomal operation groundwater levels for safety-related structures, systems and components. The design basis groundwater 12 vel is defined as the maximum groundwater level used in the design analysis for dynamic or static loading conditions (whichever is being considered), and may be in excess of the elevation for which the underdrain system is designed for nomal oper tion. This level should consider abnormal and rare events (such as an occurrence of the Safe Shutdown Earthquake (SSE), a failure of a circulating water system pipe, or a single failure within the system), which can cause failure or overloading of the permanent dewatering system. (f) A single failure of a critical active feature or component must be postulated during any design basis event. Unless it can be documented that the potential consequences of the failure will not result in Regulatory Guides 1.26 and 1.29 dose guidelines being exceeded, either (1) document by pertinent analyses that groundwater level recovery times are sufficient to allow other foms of dewatering to be implemented before the design basis groundwater level is exceeded, discuss th_e measures to be implemented and equipment needed, and identify the amount of time required to accomplish each measure, or (2) design for all system components for all severe natural phenomena and events. For example, if the design basis groundwater level can be exceeded only as a result of a single nonseismically induced failure of any component or feature of the system, the staff may allow the design basis level of the dewatering system to be exceeded for a short period of time (say 2 or 3 days), provided that (1) effective alternate dewatering means can be implemented within this time period, or that (2) it can be shown that Regulatory Guides 1.26 and 1.29 guideline: will not be exceeded by groundwater induced impaiments of safety-related structures, systems, or components. t 'See (f) for considerations of differing system types. Rev. 1 2.4.13-14 t,t y q r llJ7 N N. ia _.-....v.--.-
a (g) Where appropriate, document the bases which assure the ability of the system to with-e stand various natural and accidental phenomena such as earthquakes, torvtadoes, surges, floods, and a single failure of a. component feature of the system (such as a failure of any cooling water pipes penetrating, or in close proximity to, the outside walls of safety-related buildings where the groundwater level is controlled by the system). An analysis of the consequences of pipe ruptures on the proposed underdrain system must be provided, and should include considerations of postulated breaks in the circulating system pipes at, in, or near the dewatering system building either inde-petidently of, or as a result of the SSE. Unless it can be documentea that the poten-tial consequences will not be serious enough to affect the safety of the plant to the extent that Regulatory Guides 1.26 and 1.29 guidelines could be exceeded, provide analyses to document that (1) water released from the pipe break cannot physically enter the dewatering system, or (2) if water enters the dewatering system, the system will not be overloaded by the increased flow such that the design basis groundwater level is subsequently exceeded. (h) State the maximian groundwater level the plant structures can tolerate under various significant loading conditions in the absence of the underdrain system. (1) Provide a description of the propos d groundwater level monitoring programs for dewatering during plant constructioi and for permanent dewatering during plant opera-Monitoring information requested includes (1) the general arrangement in plan tion. and profile with approximate elevation of piezameters and observation wells to be installed, (2) intended zone (s) of placement (3) type (s) of piezametea (closed or open system), (4) screens and filter gradation descriptions, (5) drawings showing typical I installation? showing limits of filter and seals, (6) observation schedules (initial and time intervals for subsequent readings), (7) plans for evaluation of recorded data, and (8) plans for alarm devices to assure sufficient time for initiation of corrective action. Provide a comitment to base the final design of the operational monitoring program on data gathered during the construction sonitoring program (if construction experience shows the assumed operationsl program bases to be nonconservative or impractical). Changes to the operational program are to be documented in the FSAR. (k) Provide infomation regarding the outlet flow monitoring program. The infomation required includes (1) the general location and type of flow measurement device (s), and (2) the observation plan and alarm procedure to identify anticipated high or low flow in the system and the condition of the effluen* - (1) For OL reviews, but only if not previously reviewed by the staff, provide (1) sub-stantiation of assumed design bases using information gathered during dewateriiig for construction excavation, rnd (2) all other details of the dewatering system design that implement design bases established during the CP review. I i (m) For OL reviews, provide a Technical Specification for periods when the dewatering An example system may be exposed to sources of water not considered in the design. of such a situation would be the extavation of surface seal material for repair of 2.4.13-15 Rev. I t 7cIC' (D 97c s/
,. e piping such that the underdrain would be exposed to direct surface runoff. In addi-tion, where the permanent dewatering system is safety related, is not completely redundant, or is not designed for all design basis events, provide the bases for a technical specification with action levels, the remedial work required and the esti-mated time that it will take to accomplish the work, the sources, types of equiprfnt and manpower required and the availability of the above under potentially adverse conditions. [See Section V(f)]. 2 : a l 9 ',/m r- . v s. covemo,r remn ernet t :, _ ;.. >. rf:., Rev. 1 2.4.13-1f l 4
~ 421-2 / l 421.0 QUALITY ASSURANCE BRANCH I 421.4 We find your response to question 421.1 acceptable. It is requested, however, that you incorporate this response into Section 17.0 on page 17.0-1 of the FSAR. 421.5 The QA Topical Report CE-1-A, Section 2.2, provides for exceptions or alternatives to the topical report, including Regulatory Guides and ANSI standards referenced therein to be identified in the SAR and that these exceptions and alternatives take precedence over the commitments in the topical. Since Appendix A of the Byron /Braidwood application addresses certain quality-related Regulatory Guides with such phrases as " complies with," "we believe we comply," or "is discussed in Section 17.0 of the FSAR" with no reference in Section 17.0 to this Appendix A, it is.not clear as to the degree to which you intend to comply with these documents. Accordingly, CECO should provide a specific commitment in Section 17.0 to comply with the regulatory positions in those Regulatory Guides and with the requirements of the ANSI standard addressed in our question 421.2. The commitment should identify the Regulatory Guides and ANSI standard by number, title, and revision number. Any exceptions, alternatives, or clarifications Ceco believes are warranted should be clearly identified with sufficient supporting detail to allow for NRC review and acceptance. If Appendix A of the FSAR is to describe exceptions or alternatives to these Regulatory Guides and ANSI standard, a reference to this effect should be provided in Section 17.0. 2157 237 m}}