ML19261A744
| ML19261A744 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/29/1979 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Stampley N MISSISSIPPI POWER & LIGHT CO. |
| References | |
| NUDOCS 7902080098 | |
| Download: ML19261A744 (33) | |
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JAT' 2 o Q Docket Nos:
50-416 and 50-417 Mr. N. L. Stampley, Vice President Production and Engineering Mississippi Power & Light Company P. O. Box 1640 Jackson, Mississippi 39205
Dear Mr. Stampley:
SUBJECT:
FIRST-ROUND RE0 VESTS FOR ADDITIONAL INFORMATION (GRAND GULF NUCLEAR STATION, UNITS 1 AND 2)
As a result of our revier of the information contained in the Final Safety Analysis Report f >r the Grand Gulf Nuclear Station, Units 1 and 2, we have developed the enclosed first-round requests for additional information. As suggested by ou. review schedule, a copy of which was forwarded to you by our letter dated December 8,1978, additional first-round requests are being developed by other review branches.
We will forward these additional requests as they become available.
In order tr. maintain our current review schedule, we request that you amend your Final Safety Analysis Report to reflect your responses to the enclosed requests by May 4, 1979.
If you cannot meet this date, please advise us as soon as possible so that we may consider the need to revise our review schedule.
Please contact us if you desire any discussion or clarificaticn of the enclosed requests.
- incerely, Q
J- ?
/
John F. Stolz, Chief L,ight Water Reactors Branch No. 1 livision of Project Management
Enclosure:
Requests e Additional
. - ^
In formi.. ion cc:
See next page 79020800'18
9 Mr. N. L. Stampley cc:
Mr. Robert B. McGehee, Attorney Wise, Carter, Child, Steen &
Caraway P. O. Box 651 Jr;kson, Mississippi 39205 Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Mr. Adrian Zaccaria, Project Engineer Grand Gulf Nuclear Station Bechtei power Corporation Gaithersburg, Maryland 20760 e
0
ENCLOSURE FIRST-ROUND REQUESTS FOR ADDITIONAL INFORMATION GRAND GULF NUCLEAR STATION UNITS 1 AND 2 DOCKET NOS. 50-416 AND 50-417
130.0 STRUCTURAL ENGINEERING 130.1 Since the response spectra curves as shown in Figures (3.7.1.1) 3.7.1 and 3.7.2 are in variance with those in Regulatory Guide 1.60, a comparison should be made between those shown in above mentioned figures and Regulatory Guide 1.60.
Also, provide a discussion on the effects of design as a result of the differences and their significances.
Specifically, provide the assurance that Grand Gulf Plant design is as safe as if criteria in Regulatory Guide 1.60 and 1.61 were used.
130.02 In Appendix 3A, in your discussion on conformance with (3.7.1.1.2) Regulatory Guide 1.60, it is indicated that for NSSS equipment the vertical design response spect as shown in Figure 1 attached to the discussion are used.
These response spectra are different from those as discussed in the last paragraph on Page 3.7-3.
Explain why different vertical design response spectra were used for the G.E.
NSSS equipment.
130.03 In section 3.7.1.2, it is stated that the response spectrum (3.7.1.2) generated from the synthetic time-history occasionally falls below the design response spectrum.
However, an examination of Figures 3.7-8, 3.7-13, 3.7-14, reveals that the generated response spectrum falls below the des'ign response spectrur at more than five points. Provide justification for using such a synthetic time-history.
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130.04 la the mathematical modes shown in Figure 3.7-18. the (3.7.2) enclosure building as shown in Figures 3.8-112 to 3.8-114 is not included structurally and only its weight is added to mass points 8 and 9.
Indicate the rationale for not including structurally the enclosure building in the mathematical model for the containment structure and describe how the enclosure building is seismically designed.
130.05 Your description of the analysis for items supported at two (3.7.2.1 2.5.2) or more elevations and for differential seismic movement of interconnected item is not clear.
Indicate if the procedure used is the same as that described in Section 5.3 of BC-TOP-4A revision 3 dated November 1974.
The procedure of analysis described in section 5.3 of BC-TOP-4A is acceptable to the staff.
If other procedure is used, provide justifiction.
130.06 In sections 3.8.1.4.2 and 3.8.1.5.2 it is stated that the (3.8.10) analysis and design of the liner plate was done in accordance with Bechtel Technical Reports BC-TOP-1 and BC-TOP-5. However, for the liner plate forming the boundary of the suppression pool, the analysis and design should be in accordance with the ASME Boiler and Pressure Code Section III Division 1 Sub-section NE to resist the SRV negative pressure. The anlysis should consider strength, buckling, and low cycle fatigue.
-3 130.07 On page 3.8-59 under Item i, reference is made to article (3.8.3.3.
2.1 )
CC-3000 of the ASME Code,Section III, Division 2.
- However, in your listings of codes and standards, it is not mentioned.
Indicate to which edition of this code you are referring.
130.08 Describe how the loads and combinations thereof occurring in
( 3.8.1 )
(3.8.3) the suppression pool and their feedback effects in other locations resulting from (1) LOCA air clearing, (2) steam condensation and chugging and (3) SRV attuations, are considered in the design of all structures, systems and components housed within the Reactor Building.
In particular, describe:
(a) How the load combinations and acceptance criteria have been modified to include the effects of these loads.
(b) The method of combining the loads within each load combination.
(c) The procedures u:ed to generate in-structure response spectra at critical locations, such as the RV supports, including the effects of soil-structure interaction, if any.
(d) The extent, if any, to which structures adjacent to the Reactor Building will experience the effects of these loads.
130.09 Describe the analytical and design techniques utilized to (3.8.3) determine the effect of annulus pressurization loads on the shield wall trounding the reactor vessel.
Include in the description of how these pressurization loads are combined with other coincident loads including seismic loads and feedback loads of LOCA and/or SRV loads occuring in the suppression pool.
_4 130.10 Describe the loads and loading combinations used to design (3.8.3.3.
2.1) the metal portion of the lower drywell wall.
130.11 The finite element model shown in Figure 3.8.53 consists 3.8.3.4.)
of solid elements of triangular 'or quadrilateral shapes and thin si.
i elements.
Since the nodal points of solid elements has fewer degrees of freedom than those of the thin shell elements, explain how you resolve this problem for the common nodal points at the juncture of the shell and the base mat.
130.12 You have divided the drywell structure into two parts:
(3.8.3.4.1) the lower drywell wall including the vents, and the upper drywell wall including drywell roof and upper pool walls.
Separate finite element models were utilized for analysis and design for the two parts. However, the boundary conditions used in these analyses were inconsistent (i.e.,
free at top of the lower wall and vertical support only at the bottom of the upper wall). Explain why the design of the drywell wall at the interface based on these analyses is adequate and conservative.
211-1 211.0 REACTOR SYSTEMS 211.1 With regard to the floor drain sump level monitoring systems, (5.2.5) discuss sump geometry, leakage flow rate mesurement accuracy, (7.6.1.4) measurement (monitoring) interval, and other information relevant (9.3.3) to demonstrating a sensitivity of 1 gpm per hour.
211.2 The drywell equipment drain sump receives two types of reactor (5.2.5) coolant leakage--hot and cold.
Leakage from " hot" sources such as the reactor vessel head flange, vent drain, and valve packings may flash into steam which must be condensed to reach the sump.
What assurance is there that the steam will be condensed for leak detection monitoring purposes? For leakage from " cold" sources, the floor drain system is employed.
Thus, the floor drain system should be tested periodically for blocked lines. Discuss the surveillance program planned to minimize the potential for drain system blockage.
211.3 In conformance with Regulatory Guide 1.45, the radioactivity (5.2.5) monitoring channels are stated to be qualified for operation (7.6.2.4) following an SSE. Confirm that all of the remaining leakage detection methods (systems) are qualified for operation following an OBE.
(This includes the drywell equipment and the floor drain sumps, sump coolers, and associated instrumentation and piping.)
211.4 With regard to the sensitivity and response times of the contain-(5.2.5) ment airborne radiation monitoring systems, provide a detailed (7.6.2.4) discussion on the capability of these monitors to detect a 1 gpm (12.3.4) leak in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for varying containment background activity levels.
The background activity levels should be considered for the plant containing fresh, irradiated, and permissible amounts of failed fuel, and the presence of normal expected leakage rates. Also, include the assumptions used in determining response times, such as the preset alarm level for higher background leakage and the plateout factor.
211-2 211.5 Expand your discussion on how the means of correlation with di-(7.6.2.4) verse monitoring methods will be used for leakage detection system operability verification and calibration.
For example, if a radioactivity monitoring systen is checked against the sump level and flow monitoring system, how is the latter system found to be acceptable for accuracy? Secondly, confirm that calibration and operability tests will be performed during plant operation and in compliance with IEEE Standard 279-1971.
211.6 Provide leakage detection system drawings fi-1090A and f4-10908.
(5.2.5) 211.7 Table 5.2-9 shows RCIC isolation function for the variable moni-(5.2.5.1) tor of reactor low water. Verify if this is an error.
211.8 You state that major components within the drywell which are (5.2.5) sources of leakage by nature of their design (e.g., sump seals, (7.6.1.4) valve stem packing, equipment warming drains) are contained and piped to an equipment drain sump and thereby identified.
Clarify the term " sump seals." Discuss what monitors are available to the operator to identify the source of leakage between such components as the sump (or pump) seals and the equipment warning drains and all other component sources drained to the drywell equipment drain tank. Clarify if there are any sumps within the drywell which must be filled before sump drain flow occurs to the equipment drain tank.
211-3 211.9 Explain the purpose of locating the 3/4-inch piping to leak (5.2.5) detection system downstream of the testable RHR check valves F041A, F041B, and F241 (see drawings M-1085A & B).
The same question also applies to the LPCS and HPCS designs (see drawings M-1086 and M-1087).
211.10 In confonnance to Regulatory Guide 1.45 you state that provisions (5.2.5) will be made to monitor systems connected to the RCPB for signs (7.6.2.4) of intersystem leakage. Provide a detail discussion that includes identification of all potential intersystem leakage paths (in-cluding detecting leakage from primary coolant system to the RHR and ECCS injection line) and the instrumentation used in each path to provide positive indication of intersystem leakage in the af fected systen.
211.11 BWR operating experience has shown that the HPCI and RCIC systems (5.2.5) have been rendered inoperable because of inadvertent leak de-tection isolations caused by equipment roon area high differ-ential temperature signal. The events occurred when there was a relatively sharp drop in outside temperature. As noted in Sec tion 5.4.6.1.1.1 and Table 5.2 9, Grand Gulf incoroorates this type of RCIC and RHR (steam) isolation.
Provide a discussion of the modifications that have been or will be made to prevent inadvertent isolations of this type which affect the availability and reliability of the RCIC and the RHR systems.
Secondly, provide the trip settings for isolation of the RHR and RCIC systens due to high area temperature in tenns of degrees above ambient temperature.
Also, discuss the method of specification that would be applied.
Show that the setting could not be set too low and cause in-advertent isolation when the system is needed.
211.12 Describe the provisions used for protection of the RCIC and (4.6) the RHR systems from cold weather in order to assure satis-(5.4.6) factory operational performance.
Also, in the assessment (5.4.7) include the standby liquid control and the control rod drive hydraulic systems and sources of water (e.g., CST standby service water) for all the above systems.
211-4 211.13 In Section 3.5.1.2.2 you state that pressurized components, (3.5.1.2) namely, pressure vessels and pressurized bottles containing noncondensible gases with an operating pressure at or above 100 psig have been evaluated as potential sources of missiles.
Provide this evaluation which should include the CRD scram nitrogen bottles and the safety / relief valve air accumulators acting as potential missiles. Describe protective barriers available to prevent missiles from striking nearby safety systens or components.
211.14 Discuss the potential for missiles inside the containment due to (3.5.1.2) gravitational effects (of such components as electrical hoists or any unrestrained equipment) during maintenance times, reactor operation, and following a LOCA.
211.15 With regard to rotating component failure missiles, show by (3.5.1.2) analysis that the impeller fragments resulting from recirculation pump overspeed condition during a LOCA will not penetrate the pump case.
Secondly, provide or reference a study that shows the probability for significant damage to occur within the con-tainment from inpeller missiles being ejected out the open end of the broken pipe is acceptably low.
If a similar study for another plant is to be referenced, justify its appropriateness to your plant design.
211.16 Based on the review of nuclear power plant piping system design (3.5.1.2) integrity, past history has shown several failures of safety valve headers resulting in the valves becoming missiles (NUREG-0307).
Since you address only the credibility of valve bonnets and stens, justify why the safety valve header and valve is not considered as a credible missile.
Also, your statement that bonnet ejection is highly improbable and not considered credible missiles for valves of ANSI 900 psig rating and above is not supported.
Show that should a large valve component become a missile, containment penetration would not occur.
Discuss protection, such as equipment separation and redundancy, to preclude damage to the systems necessary to achieve and maintain a safe plant shutdown.
211-5 211.17 Provide information demonstrating that loss of the operating CR0
'4.6.1) pump at low reactor pressure (less than 500 psig) will not result in accumulator depressurization and loss of scram capability.
If the accumulator check valves leak following loss of the operating CRD pump, provide estimated time and basis before reactor scram capability becomes narginal. Also, present a testing program or procedure that would assure that operation of these check valves is acceptable over the plant lifetime.
211.18 We note in your response to question 110.2 regarding the concern (4.6.1) of cracking of the CRD return nozzles that you intend to fix this oroblem by cutting and capping the CRD return line.
Discuss the impact of this modification on the plant.
In oarticular, include information covering, but not limited to, the following areas:
(1) Compare reactor vessel makeup capability for one and two CRD pump operation before and after the proposed modifi-cation.
Commit to preoperational testing to verify the modified flow capability.
(2)
Commit to preoperational testing to verify individual per-formance of modified CRD components and other aspects of the CRD systen potentially affected by the cut and capped CRD return line (equalizing valves, filters, scram times, settling function, etc.).
(3)
Should new equalizing valves be added, discuss the potential lifetime effect on drive speeds; in particular, evaluate the viilnerability of the CRD system to a voiding of the drive exhaust header after a single failure.
(4) Evaluate the lifetime effect of the added flow through such conponents as the drive exhaust header and stablilizing lines; in particular, discuss the increased potential of corrosion products from carbon steel piping to deposit additional foreign matter in the drives.
(5) Discuss the potential for, and effect on, flow reversal through the directional control solenoid valve over the plant liftime.
(6) Discuss the expected effect of the CRD nodifications on theA P settling function across drives to ensure latching after withdrawal.
211-6 211.19 Appendices G and H of the La Salle and Zinner FSARs, respectively (5.4.1) provide information on the recirculation flow control systen.
(15.0)
State whether this information is applicable to Grand Gulf, If applicable, it should be referenced; otherwise, a comparable section for Grand Gulf should be provided.
Also, provide the following information:
(1) Justify the 8 degree subcooling limitation in operating the recirculation pump.
(2) Secondly, you state that if the subcooling falls below 8 degrees Fahrenheit, the 60 Hz power supply is tripped to the 15 Hz power source to prevent cavitation of recirculation pump, jet pumps, and/or the flow control valve.
The above conditions appear to initiate a two-pump trip transient.
Is the pump coastdown rate result-ing from the above condition more severe than the one used in the Chapter 15 transient analysis?
If so, then reanalyze pump trip transient with the more severe pump coastdown rate.
Also, describe the consequences of a sudden increase in recirculation pump speed (possibly due to an increase in the frequency of the power supply).
211.20 Confirm whether Figure 5.4-5 (the reference flow control valve (5.4.1) characteristic) is applicable for closing and apening conditions.
211.21 Provide assurance that the essential portions of the control (3.5) rod drive system, namely, the 1-inch supply and return oiping (4.6.1) located inside the containment (see Figures 3.6A-15, -19, and
-20) are protected from the effects of high or moderate energy line breaks such as the high pressure core spray system, or high pressure core injection feedwater system, reactor coolant pressure boundary, etc.
The concern is whether pipe whip and/or jet impingement can impair the capability to scram.
In addition to the above requested evaluation, assess damage to the cluster of CRD return and supply lines, and scram capability by postulating rupture of a single CRD supply or return line.
211-7 211.22 Per Table 1.3-8, the Grand Gulf Nuclear Station will incorporate (4.6.1.1) the " fast scram" control cod drive (FSCRD) system as it was proposed on the GESSAR docket. Provide any new available information on the FSCRD qualification test programs.
Specifi-cally, the status of field evaluation--where this unit has been installed and monitored in Peach Botton Unit 3, manufacturing qualifications, and production tests. Discuss how these latest results compare to those obtained from previous acceptance testing.
211.23 Note 4 in Figure 4.6-10 states that the charging header shall (4.6.1.1) not exceed 1510 psig and any pressure in excess of this value will damage the CRD during a scram.
However, in Section 4.6.1.1.2.4.1 you state that an accumulator hydraulic charging pressure of approxinately 1750 to 2000 psig is required.
Since the charging header is connected to the hydraulic control units, clarify whether there is a piping design pressure change within and down-stream of these units.
If so, provide the interface location between the high and low pressure boundaries.
If not, correct or discuss the discrepancy between the above two pressure values and how it affects the CRD design.
211.24 Per the Standard Review Plan 4.6 and Regulatory Guide 1.70, (4.6.1)
Rev. 2, information for combined performance and evaluation of reactivity systems is required.
In particular, address the vulnerability of the reactivity control systems (control rod drive and the standby liquid control systems) to common mode failures.
211.25 The RHR system shall be capable of bringing the reactor to a (5.4.7) cold shutdown using only safety-grade systems.
Confirm that this requirement is met.
Include in your assessment the air supply system used to operate the RCIC steam and condensate control valves located at the RHR heat exchanger when the RHR system is in the steam condensing mode.
211-8 211.26 The RHR system shall be capable of bringing the reactor to a (5.4.7) cold shutdowa with only onsite or offsite power available and with the most limiting single fa!1ure.
Describe your plans for tr. Sting the alternate shutdown cooling modes of operation.
Demonstrate that adequate passage of water through the safety /
, relief valves can be achieved and maintained when the alternate method is in use.
Include the quantity of air supplied, the source, and the time before the air is exhausted.
211.27 During the shutdown cooling moae, the " flush water" valves (5.4.7.2) are opened and closed outside the control room.
Specifically identify the operated local flush water valves and the source of flush water.
Discuss the consequences assuming the operator would omit this procedure and/or forget to close a local flush water valve and continue shutdown operations.
211.28 In Section 5.4.7.1.3 you identify the RHR relief valves and the (5.4.7.1)
RHR design pressure used as the sizing basis.
Expand your discussion by providing the relief valve capacity, nominal set points, set point tolerance, and ASME class rating of the valves and lines.
In addition, discuss the vulnerability of the RHR system to malfunctions which could result in overpressurization of low pressure piping. Support your evaluation by providing an outline of all operating procedures required to bring the plant to a cold shutdown condition from hot standby and procedures for plant startup from cold shutdown.
211.29 In section 5.4.7.2.1, you state that RHR pump start is prevented (5.4.7.2) when the suction valve (s) are not open.
Confirm that RHR pump trip would also occur assuming the pump is operating and one or both suction valves 'nadvertently close.
Show that the above punp trip feature does not compromise the ECCS function.
211,30 Provide more detailed information regarding the actuation of (5.4.7.1) the automatic minimum flow valves used for RHR pump protection against damage from a closed discharge valve.
For example, specify flow rate quantities that signal minimum flow valve opening and closure on low main line flow and high main RHR line flow, respectively.
Also, state whether the control system meets IEEE-279 standards.
Confirm that the minimum flow line valve restrictors are designed to safety-grade standards (e.g., seismic Category I, ASME Code Section III).
211-9 211.31 Per Table 5.4-3, the RHR isolation valves F008 and F009 are (5.4.7) signaled to close on reactor low water level.
Clarify whether this valve isolation signal is based on the same signal as the RHR pump actuation in the LPCI mode, which is a water level of 1.0 foot above the active core.
If not, provide vessel water level that :solates the RHR suction valves and show that core cooling can be maintained assuming a pipe break outside the containment.
Hence, provide the following additional information assuming a pipe break o".tside containment in the RHR system when the plant is in a shutdown cooling mode:
(1)
Identification of systems available for maintaining core cooling.
(2) Maximum discharge rate resulting from the break and the time frame available for recovery based on the discharge rate and its effect on core cooling.
(3)
Identify the alarms available to alert the operator to the. event, assurance that recovery procedures are available, and show that adequate time is available for operator action.
(4) Following the moderate energy line break, single failure criterion should be applied consistent with SRP 3.6.1 and BTP APCSB 3-1.
211.32 Discuss systen design provisions to prevent damage to the RHR (5.4.7)
(LPCI) pumps against pump runout conditions during ECCd and test modes of operation.
211.33 Provide a more detailed description and location of the RHR (5.4.7) pump suction strainer inside the suppression pool.
Include pipe bends and the minimum height of the suppression pool water level above the suction strainer.
Show that the NPSH at the center line of the RHR pump will be met at the pump's design condition as well as at the most limiting operating co.1dition.
Also, discuss the size of particles that could pass through the strainer and continue to the RHR pump passages.
How much material blockage would it take to significantly affect RHR pump suction flow from the suppression pool following a LOCA?
211-10 211.34 Provide pressure interlock set points used in the prevention (5.4.7.1) of opening the RHR isolation valves F008 and F009 to the low pressure suction piping, and for the initiation of valve closure on increasing reactor pressure.
211.35 Confirm that all valves performing an isolation function between (5.4.7) the high pressure and low pressure boundary in the RHR system (e.g., check valves and motor-operated valves) meet the leak testing and inspection requirements of the ASME Section XI code for Category A valves. A combination of two or more check or motor-operated valves in series should have design provision for individual leak testing of any two valves.
211.36 Commit to providing a means for pressure relief between the two RHR isolation valves F008 and F009 or show by analysis that piping integrity would be maintained assuming a LOCA or steam line break would occur cnd the trapped water between the valves would thermally expand.
211.37 Operation of the RHR system in the steam condensing mode (5.4.7) involves partial draining of one or both RHR heat exchangers and introduction of reactor steam into initially cold lines and heat exchangers. Describe the methods (e.g., valve operation, air introduction, etc.) and provisions to be used to prevent occurrence of water hammer during the initiation of operation in this mode, and the change to the pool cooling mode. When the RHR is used in the steam condensing mode with one or both heat exchangers, can the. jockey pump system fill the lines to the injection valve in the core spray and RHR lines?
If not, what procedures would be used to prevent water hammer following startup of the core spray or RHR pumps?
Pressure relief valves and lines designed to prevent overpres-surization of the RHR system are routed outside containment before being returned to the suppression pool.
Discuss design provisions made to mitigate possible water hammer in these lines.
Secondly, confirm that these relief lines are capable of taking the seismic and dynamic blowdown loads without loss of piping integrity.
211. 38 Discuss the procedures for minimizing the potential for exceed-(5.4.7) ing the allowable cooldown rate (greater than 100 degrees Fahrenheit / hour) of the RHR and the reactor coolant system when placing the plant in a shutdown cooling mode following planned normal conditions or an emergency.
211-11 211.39 Discuss the RHR pump reliability for long-term operation.
Long-(5.4.7) term reliability should be demonstrated by either operational experience or testing.
If previous operational experience should be cited as the basis for qualifying the pumps, state any pump design differences and conditions of previous pump operations.
211.40 Confirm that the RCIC electro-hydraulic system integrated with (5.4.6) the turbine governing valve is of a safety-grade design (e.g.,
seismic Category I).
211.41 Provide an RCIC pump performance curve that depicts flow rate (5.4.6) versus reactor vessel pressure.
Also, identify the most limiting operating condition and specify the hPSH margin under this condition.
211.42 It appears that it is possible for some steam condensate to (5.4.6) remain in the lines leading to the RCIC steam turbine.
(This o: curs when t.he steam isolation vahes would be temporarily closed for maintenance.) Discuss whether the amount of liquid can cause daraage to the RCIC turbine so that the system is incapable of ielivering water to the reactor vessel as required. Also, describe the design modifications you propose to prevent water hammer effects at the turbine exhaust.
211.43 An isolation signal closes a number of valves in the RCIC system.
(5.4.6.1)
In particular, the affected valves are F063 and F064 located inside and outside containment, branched off the main steam line.
However, the P&ID shows that these valves are keylocked open.
Justify this apparent discrepancy and evaluate the consequences of a postulated pipe break downstream of the first or second isolation valve for steam flow rates less than or greater than the 30') percent of the steady-state steam flow indicated in this s(ction.
211-12 211.44 The FSAR states that the accumulator sizing for the power-(5.2.2) operated relief valves is sufficient for one actuation; and (5.4.7) for the automatic depressurization system (ADS) valves it is (6.3) sufficient for two actuations. A "noninterruptible" safety-grade source of air for the ADS valves is required to terminate certain postulated transient and accident events without loss of the ADS function.
Show that an adequate supply of air will exist to operate the ADS valves for the following conditions:
(1) the alternate method of achieving and maintaining a cold shutdown following a loss of offsite power with a worst single failure in the RHR system; (2) for a small LOCA with failure of high pressure ECCS where the ADS valves would be used for reactor vessel depres-surization and maintaining long-term cooling.
Include a discussion on procedures to be used to replenish coolant inventory; and (3) for a snall steam line break disabling the RCIC concurrent with a single failure of the HPCS that would require ADS function to depressurize the reactor vessel.
Consider the air supply needs for long-term cooling (e.g., how would reactor vessel inventory be maintained when decay heat repressurizes the vessel above the shutoff head of the low pressure cooling system?).
371.0 HYDROLOGIC ENGINEERING 371.6 Prohideadetailedtop-of-roadprofilefortheaccessroadwhich crossesculYerts#1and#9. Theherticalextentofthepro-fileshouldbeatleasttoeleYation140feetmsl.
371.7 ProYideanalysesof,andthebasesfor,maximumwaterleYelsof (RSP) 128.5atculYert#9and127.7atculYert#1.
Include your assumptionsof,andbasesfc,coefficientsforflowsoYerroad-ways,entrancelosses, Manning's'n'haluesorfriction-coefficients forculherts,andanyotherassumptionsusedincomputingthe applicablewatersurfaceeleYation.
ProYidedetaileddrawingsofculYerts#1and#9inbothplanand section. The drawings should clearly show the type of inlet, projection from the roadway fill, stilling basin details, apron details, and other pertinent features.
DuetotheapparentimportanceofthesecuTYertstoth'efloodleYel
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at the site and thus to the safety of the plant, it is our position thattheeffectsofblockageoftheseculYertsbeconsideredin their design. Accordingly,proYidetheresultsofanalysesshowing thesensitiYityoffloodleYelstoblockage. UseseYeralblockage modes.
Inaddition,discussandproYidethedetailsofanymeasures i
which would be taken to minimize the transport of debris during f
flood conditions. StaffexperiencewithdebriscollectiondeYices, such as a V-shaped racks placed some distance upstream to intercept floating debris and preEant further downstrean transport, has beenfaYorable. Alternately,oroYideanalysestodocumentthat debrisblockagewillno'tadYerselyeffecttheparticularchannel and culvert design used. The staff notes that highway culverts andembankmentshaYehistoricallycollapsedduringfloodsand subseqbentsaturationoftheupstreamsideofthefill. The
~. _.. _. _ _.
collapses have led to blockage of flow through the culverts.
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Accordingly, document that the 12' and 15' diameter culverts and embankment are capable of resisting the loads imposed on them by the PMF from both streams A and B.
Alternately, assume complete blockage of the culverts and recompute the flow profiles over the accessroadandupstreaminthehicinityoftheplantandusethis leYelasthedesignbasisfloodleYe1forsa.fetyrelatedbuilding, structures and components.
371.8 Provide analyses to document the computation of water levels in theYicinityoftheplantduettoanoccbrrenceofthePMF.
It is notclear,forexample,howa,waterleYelof132.7wasderiYed for Area "H" or Area "L" as shown on Figure 2.4-7A.
If flow is assumed to occur'over the railroad tracks, provide detailedtop-of-railprofilplsateachsubarea.
Provide the weir coefficients used for flow over the railroad tracks.
For each subarea, docur_at that the railroad tracks are, in fact, the control location for flow and that control does not occur at some downstream location with resulting backwater effect at the tracks.
ProYidefurtherbasesforyobrcompbtationoftheflowprofile 371.9 shown on Figure 2,4-16.
Provide additional information regarding
'n' values used, expansion and contraction coefficients, method of computation (e.g., standard tepb$ckwater),andotherdetailsof the analyses.
371.10 ProYideadditionalcross-sectionsforbothStreamsAandB. The-sections should be taken upstream of the access road, beginning atculYerts#1and#9,andshouldbespacedabout100'feetapart foradistanceofabout1000feetbpstream.Ifcontroloccbrsdownstream of the culvert, provide downstream cross-sections, also.,
371.11 Provide additional details of ditches constructed to carry'the PMPrunoffinthehicinityoftheplant. ProYideinformation regarding the profile, composition, size, and drainage area at various location along the ditches. For example, the paved ditch which carries runoff around the Unit 2 cooling tower should be defined more clearly ~so that its flood-carrying capability can be Itis lnotimmediatelyobvious,forexample,what assessed.
I dischargeshobldbeusedandwhattheresultingdepthofflow wobidbeatthepointwheretheditchelevationis1Ei.0,since the ditch is apparently receving runoff from several drainage subareas.
371.12 ProYidethedetailsof,andbases.for,theerosionprotectiontobe
'placedaboYethe-concreteliningonStream8,asshownon Figure 2.4-16.
ProYide the median riprap size (D50) ad layer gradation, and the design criteria for its selection-OnFigure2.4-16,yobindicateanapparenthydrablicjumpinthe 371.13 vicinityofcblYerts#5. ProYidethedetailsofthecomputation of the jump and its character (e.g., length, energy dissipated, location). Document that the channel and its associated erosion protection _is_ capable._of_r.esisting a jump at this location.
371.14 ProYidetheanalyticalmodelusedtodeterminethewaterlevel and the time required for ground water to risewhich you used in theassessmentoftherbptbreofthecircblatingwatersystempiping (seeresponsetoQbestion371.3). ProYidethebasesforUsinga permeability of 200 gpd/ft. ProYidethebasesfordeterminationof 2
a t me of at least sehen days for ground water to rise.
(Please note that a permeability of less than 200 ft/ week does not necessarily implyaresponsetimeofatleastaweekforastrbcture2b0ftaway if the effective hydraelic gradient is greater than one, such as in the case of a pipe break.)
371.15 ProYide detailed analyses,and bases therefor,to document that rainfall in the plant area and recharge areas will not significantly affectgroundwaterleYels. Theanalysesshobldincludeconsideration of rainfall and infiltration rates greater than those observed in theshortperiodthatobserYationswellshaYebeenemployed (since1972).
If credit is taken for reduced infiltration by the clay seal layer, document the adequacy of the clay seal layer if the layer is in a degraded (e.g., cracked, dessicated) condition.
Alternately,discusstheconditionsorprocedurestobebsedto oreclude such degradation.
Describe in detail the groundwater leYel monitoring that will be 371.16 Includeinyobrdiscbssionthe performeddbringplantopwa' wells or boreholes to be used, their location, depth, frequency of sampTing, and other related-parameters.
371.17 You state that pumping of wells may be required to lower groundwater levels. Provide the technical specifications to be employed if the designbasisgroundwaterleYelisexceeded. ProYidethebases forselectionofwells,fregbencyofmonitoring,methodsof monitoring, etc.
The staff's position on dewatering systems is identified in the attached branch position.
371.18.
Provide a detailed description of the models used to analyze the (9.2.5) ultimate heat sink cooling tower performance which is shown in Tables 9.2.5 and 9.2.6.
371.19 We require that you meet the criteria suggested in Regulatory Guide (RSP)
(9.2.5) 1.27, Rev. 2, January 1976, with respect to the transient analysis
_. of water-supply-and temperature The_ guide specifically states,
_.. ~ ~_
"The above analysis related to the 30-day cooling supply and the excesstemperatureshouldincIudesufficientinformationto.
sbbstantiatetheassbmptionsandanalyticalmethodsused. TMs infor;ation should incibde actbal performance data for a similar coolingmethodoperatingbnderloadnearthespecificdesign conditions,orjustificationthatconserYativedriftlossandheat transferYalueshavebeenused."
Ifyobrcoolingtowerdesignor
/
specifications-hahe noL prugre.>>ed-to the stage where_a predictive modelcanbedehelopedandYerifiedbyhigh-qbalityperformancedata from existing towers of similar size and type, then you should commit to a pre-oper;ational testing program to validate the model and tower performance. Stateyobrintentwithregardtothisposition.
371.20.
Prohide'thebasesforyourestimateofmaximumevaporrivelossof (9.2.5) 11,000,000 gallons over the 30-day period.
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.I C.
1r
- P BRANCH TECHNICAL POSITIONS HMB/GSB 1 SAFETY-RELATED PERMANENT DEWATERING SYSTEMS I.
Surrna ry This positic, has been fomulated to minimize review probless cocoon to pemanent dewater-ing systems tnat 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 necenary to support safety-related functions is defined. This position applies 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 criteria.
II. Background The staff has reviewed a number of pemanent dewatering systems, including McGuire 1 & 2 Cherokee 1 & 2 Perkins 1 & 2, P_erry 1 & 2 WPPSS 3 & 5, Douglas Point 1 & 2, and Catawba 1
& 2.
Perry, beginning in 1975, was the first plant reviewed with such systems, and was reviewed very late in.the CP process. Only WPPSS 3 & 5 and Douglas Point use a passive
(
systen (no pumps).
Pemanent dewatering systems lower groundwater levels to reduce subsurface water loads on plant structures. In addition, they can increase plant operational dependability and reduce costs. These effects are accomplished by providing added means of keeping seepage water out of lower building levels during the later stages of plant life when nomal water-
. proofing provisions may have deteriorated, and reducing radwaste sy"os operating costs by minimizing.the amount of drain water that must be *reated. BenefiS 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-related building areas, which 'underdrains are expected to minimize, is estimated to be in the range of $100K to $200K per year per rei.ctor. The construction costs of a*.ternati:!s to underdrains for structural purposes alone (exclusive of inleakage treatment) is estimated to range upward from $300K per unit and is highly dependent on site conditions. Structural alternatives to pemanent underdrains include additional concrete and steel in the lower portions of buildings, and the use of anchor systems to resist floatation.
Dewatering systems are generally composed of three components; the collector system, the drain system, and the discharge systen. Water is first collected in collector drains F
2.4.13-g Rey. I k
adjacent to buildings or excavations. Interceptor drains or piping are then used to ccnvey this water to a final 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 t such as reactor or auxiliary buildings) to discharge collected water. Finally, permanent 4 atering systems are not a required feature at any plant, but may be proposed as a cost effective feature.
Many pemanent 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 tnan the probability of occurrence of a nuclear power plant design basis event such as a Probable Maximum Hurricane, 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 have concluded that safety analyses of such systems should consid&r 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 imediately following a severe earthquake, or if the system can be expected to perfom an underdrain function in a degraded condition. Certain portions of such systems, however, may be required to regularly perform other safety functions (e.g., porous concrete base mats) and should be designed for severe earthquakes. Failure of a dewatering system could cause groundwater levels to rise above design levels, resulting in overic,eding concrete walls and mats not designed to withstand the resulting hydrostatic In addition to causing potential structural and equipment damage, groundwater pressures.
could enter safety-related buildings.and flood components 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 corresponding degree to which the safety of the structures and systems rely on the integrity of the dewatering sy' stem, particularly with a dewattring system in a degraded situation. For example, if structures can accomodate hydrostatic 4 oads that would result with a total failure of 1
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, however, 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.
asstaning degradation of the underdrain system, many of the staff's concerns may be eliminated from further consideration because of the time available for remedial action after detection of system degradation.
Rev. 1 2.4.13-10
III. Situations Identified During ' Previous Reviews Four generai categories of situations have been identified during case reviews as follows:
e (a) Estimating and Confiming Permeability Yalues 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 pemeability 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 by foundation excavation. In addition, geological and foundation conditions that were not detected in site explorations may affect flow conditions and cause the estimated permeability values and flow regimes to be substantially different from those assumed These conditions are often first detected during at the CP preliminary design stage.
s Therefore, we have required a corrittment to consider con-construction dewatering.
struction excavation and dewatering data in the final design of underdrain systems.
(See situation (d) below.)
(b) Operational Monitoring 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 failures, and contingency Since measures for these failures must be well defined prior to plant operation.
drain systems are usually buried and concealed and there may be no direct way of inspecting them, reliance must be placed on piezometers, 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 underdra'" 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 monitoring; and (c) competent evaluation of observations during both construction and operation. In addition, the bases for acceptable contingency measures suitable for coping with various possible hazards must be established at the CP stage.
(c) Pipe Breaks A dewatering system might be overloaded by such conditions as leaks or breaks in eithe.- 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 d ains, pipes and pumps in the dewatering For example, a complete failure of cir:ulating water system piping has been system.
This requirement required in the design of the dewatering systems reviewed to date.
was made to assure that such abnomal occurrences do not adversely affect the integ-rity of safety-related structures, systems, and components.
(d) Seauence of Review Underdrain systems are usually one of the first items constructed and, after back-filling and construction of subsurface facilities, are then no longer visible for Rev.1 2.4.13.11
[h
regular inspection. In most cases, these systems are in'itially designed based on rather limited information from preconstruction field activities, and are tailored specifically for the site and fatilities. By necessity then, final review and approval by the staff of the design must rely in some part on infomation gathered during construction. Therefore, the review and approval can be accomplished in two ways:
(1) design details of the pennanent underdrain system, tha operational monitoring program and plans for construction dewater,ing 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 dewateri. 3 requiring. review and approval prior to actual construction. Review and approval i J unique designs as post-CP setters is based upon 10 CFR Part 50 Subsections 35 b) and 55(e)(1)(iii). To t
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 submitted.
IV. Proposed Staff Position We have reviewed and approved the design of a limited number of pemanent dewatering systems, However, because of the impc-tance of these systems to plant safety, we have always required that they be designed and used in a conservative manner. The following is a list of required design provisions which are consistent with requirements in recent CP reviews:
(a) if the dewatering s;. tem is relied upon for any safety-related function, the system must meet the appropriate criteria of Appendix A and Appendix B to 10 CFR Part 50.
In addition, guidance for structural, meenanical and electrical design criteria is 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 accommodate 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 important to safety; nevertheless, a clear demonstration of the effectiveness of a backup system and the timeliness '
- its imolementation must 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 must be considered; (d) the potential for piping fines and clogging of filter and drainage layers must be considered; Rey, 1 2.4.13-12
(e) assurance must be provided that the system as proposed can be expected to reliably perfom 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, n:cessary remedial action can be implemented before design basis conditions are exceeded.
SAR's (Std. Format & Content Information, Sections 2.4 & 2.b) for each of the plants with V.
permanent dewatering systems should include the following infomation:
(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-tion features, and placement and quality control measures. (Note structural, mechanical and electrical information rr - 5 described elsewhere.) Where the dewater-ing system is important to safety, e a discussion of its expected functional reliability. The discussion of tb
..ses for reliability should include comparisons of proposed systems and componen*
'th the perfomance of existing and comparable systems and components for appli.uions under site conditions similar to those proposed.
Where such inforniation 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 aaalyses to demonstrate the conservative ns 4re 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.
(b) Provide estimates, and their bases, for soil and rock permeabilities, total porosity.
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 detemined 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 pumping tests and other in-situ tests performed to estimate the pertinent hydrologic parameters of the aquifer. Monitoring of pumping rates and flow patterns during dewatering for the construction excavation is also necessary to verify assumed In design bases relating tn such factors as permeability and aquifer continuity.
addition, the final design of the system should be based on construction dewatering data and related observations to assure that the values estimated from site exploration data are conservative. Lastly, the 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
If such information fails to support the conservatirm'of design information previously reviewed by the staff, the changed information should be reviewed under 10 CFR Pa 50, Subsections 35(b) and 55(e)(1)(iii).
(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 body affects, or is affacted 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 in alternate dewatering system that would become operational before the design basis groundwater level is exceeded.
(e) Provide both the design basis and normal operation groundwater levels for safety-related structures, systems and components. The design basis groundwater level 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 nonnal operation. 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 analyset that groundwater level recovery times are sufficient to allow other forms of dewatering to be implemented before the design basis groundwater level is exceeded, discuss the 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 guidelines will not be exceeded by groundwater induced impairments of safety-related structures, systems, or components.
- See (f) for considerations of differing system types.
Rev. I 2.4.13-14 a
Where appropriate, docu[nent the bases wuich assure the ability of the system to with-(g) stand various natural and accidental phenomena such as earthquakes, tornadoes, 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-Unless it can be documented that the poten-pendently of, or as a result of the SSE.
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 documert 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 s
level is subsequently exceeded.
State the maximum groundwater level the plant structures can tolerate under various (h) significant loading conditions in the absence of the underdrain system.
Provide a description of the proposed groundwater level monitoring programs for (1) dewatering during plant construction and for permanent dewatering during plant opera-lionitoring infomation requested includes (1) the general arrangement in plan tion.
and profile with approximate elevation of piezometers and observation wells to be installed, (2) intended zone (s) of placement, (3) type (s) of piezometer (closed or open system), (4) screens and filter gradation descriptions, (5) drawings showing typical
(
installations showing limits of filter and seals, (6) observation schedules (irtitial and time intervals for subsequent readings), (7) plans for evaluation of recorded data, and (8) pleas for alam devices to assure sufficient time for initiation of corrective Provide a cocnitment to base the final design of the operational monitoring action.
program on data gathered during the construction monitoring program (if construction experience shows the 'ssumed operational program bases to be nonconservative or to the operational program are-to be documented in the FSAR.
impractical). Char The infomation (k) Provide infomation regarding the outlet flow monitoring program.
required includes (1) the general location and type of flow measurement device (s),
and (2) the observation plan and alarm procedure to identify unanticipated high or low flow in the system and the condition of the effluent.
For OL reviews, but only if not previously reviewed by the staff, previde (1) sub-(1) stantiatinn of assumed design bases using information gathered durirg dewatering for construction excavation, and (2) all other details of the dewatering system design that implement design bases established during the CP review.
For OL reviews, provide a Technical Specification for periods when the dewatering (m)
An example system may be exposed to sources of water not considered in the design.
I of such a situation would be the excavation of surface seal material for repair of 2.4.13-15 Rev. 1
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-5 mated time that it will take to acccmplish the work, the sources, types of equipment and manpower required and the availability of the above under potentially adverse conditions. [See Section V(f)].
e U. S. GOVE1tNMOET PapeTU60 OFFICE : 1973 720 137/277 Rev. 1 2.4.13-16