ML20055B327

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Requests Addl Info Re Containment Sys Branch Review of Fsar. Schedule for Response Needed within 7 Working Days from Receipt of Ltr
ML20055B327
Person / Time
Site: Limerick  Constellation icon.png
Issue date: 07/09/1982
From: Schwencer A
Office of Nuclear Reactor Regulation
To: Bauer E
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
References
NUDOCS 8207210380
Download: ML20055B327 (30)
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JUI. 9 12 DISTRIBUTION:

Docket File bcc: NRC PDR LB#2 File Local PDR HAbelson NSIC EHylton PRC Deckt.c:Nos. :~ ; 50-352/3531/ ACRS (16)

Region I I&E Slewis, OELD Mr. Edward G. Bauer, Jr. F. El tawila Vice President & General Counsel Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101

Dear Mr. Bauer:

Subject; Request for Additional Information - Limerick (Containment Systems)

The Containment Systems Branch has reviewed the appropriate sections of the FSAR and the DAR for the Limerick Generating Station. This review has indicated a need for the additional information delineated in Enclosure 1.

To expedite the review, we would like to meet with marbers of your staff in the near future to discuss the positions of the Containment Systems Branch with regard to this infonnation request.

Please provide us, within 7 working days from receipt of this letter, with the date(s) on which you plan to respond to the above. Any questions concerning this infonnation request should be directed to Dr. Harvey Abelson, (301) 492-9774, the Licensing Project fianager.

Sincerely, A. Schwencer, Chief i Licensing Branch No. 2 3 Division of Licensing L

Enclosure:

As stated i cc: See next page $

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, ho Lime rick Mr. Edward G. Bauer, Jr.

Vice President & General Counsel Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 -

cc' : Troy B. Conner, Jr. , Esquire Mr. Marvin I. Lewis Conner and Wetterhahn 6504 Bradford Terrace 1747 Pennsylvania Avenue, N. W. Philadelphia, PA 19149 Washington, D. C. 20006 Frank R. Romano, Chainnan Mr. Robert W. Adler Air & Water Pollution Patrol Assistant Counsel 61 Forest Avenue Commonwealth of Pennsylvania, DER Ambler, PA 19002 505 Executive House P. O. Box 2357 Charles W. Elliott, Esquire Harrisburg, Pennsylvania 17120 Thomas & Hair 123 North Fifth Street Honorable Lawrence Coughlin Allentown, PA 18102 House of Representatives Congress of the United States Judith A. Dorsey, Esquire Washington, D. C. 20515 Limerick Ecology Action 1315 Walnut Street, Suite 1632 Roger B. Reynolds, Jr., Esquire Philadelphia, PA 19107 '

324 Swede Street Norristown, Pennsylvania 19401 Mr. Karl Abraham Public Affairs Officer Lawrence Sager, Esquire Region I Sager & Sager Associates U.S. Nuclear Regulatory Commission 45 High Street 631 Park Avenue -

Pottstown, Pennsylvania 19464 King.of Prussia, PA 19806 Joseph A. Smyth Mr. Jacque Durr Assistant County Solicitor Resident Inspector County of Montgomery U.S. Nuclear Regulatory Commission Courthouse P. O. Box 47 3 Norristown, Pennsylvania 19404 Sanatoga, PA 19464 Eugene J. Bradley James M. Neill, Esquire Philadelphia Electric Company Associate Counsel for Del-Aware Associate General Counsel . Box 511 2301 Market Street Dublin, PA 18917 Philadelphia, Pennsylvania 19101 Joseph H. White III Mr. Vincent Boyer 11 South Merion Avenue Senior Vice President Byrn Mawr, PA 16801 Nuclear Operations Philadelphia Electric Company 2301 Market Street Phi idelnhia, Pennsylvania 19101 7

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Thomas Gerusky, Director Sugarman & Denworth Bureau of Radiation Protection Suite 510 Dept. of Environmental Resources North American Building 5th Floor, Fulton Bank Bldg. 121 South Broad Street Third & Locust Streets Philadelphia, PA 19107 Harrisburg, PA 17120 -

Donald S. Bronstein, Esq.

Director, Pennsylvania Emergency The National Lawyers Guild Management Agency Third Floor -

Basement, Transportation & 1425 Walnut Street Safety Building Philadelphia, PA 19102 Harrisburg, PA 17120 Lawrence Brenner, Esq. , Chainnan*

John Shniper Administrative Judge Meeting House Law Bldg. & Gallery U.S. Nuclear Regulatory Commission Mennonite Church Road Washington, D.C. 20555 Schuykill Road (Rt. 724)

Spring City, PA 19475 Dr. Richard F. Cole

  • Moylan, PA 19065 Administrative Judge U.S. Nuclear Regulatory Commission W. Wilson Goode Washington, D.C. 20555 Managing Director City of Philadelphia Philadelphia, PA 19107 William A. Lochstet -

119 E. Aar'on Drive State College, PA 16801 Malter W. Cohen Consumer Advocate Office of Attorney General 1425 Strawberry Square Harrisburg, PA 17120 Steven P. Hershey, Esquire Consumers' Education 2 Protective Association Sylvania House Juniper & Locust Streets Philadelphia, PA 19107 -

Alan J. Nogee The Keystone Alliance 3700 Chestnut Street Philadelphia, PA 19104

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, ,o, ENCLOSURE 1 REOUEST FOR ADDITIONAL INFORMATION FOR CONTAINMENT SYSTEMS REVIEW OF THE LIMERICK GENERATING STATION FSAR 480.1 (6.2.1.1) The following additional information is required for our confirmatory analyses of the drywell and wetwell temperature and pressure responses to the design basis loss-of-coolant accidents (LOCAs):

o. Provide the volume of the prirnary system below the elevation of the '

recirculation line break and the main steam line break (MSLB).

b. Provide the horizontal cross-sectional crea of the teactor vessel,
c. Provide the surface crea of the liquid pool formed on the drywell floor following a loss-of-coolant occident,(LOCA),
d. Verify that the blowdown' dato provided in Tables 6.2-10 cnd 6.2-11 incorporate ECCS additions prior to the end of the. blowdown. Also, provide the reactor vessel water level at the end of blowdown.
e. Verify that the MSLB steem flow at 0.73 seconds in Table 6.2-11 is correctly listed cs 84640 lbm/sec.
f. Verify that the MSLB liquid flow at S8.87 seconds on Table 6.2-11 is correctly given as 4042 lbm/sec.
g. Provide o," Accident Chronology" for the MSLB accident.

480.2-l (6.2.1.1) Justify why for the 0.1-square-foot intermediate size break discussed in FSAR Section 6.2.1.l.3.3.4 the break was cssumed to occur below the reccior pressure vessel water level instead of above it. Also list end justify the initial reactor and containment conditions utilized in the analysis.

2g June 24,1982

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480.3 (6.2.1.1) Provide o detailed description of the administrative procedures that will preclude the cetuation-of both drywell spray networks whenever the suppression pool temperature is below 10S F (see FSAR Section 6.2.1.1.4).

1 480.4 '

(6.2.1.1) Provide additional information regarding the containment overpressure relief (COR) system (reference FSAR Section 6.2.1.1.2.7) including its design bases, system design description, design evoluction, tests and inspections, instrumentation requirements, and containment isolation provisions. '

480.S s (6.2.1.1) In FSAR Section 6.2.l.l.3.3.2, " Main Steam Line Breole," it is stated:

The main steem isolation volves (MSIVs) are assumed to stort closing at 0.S seconds and are fully closed in the _

maximum time of S seconds following closure initiation.

By assuming slow closure of these valves, a large effective break crea is maintcined for a longer period of time. The peak drywell pressure occurs before the reduction in effective break 'crea and is therefore insensitive to any additional delay in closure of the isolation valves.

The last statement appears to be incorrect since the peak drywell pressure l

for the main steam line break occurs at 2032 seconds (see FSAR Table 6.2-5 cnd Figure 6.2-12). Therefore,. justify the S-second time to closure cs conservative instead of some longer time to closure that accounts for the delcy between the occident cnd receipt by the MSIVs of the isolation signol.

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2g June 24,1982 l _ ._ __

480.6 (6.2.1.1) Appendix l to SRP Section 6.2.1.l.C provides criterio designed to upgrade the steam bypass capability of the Mark 11 containment design and to assure that the bypcss leakage is not substantia!!y increased over the life of the

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plant. Provide the following information to demoristrate cornplian'ce with Appendix ! to SRP Section 6.2.l.l.C:

a. The cnalysis of the Limerick steam bypass capability for small breaks presented in FSAR Section 6.2.1.1.S is unccceptable. Provide on analysis that shows the suppression chamber design pressure is not exceeded when o leakage crea of A/-[K equal to 0.0S f t2 is assumed and a minimum of 30 minutes.is cssumed for operator action to terminate the suppression chamber pressure transient following indication in the main control room that a bypass leakage path exists. Specify the picnt parameter that will indicate the~ existence of a bypcss leakage path, and commit to providing main control room annunciation of this condition. Also specify the specific operator action that will be taken to terminate the suppression chamber transient. If this analysis shows the suppression chamber design pressure is exceeded prior to the time when operator action ccn be assumed, then NRC's position is that the

.wetwell spray must be automatically actuated. If the wetwell sprays must be cutomatically actuated, the consequences of automatic cctuation of the wetwell sprays on ECCS function cnd long-term pool cooling must be evolucted to show that the minimum ECCS cnd pool cooling requirements cre met.

b. Provide a complete description of the transient analysis requested in part (c) including all cnolysis assumptions; initial co'.ditions; the pressure history in the drywell cnd wetwell; wefwell sprcy capacity, efficiency, coverage, start time cnd temperature history; 'cnd identification cnd quantification of heat sources. In addition, for the wetwell spray nozzles, provida the spectrum of drop sizes cnd mecn drop size emitted from the nozzles os a function of pressure drop across the nozzles cnd describe how this data was obtained (e.g., a spray nozzle test progrcm). Also, discuss the consideration given to evaporation due to impingement of sprcy water on the hot downcomer surfaces.

2g June 24,1982

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c. If the wetwell spray system is to be used to mitigate the consequences of suppression pool steam bypass either manually or automatically, it is our position that the wetwell spray headers must-meet Quality Group B standards rather than the Quality Group C standards sHown in FSAR Table 3.2-1. Provide information on how you will comply with this position.
d. Per the guidance of SRP 6.2.1.1.C (Appendix I) it is our position that a preoperational high-pressure leakage test and postoperational

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low-pressure leakage tests should be performed to detect leakage from '~

the drywell to the suppressioTchamber. Th'e high-pressure test should be performed at approximately the peak drywell to wetwell differential pressure. The low-pressure test should be performed at a differential pressure corresponding to approximately the submergence of the vents during each refueling outage. Acceptance criteria for both tests shall be a measured leakage less than 10% of the capability of the containment to accommodate bypass leakage at the test pressure.

Verify that the above testing requirements will be met for Limerick.

e. Verify that a visual inspection will be conducted during each refueling outage to detect possible leakage paths and to check each vacuum

. relief volve and associated piping to determine that it is clear of foreign matter.

f. ' Demonstrate that the vacuum relief valve position indicator system has adequate sensitivity to detect a total valve opening, for all valves, that is less than the bypass capability for a,small break. The. detectable valve opening should be based on the assumption that the valve opening is evenly divided among cl! the vacuum breakers. -
g. Verify that the vacuum breakers will be tested for operability at monthly intervals.

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480.7 (6.2.1.1) Provide information discussing how the requirements of NUREG-0737 Item II.F.l Attachments 4, " Containment Pressure Monitor," and 5,

" Containment Water Level Monitor," have been rnet.- The responses should specifically address all of the position requir ments find points of .

clarification for each NUREG-0737 item (see also Regulatory Guide 1.97, Rev. 2).

480.8 (6.2.1.1) Provide the range, accuracy, and response time of instrumentation provided

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ct Limerick capable of operating in the post-accident environment for *~

monitoring primary containment atmosphere temperature (reference SRP Section 6.2.1.1.C ltem II.8 and Regulatory Guide 1.97,'Rev. 2).

480.9 (6.2.1.2) Provide additional information demonstrating why a main steam line break ,

(MSLB) will not create local pressurization and resulting forces and moments on the reactor pressure vessel. The staff concern is with the size of the vent paths leading away from the location of a MSLB (i.e., upper drywell creo).

480.10 '

(6.2.1.2) In your differential pressure analysis of the recirculation outlet line break,

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the postulated break location is stated to be within the RPV shield annulus, near the shield wall penetration. Provide a description end/or illustration of the exact location of this break and provide justification of your assumption that 50% of the blowdown mass is released to the drywell atmosphere rather than all of the blowdown being distributed to compartment volume numbers 13 and 24 which surround the break.

480.l{

(6.2.1.2) Justify the adequacy of treating the entire drywell regior) (volume number 73) as a single compartment in the RPV shie!d cnnuius subcompartment analysis. Also verify that the 235,200-ft3 size of the drywell region is a " free volume". Our concern is with the considerable amount of equipment, structures, and floors (i.e., obstructions) contained within the drywell.

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June 24,1982

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480.12 (6.2.1.2) In your RPV shield annulus and drywell head region subcompartment analyses, justify the assumed initial conditions of,P = 15.45 psia, T = 135 F, and RH = 30% instead of the more conservative values of minimum absolute pressure, maximum allowable temperature, and zero percent relative humidity (see'5RP Section 6.2.1.2 Item 11.8.1).

480.13 (6.2.1.2) Justify why, for the drywell head region subcompartment analysis, you did not consider the possibility that insulation could break loose and block the '

vent paths. .

J 480.14 ,

(6.2.1.2) Provide a comparison of the reactor primary system blowdown flow rates and fluid enthalpy for the recirculation line break cnd feedwater line break used in the subcompartment differential pressure analyses calculated using GE's " Generic Annulus Pressurization [Aass-Energy Release Methodology" (MFN178-78) and " Technical Description Annulus Pressurization Load Adequacy" (NEDO-24548/78NED 302) (FSAR Tables 6A-1 and 6A-3,

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respectively) versus the same mass cnd energy release rates calculated using the methodology in NEDO-20533.

480.IS, .

(6.2.1.2) Provide the design pressure differential of the RPV shield wall.

480.16 (6.2.1.2) Concerning the data provided in FSAR Tcble 6A-7, provide a revised table

!. that corrects the below problems:

a. On page 3, flow path 49-50 is listed twice, while 49-60 is not listed.

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b. On page 4, flow path 71-72 is listed twice.
c. There are no listings for 1-73, 4-73, 7-73,10-73, 50-62, 51-63, 58-70, or 59-71. Verify that there are no vent areas between these compartment volumes.

29 June 24,1982

480.17 (6.2.1.2) The paragraph at the bottom of FSAR page 6A-3 states there ore no 4

. external moments generated by the nonuniform load distribution on the reactor pressure vessel frorn a recirculation outlet or feedw'ler c line break inside the RPV shield wo!! cnnulus. Provide additional explanation justifying this statement.

480.18 (6.2.1.2) Provide additionc! information that demonstrates that your selection of the RHR head spray line breck as the worst postulated break for the drywell '

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head region subcompartment cnol5s'is is ceceptable.

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(6f.' FSAR Table 3.2-1 indicates that the RHR heat exchanger primary cnd secondcry sides are Quality Groups B and C, respectively. However, FSAR Section S.4.7.2.4 lists the primcry side os ASME Section 111, Class C and the secondary side os ASME Section Vill, ' Class 1. Verify that the system cicssifications for the RHR heat exchangers cre es shown in. FSAR Table 3.2-1. .

480.20 '

(6.2.2) Provide a description of the RHR intcke strainers including detail drawings

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of'the construction of the strainer assembly, material of construction, and the method of attachment to the RHR suction piping "T" crrangement (reference SRP Section 6.2.2 Item til).

480.21 (6.2.3) FSAR Section 6.2.3.2.I, page 6.2-40, states "An analysis of the post-LOCA pressure transient in the secondcry containment will be. performed to determine the length of time following isolation signal initiation of the SGTS that the pressure in the secondcry containment would exceed minus 0.2S in. wg." Provide the results of this cnclysis of the pressure cnd temperature response of the secondary contcinment to a loss-of-coolcnt accident (LOCA) occurring inside the primcry containment, cnd describe specifico!!y how each of the guidelines of SRP Section 6.2.3 Item II.I hcs been followed.

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June 24,1982

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480.22 (6.2.3) Pro'v ide the following information concerning your FSAR Section 3.6 analysis that shows that the secondary containment structure is capable of withstanding the effects of a high energy pipe rupt0re occurring inside the secondcry containment without loss of integrity (reference SRP Section 6.2.3 Item 11.2):

a. The analytical model that was utilized to calculate the mass and

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energy release rates presented in FSAR Table 3.6-6 for postulated high energy pipe ruptures outside primcry containrnent. '

b. Provide and compare the compartment design pressures with the pressure-temperature transient analysis results e presented in FSAR Table 3.6-7 to support your conclusion that the compartments inside the secondary containment where high energy pipe breaks can be postulated are designed to withstand the maximum pressure developed.

480.23 (6.2.3)

The FSAR states in Section 6.2.3.2.1 that the reactor enclosure is designed to Jimit the inleakage to 50% of the. reactor enclosure free volume per day at a negative interior pressure of 0.2S in. wg. while the SGTS is operating.

Verify that this stated inleakage limit applies also to the refueling crea, or provide a sepcrate inleakage limit for the refueling crea in terms. of a percentage of the refueling crea free volume per day (reference SRP Section 6.2.3 Item !!.3.b cnd BTP CSB 6-3 Position B.4).

480.24' .

(6.2.3) Verify that the railroad access door (secondary containment access opening number 194) is provided with position indicators and c!crms having readout end alarm capability in the main control room, or describe in more detail the administrative procedures that ensure all the access hatches cnd doors from the railroad access shaft to the secondary containment will be sealed closed unless the railroad access door is sected closed.

2g June 24,1982

480.25 (6.2.3) Supplement your description of containment bypass leakage barriers listed in FSAR Table 6.2-15 in -the following crecs:

o. Penetrations X-23, 24, 53, 54, 55, and 56 rely on closed'iystem3 inside containment to preclude bypass leakage. Verify that each requirement listed in Branch Technical Position (BTP) CSB 6-3 Position B.9 is met for these systems.
b. Provide adoitional information describing the feedwater fill system and

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demonstrate that it conforms to the NRC guidelines commensurate to its safety function of. eliminating bypass leakage (reference

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BTP CSB 6-3 Podtson B.8).

c. For all containment penetrations in FSAR Table 6.2-15 that rely on a water sect to prevent bypass leakage, provide additional information that demonstrates the water seal will be maintained for 30 days

, following a LOCA, i.e., that the water seal is maintained at a pressure greater than the peak containment accident pressure end that a sufficient water seal inventory to lost at least 30. days following a

. . design basis accident is provided..
d. .The recirculation pump seal purge lines (Penetration X-61) are vented to secondary containment by use of vent lines located before two block valves and the secondcry containment (see FSAR Figures 5.4-2 and 4.6-5). The P&lD shows no automatic actuation signal to the volve operators for the normally closed vent valves (HV125 and HV126) or the normally open block valves (HV127 and HV128). Verify that.these valve

, operators will receive containment isolation signals to open and close, respectively. -

a.

e. FSAR Table 6.2-15 indicates that the HPCI cnd RCIC vacuum relief lines (Penetrations X-228D and 241) contain temporary spool pieces that cre removed during normal operation and replaced by blind fienges. These spool pieces are not shown on the P&lDs (FSAR Figures 6.3-7 cnd 5.4-8). Provide revised P&lDs that include the spool pieces, or describe alternative provisions to preclude bypass leakage.

2g June 24,1982

480.25 (6.2.3) continued NOTE: Any penetratiorrthrough which bypass leakage cannot be precluded by cn acceptably described bypass leakage barrier must be considered as a bypass leakage path and treated cs described in BTP CSB 6-3 Positions B.6 and 7.

480.26 (6.2.3) Provide a time line of events for actuating the SGTS following a LOCA that conservatively maximizes the time it takes the SGTS to reach full operational status cssuming loss,of off-site power and the most severe single active failure. Any delay, due to system-design, in actuating the SGTS should be considered. The maximum delay time for SGTS actuation should be utilized in the analysis of the pressure and femperature response of the secondary containment to a LOCA requested in Question 480.21.

Also clarify whether upon receipt of an automatic actuation signal both SGTS fans stort es stated in FSAR Section 6.S.1.1.2 or only the lead fcn starts cnd the other fcn remains on standby as stated in FSAR Section 7.3.l.l.7.

480.27 (6.2.3) Provide additional information cnd/or cnolyses that demonstrate the external design pressure of the secondary containment structure (i.e., the reactor enclosures, Zones I cnd !!, and the refueling crea, Zone fil) assures on adequate margin above the maximum expected external pressure (reference SRP Section 6.2.3 Item II.3.e).

480.28' -

(6.2.3) Verify that periodic tests will be required (i.e., technical specification surveillance requirement) to determine the time it takes the SGTS to deprusurize the secondary containment to a negative pressure greater than 0.25 in. wg. In order to confirm the cnolysis requested in Question 480.21.

480.29 (6.2.4) The containment isolation provisiens for each of the following lines consist of two volves in series located outside the primary containment. In each

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case the valve nearest containment and cny piping between the containment and the volve are not enclosed in a leck-tight or controlled leakage housing and therefore, in accordance with SRP Section 6.2.4 Item !!.d, the valve and piping must be conservatively designed to preclude a break of piping integrity. Justify why it is not feasible or practical to locate a containment isolation volve inside containment, and verify that the design of the volve nearest containment end any piping between containment and the volve conforms to the guidelines

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of SRP Section 3.6.2 for each of the following lines:

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Penetration Line ,

_ Valve X-25 Drywell pur~ge supply 122 X-26 Drywell purge exhaust , 113 X-28A Drywell H2 /02s niple 134, 132 X-288 Drywell H2 /02s mple 133

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X-39A, B Containment spray FO21.A, B X-201A Suppression poc! purge 125 supply .

X-202 Suppression pool purge 103 exhaust -

X-221A Wetwell H2 /0 2sample 181 ,

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X-2218. Wetwell H2 /02sample 183 X-22'S RHR vacuum relief 130 X-228D. HPCI vacuum relief F095 1

X-241 RCIC vacuum relief F084' l X-231 A, B Drywell sump drains ,

110, 130 i

X-406 ILRT data acquisition 1057, 1071 X-227 ILRT data acquisition 1073 l

!' X-62 H2/0 sample 2 return; 150 crywell purge makeup X-220A H2/0 sample 2 return; 190 wetwell purge makeup 29 June 24,1982

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480.30 (6.2.4) The containment isolation provisions in each of the following lines consist of a single isolation volve outside containment and a closed system outside containment. In each case the volve and any piping"between the containment and the valve are not enclosed in a leak-tight or controlled leakage housing and therefore, in accordance with SRP Section 6.2.4 Item II.e, the valve and piping must be conservatively designed to preclude a break of piping integrity. Verify for each of the following lines that the

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design of the valve nearest containment and cny piping between contcinment and the volve conforms to the guidelines of SRP Section 3.6.2. Also, verify for each line that the closed system outside containment is protected from -missiles, designed to Seismic Category I standards, classified Safety Cicss 2, and has a design tempe.ature and pressure rating at lecst equal to that for the containment, and verify either that the closed system outside containment will be leak tested or that the system integrity con be shown to be maintained during normal plant operations (reference SRP Section 6.2.4 Item II.e):

Penetration Line Valve X-203A,B,C,D RHR pump suction, F004A,B,C,0

. X-204A,B RHR pump test end minimum 12SA,B flow X-20SA,8 Suppression pool spray F027A,B X-206A,B,C,D CS pump suction F00l A,B,C,D X-207A,8 CS pump test and flush F0lSA,8 X-2088 CS pump minimum flow F031B X-209 HPCI pump suction F042' X-210 HPCI turbine exhaust F072 X-212 HPCI pump test cnd flush F071 X-214 RCIC pump suction F031 X-2lS RCIC turbine exhcust F060 29 June 24,1982

. _ _ _ _ _ , _ _ . . - . - - . . _ _ . 17 - - - - - - - - - -

X-216 RCIC minimum flow F019 X-217 -

RCIC--vacuum pump F002 discharge

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X-226A,B RHR minimum flow ~

10SA,B -

X-23S CS. pump minimum flow F031A X-236 HPCI pump minimum flow F012 X-238 RHR relief valve 1068, F1038, discharge FOSSB,1018 X-239 RHR relief valve discharge 101 A, FOSSA, '

106A, F103A X-240 RHR relief volve discharge F097 480.31 (6.2.4) Justify the use of a check valve (F028) in Penetration X-217 as the second isolation valve outside containment.

480.32 (6.2.4) Verify that cil normally closed manual valves in test, vent, drain, and similar types of branch lines which serve os containment isolation barriers will be sealed closed, and will be under administrative control, cs defined in SRP Section 6.2d Item II.f.

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480.33 (6.2.4) Provide the relief setpoint for the suppression pool elecn-up pump suction relief valve, PSV127 (reference SRP Section 6.2.4 Item II.g).

480.34 (6.2.4) NUREG-0737 Item II.E.4.2 pertains to containment- isolation

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dependability. Describe how each paragraph of item li.E.4.2 is satisfied.

Specifically indicate whether each valve listed in Tcble 6.2-17 is essential or nonessential. Concerning Position (3) cnd Clarification (2), explain why the following valves in systems which cre not listed as ESF systems in Tcble 6.2-17 do not receive automatic containment isolation signals:

29 June 24,1982

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Penetration Line Valve X-35C-G - TIP drives 141A-E X-217 RCIC vacuum pump discharge - F002 480.35 (6.2.4) Provide the length of piping from containment to the outside cohtainment isolation volves missing in FSAR Tobie 6.2-17. Also, the review of the length of piping from the containment to the outside isolation valves provided in FSAR Tcble 6.2-17 shows that some of the distances appect to be excessive. Verify that all outside containment isolation volves have been located cs close to containedsnt cs practical, cnd specifically justify the location for volves greater then 1,0 feet from containment.

480.36 (6.2.4) Verify that Vcive SV101 (Penetration X-2208, instrumentation - suppression pool) fails closed upon loss of power (Note: FSAR Figure 9.4-5, Sheet I of 2, indicates that this valve fails open). ' -

480.37 (6.2.4) Justify why diverse pcrameters are not sensed for the automatic initiation of containment isolation for the following volves in lines which are not listed as ESF systems in FSAR Toble 6.2-17 (reference SRP Section 6.2.4 Items ll.1.and m):

Penetration Line Valve X-9A, 8 Feedwater 109A,B X-12 RHR shutdown cooling supply F009;F008 X-13A, B RHR shutdown cooling return 151 A,8, F015A,B X-17 RPV hecd spray F022,,F023 X-28A Recirculation loop semple F019,F020 X-43B Mcin steem sample F084,FOSS 29 June 24,1982

480.38 (6.2.4) Justify why closure times greater than 60 seconds are acceptable for the following valves (reference SRP Section 6.2.4 ll.n and !!!):

Penetration Line Voive X-9A,8 Feedwater 109A,B X-12 RHR cooling supply F009,F008 X-17 RPV head spray F023 480.39 (6.2.4) It is the NRC position that the requirements of GDC 56 and not GDC 57 must be met for the instrument gas supply lines (Penetration X-38, X-3D,.

X-40H, and X-218), the recirculation pump cooling water supply and return lines (Penetrations X-23 and X-24) and the drywell chilled water supply and return lines (Penetrations X-53, X-54, X-55, and X-56) because the systems inside containment to which these 1,ines connect do not meet the requirements of a closed system (see SRP Section 6.2.4 Item II.o and FSAR l Section 6.2.4.3.1.6). Therefore demonstrate for these penetrations how the containment isolation requirements of GDC 56 will be met. (Note: For the instrument gas line penetrations, CDC 56 requirements will be met if both the outside valves (HV1298, HV151, .HV129A, and HVl35) and the inside

, check valves (10058, i112, 1005A, and 1001) are included, in FSAR Table 6.2sl7 as containment isolation valves for Penetrations X-38,' X-3D, X-40.H, and X-218, respectively.)

480.40 (6.'.4) 2 Verify that the containment isolation barriers.cnd the piping between them, or the piping between the containment and the outermost isolation barrier, have Group B quality standards cpplied cnd are designated Seismic Category I, for the following penetrations (reference SRP Section 6.2.4 II.p):

2g June 24,1982

___ _ __, 19 . .. . __ . . _ - -

Penetration Line X-35C-G TIP drives X-Il78 Drywhil radiction sampling supply cnd return X-220B Instrumentation-suppressio'n pool' pressure

Also justify why the containment isolation barriers at Penetration X-2318 (drywell equipment drain tank outlet) do not have Group B quclity stenderds applied, cs indicated by FSAR Figure 9.3-4; and justify why the piping cssociated with the containment isolation valves at Penetrations X-53, ,

-55, cnd -56 (drywell chilled water system) do not have Group B quality s stcndards applied, cs indicated by'FSAR Figure 9.2,27 cnd Tcble 3.2-1.

480.41- ,

(6.2.4) Provide the following information concerning compliance of your containment purge system with Branch Technicci Position (BTP) CSB 6-4:

a. Describe cnd justify all enticipated uses of the containment atmosphere control systems to vent or purge the primary containment during the operational modes of stcrtup, power operation, cnd hot shutdown.
b. Commit to limiting the number o~f lines in use at cny one time during the op,erational modes of stcrtup, power operation, cnd hot shutdown to one supply line and one exhaust line in accordence with BTP CSB 6-4 Position B.I.b.
c. Provide detailed justification for the size of cny containment atmosphere control system line exceeding eight inches in dicmeter if it is to be used during the operational modes of stcrup, power operation, and hot shutdown (reference BTP CSB 6-4 Position B.I.c).
d. Explain why diverse pcremeters (besides the high radiction signal) cre not sensed for the cutomatic initiction of contcinment isolation for the following vcives in accordcnce with BTP CSB 6-4 Position B.I.e (see also SRP Section 6.2.4 Items li.I cnd m):

_ _ . . _ _ ._ - 1 6 _ - . ._ _ _ . _

1(r tration- Line valve  ?~

X-25 Drywell purge supply 163 X-26 Drywell purge exhaust 145,161 7 Xo H2 /0 2sample return; ~ 150,159,!!6 drywell purge makeup

^' e X.241A Surpression pool purge 164 -

supply s

~~

, ~

s y i X-202 Suppression peol purge 185,162 k , s. - O-exhaust

~ C*[N X-220A H2/0 sample 2

return; 190,191,i16 wetwell purge makeup l

e. Explain why the following valves bove closure times greater then -

<>~-

5 seconds (reference BTP CSB 6-4 Position B.I.f and SRP Section 6.2J4 II.n)- ,

's-ys

-gn Penetroficn Linc. Valve -

4.

s s s

, g X-25 Drywell purge supply 122,163 1 'g

~

<c s X-26 Drywell purge exhaust I13,III,161-x Nx X-201A Suppressio'n pool purge 125,164 _

supply .

.,t '~ '

N X-202 Suppression pool purge 103,:162, 105 exhaust ,

f. Provisions (i.e., debris screens) are- required to ensure that isolation , .Il ,

volve closure will not be prevented by debris which could pe%ntallyq .

becerne entrained in escaping air end stecm at Penetration.s X-25, j'1 X-26, X-62, X-201 A, X.202, and X-220A in accordcnce with BTP CSB l-6 4 Position B.I.g. Provi.lc enc,ineering drawings showing the materials; and dimensions of the re imd debris screens cnd the position of each; ,

Jehnt screen relative to, es:d its distance from, the nearsst w (inboard)n isolati.;n valve. Also dernenstrate complicnce with 'the' fol!6wina criteria:

I

' es

  • *9 % ,,

~;

g

s- . ,; j -. . -

x, .

.s

.. N Ns ,

~

% c,, R, 8 -

y .

A.- \' ' '

(1) s ,

The debris' screen'dhuld be of Seismic Category I design.

, _ v ., ,

s t 4 &

R { .

s-s .

S '

,tv' N(2). Any pif;ing'I>etween the debris screen and the nearest (inboard)

T ,,t

,, , 'Yx ' f<$. iisoja, tion valve sb0ld also be of Seismic. Category I design.

....; s _y, , . .x, - .

s

r. s- .g 3 x -

, g s. W

^

M 3.5 " ' ,

JC w g A r E Q). TSe debris scree 6'should be designed to withstand the LOCA m- m.g ,

,* , ~-

differerttial ' presser' e.

n.s. f,. ~ N.N _- . s gi Verify Tthat fission' products were assumed to be released through the s ,.

\,

N V ,.u 'normoily'open purge valves during the maximum interval required for s s

\ volve.,s cl.osure y In the analysis of the radiological consequences of a s .

loss-ofUcoolonf^ accident. in 'occordance. with BTP CSB 6-4

+ . .. m .. x +-

4h

' # Position B'.S.' -c. . >

_,~ <- 'n s, .

1 -: < p ~%

' ~

s 4.- h. Provide a o,sseription enddesults of an analysis which demonstrates the

% s.

.s _ s y- '

s ' acceptobility' tof ~ 'the provisions made to protect structures and

\

M .s1 .

\

, safety-related equipment (e.g., fcns, filters, and ductwork) located x s p beyond the purge system ~ isolation valves against loss of function from N

/ )

( the environment created-by escaping air and steam, in accordance with e- s BTP C58 y. 6'-4 Position 8.5.b. s

,5

% ,w I\ 3 ,.._

Qi. .s Specif[p.the maximum allowchie i.eck rate of the purge isolation valves giving'oppropriate., consideration to valve size, maximum allowable w" W leakage rate for the containment (os defined in Appendix J to 10 CFR

._ dv '. ..' y-- .

Part SO),' tend the max' mum allowable bypass leakage fraction, in N cecordance- with BTP CSB 6-4 Position B.S.d.

% 7. .

q .

-t s a

% ~

1 480.4k .  ?. 1 y# '<

-- x s w

, (6.2.4) Describe the conta inment _ isolation provisions for the following penetratioris which are not inniuded in Table 6.2-17:

c s -

Penetratien . f Line -

. s r

_ + ,

~

S.

'y[ X-ly ; 2 Equipment Access Door s ~% .

.'m X-2 x .

Equipment Access Door cnd Personnel Lock t

s s's e

..(

. .,v, . +

s

- , s 29 June 24, I?82 t - . - .

-is. .

1

" 4 - - , - ._. "W -*9~g ,,

X-4 Head Access X-6 CRD Removal Hatch X-200A,B Access Hatches (Suppression Pool) 480.43 (6.2.4) The following volves are containment isolation volves that cre not listed in FSAR Table 6.2-17. Provide for these volves the information listed in Table 6.2-17:

~

Penetration Line Valve X-9A,8 Feedwatei- '

154A,130B X-44 Alternate RWCU return , PSVI12 X-62 H2/0 2s mple return; I16 crywell purge makeup X-237 Suppression pool clean-up 139 pump suction ~

X-238 RHR relief volve discharge F1048. ~

X-239 RHR relief valve discharge FIO4A 480.44 (6.2.4) Since the TIP drive penetrations (X-35C-G) are classified as nonessential,

~

isolation. valves should be provided both inside and outside containment to meet . GDC 56 requirements. Provide information to either justify the existing isolation provisions or show how the isolation provisions will be upgraded to . meet the requirements of GDC 56. If your position supports location of both containment isolation volves outside containment, verify that the design of the valve nearest containment and cny piping between containment and the vc!ve conforms to the requiremen'ts of SRP Section 3.6.2 (see Question 480.29). Also justify the fail-as-is de~ sign of the volves and demonstrate that a single active failure will not defeat the containment isolation function for the TIP drive penetrations.

480.45 (6.2.4) Provide the power source to isolation valves 133A,B missing from FSAR Table 6.2-17 for the feedwater penetrations (X-9A,B).

480.46 ~ -

(6.2.4) Concerning Penetration X-9A, FSAR Figure 5.1-3 shows an incoming line (8"DBB-103) from the HPCI system tying into Feedwater Line "A". This line is not shown on FSAR Figure 6.2-36 Detail (2), nor is it traceable to the HPCI P&lD (P&lD No. M-55, FSAR Figure 6.3-7). Provide a description, including the correct reference drawing, of the isolation provisions for Line 8"DBB-103. '

480.47 s (6.2.4)' Concerning Penetration X-237, FSAR Figure 6.3-9 shows a branch line (l"HBB-134) from the suppression pool cleanup system pump suction which is not shown on FSAR Figure 6.2-36 Detail (26) or traceable to P&lD No. M-55 (FSAR Figure 6.3-7). Provide a description, including the correct

, reference drawing, of the isolation provisions for this brcnch line.

480.48 -

(6.2.4) Since the drywell radiation sempling supply and return lines (Penetration i178) are classified as rionessential, isolation valves should be provided both inside end outside containment to meet GDC 56 requirements. Provide information either justifying the existing isolation provisions or state how the isolation provisions will be upgraded to meet the requirements of GDC 56. If your position supports location of both containment isolation volves outside containment verify that the design of the volve nearest containment and any piping between containment and the volve conforms to the requirements of SRP Section 3.6.2 (see Question 480.29). Also provide the missing information in FSAP Table 6.2-17 and a P&lD showing the containment penetration isolation provisions (FSAR Figure 11.5-1 does not adequately show the containment isolation provisions).

2g June 24,1982

.._ _ . _ _ _-20 . .

480.49 (6.2.4) FSAR Section 6.2.4.3.1.5 invokes the provisions of Regulatory Guide 1.11 for instrument lines that penetrate the primary containment. Inspection of FSAR Table 6.2-17 and the P&lDs indicates that the criteria outlined in Regulatory Guide 1.11 Regulatory Position C.I ha've been 'applieff to the containment isolation provisions for the following instrument lines, which include excess flow check valves outside containment:

Penetration Line

~

X-3A Instrumentation - main stearn line D flow X-3C Instrumentation - HPCI steam' flow X-3D instrumentation - steam line A flow X-7A-D /dain steam (MSIV leakage control)

X-20A,B Instr'umentation LPSI A P X-20A,8 Instrumentation - RPV level X-27B instrumentation - HPCI flow X-288

  • Instrumentation - LPCI A P X-29A Instrumentation - RPV flange leakage X-29B Instrumentation - CS A P .

X-30A Instrumentation - main steam line D flow X-308 Instrumentation - main steam line C flow X-31 A,8 Instrumentation - jet pump flow X-32A,B Instrumentation - jet punip flow X-33A Instrumentation - pressure above and below' core plate X-338 Instrumentation - RCIC steam flow X-34A Instrumentation - main steam line C flow X-34B Instrumentation - recirculation flow X-40A,B,C Instrumentation - jet pump flow X-40D Instrumentation - pressure below core plate 2g June 24,1982 y- , . _ . _ _

X-40F Instrumentution - RCIC steam flow X-40H Instrumentation - recirculation pump cooler X-41 Instrumentation - RWCU flow . .-

X-41 Instrumentation - LPCI A P X-43A instrumentation - recirculation loop A A P X-43A Instrumentation - recirculation pump seal pressure X-47 Instrumentation - RWCU flow -

X-48A Instrumentation - core spray A P '

X-48A,8 Instrumentation - RPV level -

. X-49A,B Instrumentation - main steam line A & B flow X-50A Instrumentation - recirculation flow X-50B Instrumentation - recirculation pump seal pressure X-508 Instrumentation - recirculation pump cooler flow X-SIA instrumentation - recirculation line flow X-518 Instrumentation - jet purnp flow X-52A Instrumentation - main steam line B flow

~

X-528 Instrumentation - recirculation line flow

~

X-57 Instrumentation - RWCU flow X-58A Instrumentation - recirculation loop B A P X-63 Instrumentation - recirculation loop A P X-63 Instrumentation - recirculation pump seal pressure X-65A,8 Instrumentation - RPV pressure X-66 A,B Instrumentation - RPV ievel X-66 A,8 Instrumentation - LPCI AP X-67A,B Instrumentation - RPV level; RPV pressure Indicate whether each of the cbove listed lines is a sensing line for- an instrument that is part of the protection system. Veniy that those lines 2g June 24,1982 y-.

. .e -

I' -

1 that are not sensing lines for instruments that cre part of the protection system will be provided with one automatic isolation valve inside and one automatic isolation valve outside containment in accordance with Regulatory Guide 1.11 Position C.2.b, or provide.an acceptable citernative

~

basis for the containment isolation provisions of these lines.

480.50 (6.2.4) The containment isolation provisions shown on the P&lDs for the f'ollowing instrument lines connected to the RCPB are inconsistent with the evaluation presented in FSAR Section 6.2.4.3.1.5 (i.e., restricting orifices are miss..,g on the P&lDs for these lines):

FSAR Fioure Penetrations 5.!-3 X'-7A-D 5.4-2 X-34'B, 43 A, 50 A, 51 A, 528, 58A, 63 Either commit to adding a restricting orifice located inside the drywell for these lines or otherwise describe how the requirements of GDC 55 will be met for these lines.

480.51 .

(6.2.4) Describe the leak detection provisions and their location that will provide prompt and unambiguous notification of a leak in one of the CRD lines or in the scram discharge volume piping.

480.52 (6.2.5) Explain physically why in FSAR Figure 6.2-44, "Short Term Containment Oxygen Concentration After LOCA," the oxygen concentration initia!!y drops in both the drywell and suppression chamber. -

c 480.53 (6.2.5) Confirm that the fission product decay energy used in the calculation of hydrogen and oxygen production from radiolysis is equal to or more conservative than the decay energy model given in Brcnch Technical Position ASB 9-2 in SRP Section 9.2.5 (reference SRP Section 6.2.5 Item li.2).

9 '

480.54 (6.2.5) Verify that the combustible gas cnolyzer system is capable of providing mecsurement of hydrogen and oxygen concentration under both positive end negative pressure (reference NUREG-0737 Item II.F.! Attachinent 6).

480.55 (6.2.5) The existing containment hydrogen recombiner system d sign described in FSAR Section 6.2.5.2.1 cnd shown in Figure 9.4-5 does not comply with the requirements of NUREG-0737 Item II.E.4.1 " Dedicated Hydroge'n

~~

Penetrations." Provide additionalinformation that demonstrctes how you '

~

will comply with NUREG-0737 item II.E.4.1. Specifically address in your response all position requirements and points of clarification.

480.56 (6.2.5) Provide the cnolyses or.other cppropriate justification demonstrating the capability of the drywell air cooling system (i.e., drywell unit cooler

, housing and ductwork) to withstand the dynamle effects, such cs trcnsient differential pressures, that would occur early in the blowdown phase of a loss-of-coolcnt accident (reference SRP Section 6.2.5. Item 11.5).

480.57 .

(6.2.5) Confirm that there is no enclosed compartment or cubicle in the primcry containment where oxygen could accumulate in a concentration in excess

~

of 5 volume percent (reference SRP Section 6.2.5 Item 11.3).

480.58 (6.2.5) Identify the computer program used to conduct the primar, containment l

1 hydrogen cnd oxygen enclyses whose results cre presented in FSAR

! Section 6.2.5.4.

480.59 l_ (6.2.5) Provide the maximum cllowable operating pressure end inlet temperature l limit for the hydrogen recombiners. Provide a long-term cnolysis of the conservative containment pressure end temperature response follow'ng a postulated LOCA to cssure these operating limitations are not exceedec ,t j the time hydrogen recombiner operation is required. .

29 June 2'i,1982 l ._ _ _ _ . . - __ _ _ -_.. - 2 4 -

480.60 (6.2.5) Verify that the drywell- unit coolers are designed such that failure of non-safety-related equipment (e.g., cooling coiIs) will not prevent the drywell unit coolers from accomplishing their safety ~ function.'

~

480.6i (6.2.5) Verify that all zine (as galvanized steel) listed in Table 6.2-19 is located in the drywell.

480.62 (6.2.2) Although FSAR Section 6.2.2.2 Etates that the RHR intake strainers cre designed to withstand all hydrodynamic loads postulated to occur in the suppression pool, concerns crise due to the close proximity of the downcomer discharged to the intake ' strainers. Provide a list of all loads used in the design of the strainers cnd also provide additional information on your analyses that demonstrate the capability of the strainers to accommodate the hydrodynamic loads from downcomer discharges.

l 4

2g A 'e 24, , yd2

-~

48'0.63 Notes 1,11 and 21 to Table 6.2-25 indicate that no Type C test will be performed for the isolation valves in instrumentation lines. State your position whether or not these valves will be opened during the Type A test. .

~

480.64 Note 15 to Table 6.2-25 states..that the feedwater and the RCIC va-

. cuum pump discharge isolation valves will be tested with water; it is our position that these valves be pneumatically tested with air or nitrogen.

480-65 Note 14 to Table 6.2-25, is not an accepta.ble substitute to per-forming Type C tests on the containment isolation valves listed in Table 6.2-25 and remarked by note 14. It is our, position that all these valves should be subjected to Type C testing. State your intention regarding out positionJ

~

480.66 With regard to the Traversing Incore Probe, it is our position that the following should be perfomed on:the shear valves to assure that they will p. -fom their intended function:

(1) Verify the conti6uity of the explosive charge at least once

~

per 31 days.

(2) Initiate one of the explosive squib charges at lease. once per 18 months. The repircement charge for the explosive valve shall be from he same manufacturing batch as the one fired or from another ich has be' 1rtified by having one sample of that batch successfull. tired.

(3) All charges should be replaced according to the manufacturer's recommended life time. ~

480.67 Chapter 8 of the Des'ign Assessment Report (DAR) that addresses the T-quencher verification test (proprietary) has not been sub-mitted. We request that a copy of this chapter be submitted' for .

our review.

480.68 Provide the pool temperature analysis for the transient involving the actuation of one, or more SRV's. For additional guidance, your attention is directed to NUREG-0873, " Pool Temperature Transients for BWR."

480.69 Table 1,3-2 of the DAR indicates that the quencher arm loads, .the total quencher loads during SRV opening, and loads during irregular condensation are under evalution. Provide these load specifications.

480.70 Concerns regarding the capability of the vacuum breaker to perform its function during the pool swell and chugging phases of LOCA have been raised. Provids the design changes, if any, that have been fr-plement'ed to resolve this concern.

O Y

a 4

'l

.;.... -.-... w.

_.~.-- . --w., .

. _. -. . ..#..p._. - . . . . . - - . , . _

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