ML19206A111
| ML19206A111 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/21/1974 |
| From: | Kniel K US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7904180051 | |
| Download: ML19206A111 (81) | |
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,.3. "T' UNITED STATES
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ATOMIC ENERGY COMMISSION I
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Locket No. 50-320 Metropolitan Edison lompany ATTN:
Mr. R. C. Arn31d Vice Preside.it
?. O. Box 542 Reading, Pennsylvan a 19603 Gentlemen:
1 Inorderthatweruahcontinueourreviewofyour application for a 1;icense to operrte the Three Mile Island Nuclear Station, Unit 2, additional informa-tion is required.
The information requested is described in the enclosure and pertains to the sec-tions of your Final Safety Analysis Report.
The majority of the items in the enclosure repre-sent "first-round" requests for additional informa-tion.
Also included are clarification questions requesting additional information on certain unsat-isfnctory responses to our _aceptance' review request for additional information.
The enclosure conforms to our new format for lists of requests for addi-tional information and Regulatory staff positions.
The new standardised item numbering systen, which differs from that used in the acceptance review request for additional information, is described in the enclosure.
Final or approved for construction versions of de-tailed electrical instrumentation and control system drawings for all systems which you classify as safety related should be submitted in accordance with the procedure detailed in our letter of May 9, 1974 to Mr. T. M. Crimmin's, Jr.
Your projected schedule in Amendment 15 for completion of these drawings should be improved, in particular, we should have a partial submittal on October 18, 197h of those systems which are complete at that time and, if possible, your projected completion of these sub.ittals should be February 1, 1975.49-020 7904180061 g
An In order to maintain our licensing review schedule, we will need a completely adequate response to all enclosed items by October lo, 1974 Please infom us within seven days after receipt of this letter of your confirmation of the schedule date or the date you will be able to meet.
If you cannot meet our specified date or if your reply is not fully responsive to our request, it is highly probable that the overall schedule-for completing the li-censing review for the project will have to be extended.
Since reassignment of the staff's efforts will require completion of the new assignment prior to returning to this project, the extent of the extension will most likely be greater than the delay in your response.
Please contact us if you have any questions regard-ing the information requested.
We will be avail-able to meet with you and to discuss these matters if you so desire-Sincerely, Karl Kniel, Chief Light Water Reactors 3 ranch 2-2 Directorate of Licensing
Enclosure:
Request for Additional Lformation ces:
See next page 4 L a 2-2
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Form MO 318 Gev. ? 53) AICA 024
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.!arrepolitan 7.dison Cc=panr ces:
Gecrge F. Troweridge, c.a q ui re Shaw, Pitt=c.n, Fotts 5 Trewbridge 910 17th Street, S. U.
Uashington, D.
C.
20006
- !r. Richard W. Eevr. d Project Manat;er CPU Service Corporation 250 Cherry dill Road Parsippany, ::w Jersey 07034
'tr. Locas M. Cri== ins, Jr.
Safety and Licensing Manager GPU Service Ce rporation 260 Cherry :lill i;oad Parsippany, ::e.r Jersey 07054 Chauncef R. depford, Esquire Chair =an York Co==ittee for a Safe Environ =ent 10 S 3. Pershing Avenue York, Pennsylvania 17403 DISTRIBUTION:
AEC PDR Local PDR Doci<et File LWR 2-2 VAMoore FSchroeder AKenneke DEisenhut RKlecker CGC R0 (3) 3Washburn MService TR 3 ranches _
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LWR 1 & 2 BC JPanzarella ACRS (16) bec:
J. R. 3uchanan, ORNL Thomas 3. Abernathy, DTIE oe ei:a *
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W ASHINGTON. D.C. 20545 AUG i s4 rnu Docket No. 50-320 Metropolitan Idison Company ATTN:
Mr. R. C. Arnold Vice President P. O. Box 542 Reading, Pennsylvania 19603 Gentlemen In order that we may continue our review of your application for a license to operate the Three Mile Island Nuclear Station, Unit 2, additional infort.
tion is required.
The information requested is described in the enclosure and pertains to the sec-tions of your Final Safety Analysis Report.
The majority of the items in the enclosure repre-cent "first-round" requests for additional informa-tion.
Also included are clarification questions requesting additional information on certain unsat-isfactory responses to our acceptance review request for additicnal information.
The enclosure conforms to our new format for lists of requests for addi-tional information and Regulatcry staff pocitions.
The new standardized item numbering system, which differs from that used in the acceptance review request for additional information, is described in the enclosure.
Final or approved for construction versions of de-tailed electrical instrumentation and control system drawings for all systems which you classify as safety related should be submitted in accordance with the procedure detailed in our lotter of May 9, 1976 to Mr. T.
.t!.
Crimmins, Jr.
Your projected schedule in Amendmtat 15 for completion of these drawings should be improved in particular, we should have a partial submittal on October 18, 1974 of those systems which are complete at that time and, if possible, your projected completion of these submittals should be February 1, 1975.
49 023
. In order to maintain our licensing review schedule, we will need a completely adequate response to all enclosed items by Cetober 13, 1976 Please inform us within seven days af ter receipt of this letter of your confirmation of the schedule date or the date you will be able to meet.
If you cannot meet our specified date or if your reply is not fully responsive to our request, it is highly probable that the overall schedule for completing the li-censing review for the project will have to be extended.
Since reassignment of the staff's efforts will require completion of the new assignment prior to returning to this project, the extent of the extension will most likely be greater than the delay in your response.
Please contact us if you have any questions regard-ing the information requested.
We will be avail-able to meet with you and to discuss these matters if you so desire.
Sincerely, l
/
O Karl Eniel, Chief Light Water heactors 3 ranch 2-2 Directorate of Licensing
Enclosure:
Request for Additional Information ces:
See next page 4{3-024
Metropolitan Edison Co=pany ces:
George F. Trewbridge, Esquire Shaw, Pitt=an, Po tts & Trowbridge 910 17th Street, N. W.
Washington, D. C.
20006 Mr. Richard W. Heward Project Manager GPU Service Corporation 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. Thomas M. Cri== ins, Jr.
Safety and Licensing Manager CPU Service Corporatien 260 Cherry Hill Road Parsippany, New Jersey 07054 Chauncey R. Kepford, Esquire Chair =an York Co==ittee for a Safe Environ =ent 108 N. Pershing Avenue York, Pennsylvania 17403 49-025
ENCI.CSURE REQUEST FOR ADDITIONAI. INFORMATION FIRST-ROUND QUESTIOU, THREE MILE ISLAND NUCI. EAR ST/ TION, UNIT 2 FINAL SAFETY ANALYSIS REFORT 49~C26
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The follcwing branch nu=cers, which are the Blue "ack/ Level D cranch ccda nu=bers, chall be used* by the contributing branches:
B ranch '?o.
Branch ':r.e 01.
Effluent Treatment 02.
Auxiliary & Power conversion Syste=s 03.
Ccatainment Systems 11.
Mechanical Engineering 12.
Materials Engineering 13.
Structural Enqineering 21.
Reactor Syste=s 22.
Electrical Instrumentatica &
Control Systems 23.
Core Perfor=ance 31.
Accident Analysis 32.
Site Analysis n....a.,
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m.
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34.
Ccs: Eenefit Analesis 33.
Environ =entar apeciaals;a 41.
Quality Assurance 42.
Industrial Security & Z,ergency Planning 43.
Operator Licensing Branch No. 00. will be used by the responsible RP L R or A2 Branch for =iscelaar.cous ite=s and will be entitled "GE"ERAL".
3 ranch No. 40. will be used for ite=s on financial catters.
3.
INDEX SYS~E'i The ite=s within a~ branch will be nu=bered as indicated below for 3 ranch 21:
21.1 21.2 21.3 21.4
=
21.107
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I:!DE:( SYST21 (continued)
~~he sys:c= to be used for sub-iters within a given ite : will be as indicated below for item 21.17:
21.17 (1)
(2)
(a)
(b)
(c)
(1)
(ii)
(iii)
(3) e e
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01-1 ROUND ONE QUESTIONS FOR THREE MILE ISLA'!D, UNIT 2 Docket No. 50-320 01.0 EFFLUENT TREAT'!ENT
"'31.1 (11.2.3)
Describe the provisions =ade to assure representative sa=pling of processed liquid radwarte prior to disposal in conformance with Regulatory Guide 1.21.
01.2 (11.2.2)
The Miscellaneous Waste Chain wastes from Three Mile Island Uni: 2 are transferred to the Three Mile Island Unit 1 Miscellaneous Maste Chain for treatment by evaporation and demineralization.
Since only one 12.5 gpa evaporator has been provided for the wastes fro = both plants, discu ss your contingency plans for the treatment of this waste strea=
in the event of a prol;nged outage of the evaporator unit and/or for periods of abnormally high radwaste input flows.
01.3 (11.2)
Provide a table listing tanks outside reactor containment which contain potentially radioactive materials.
The table should include tanks located both inside and outside of plant builcings.
For each tank indicate the provisions incorporated to monitor tank levels, to annunciate potential overflow conditions, and to collect and process liquids in the event of an overflew.
Acceptable provisions.aciude dikes areunc tanks, retentien basins, and elevated thresholds to contain liquids in boys containing the tanks.
01.4 (11.3)
For HEPA filter and charcoal adsorber installations in building ventilation systems, provide the following infor=ation:
(1)
Acceptable design practice in the use of HEPA filters limits individual filter systens to a capacity of 30,000 ef= (11=it 3 HEPA filters high and 10 HEPA filters wide).
Justify your operating the Auxiliary 3uilding HEPA filter syste=s (2 banks of 30 HEPA filters) at a design flow rate of 41,945 cf= per syste=.
Describe the layout of the filters for each syste=.
(2)
List the assumed residence time for charcoal adsorbers
.uo the efficiency of iodine retentien. Justify your assumed efficiency for iodine retention by test data (te=peratuce, pressure, hu=idity, expected iodine concentrations, anc flow rate).
Descrice tne erfects of aging anc poisenia; by =oisture and airborna contaminants and support your assu=ptions by test cata.
4r 029
01-2 01.5 (11.3)
Describe the Waste Gas Filter Syste= identified as WDG-F-l'on Figure 11.3-1.
Discuss your reasons for not croviding HEPA filters after charcoal adsorbers in the Waste Gas Filter Syste=.
01.6
( 11. 3. 2.1)
In describing your provisions to eli=inate the 1,ossibility of a hydrogen explosion, you state that the Radwaste Gas Syste= is periodically =onitored at.several points for oxygen and hydrogen. Show the locations of these =onitors on the P&ID's, and justify why continuous =enitoring is not needed.
01.7 (11.5.6)
Specify the di=ensions and location of the Additional Storage Area provided for packaged solid radwaste on the 305' elevation of the Fuel Handling Building.
Provide the nu=ber of dru=s and types of wastes to be stored in this location and the provisions for =onitoring and surveillance of stored wastes.
Describe the storage facilities provided for the decay of short-lived radionuclides prior to ship =ent and give the anticipated storage time prior to ship =ent and the bases for'the value selected.
01.8 (11.5)
Describe =ethods that will be used to ensure the absence of free water in solid waste dru=s.
01.9 (15.24)
Provid, an analysis indicating the radionuclide concentrations which could occur in both 1) the nearest potable water supply, and 2) the nearest surface water in an unrestricted area as a result of leakage based on single failures of ce=ponents located outside reactor contain=ent containing radioactive liquids.
Assu=e 1% of the operating fission product inventory is released to the pri=ary coolant, failed tanks release 80% of their design capacity, and all liquids fro = failed co=ponents enter the groundwater; i.e.,
do not assu=e liquids are retainea by building foundations.
Credit for radionuclide re= oval by the plant process syste=s, consistent with the deconta=ination factors in WASH-1253 should be assu=ed.
List all para =eters and provide justification for the values assu=ed in your calculations, including liquid dispersion and transit ti=e based on distance, the hydrculic gradient, per=eability and effective porosity of the soil, and the assumed decontamination due to ion exchange by the soil.
43 030
02-1 02.0 AUXILIAEY AND PCWER CONVERSION SYSTEMS 02.1 Discuss the effects of aircraft impact on the reofs of the (3.5) buildings listed in Table 3.5-1 with regard to consequences that =ay affect safe plant shutdown capability.
Include in the discussion a description of the added effects of burning fuel on the roof tops.
02.2 Provide a cocplete tabulation listing all =ederate and high (3F) energy syste=s outside containment.
(For definition see Appendix A of Mr. J. F. O' Leary's letter dated July 12, 1973.)
Describe all systems that have not been already described in the FSAR.
For those syste=s that have been described, identify (on the tabulation) the section(s) where the descrictica may be found.
02.3 Describe the location, physical separation, or protective (3F) barriers provided for the auxiliary feedwater pumps to en-sure their operation if flooding or gross failure of adjacent components or piping were to occur.
02.4 Discuss the potential for flooding safety related equipcent (3F) in the event of a failure in any line carrying high energy fluid.
02.5 Provide the results of an analysis which demonstrates that (3F) critical cracks in high energy lines will not have adverse ef fects on essential equipment.
02.6 In order to =itigate the consequences of many of the postulated (3F) high energy breaks, equipment which is non-seis=ic or dependent on off-site power is relied upon.
Provide the results of a high energy linebreak analysis assuming the non-seismic equipment is not available and off-site power is lost.
Provide plan and elevation drawings which de=enstrate that pipe res train ts,
walls or the relocation of equipment will avoid damaging safety related structures, equipment or cocponents.
02.7 Provide the seismic design requirements for the diesel generator (8. 3) e xhaus t system.
02.8 Provide a discussion of the astalled protective type devices (8. 3) that are incorporated in,ae design to protect the diesel gene rato rs from exceeding operating limits or otherwise pre-vent them from performing their intended function during a DBA.
'ahat measures will be taken to minimize the possibility of the above devices from needlessly preventing the diesel f rom operating when required.
49~():32.
02-2 02.9 The FSAR states the diesel engine day tank capacity is sufficient (3.3) for approxi=ately 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at full load operation.
Provide a discussion of the factors considered in arriving at this capacity.
Include in the discussion the range of malfunction ccasidered, and the time interval between the low level c 1 arm and when the day tank will be empty.
Relate the time period required to carry out the various remedial actions of the =ain fuel storage system to the ti=e period available.
02.10 In regard to potential failures or malfunctions occurring due (9.0) to freezing, icing, and other adverse environ =cntal conditions for those components not housed within temperature controlled areas and which are essential in attaining and naintaining a safe shutdown, identify and discuss the protective measures taken to assure their operation.
02.11 Provide a tabulation of all valves in the reactor pressure boundary (9.0) and in other seismic Category I systems, as reco== ended in Regula-tory Guida 1.29, e.g.,
safety valves, relief valves, stop valves, stop-check valves, control valves, whose opertion is relied upon either to assure safe plant shutdown or to mitigate the consequences of a transient or accident.
The tabulation should identify the system in which it is installed, the type and size of valves, the actuation type (s), and the environ = ental design conditions to which the valves are qualified.
02.12 For all vessels that contain gas under pressure (such as nitrogen, (9.0) chlorine, hydrogen, oxygen, air and CO t nks) pr vice the 2
following:
(1) the design and operating pressure, (2) the maximum pressure cf the gas supply, (3) the location of the vessel,
~(4) the total energy released if the largest pipe connected to the vessel should rupture, and (5) the protective measures taken to prevent the loss of f unctica of adjacent equipment essentici for a safe and =aintainad reactor shutdown 02.13 Provide the results of an analysis which demonstrates that failure (9.0) of any non-Category I auxiliary system or ccmponent ( f.ncluc _ a; associated turbine systems and components) will not as e a detrimental effect (such as flood, spray, leaks) on safety related systems or prevent safe shutdown of the plant.
@ ~ 0N
02-3 02.14 According to Section 9.1.2 the key-interlocks which prevents (9.1. 0) the fuel handling crane from passing over the spent fuel pool are imposed only when the crane load exceeds 15 tons.
Provide the results of an analysis of the effects of dropping a crane load approaching 15 tons into the spent fuel pool.
Assur.e the object is dropped from the maximum lifting h2ight attainable by the crane hook.
02.15 Provide a description of the scismic Category I water source (9.1.3) and =ake-up system to the spent fuel pool.
Include in the description the redundancy or backup provided for the make-up system.
07.16 According to Figure 9.2.3 the spent fuel pool cooling system is (9.1.3) interconnected with the decay heat removal system.
Provide a description of the means of isolating the two systems and the restrictions under which the DHR system will be tied into the spent fuel pool cooling system.
02.17 According to Figure 9.2.3 the outlet of the spent fuel pool (9.1. 3) cooling system is well below the normal fuel pool water level.
Provide a description of the anti-siphoning devices incorporated in the design of the pool cooling system to prevent draining of the fuel pool assumitga pipe or co=ponent failure.
02.18 Present information to demonstrate that a power failure during (9.1.4) refueling or spent fuel handling operations will not cause dropping of the fuel assembly.
02.19 Describe and discuss the plans and means provided to acsorb the (9.1 4) resulting impact should the spent fuel cask be dropped in the cask pool.
The discussion and analysia should inclu'c:
-(l) An outline drawing of the cask, cask dimensions, and center of gravity.
(2)
The cask weight, assumed drop height, deceleratien diste..c, deceleration force versus stroke, velocity at impact.
(3)
The maximum possible drop height.
(4)
The = cans, aside from administrative control, to limit the drop height to that assumed in the analysis.
(5)
Information which S i.onstrates that the cask cannot be tipped before bain' dropped or if tipped what ptovents the cask from danau-., the stored fuel or other safety related equipucat.n c1:. :... those that may be below the operatir.'
.;ov..
49-C33
02-4 02.20 Provide a list of all major tools and servicing equip =ent (9.1.4) including cranes necessary to perfor= the various reactor vessel servicing and refueling; " unctions and indicate whether each is designed to seis=ic Category I requirements or their storage locations are designed to these requirements.
02.21 Descrite in detail the applicable codes and standards used in (9.1. 4) the design, fabrication, installation and testing of crane, rails, supporting structures, bridge, trolley, hoists, cables, lifting hecks, special handling fixtures and slings.
02.21 For each crane, list its design load rating preoperation test (9.1.4) load, =aximum operating loads and the test loads that will be used throughout the life of the facility.
02.22 Describe the modes of failure that were considered in the design (9.1. 4) of the spent fuel cask crape and reactor polar crane such as breaking of cables, lffting slings, sheared shafts, keys, stripped gear teeth, and brake failures.
Also discuss the limitations and contrcl that will exist in handiing objects cver an opened reactor vessel.
02.23 What are the geometric changes of load position that =ay occur (9.1.4) in the event of talfunction or failure in the hoisting syste (the hoisting syste= includes the load and all items of mechanical and structural support on the bridge trolley).
Provide an evaluation of the ef:ects et these geometric changes on the fuel handling and storage area and any other safety related equip =ent.
02.24 Provide an outline of the cask handling procedure including a (9.1. 4) sketch or draving which shows the routinc of the spcr.: fuel hand-ling cask from receipt to the pool for loading with cpeut f :el
,to its return to the transporting car ready for ship =ent fro:
the nuclear plant.
02.25 Describe the protective censures provided to avoid dacage to a (9.1. 4 )
fuel element due to the fuel transfer cart moving through the fum_
transfer tube while the fuel element is in the vertical positicn.
02.26 Since the fuel handling building crano is shared with Unit I, (9.1.4) discuss the means of preventing a heavy object taken fro Uni: I from being carried over the fuel storage areas of Unit II.
49~034
02-5 Provide a description including drawings of the use of the 02.27 outlet water from the nuclear service river water system for (9.3.1) deicing the river water intake structure.
In particular dis-cuss the potential of short circuiting the cooling effect of the service water system due to the warm outlet water from the Nuclear Service River Water System (NSRWS) being drawn into the river intake structure.
Figure 1.2-20 shows the river water pump house contains four 02.28 nuclear services water pu=ps and three non-seistic secondary (9.2.1)
Provide the results of an analysis which service water pu=ps.
demonstrates that the failure of any component in the pu=p house will not da= age safety related equipment due to internally generated missiles or flooding.
02.29 It is apparent from Figure 9.2-1 that the failure of a single line of the NSRWS can result in the loss of cooling water to (9. 2.1) the following components:
(1) control builcing liquid chiller condenser (2) control building river water booster pumps (3) control building mechanical room fan cooling units (4) control building area fan coil units Discuss the effects on the safe shutdcun capability of the plant due to the loss of cooling water to the above.entionel systems and describe any modification that could be made to eliminate the single failure.
According to Section 9.2.1.1.2 a non-seismic techanical draft 02.30 Discuss (9.2.1) cooling tower is employed to cool the NSRWS discharge.
the means of preventing the debris due to the collapse of the cooling tower from restricting the flow of the NSRWS.
Provide the nuriber of Nuclear Services Closed Cooling Water 02.31 (9.2.2.3)
System (NSCCWS) pu=ps required for a nor=al cold shutdown cf the plant.
According to Section 9.2.2.3.3, a backup supply of cooling water 02.32 (9.2.2.3) to the nuclear orientei equip = cat ecclers which are essential for post-LOCA cooling recuire=ents is availa' ale from the N51WS.
Identify the intercennections and valving arrangements of the NSRWS with all the eccential coolers shown on Figure 9.2-3.
49~035
02-6 02.33 It is apparent from Figure 9.2-5 that a failure of a single (9. 2. 2. 3) line in the NSCCWS can result in the loss of cooling water to any one of the essential coolers.
Discuss the effects on the safe shutdown capability of the plant due to the loss of cooling water to each of the essential coolers.
02.34 Identify all ce=ponents that have a single barrier between the (9.2.2.3)
MSCCWS and the reactor primary coolant system and describe the isolation and/or pressure relief provisions provided to preclude da= age to the NSCCWS.
02.35 Damonstrate that in the event of a system leuk or rupture, the (9. 2. 2. 3) component cooling surge tank capacity is adequate to assure a continuous supply of component cooling water to equipment re-quired for safe shutdown until the leak can be. isolated.
Describe any automatic devices provided to mitigate the effects of system leakage or rupture.
02.36 Discuss the design provisions provided to prevent the floeding (9. 2. 3) of essential. equipment in Unit 2 due to the rupture of the denineralized water storage tanks of Units 1 or 2.
02.37 Provide a legible plot plan of the facility indicating and (9.2.5) identifying all essential lines (cooling, pcwer, sensing, and control) that pass betweta seismic Category I structures.
Dis-cuss the measures taken to prevent the loss of those lines re-quired to attain and maintain a safe shutdown due to seismic event, missiles from rotating equipment and tornadoes, fires, floods and the collapse of non-seismic structures.
02.38 List all air operated valves whose =alfunction can affect plant (9. 3.1) safe shutdown, provide their failure mode and demonstrate that their failure mode will not compromise safe shutdown of the
. plant.
02.39 Frovide additional e:<planation and assumptions used in determining (9.3.3) the drainage systems adequacy for precluding backflooding frcn one compartment to another containing safety related equip :n:.
02.40 Provide the seismic category for sump pump discharge systems (9.3.3) that are required to prevent the flooding of areas containinz essential equipment.
02.41 Provide the maximum allowable temperature for the take-up (9.3.4) purification demineralizers mixed bed and cation bed resin and the consequences of c:<.;ecding this temperature.
4r036
02-7 02.42 The letdown temperature in the letdown line downstream of the (9. 3. 4) csolers is alarmed and provides an interlock for isolation to protect the purification system.
Is the letdown temperature always indicative of the pressure associated with the letdown system?
Discuss the effects that the interlock failure would nave on the purification system.
How would excessive te=pera-ture and pressure otherwise ba detected?
02.43 Letdown flow rates are controlled by a fixed block orifice, a (9.3.4) parallel remotely operated valve, and a second manually positioned valve also parallel with the block orifice.
Discuss the operation of these valves and describe the associated conditions required for operation.
Consider also the effects on letdown flow and system pressure with either one of the va)ves open and with both valves open.
02.44 In addition to the nor=al mahe-up line, two alternate paths (9.3.4) for adding boron to the reactor coolant system are identified.
Determine the limiting condition for boration and provide the margin associated with the alternate injection method to maintain suberiticality during reactor cooldown or accident conditions.
02.45 Provide a discussion of the means of isolating the chemical (9. 3. 4) addition systems from safety related systems.
Include in the discussion the number and seismic classification of the isolation valves and identify them on the appropriate PSI diagrams.
02.46 Provide the seismic classification of the fire danpers in the (9.4.0) air intake tunnel.
02.47 Provide the temperature and humidity of the outside air used in (9.4.0) the design of the various heating and air conditioning systems.
02.48 (RSP)
.According to Section 9.4.1.2 make-up water for the chilled water (9. 4.1) loop is supplied from the non-seismic demineralized water syste=.
It is our position that a seismic Category I make-up source be provided for the control room chilled water system.
02.49 F"rovide a descr'iption of the instrumentation used to detect fire, (9.4.1) snoke and radiation in the air intake and in the ducting asse-ciated with equipment that must be protected.
02-8 Discuss the consequences of the single exhaust damper for 02.50 (9.4.1) untrol room kitchen or toilet f ailing open during 4ccident which could cause the intake of radioactive nor n.y.lous gases to the control room.
Discuss the consequences of the damper in the irtake of the by-02.51 supply system failing closed during an accident which (9.4.1) pass requires the isolation of the control room.
02.52 State the design ventilation capacities required for the (9.4.1) control room, equipment and cable room ventilation systems.
This should include flow rates, cooling and heating require-ments.
Discuss the consequcnces of the loss of off-site power on the 02.53 (9.4.1) control room air conditioning, heating and ventilation system.
02.54 Provide the,results of a failure modes and effects analysis (9.4.2) of the auxillery building heating and ventilation system including loss of off-site power.
In particular discuss the effects of a failure in the intake supply da=per and supply fan on the temperature of compartments containing essential equipment.
02.55 Discuss the effects on essential equipment contained in the (9. 4. 2) auxiliary building when the non-seismic heating and ventilation system components do not function or are isolated due to an earthquake.
02.56 Identify on Figure 9.4.3 the ducting and isolation da pers in (9.4.2) the auxiliary building ventilation system that are seismic Category I.
49-J38
02-9 02.57 Provide the results of a f ailure mode aad ef fect analysis (9. 4. 3) for the fuel handling area ventilation sys tem, including the effects of the inability to maintain preferred air flow patterns.
02.58 Since there are o radiation monitors at the spent fuel poci (9. 4. 3) surface provide t_e results of an analysis which demonstratas that the =enitor upstrea= of the exhaust filters can detect the radiation fast enough to divert the exhaust air through the filters before the contaminants can escape to the environ-ment.
02.59 Identify on Figure 9-4.4 the ducting and isolation dampers (9.4.3) in the fuel handling building that are seismic Category I.
02.60 Provide the seismic classification of the isolation dampers (9. 4. 3) and the exhaust system radiation monitor.
Also discuss the consequences of a failure in the radiation monitor during a fuel handling accident.
02.61 According to Section 9.4.6-3, "If combustible fu=es enter the (9.4.5) 200 CFM ventilation air line, they will be detected by an infrared analyzer in the pump house which shall cause the butterfly valves in the air supply lines to close and the ventilation fans to stop."
Discuss a) the consequences of a failure in the inf rared analyzer, and b) the effects on operation of the nuclear service water pumps with the ventilation fans turned off.
02.62 Since a portion of the river water supply to the pu=p house (9.4.5) cooling coils is non-seismic, discuss the effects of the loss of the coolin; coils en the nuclear services cooling water pumps.
02.63 Provide a description of the enhaus t portion of the river (9.4.5) water pu=p house ventilation system.
02.64 Provide a failui; modes and effects analysis, including les:
(9.4.7) of off-site power, for the cable room ventilation systen.
W~ COO
02-10 Provide the seismic dlassification and a f ailure modes and 02.65 ef fects analysis of the isolation dampers for the service (9. 4. 8) building ventilation system.
According to section 9.4.6.3 combustible fumes from the 02.66 contral building area will be detected by an infrared analyzer.
(9.4.9)
Discuss the quantity and type of combustible fu=es that can be generated in the control building area and the means by which they are generated.
Discuss the consequences of a seismic event disabling the 02.67 non-seismic control building heating and ventilation system (9.4.9) that could ef fect safe plant shutdown.
If one of the two fan coil units were down for maintenance, 02.68 the other unit will satisfy the heating load temporarily.
(9.4.9)
Provide the duration that the one 50% fan coil unit could satisfy the control building area heating load.
According to section 9.5.1.2 a halogen deluge system protects 02.69 (9. 5.1) the air intake tunnel.
Discuss the consequences of the halegen deluge system leaking into the air intake tunnel and the Describe resulting fu=es being drawn into the control room.
the means by which the control rocm personnel would become aware of a leak in the deluge sys tem.
Demonstrate how equivalent Jafety is achieved by the present 02.70 (9.5.1) fire protection system in order to meet the positions set forth in Regulatory Guides 1. 70.4 and 1. 73.
Provide the results of a failure mode and ef fects analysis for 02.71 the fire'p itection system, including an analysis of potenti:1 (9.5.1) adverse ef fects caused by operation of the system.
.sise pro-vide a discussion relating to the reliability of the fire detection equipment in te rms o f s ens itiv it y, me an time b e twee n failures, and other operational experiences.
02.72 Eascribe and identify the location of any applicatica of (9.5.1) polyvinyl chloride (PVC) in the construction of the plant.
Discuss protective features provided to prevent or control burning or overheating of such material in the plant.
02.73 One deluge system protects thc tuo cooling air intahe openins" (9. 5.1) to the emergency diesel generator building.
Discuss the effects on the perfor-.tce of the diesels if water is drawn into the diesel gr cra ac racms through the cooling air intakes.
Mr 040
02-11 02.74 Describe the potential fire hazards in each plant area, (9. 5.1) fire protection requirements and the fire risk evaluation utilized in the design of the fire protection system.
02.75 Discuss the potential of a fire protection system storage (9.5.1) tank rupture and the ef fects upon safety related systems.
02.76 Demonstrate with elevation drawings that the fire pump (9. 5.1) locations are compatible with minimum and maximum supply source levels.
State the required and available NPSH at mini =um supply levels.
02.77 Provide a discussion of the precautionary measures taken to (9. 5.1) prevent the buildup of flammable mixtures of hydrogen given off by the batteries in the battery room.
02.78 Provide a description of the protection metheds provided (9. 5. 4) for the fuel oil system piping and tanks from tornado udssiles.
02.79 Discuss the ef fects that a major leak of a fuel oil storage (9. 5. 4) tank would have on essential equipment.
Discuss the =eans of containing the spilled fuel oil.
For the diesel generator cooling water system, air starting 02.30 (9.5.5) system and lubrication system, ' provide a discussion which (9.5.6) demonstrates the capability of the systems to satisfy the (9. 5.3) design basis including a failure modes and effects analysis In addition provide a description of the testing and inspec-tion programs to be perfor=ed on the systems.49-041
02-12 of normal turbine speed at which the 02.80 Provide the percent various turbine overspeed trips operate.
(10.2)
Provide elevation drawings showing the water level in the turbine builfing at various times after a complete rupture 02.81 of the main condenser circulating water rubber expanaion (10. 4. 5) discuss which, if any, joint.
For each time increment essential systems and compon2nts could be rendered inoperable.
Include in your discussion the coneideration given to passage-ways, pipe chases, cableways, and all other possible flow joining the flooded space to other spaces containing pat?3 Discuss the effect of the tial systems and components.
waters on all submerged essential electrical systems ess fic-and.c=ponents.
Desc ribe the means provided to detect a f ailure in the ti=e interval 02.82 circ 21ating water system and how and in what (10.4.5) flow will be stopped, considering all f actors, e.g., cpe rator time f or control circuitry and coas t-reaction time, drop-out down.
Referring to Figure 10.1-2, discuss the consequences of a 02.83 (10.4.7) rupture of the high energy emergency feedvater line between the steam generator and the valve EF-V123 coupled with a Such a sequence of single active f ailure in valve EF-V12A.
the ability of the plant postulated events should not effect to be safely shutdown.
e
03-1 CONTAB XUT SYSTEMS 3RANCE 03.0 The response to Question 9-5 in the prelininary review 03.1 (6.2.1.1) report concerning the contain=ent subce=partment pressure response is inco=plete.
Provide the following inf ormation:
As a minimum, the diff erential pressures resulting from (1) the f ollowing pipe breaks and break locations should be analyzed:
(a)
Hot and cold (pump suction and discharge) leg RCS ruptures in the steam generator compartment, reactor cavity and primary shield wall pipe penetrations.
(b) Pressurizer spray and surge lines in the pressuri-zer co=part=ent.
(c)
Steam, f eedvater, and/or any other lines carrying high energy fluid with the potential to over-pressurize compart=ent walls, barriers and floors the f ailure of which might eff ect the perfor=-
ance of equipment necessary for the safe opera-tion and shutdown of the plant or containment integrity.
(2) For each co=partment analyzed:
(a)
Describe the nodalization sensitivity studies per-for=ed to deter ine the mini =u= nu=ber of volu=e nodes required to conservatively predict the The maximum pressure for each subecmpart=ent.
nodalization sensitivity studies should include consideration of spatial pressure variation; i.e.,
pressure variations circumferentially, axially and radially within the subcompartment, par-ticularly in the reactor cavity.
(b)
Provide schematic drawings showing the nodali-sub-zation of each subec=part=ent or co=partment division indicating nodal net free volumes and interconnecting flow path areas.
(c)
Provide sufficiently detailed plan and section drawings f or several views showing the general arrangement of subcompartment structures, components, piping, and other major obstructions from which subconpartnent vole =es and flow areas cca E J_t c=1.al.
d8'C_QJ
03-2 03.1 (a)
Provide and justify the break type and area (6.2.1.1) used in the analysis.
(e)
Provide and justify values of vc.nt loss co-efficients and/or friction factors used to cal-culate flev between nodal volumes. When a loss coefficient consists of more than one ce=ponent identify each component, its value and the flow area at which the loss coefficient applies.
(f)
Discuss the manner in which scvable cbstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated.
Include cnalytical justification if credit is taken for the removal of such ite=s to obtain vent area.
Provide assurance that vent areas will not be partially or completely plugged by displaced obj ects.
(g) Provide a table of blowdown = ass flew rate and energy from that postulated that would result in the highest differential pressure for each compar t=ent.
(h)
Provide a curve of differential pressure as a function of tire indicating spatial response for the analysis of the reactor cavity.
(1)
Provide the design differential pressure, peak calculated diff erential pressure, and ti=e of peak pressure for each cc=part=ent.
(3)
With regard to method of analysis for subcc=partments:
(a)
Provide the na=e and a detailed description of the blowdown and pressure transient code (s) used in the analysis.
The description shculd include all =athematical correlation to determine the subsonic and sonic vent flow.
(b)
Justify the blowdown model~used showing that it maximires the mass and energy release rate.
(c)
Provide and justify the critical flow model used in the blowdown analysis and the break discharge coefficient applied.
(d)
Provide and ustify, preferably by comparisen with experine:: t al c e t a, the c pation or correlation
--.us;mte fim, m_twaea cospartments.
esca
.s 48-C:44
03-3 Include a discussion of the critical flow 03.0 (6.2.1.1) model and discharge coefficient applied to critical flow.
(e) Discuss the method of treating the air-steam-water mixture in subco=partnent ther=odyna=ics.
03.2 If sand plugs are utilized as radiation shieldings inside (6.2.2.2) the containment, provide the following information:
(1)
Describe the sand plugs identifying the materials of construction.
Provide drawings to indicate design details of the structures.
(2)
Describe and discuss the design provisions made to ensure that the sand plugs do not becc=e damaging missiles following a LOCA.
(3)
Discuss ?.he design provisions of the containment sumps, spray no::les, pumps and other safety equip-ment to acco modate the release of sand to the cen-tainment atmosphere.
03.3 Table 9.2-2 shows that a significant amount of energy (None) stored in the steam generators has not been released to the containment during post-reficod period.
Justify that this energy re=ains in the steam generators.
We find that nucleate boiling will exist on the steam generators primary side for reverse heat flow during the blowdown, reflood and post-reflood periods.
03.4 Provide the following infor=ation for a double-ended pu=p (6-2.1.3) suction break:
(1)
Analyses of mass and energy rates released to the contain=ent to include the following:
(a) multinoding in pri=ary metal slabs; (b) no quenching f rom LPI pump flow; (c) no quenching from HPI pump flow or flood tank flow af ter blowdown; (d) nucleate boiling on primary side of steam generators and condensation on secondary side for reverse heat flew; (c) all avail bIc ctaam generator energy using the assumption that nucleate boiling exists in the secan generator tubes for reverse heat flow.
4r C45
t 03-4 03.4 (2) The results of the contain=ent pressure transient analyses to include the above energy release.
03.5 The FSAR indicates that the cc=puter programs, FLASH and (6.2.1.2)
PRIT were used for hot leg bleak analyses and CRAFT was used for cold leg break analyses to predict the = ass and energy releases to the contain=ent during blevdown and post-blowdown.
Because CRAFT has been found by us to be acceptable for these analyses, justify the use af the FLASH and PRIT codes for the hot leg break = ass and energy release rates to be adequately conservative for contain=ent pressure analyses.
0?-6 Describe and justify the mesh spacing used to determine the (o.2.1.3) heat transf er to the containment concrete heat sinks.
We currently believe that about 0.1 inch mesh spacing should be used for the initial 3 to 4 inches of concrcte.
Describe and justify the assu=ptions used to =odel the heat 03.7 (6.2.1.3) transfer at the contai==ent liner to concrete interface; i.e., consideration for contact resistance or effects of air gaps.
The max 1=u= heat transf er coefficient of 620 Stu/hr-f t
- F 03.8 (6.2.1.3) obtained frc= Kolflat's work was utilized in your contain-
=ent pressure response analysis during the blowdown phase of a LOCA.
This initial high heat transfer coefficient for a large contain=ent was not substantiated by the work reported by Taga=1.
Provide an analysis of the containment peak pressure utill:1ng heat transfer coefficients during the blevdewn phase of the accident which are consistent with the data reported by Taga=1.
In addition, provide the heat transf er coefficient which was used in your pressure response analysis as a function of ti=e af ter LOCA.
03.9 With regard to the reactor building spray syste=, provide (6.2.2.2) the following' infor=ation:
(1)
A discussion of the NPSH requirements for the R3 spray pumps with supporting infor=ation (i.e., static elevation head, the f riction head loss in the suction piping, the vapor pressure of the fluid and the reacter building pressure) to show the cargin between the
_equired and available NPSH to demonstrate confor=-
ance to the guidelines of Regulatory Guide No. 1.1.
(2)
A dire'tetic-ef t'^ deri:n pr-"isic-s chich perr.it the spray vater that may enter the refueling cavity 43 - CN
03-5 and the reactor cavity following a loss-of-coolant 03.9 (6.2.2.2) accident to be drained to the containment sump.
With regard to the housing and ductwork f or the reactor 03.10 building f an cooler system, provide the following infor=a-(6.2.2.2) tion:
(1)
An analysis of the pressure differential for the housing and ductwork.
(2)
The design pressure and the calculated differential pressure as a functicn of time for the hcusing and ductwork.
(3)
A detailed description of the analytical model, assumptions and appropriate bases used in calcu-lating the pressure differentials.
03.11 Analy:e the snount of contain=ent atnosphere that will be (9.4.15.2) released to the environ =ent if the containment purge vent valves are open at the time of a loss-of-coolant accident.
Describe the essential specifications used in the design 03.12 (9.4.15.2) of the contain=ent vent and purg2 valves (i.e., design radiation, te=perature, pressure, dif f erential pressure,
and dynamic forces).
Describe analyses or tests that are conducted to de=onstrate that the valves will operate as specified.
03.13 Describe the provision for =enitoring hydrogen in the (6.2.5.2) reactor building following a LOCA.
The discussion should include the following design considerations:
(1) the number and location of samplin; points within the reactor building; (2) seismic design classification of the system.
03.14
- Identify those systems or portions of systems which vill (6.2.4) be open or vented to atmosphere and/or drained of fluids to assure that isolation valves will be exposed to contain-ment air pressure as required by Appendtc J.
Those systems not vented or drained and vnich form a part of the contain-ment boundary should be identified and the reason with justification for not venting and draining should be stated.
03.15 Airlocks should 'c. pr:ceuri:cd to ? following the sa=e schedule as preopern:ional and peri $dic Ty7e A tests.
(6.2.1.4)
@ ~ CN
03-6 03.15 Every six =nnths during periods when airlock doors opened the airlock door and penetration seals (6.2.1.4) are not should be tested at the door manufacturers recocmended test pressure.
During periods of door use when the reactor contains fuel, the door seals should be leak tested every three days.
03.16 Specify the acceptance criteria for the measured contain-leak rate (L_ ) at peak pressure and the acceptance (6.2.1.4) ment criteria for the db,atainment integrated leak rate verifi-cation test.
03.17 You state in Section 6.2.1.4.2.3 that the measured leakage shall not exceed the maximum allowable leakage (E'). We will require in accordance with Appendix (6.2.1.4) rate (L
)
- rate, J that (t ) shall not exceed 0.75 L.
g 03.18 For the Type A test you indicate that (16) temperature (6.2.1.4) detectors and (2) dewcells will be utilized.
To provide more accurate temperature and humidjty values, one of contain=ent volume temperaturedetectorper100,0g0ft and one deweell per 500,000 ft of volume should be provided.
03.19 You indicate that for the cent.ainment integrated leak test (6.2.1.4)
(Type A) the reduced pressure test will be performed sub-sequent to the peak pressure test.
In order to mininize stabill:ation ti=e and reduce off-gassing due to entrain-the following test sequence should be provided:
ment (1) reduced pressure test; (2) structural integrity test; (3) peak pressure test.
03.20 With ref erence to Techni;al Specification Section 4. 4.1. 2. 3, (16.4.4.1) you state that local leak testing shall be perforced at each refueling.
The frequency ct local leak testing shcu.c not exceed two years as stated in A7pendix J.
03.21 The Technical Specifications should state compliance wicS (16. 4. 4.1)
Appendix J of 10 CFR 50.
03.21 With reference to Section 4.4.1.1.5.C, you state that if (16.4.4.1) rwo consecutive Type A tests fail to meet acceptance cr:;eria, a Type A test shcil be performed at each unit shutdotin for refueling.
The follewvp cent is required by Appencix J to be perfor=ed au cach plant shutdown for refueling or mouths wuicuevet occurs..uo-.
appcexicaseiy u
,y
.s 4W C48
11-1 11.0 MECHANICAL ENGINEERING 11.1 Reconcile (a) the statements made in 3.6.2.1 regarding postulated (3.6) pipe break criteria ins:Ae containment, (b) the entries in Table 3.6-2, and (c) the locations shown in Figures 3.6-8 and 3.6-11.
In particular, notice that the figures do not indicate postulated breaks at the terminal ends whereas the criteria given in the_ text in 3.6.2.1 do call for such breaks.
The table appears to agree partially with the criteria in the text although the agreement is not ccepletely una=biguous.
In making this reconciliation, extend it to any other figures and tabular entries which may be similarly involved.
11.2 In the discussion under 4-2/3.8 of Supplecent 1 of the FSAR (3.6) regarding Safety Guide (Regulatory Guide) 1.46, clarify the meaning of the expression "more than the mini =um of 2".
Recall that the Regulatory Guide 1,46 calls for a minimum of twt intermediate locations in addition to the two terminal ends of a piping run or branch run.
The discussion under 4-2/3.8 of Supplement 1 appears to state that pipe breaks have been considered in even more piping runs than are called out by Regulatory Guide 1.46.
If justified, previde a specific stater.ent that all high encr;y lines inside containmant which would requite postulating of breaks by virtue of precise and complete application of Regulatory Guide 1.46 have been discussed with regard to postulated pipe breaks in the FSAR.
If this is not the case, describe and justify the exceptions.
11.3 Provide specific assurance that the piping systems incide (3.6) containment hriebeen properly examined from the standycint of the physical impact of postulated broken lines en saf ety-related lines of smaller diameter or wall thickness.
State the criteria used to insure the preventien of unacceptable dacage.
11.4 Submit the design criteria for pipe runs which extend frca (3.6) the containment penetrations to the first isolation valve outside containment.
11.3 Basing your response en a general survey of the plant, identify (3.6) those safety-rclated structures, systems, and components where you might anticipate a problem in their survival and proper functioning if the criterion for the conical spread of an i= pinging j et f cca a postulated pipe break or crack S
'l.. ' f' A n AC
< sU
11-2 were a ten degree half angle or, alternatively, a twenty-two-and-a-half degree half angle out to five pipe diameters from the centerline of the source pipe followed by tero spread beyond that point.
11.6 You have specified the Oconee Unit I reactor internals as the (3.9) prototype of the 177-fuel assembly units, which include TMI-2.
C1carly state your intention to perform confirmatory testing of the TMI-2 reactor internals using procedures delineated in Regulatory Position D of Regulatory Guide 1.20.
Indicate which option of said position you plan to use and briefly describe your proposed progrr.m of testing.
Identify any design differences between the designated prototype and TMI-2 which may lead to different response behavior of the reactor internal structures resulting from flow-induced vi-bration and evaluate the significance of these differences.
11.7 Expand the description of the preoperational piping vibration (3.9) testing program in 3.9.1.1 of the FSAR.
Supply a more explicit statement regarding the flow modes of operation and transients to which the system components will be subjected and the criteria for establishing acceptable limits of response.
Essential elements of an acceptable program arc outlined in Attach =ent A, "Preoperational Piping Dynar.ic Ef fects Test Program."
11.E In FSAR 3.9.2.2, the use of the AEME 3 & PV Code Section III (3.9)
" Hopper Diagram" is described in connection t:ith the desien of Architect / Engineer Class 2 and 3 components to which emergency and faulted conditions were applied.
Provida assurance that these considerations regarding emergenc/ and faulted conditions or their equivalent were applied c: cil such components which are classified Category I or whose failure could result in the loss of structural and functional integrity of a Category I component.
11.9 Clarify the statements in 3.9.2.4 regarding the design of (3.9) pumps for dynamic loads.
Whether an equivalent static load was derived frem a dynamic analysis usi:ig input based on response spectra curves or was inferred diructly from such curves, indicate explicitly that the reupensa spectra curves in were applicable to the '5vaical location cf :he cor.ponent or that response-the fluid piping system or in the structure spectra curves of equivslant or greater conservatism were used.
- 48) 43fi()
11-3 11.10 In addition to the infor=ation supplied in ;.o.3.2, sub=it a (3.9) list of all items of Category I =echanical equipment which have been seis=ically qualified.to assure their ability to withstand the effects of the SEE and any plant operating condition including f aulted (includes SSE) without loss of safety function.
For each ite=, indicate whether the qualification was perfor:ed by testing or analysis or a co=bination.
The term mechanical equipment in this context applies to ite=s such as fans, filters, other ventilation, gas control, and safety related cleanup equipment, pump drives, valve operators, and heat exchanger bundles.
For guidance see Attachment 3.
11.11 In connection with the design criteria for the mounting of (3.9) the pressure relieving devices provided for overpressure
--protcetion of Category I, ASME Code Class 2 systc=s and
~
components, describe the methods used to account for possible dyna =ic amplification of the effects of the blow-off loads.
Also provide assurance that the design of the valve counting is adequate for a postulated sequence of discharges of valves on a co==on header which are such as to produce the greatest ratio of estimated instantaneous stress to allowable stress at any point, independent of the relative set-points of the valves.
11.12 In addition to the infor:ation already provided in Tables (3.9) 3.2-1 and 5.2-3, submit a list of all active ASME Code Class (5.2) 1, 2 and 3 pumps and valves, that is a list of all c 2penents whose functional as well as structural and pressure-retaining integrity =ay be required for safe shutdown under any plant operating condition.
For each item, indicate whether cper-ability is assured by testing, by design analysis, or by a combination of testing and analysis.
Describe the basis for the assurance of operability (conservative design stress 11=its or pressure rating appear to be enc of the bases stated in this FSAR in =ost cases.)
11.13 In the case of activ s safety / relief valves, if any, and (3.9) active valves with physically extended va]ve operators, (5.2) describe the measures taken in design analysis to include the effects of dynamic a=plification of seismic or other vibration induced loads in the cantilevered portions when equivalent static load techniques are used.
49-C51
11-4 11.14 Expand the discussion in 3.10.1.3 of the use of shock testing (3.10) of safety related electrical equipment as an alternative to vibratory testing.
That is, submit facts and details to substantiate a conclusion that the shock testing employed is adequate to represent the specified dynamic input in terms of its resultant effects.
In the case of equipment and supports with essentially linear force-displacement characteristics, recourse to harmonic spectrum arguments is acceptable provided due consiceration is given to the effect of =odal participation factors and = ode shapes.
Where the equipment has sufficiently non-linear characteristics to limit the validity of the harmonic spectrum approach, indicate and justify your degree of confidence that the shock signature has adequately exercised the equip =ent.
Discuss the adequacy of the shock tests with respect to the possible~ build-up of low cycle fatigue damage during the postulated plant operating condition.
11.15 The discussion of active RCPB components in Section 5.2.1.
(5.2) of the FSAR does not provide convincing assurance of the operability of these components under faulted plant conditions.
Acceptable programs for assuring operability of active compo-ncnts are delineated in Attachments C and D " Class 1, 2 & 3 Pump Operability Assurance Program" and " Class 1, 2 & 3 Valve Operability Assurance Program." Supply assurance that the procedures enployed in designing and furnishing the active components in the RCPB provide an acceptably equivalent degree of confidence in their operabilit/.
11.16 Supply a description of the finite element computer progr:r (5.2}
mentioned in the discussion of pv=p casing design analysis in Section 5.2.1.7 of the FSAR.
Indicate the contrcl measures used to verify the validity of the program and tts applicati~c Adequate procedures are delineated in Attachment E "Acce,t bility of Computer Programs Analysis of Mechanical Co=ponents and Equipment."
11.17 In connection with the design and furnishing of piping (5.2) connected to pressurizer safety valves in a manner so as not to. exceed the maximum permissible loads on the valves, as you have stipulated in f action 5.2.2.2 cf the FSAR, explain the choice of stress allowables and the control of pipe dimensions e= ployed to assure such a result.
4r 052
6/6/73 Attachment A PRT.0""."*.TIO_: aL ?!p G Dv':' i'C C T: CTS TCST P:'.0GM:1 e -.1 pipt vf.bruienni and dyna.ic ef fects testina should be i
Preoperat r.r ' ctup f inc tic. a.1 testinr, on pipin;; sys t ems cad rest ra.nt',
c conducted v.rin, classified cs /,S'11', Class 1 and Class 2 components.
The purpose of these tests is to cc: P.-
- r these com;cr.unts have been desic,nul to withutr-'
the dyna::ic J ccid in ;.:, ro-ops rational transient corlitions that will be
~
encountered durin; service as required by AE!!E Code Section llI, par.
NB-3622. 3 and ':C-3522 -
An acceptable test progra= to confirm the ade-quacy of the dcai.;ns shoul.J ceasist of the follouing:
a.
A listing of the different flou nodes of operation and trranients such as pump trips, valve closures, etc. to which the cons.acnts will be. subjceted during the test.
- For c:: ample, the trancients acsociated with the Reactor Coolant Systca heatup tests should include, but not necessarily be limited to:
(1) Reactor coolant punp start (2) P,eactor ecolant pump trip (3) Operation of pressure-relieving valves b.
A list of selected locations in the piping system that will be subjected to visual inspection and =casurements (if needed) as For c ch perforned by the piping deslanar during there tents.
of thec2 c.e'.cc:.:J 1ccc. tie s, r ic :.llo =bi c defier-ice ( '. <
-to-peak) criteria that will. be appi tud to estai>1tsh that tiie stress limits are within the design levels should be provided.
If vibration ic a 3ted bevoad the acceptnnec levelc ret in c.
criteria es b. above, corrective restrajnt.s chould be 3:c!,
incorporated in the piping syst...i analysi.i sad ine.talled.
If during the test, the pipin;, cyste.- s restraints are deter v..i te be inadequate or da.na;,el, corrective restraint s should ;-..stalle:
and another test should be perfotaca to dcternine that the.ix ;ti
.s have been reduced to an acceptable level.
Jef erence AS:1E Code Section Ill, "':uclea r Pc.*er Plant Ccmponents' i:%! !it i on,1 qu i A- '. e f mc ' Sc e t i o1 of ecch trcccients is pr-d-e e eu.
e e
'l I-e g
ww w s,
% e e
b
/lfl~ (?fi3
12/5/73
- At cac h=ent 3 ELEC 11 CAL AND MECHANICAL ZGU1FME!.7 SIISMIC CUALIFICATION PROO?aM I.
Seis=ic Test for Ecuirmen: 0:erability 1.
A test program is required to confir the functional operability of all Seis=ic Category I electrical and mechanical equipnent and instrumentation during and after an earthquake of magnitude up to and including the SSE.
Analysis without testing =ay be acceptable only if structural integrity alone can assure the design intended function.
When a complete seis=ic testing is i= practicable, a combination of test and analysis may be accept-able.
2.
The characteristics of the required input motion should be snacified by one of the following:
(a) response spectrum (b) power spectral density function (c) ti=e history Such characteristics, as derived from the scructures or systc=s seis=le analysis, should be representative of the input =otion at the equip =ent mounting loca: dons.
3.
Equip = eat should be tested in the operational condition.
Oper-ability should be verified during and after the testing.
4.
The actual input =otion should be characterized in the same manner as the required inpuc =otion, and tne censervatis: in a=plitude and frequency corrent should be de=onstrated.
5.
Scis=ic excitation generally have a broad frequency content.
Random vibration input motion should be used.
However, single frequency input, such as sine beats, =ay be applicable pro ridad one of~ the following conditions are =et:
(a) The characteristics of the required input =otion indicate that the motion is dc=inated by one frequency (i.e., by structural filtering effects).
(b) The ant'icipated response of the equip =ent is adequately represented by one mode.
(c) The input has sufficient intensity and duration to excite all =edes to tha required magnitude, such that the tesiing response spectra will envelope the corresponding response spectra of the indiviu-al = odes.
49 Ofidt
. 6.
The input cocion should be applied to one vertical and one principal (or two o: ho;;onal) horizontal axes si=ultaneously unless it can be deconstrated that tha equipment response sensitive to the vibratory along the vertical direction is not The
=otion along the hori:catal direction, and vice versa.
ti=e phasing of the inputs in the vertical and horizontal direc-tions must be such that a purely rectilinear resultant input is avoided.
The seceptable alternative is to have vertical and horizontal inputs in-phase, and then repeated with inputs 180 degrees out-of-phase.
In addition, :he test must be repeated with the equip =ent rotated 90 degrees horizontally.
7.
The fixture design should =eet the following requirements:
(a) Simulate the actual service counting (b) Cause no dynamic coupling to the test item.
8.
The in-situ application of vibratory devices to superimpose the seismic vibratory loadings on the co= plex active device for operability testing is acceptable when application is justifiable.
9.
The test program cay be based upon selectively testing a repre-load sentative nu=ber of =echanical components according to type, level, size, etc. on a prototype basis.
II.
Seismic Desien Adecumev of Suecorts Analyses or tests should be perfor:cd for all supports of 1.
electrical and mechanical equipment and instrumentation to ensure their structural capability to withstand seismic excitation.
2.
The analytical results cust include the following:
e.heuld (a) The required input motions to the mounted equipment be obtained and characterized in the =anner as statad in Section I.2.
(b) The combined stresses of the support structures should be vithin the limits of ASME Section III, Subsection NT -
" Component Support Structures" (draf t version) or other comparable stress limits.
3.
Supports should be tested with equipment installed.
If the equipmeat is inoperative during tha supper: test, the resp r 2 the equipment ecunting locations should be coni ored and at characterized in the nar.ner as stated in Section 1.2.
In such a case, equipment ah; >: J ce testad separately and the actual input to the ecuir ent -0.ould be core conservative in acplituce
.., a 5,... n -.. :, ~..
a. n c y, w.c. a, n e.n u 4
The r ec.. im- :- t -
sc.:-fen;
,2, 7,u,
,, 7,5 3.
applicable when tests are conducted on the equipcent suppcr:s.
49 '()[5U
Attach =ent C 2 & 3 PUMP OPDA3ILITY ASSURANCE PROGRAM class 1.
The operability of ASME Class 1, 2 & 3 " active" pumps under plant conditions uhen their safety function is relied upon to effect either a plant shutdown, or to mitigate the consequenecs of an accident may be demonstrated by either of the following programs:
An individual pump, selectet' as a prototype pump, may be 1.
tested in the manuf acturer's shop, provided the test conditions imposed are equivalent to the combined plant conditions which the pump is expected to withstand at the time when the " active" function is required.
An individual pump, selected as a prototype pump, may be 2.
tested in-situ following installation in the system during the performance of the system preoperational functional tests, provided the test conditions duplicate those conditions when the " active" function is required.
In either of the test programs, 1 and 2 above, vibratory excitation of the pump to simulate seismic loading may be demonstrated (a) by under conditions sufficiently severe to provide a separate test adequate margins for assurance of operability under combined plant loading conditions or (b) by seismic dynamic analysis of critical pump components.
3.
An individual pump, sele ed as prototype pump, may be tested partially (a) in the.
.c#acturer's shop under those test conditions as limtied by the test facility, (e.g.,
pressure temperature loadings) (b) in a testing laboratory for simulated seismic excitation loadings, and (c) in the plant after pump installation for confirmatica of operability under flow conditions during system preoperational hot The distribution of test partmeters handled functional tests.
by each testing group =ay have variations depending upon the pump testing requirements.
Such a test program should be supplemented by analyses as required under test program 2. above.
An individual pump, selected as a prototype may be tested 4
completely in-situ following installation in the system during the performances of the system preoperational functional tests.
(18' $ [i6
Attachment D CLASS 1.
2 & 3 VALVE OPERA 3ILITY ASSURANCE PROGRAM The operability of ASME Class 1, 2 & 3 " active" valves under plant conditions when their respective safety function is relied upon to effect either a plant shutdown or to mitigate the consequences cf an accident may be de=onstrated by any one of the following acceptable programs:
1.
An individual valve, selected as a prototype valve, may be tested in the manufacturer's shop, provided the test conditions imposed during the demonstration of valve opening and/or closing are equivalent to the combined plant conditions which the valve is expected to withstand at the time when the " active" function is required (such a test psogram may be practical for small size valves).
2.
An individual valve, selected as a prototype valve, may be tested in manufacturer's shop under test conditions which simulate separntely each of the plant loadings which the valve is expected to withstand in co=bination durir.g valve opening and/or closing.
Such a test program should be supplemented by analyses which demon-strate that the individual test loadings are sufficiently higher than the plant loadings, to provide adequate margins for assurance of operability under combined loading conditions.
In addition, the analyses should demonstrate that the strains in critical component parts of the valve under individual test loadings are greater, by a substantial margin than those which the valve may experience under the combined plant loading conditions.
3.
An individual valve, selected as a prototype valve, may be tasted partially (a) in the manufacturer's shop under those test conditiens as limited by the test facility, (e.g., pressure temperature 12:: dings)
(b) in a testing laboratory for simulated seicnic exc itation leadings, and (c) in the plant af ter valve installation for confirmation of operability under flow conditions during system preoperational hot functional tests. The distribution of test parameters handled by each testing group may have variations depending upon the valve testing requirements.
Such a test program should be supplemented by analyses as required under test program 2. above.
4.
An individual valve, selected ar a prototype may be tested completely in-situ following installation.a the system during the performances of the system preoperational f,metional tests.
49~057
o s.
Vibratory excitation of the valve under such test conditions to represent seismic loadings may be induced by mounting devices which will vibrate principally the valve operator and the controls mounted on the valve system.
5.
Valves that can be demonstrated to be ectivalent to a prototype valve, which has successfully met the te-;t requirementa of a valve operability assurance program, may be exempted from testing provided:
(a) the test results of the prototype valve are documented and available and (b) the loading conditions for the exempted valve are equivalent to those imposed during testing of the prototype valve.
The prototype valve may be selected from a group of similar valves which will be used in the plant.
A prototype valve used in one nuclear power plant qualifies as a prototype valve for another plant provided the system operating conditiens of both plants, and the valve loading conditions at the time when the " active" function is required are equivalent.
I e
Le
5/24/73 o
Attachment E ACCEPTABILITY OF COMPUTER PROGRAMS ANALYSIS OF MECHANICAL COMPONENTS AND ECUIPMENT_
A list of computer programs that will be used in dynamic and static l.
analyses to determine mechanical loads and deformations of Seismic and the analysis to Category I structures, couponents and equipment determine stresses should be provided including a brief description of each program and the extent of its application.
The design contrci =easurec as required by Appendix B - 10 CFR Part 2.
50 that will be e= ployed to de=onstrate the applicability and valid-ity of the above co=puter programs should be described by any of the following criteria or procedures (or other equivalent procedures).
(a) The computer program is a recognized program in the public domain, and has had sufficient history of use to justify its applicability and validity without further demonstration.
The dated program version that will be used, the software or operating syteem, and the computer hardware configuration must be specified to be accepted by virtue of its history of use.
(b) The co=puter program's solutions to a series of test problems, with accepted results, have been demonstrated to be substantially identical to those obtained by a similar, independently written program in the public domain. The test proble=s should be demonstrated to be similar to or with the range of applicability for the problems analyzed by the computer program to justify acceptance of the program.
(c) The program's solutions to a series of test problems are substan-tially identical to those obtained by hand calculations or frem accepted experimental test or analytical resul:s published in technical literature. The test problems should be demonstrated to be similar to the problems analy:ed to justify acceptance of the program.
3.
Provide a su= mary comparison of the results obtained frem each ccmputer program with either the results derived from a similar program in the public domain, on a previously approved computer program or results from the test problems.
Include typical static and/cr dynamic response loading, stress, etc. comparisons pref erably in graphical form.
48 059
12-1 12.0 MATERIALS ENGINEERING 12.1 Provide a list of any cold worked austenitic stainless steels, (4.2.2) precipitation hardening stainless steels or hardenable martensitic stainless steels having yield strengths greater than 90,000 psi used for components of the reactor vessel internals.
If any such steels are used, provide assurance that they will be compatible with the reactor coolant.
12.2 In Section 5.2.3.3 of the FSAR you state that the reactor (5.2.3.3) coolant insulation is all =etal reflective insulation.
Describe the requirements for non=etallic insulation for stainless steel components i=portant to safety, particularly with respect to chloride, fluoride, and silicate content.
Indicate the degree of confor=ance with Regulatory Guide 1.36, "Non=etallic Thermal Insulation for Austenitic Stainless Steel," dated February 23, 1973.
Include detailed infor=ation on the nature of the control (s) of installation.
Of specific interest in this respect are the types of cements that are field =ixed for application.
12.3 In Section 5.2.4.3 of the FSAR you state that Appendix G to (5.2.4.3)
Section III was used as a guide in establishing operating pressure-te=perature li=itations of the reactor coolant systen and that deviations fro = Appendix G are described in report BAW-10046.
Provide us a copy of the report, 3AW-10046, or the information contained in the report, for our review and evaluation.
In addition, the te=perature-pressure li=itations for core operation =ust conform to the requirements of Appendix G, 10 CFR 50.
12.4 Provide the extent of confor=ance of the reactor vessel surveil-(5.2.4.4) lance progra= to ASTM E-135-73 and Appendix H, 10 CFR 50 and provide juscification for any deviations.
12.3 the discussion of paragraph C-1c of Regulatory Guide 1.14 in (5.2.6)
Supplement 1 of A=end ent 14 is unsatisfactory.
The mini =un fracture toughness of the flywheel material at normal operat-ing temperature should be equivalent to a dynanic stress intensity factor (K dynamic) of at least 100 ksi vin. or to 7
a static stress intensity factor (K static) of at least 7
150 ksi /in.
12.6 The inservice inspection'progra= for pump flywheels presented (5.2.5) in Supplement 1 is unsatisfactory.
Provide sufficient information about your inservice inspection progran for the coolant pump flywheels to assure the degree of flywheel integrity comparable to the progra= reco== ended in Regulatory Guide 1.14, " Reactor Coolant Pu=p Flywheel Integrity,"
October 27, 1971.
49'CSO
12-2 12.7 You state in Section 5.2.7.3 of the FSAR that the max 1=um total (5.2.7.3) leakage rate is 30 gpm. The total allowable leakage rate should be changed to 10 gpm to conform to the value stated in Technical Specification 3.1.6.1.
12.8 Your response to Question 3-1 in Amendment No.19 is incomplete.
(5.2.8)
The following items should be addressed regarding your inservice inspection program for Code Class 2 and 3 co=ponents.
(1) The criteria for Class 2 components requiring inservice inspection should be in accordance with ISC-261 of 1972 Winter Addenda to Section XI.
This article requires inspection of portions of the main stems and feedwater systems including Class 2 parts of the steam generators.
(2) The pressure required for testing closed Class 3 systems should be 110 percent of design pressure as stated in position C.3.a(1) of Regulatory Guide 1.51.
(3)
Provide a statement that inservice inspection of Code Class 2 and 3 components will be conducted in accordance with the 1972 Winter Addenda to Section XI and Regulatory Guide 1.51 to the extent practicable and consistent with the existing design.
12.9 Provide sufficient infor=rtion about your plans for periodic (5.5.2.5) inservice inspection of steam generator tubes to indicate that your program will provide assurance of steam generator tube integrity comparable to what would be achieved by follow-ing the reco=mendations of Reptlatory Guide 1.83, " Inservice Inspection of Pressurized Water deactor Steam Generator Tubes,"
issued in June 1974.
12.10 Temperature-pressure limitations for heatup, cooldown, core (16.3.1.2) operation and inservice hydrotests should be determined in accordance with the requirements of Appendix G of the Su==er 1972 Addenda of Section III of the ASME Code and Appendix G of 10 CFR 50.
The following information is required before we can evaluate your Technical Specifications in this area.
(1)
Provide the basis for the heatup and cooldown limit curves, Figures 3.1.2-1 and 2.
This should include specific fracture toughness data used to determine the limiting initial RT and the a=ount of radiation induced te=perature shikkI The chemical analysis of the vessel beltline materials, particularly those elements such as copper, phosphorus, and vanadium which affect fracture toughness properties, should be provided.
'19/ (26I.
~
12-3 (2)
Specification 3.1.2.1, "Hydrotests," refers to Specification 2.2 for pressure limitations when there are fuel assemblies in the vessel.
Specifications 2.2 and 3.1.2.1 should also provide temperature limitations calculated in accordance with the requirements of paragraph G-2410 of Appendix G to ASME Boiler and Pressure Vessel Code,Section III.
) j,
13-1 13 STRUCTURAL ENGINEERING 13.1 Justify your statement that the Petry formula is less con-(3.5.4.2) servative than the Army Corps of Engineers and the Ballistic Research Laboratory formulas and that Ballistic Research Laboratory formula is the most conservative.
13.2 The seismic instrumentation program given in Section 3.7.4 is (3.7.4) not in accordance with Regulatory Guide 1.12 (Rev.1).
As a minimum requirement, one multi-element seismoscope should be installed at the basement of the containment for a rapid determination of the ground input response spectra during an earthquake.
Provision for time history accelerographs alore is not adequate.
State your intent to comply with the re-qui remen t.
13.3 Your use of concrete interlock approach for tangential shear (3.8.1.4) is not suitable for prestressed concrete.
Demonstrate that the design for tangential shear is in accordance with ACI-318 or ACI/ASME (Committee-359) codes, which are applicable.
13.4 Justify the use of a reduction factor of 0.95 for prestressing (3.8.1.5) steel instead of 0.9, as recommended in the ACI/ASME (Committee-359) Code.
13.5 As rcquested in our preliminary review, state that the design (3.8.3 &
criteria and design methods are in accordance with Document (B) 3.8.4) prepared by the Structural Engineering Sranch, Directcrate of Licensing " Structural Design Criteria for Evaluating the Effects of High-Energy Pipe Breaks on Category I Structures Outside the Containment".
This document, prepared originally for structures outside of the containment is also applicable to interior structures.
13.6 Your response to our question on degree of conformance with (3.8)
Regulatory Safety Guide 1.18 is not cc plete.
Provide a com-prehensive discussion of degree of conformance.
13.7 Re'ference in Section 3.8 your discussion of conformance with (3.8)
Regulatory Safety Guide 1.13.
21-1 21.0 REACTOR SYSTEMS BRANCH 21.1 In Table 3.2-1, the borated water storage tank is classified (3.2.2)
Quality Group B.
However, the suction line to the decay heat removal pumps is classified Quality Group C from the B'iST to valves DH-V5A/B.
This line should be Quality Group B in its entirety, and the Three Mile Island-Unit 2 (mI-2) applied code should be ANSI B31.7, Class II.
21.2 In Table 3.2-1, several lines connected to the borated water (3.2.2) storage tank and the sodium hydroxide storage tank are classified N-3.
Since these tanks are classified N-2, all lines connected to the tanks below the normal fluid level should be classified N-2 out to and including the first normally closed isolation valve or a valve capable of automatic closure.
21.3 In Table 3.2-1, the line from the sodium hydroxide storage tank (3.2.2) to valves DH-V134A/B is classified as Quality Group C.
This line should be Quality Group B and the TMI-2 applied code should be ANSI B31.7, Class II.
21.4 In Table 3.2-1, the diesel generator tuel oil system and all (3.2.2) of the diesel generator auxiliary system should be classified Quality Group C.
21.5 In Table 3.2-2, you state that auxiliary piping ersential for a (3.2.2) safe shutdown is classified as Quality Group D.
ihis is inconsistent with the classification you have assigned essential systems in Table 3.2-1 and on the system flow diagrams.
Quality Group D is not applicable to systems essential to a safe shutdown.
Revise Table 3.2-2 accordingly.
21.6 In Table 3.2-1, the reactor buildinc penetrations of the reactor (3.2.2) building emergency cooling water system and the intermediate closed cooling water system are incorrectly classifiec as Quality Group C.
This portien of these systems should be classified Quality Group B and the TMI-2 applied code should be ta accordance with note 2 of Table 3.2-1.
21.7 In Table 3.2-1, the spent fuel pool pumps, the borated water (3.2.2) recirculation pumps and the spent fuel coolers are incorrectly classified as Quality Group D.
This equipment should be classified Quality Group C.
AO'C(it]
21-2 21.8 In Table 3.2-1, the condensate storage tanks are classified (3.2.2)
Quality Grouc C and the TMI-2 applied code is ASME Section III, Class C.
Since the scope of the 1968 edition of ASME Section III does not include atmospheric tanks, describe how you intend to apply this code to the condensate storage tanks.
Acceptable standards for the construction of these tanks are API-650, AWWA-0100 or ANSI S96.1 with the folicwing non-destructive test requirements for tank welds:
Sidewalls Liquid Penetrant or Magnetic Particle Roof Visual Roof to Sidewall Visual Bottcm Vacuum Box Bottom to Sidewall Vacuum Box Nozzles Liquid Penetrant or Magnetic Particle 21.9 In Table 3.2-1, ycu indicate that the Quality Group C piping (3.2.2) will be constructed in accordance with ANSI B31.1.0.
In addition to the requirements of ANSI S31.1.0 and the design features indic'ated in the notes to Table 3.2-1, all welds in Quality Group C piping larger than 4 inches shall be exanined by either the magnetic particle or liquid penetrant method in accordance with ANSI 931.7, Class III.
21.10 In Table 3.2-1, you incorrectly state that the dccay heat closed (3.2.2) cooling water system coolers, pumps, surge tanks, piping, and valves will be constructed in accordance with ANSI B31.1.0.
This equipment should be constructed in accordance with ASME Secticn VIII, AtlSI B31.7, or the Draft ASME Code for Pumps and Valves, as applicable.
21.11 In Table 3.2-1, you indicate that the borated water storage tanks (3.2.2) will be constructed in accordance with AWWA-D100-67.
In additien, non-destructive testing should be performed on the BWST welds in accordance with the.folicwing:
Sidewalls 100" Radiograpnic Roof Visual Roof to Sidewall Visual Bottom Vacuum Box Bottom to Sidewall Vacuum Sox and Magnetic Particle or Liquid Penetrant Nozzles Magnetic Particle or Liquid Penetrant 21.12 In Table 3.2-1, for equipment which is tc be constructed in (3.2.2) accordance with ANSt 531.7 or the Draft ASME Coce for Pumes and Valves, indicate the code class where you have not already done sc.
49 CGS
21-3 21.1 3 In Table 3.2-1, all equipment constructed in accordance with (3.2.2) the Draft ASME Code for Pumps and Valves, Class I or Class II, shall be examined as follows:
All pressure retaining cast parts shall be radiographed (or ultrasonically testec to equivalent standards).
Where size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted.
Examination procedures and acceptance standards shall be at least equivalent to those specified in the applicacie class in the code.
21.14 In Table 3.2-1, valves MU-V8 and V9 of the Makeup and Purification (3.2.2) system are incorrectly classified as Draft ASME Code for Pumps and Valves, Class III.
These valves should be Class II of the Draft ASME Code.
21.1 5 In Table 3.2-1, you indicate that the piping frca the sodium (3.2.2) thiosulfate tanks to valves BS-V4A/B is classified Quality Group C.
This portion of the system should be Quality Group 8 and the TMI-2 applied code should be in accordance with note 2 of Table 3.2-l'.
21.16 In Table 3.2-1, you indicate that the piping frcm valves (3.2.2)
BS-V105A/B to the reactor building spra, header nozzles is classified Quality Grouc C.
Th u portion of the system should be Quality Grcup B and the TMI-2 applied code should be in accordance with note 2 of Table 3.2-1.
21.1 7 Figure 6.3-1 does not provide adequate information to relate the data in Table 3.2-1.
Revise Figure 6.3-1 to indicate (3.2.2) piping classifications and provide valve numbers which relate to the information in Table 3.2-1.
49 CGG
21-4 21.18 Provide a schedule for the s'ubmittal of a review of (4.3.1.3) the shutdown system design, plans for any proposed plant changes required to make the consequences of an anticipated transient without scram acceptable and the results o# supporting analysis as recuirec Dy paragraph II-B of appendix A to '! ASH-1270.
21.19 The thermal-hydraulic design basis requires that the minimum DNSR under operating conditions ar.d transients does not fall below 1.3.
At a given value of maximum linear heat generation rate, the radial power ratio affects DNBR more than the axial cower ratio.
There-fore, merely specifying the maximum linear heat gen-eration rate and the product of axial and radial peaking values does not guarantee that the minimum
'! hat assurance is DNBR limit will not be exceeded.
there that the radial ceaking factor will not exceed the values listed in Section 4.4 and the values used for the safety evaluation of the plant in Section 15.
21.20 Table 4.4-1 lists Rancho Seco as an essentially identical (4.4.2.1)
NSSS. Rancho Seco was granted a license limiting its power to 2568 MWt subject to later review of startup reports and initial operating experiences.
- Also, satisfactory operating experience of the prototyce Oconee Unit 1 is required.
Summarize this experience an:
show how it justifies the design thermal rating of Three Mile Island Unit 2.
21.21 Identify reactor internal elements critical to the safe (4.4.2.7) operation and control of the reactor.
Tabulate for these elements the limiting design loads along with the aost~ severe up, down and horizontal loads predicted i
during transient analysis.
Identify the events creating 6
the most severe loads.
21.22 Provide the results of the calculation of maximum fuel (4.4.3.5) clad strain for operational transients to end of life.
W'CG?
21-3 21.23 In addition to load changes at constant pump combinations, (4.4.3.5) the reactor ccolant system must be demonstrated to be free of undamped oscillations or other hydraulic instabilities for all conditions of steady state operation, for all oper.aticnal transients, for all load follcuing maneuvers and for partia! loop oceration.
Provide analysis, operational experience anc experimental results proViding this assurance.
21.24 Discuss experience in observing crud or scale build-up (4.4.4)
(or absence of) during the life of a plant.
Discuss hc.i crud build-up is considered in heat transfer analysis and component design.
21.25 Describe and discuss instrumentc tion for vibration and (4.4.5) loose parts monitoring in the i 3 actor coolant system.
21.26 Identify the margins in net positive suction head for the (5.2.1.1) operating main ccolant pumps when operating with one or two pumps shut down.
21.27 What is the allowable back prassure for the safety valves?
(5.2.2.2)
What is the basis for this limit?
Provide the methed used, including c::perimentcl verification, in u=cermininc that the back pressure limit is not exceeded.
If this limit is e::ceeded,,; hat would be the effect on safety valve relief capacity?
21.28 Show that all the assumptions and initial conditions used (5.2.2.3) in the GAU-10043 analysis are applicable to the Three Mile Island # 2 plant.
21.29 BAW-10043 does not provide the basic plant carareter such (5.2.2.3) as plant gecmetry and power level.
Further, BA'!-10043 does not provide the set points for bath tne primary and the secondary safety valves.
Provide thi information.
21.30 BAW-10043 does not address the severity of a com:lete loss I
(5.2.2.3) of feedwater on the overpre:sure protection capacity provided.
Provide the analysis to substantiate the adequacy of safety valve discharge capaciti'es for a complete loss of feedwater transient.
21-6 21.31 In the EAW-10043 analysis, pressurizer spray is assumed (5.2.2.3) not to operate although the high pressurizer pressure signal was used to scram the reactor.
Provide an analysis where the spray is assumed to operate and tnerefore scrar.; is delayed.
21.32 Show that the pressurizer does not go solid for any over-(5.2.2.3) pressure transients.
Otherwise, provide the bases for the Water discharge rates through the safety valves.
21.33 The capability of the RHR system to perform its shutdown (5.5.7) cooling function assuming the most restrictive single active failure in the RHR system has not been demonstrated.
The RHR system is not single failure proof and, therefore, violates the intent of AEC General Design Criterion 34.
An example of a single failure which could render the RHR systen inoperable is a failure-to-open of one of the isolation valves in the RHR line leading frcm its associated recirculation loop.
Such a single failure could place the reactor in the position of not being able to achieve a cold shutdo.in ccr.dition within a reasonabic period of time.
It may be possible that some "bcotstra;;"
type of operation outside of the RHR system coula be effective in achieving a degree of shutdown capability (such as with the ECCS), however, it is the intent of GDC 34 that the system normally utilized to place the plant in a cold shutdown condition (the RHR systera) be single failure proof.
Also, since the RHR system is a lcw pressure system for which overpressure protection is recuired, any design modifications should not reduce the level of protection against overpressurizatior The RHR system should be modified so as to be immune to single active failures before final Regulatory staff approval.
21.34 The AEC " Interim Acceptance Criteria" has been superseded
( 6. 3.1.1 )
as stated in the Federal Register, Vol. 39, l!o. 3-Friday, January a,1974.
It is required.for Three Mile Island 2 that analysis and ' evaluation of ECCS cociing performance follcwing postulated loss-of-coolan; accidents shall be performed in accordance with the recuirements of 10CR 50.46 using an Evaluation 6:odel in conformance witn Accendix K.
A commitment is required from the applicant identifying when the Safety Analysis Report will be revised and re-submitted so that the review may proceed. This comment applies to ChaptersG and 15.
7-
.,. y
21-7 21.35 Refer to Figure 6.3-1, the ECCS'P&ID.
Starting at either ECC (6.3.2) vessel injection nozzle, trace back along the piping toward tne low pressure system, through two check valves, the reactor building boundary, and a normally open motor coerated valve.
At this last valve there is a transition from hign pressure to low pressure piping.
Our concern is that no means are provided to detect leakage from the reactor coolant system back through the first (relative to the RCS) check valve, or from the core flooding tank (CFT) back throuch the second check valve.
In the latter case, a decreasing CFT level may be sufficient indicati]n of leakage for the aparator.
However, in the former cas e, undetected leakage from the RSC could pressurize the line between the two check valves for an undetermined period of time.
Subsequent failure of the second check valve or the CFT check valve would result in a LOCA (outside containment in the first case, in' side in the second) with diminished ECCS capability.
Thus, the failure of one check valve could lead to a LOCA and a degraded ECCS.
A second concern is that no pressure relief devices are shown on the Figure.
A change in design nr monitoring should be made so that full credit can be taken for both check valves as ::rotection against back leakage from the RCS.
Such a change could take the form of a pressure indicator between the check valves of high pressure piping throughout, additional valvias,, use addition of safety valves, different valve administrative alignment, or a combination of these.
21.36 The ECCS design shown in Figure 6.3.1 does not meet the (6.3.2) requirements of GDC 35.
A failure of one injection line resulting in a LOCA coupled with a single failure in the Other injection train would incapacitate the ECCS.
The proposed design for such S&W plants as " orth Anna 3/4, Bellefonte, Greenwood, and WPPSS are examples of acceptable designs with respect to low pressure injection.
Provide a description of the re-designed ECC system which fully complies with GDC 35.
Include a discussion of the design basis and an evaluation of the operation of the system.
21.37 Provide an enlarged (legible) Figure 10.1-2 (10.1) 4T C70
21-8 21.38 Provide a curve showing the effect of reactivity insertion (15.1. 2 )
rate on minimum DM3R.
Forthis insertion _ rate which gives the minimum D' BR provide a curve of DN6R vs. time.
Assume that the transient starts at 102", rated power, 2132 psia and 559 F inlet temperature.-
21.39 Uhat values of radial peaking factors and enthalpy rise (15.1. 2 )
factors (F~H) were used in the analyses presented in Chapter 15?
21.40 For the loss of Coolant Flow Analysis, crovide curves of (15.1. 5 )
DNSR vs time and hot spot heat flux vs. time for the four pump shutdcwn.
21.41.
What is the DNCR for steady-state operation at the conditions (15.1. 5 )
assumed for the start of the flow coast-down transient (102", rated power, 2135 psia and 559 F inlet temperature)?
21.4 2 Provide evidence that the uncertainty in inlet temperature (15.1.5) is only 2 F.
Reference:
Table 15.1.5-1.
21.43 Determine if one motor driven emercency feeri pump (U0 GFN (15.1.8) is sufficient to bring the plant to a safe shutdown ccnditur.
The event postulated is as.follcus:
A rupture occurs in -he high energy steam supply line to the emergency feedwater pump turbine.
This is coupled with the active failure of one motor driven emergency feed pump.
Provide analysis and discussion of this postulated event.
21.44 Provide an analysis for a feedwater line rupture.
In the (15.1.8) analysis justify the method used to calculate break fic.,
the sizes and locations of breaks.
Show that the. single failures considered in the analysis are the most limiting ones.
Further, if the pressurizer goes solid as a result of this accident, provide bases for water discharge rates through safety valves.
21.45 The set points for the various overpressure protection (16.2.2) devices should be stated under " Specifications."
4W C71
22-1 22.0 Electrical, Instrumentation & Control Systems 22.1 Supplement the informatier. contained in FSAR Section 3.10, (3.10)
Seismic Design of Category I Instrumentation and Electrical Equipment, as folicws:
(1) Provide a summary listing (tabulation) correlating all safety related electrical equipment, equipment locations, seismic design bases at each location, seismic qualification method used (test and/or analysis) and seismic test and/or analysis results.
This should include the 4.16 kV switchgear, 480 V switchgear, unit substations, and motor control centers; 120 v a-c system components; 125 v d-c system components including batteries, battery racks, battery chargers, distribution centers, and panelboards; static inverters; process control equipment; protection and safeguards actuation racks; nuclear instrumentation; electrical penetration assemblies; motor operated valves; diesel generators etc.
(2)
Provide a more detailed" description of the seismic qualification method (test and/or analysis) used for each Class I component and module.
This description may be incorporated in the summary listing of (a) above.
(3)
Confirm that the seismic qualification testing demonstrated the capability to change state or operate during a SSE for all components and modules which are required to so operate in perfornance or~ their design safety functions.
Provide the bases for the methods of simulating the net effect of the desi n basis J
seismic event which were used in the qua ification tests.
(4) Verify that the auxiliary equipment (local control panel, lube oil system, etc.) which is required for the operation of the emergency diesel generators has been seismically qualified.
Describe the testing and/or analysis performed to seismically qualify this equip-ment, and state the results of this testing and/or analysis.
(5) With reference to the seismic qualification methods described in BAW-10003 Revision 3, confirm the items stated below:
(a) The responses of cabinet assemblies at the various instrument or device mounting locations due to actual earthquake distur-bances are determined either by testing or analysis.
48~C72
22-2 (b) The maximum response detemined in (a) above at the instrument or device mounting locations are less than 1.
9 (c) The instrument or device responses resulting from a continuous sine wave test input are shown to be equivalent to the responses due to actual-earthquake disturbances in magnitude and frequency content.
(6) Table 5-2 of BAW-10003 provides a list of instruments to be used in the Nuclear Instrumentation, Reactor Protection and Safety Features Actuation systems.
Identify the supplier of the cable assemblies and cable connectors to be used with these instruments and p.rovide a detailed description of the seismic qualification (test and/or analysis) for these assemolies and connectors.
22.2 Supplement and revise the information in Section 3.11 (3.11) regarding environmental quafification of safety related components located in the primary containment to include the following:
(1)
Provide a concise statement of the limiting DBA environmental conditions in the containment.
This should include temperature, pressure, humidity, radiation, and chemical environment.
(2)
State the length of time from occurrence of the CBA that each component is required to operate in order to perfom its design safety function.
(3)
Describe the environmental qualification testing performed for each component and identify the applicable test documentation.
Provide this test documentation if it has not been previously submitted.
Your response snould (a) state whether the tests were perfomed on prototype equipment, (b) contain sufficient detail to permit a direct comparison between the test conditions and the limiting DBA conditions (superimposed on normal aging) for all parameters, and (c) discuss the adequacy of the environ-mental qualification for each component.
(4)
If environmental qualification is based (in whole or in part) on the analyses or on use of data from tests performed on other than prototype equi. ment, describe and justify each instance of the use of these methods and justify the applicable docume.ation.
4F C73
22-3 22.3 The system design for the combustible gas control systems (6.2.5.2)
(hydrogen recombiner system and as a backup, the hydrogen purge system) as described in Section 6.2.5.2 does not contain sufficient information to ascertain the adequacy of the instrumentation, control and electrical equipment for these systems.
Therefore, provide this description in sufficient detail to permit an evaluation.
This description should include the following:
(1)
Explicitly, identify the source of power for the flydrogem recombiner unit or,assumina this unit is inoperable,the source of power for the hydrogen control exhaust fan.
(2)
Identification of any instrumentation, control or electrical equipment which may be ccmmon to both the hydrogen recombiner system and the hydrogen purge system.
22.4 The design of the motor-operated isolation valves for the (6.3.2.15 )
core flooding tanks (as described in Section 6.3.2.15) does not provide s1fficient assurance that these valves wi11 open when required.
An acceptable variation of your design would include the following features:
(1) Valve position visual indication (open or closed) for each valve which is not dependent on pcwer being available to the valve controller.
(2)
Redundant visual and audible alams for each valve when the valve is not fully open and reactor coolant pressure is above a preset value.
These alarms shall be actuated by redundant and independent valve position sensing circuiLy including at least one position sensor sensing actual valve position, and by redundant and independent pressure signals.
(3) A Technical Specification requirement that the reactor shall r.ot be made critical or shall be shut down unless the motor operated isolation valve in the discharge line of each core flooding tank is open and the breaker supplying power to the valve operator is locked open and tagged.
Please indicate your plans and schedule to modify the design of the isolation valving to include the preferred features stated above or to confom to other criteria that provide equivalent assurance that these valves will be open when required.
49 C74
22-4 22.5 (RSP)
Based on the infomation presented in Section 6.3.2.17, we (6.3.2.17) believe that:
(1) The proposed design of the circuits used to change over to the cross-over uoda (using LPI pumps as boosters for the HPI Pumps) and the recirculation mode of operation following a LOCA does not conform to the requirements of IEEE Std 279-1971, and (2)
The complexity of the proposed change-over may not provide adequate assurance that the operator will correctly perform the required actions.
The Staff's position is that the instrumentation and controls provided to accomplish the change-over to the recirculation mode and the cross-over mode should be designed to meet the requirements of IEEE Std 279-1971 including the require-ments for automatic and manual initiation of protactive actions at the systems level.
In addition, the procedures should be of such simplicity as to provide a high degree of assurance that the operator will perfom correctly all actions that are necessary to protect the health and safety of the public.
Therefore, either modify your design to show conformance with the Staff's position, or justify your design, as opposed to a design which provides automatic and system level manual initiation of switch over as required by the literal interpretation of IEEE Std 279-1971.
Justification of the present design should include the following:
(a)
Define the time available for the operator to complete the necessary monitoring and switching functions for the cross-over mode (i.e., using LPI pumps as boosters for HPI pumps) and the recirculation made prior to onset of conditions (assume and define worst case) that are unacceptable from the standpoint of plant safe +.y.
(b)
Provide a description of the control panel arrangement of the indicators and switches which the operator must monitor and operate to effect switchover.
(c)
Describe the pernissive interlocks that are provided (for equipment protection or other reasons)between the various ECCS components that are operated during switchover.
(d)
Identify the system conditions that require the use of the cross-over mode prior to or coincident with the initiation of the recirculation mode of operation.
Y??{i
22-5 22.6 Supplement the information contained in Section 7.0 (7.0 and 15.0) with regard to the Reactor Protection System (RPS),
Control Rod Drive Control System (CRDCS) or where appropriate to the Safety Feature Actuation System (SFAS) as follows:
(1) Table 7.1-2 indicates that the RPS meets the require-ments of IEEE Std 279-1968.
Identify those features of the RPS design which do not meet the requirements of the 1971 version of this. standard.
Your response should specifically address confomance to Sections 4.7, 4.17 and 4.22 of IEEE Std 279-1971.
(2)
Section 7.1.1.2 of the FSAR indicates that no supporting systems are necessary for the safe operation of the RPS.
Discuss the failure of various Heating, Ventilating and Air Conditioning Systems located throughout the plant and identify those (if any) who's failure could result in the RPS not performing its design function.
The portions of the RPS considered should include (as a minimun) the process-to-sensor coupling, sensors, instrument channels, decision logic, and actuation devices.
(3) The infomation presented in Sections 7.1.2.8 and 7.2.2.4 is not sufficient to demonstrate conformance of your RPS design with all of the regulatory positions of Regulatory Guide 1.22 (Safety Guide 22), particularly Positions 1(a),1(b) and 3(a).
This information should be provided in the detail required for this evaluation.
Your response should address conformance of your design with all the positions of Regulation Guide i.22 and particularly including a description of the interlocks which preclude the bypassing of more than one RPS channel.
(4) Sections 7.1.2.7 and 7.2.2.2 state that the RPS does not comply with IEEE Std 338-1971, Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems, because this standard was issued subsequent to equipment procurement.
This standard is primarily concerned with periodic testing, not with system design.
Define the degree of confomance of the test and surveillance program for the RPS and SFAS embodied in the Technical Specifications (Section 16.0) with the provisions of IEEE Std 338-1971.
Identify any system design features that preclude testing in confomance with this standard.
49 C7c
22-6 (5)
Provide a more detailed description of the RPS manual trip switches including their installation.
Your response should address those features of the design which implement the separation and independence requirements for redundant safety circuits.
(6) Table 15.1.1-1 of the FSAR indicates that the maximum rod withdrawal speed (used for the analysis of the startup accident) is 30 in/ min.
However, it appears from Figure 7.7-1 of the FSAR that if the programmer motors were to overspeed, this maximum withdrawal speed could be exceeded.
Provide a discussion of the design features, other than the use of synchronous programmer motors, that would limit rod withdrawal speed to 30 in/ min.
(7) Describe the methods for periodically t' sting the RPS response time for the trip parameters.
Of concern is the time history changes for these responses times.
Include a discussion of the response times in relation to the safety limits and state the worst case margin in terms of time.
(8)
Describe the means by which a control room operator is apprised of a reduction in engineered safety features redundancy on an overall systems basis.
22.7 With regards to IEEE Std 336-1971, Installation, Inspection (7.1.2.6) and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations, it is not clear from the discussion presented in Section 7.1.2.6 whether the requirements of this standard have been or will be met for the installation, inspection, and testing procedures for instrumentation, sensing lines, electrical,and instrumentation penetrations, cabling and raceways, switchgear and panels.
Provide an additional discussion defining the degree of conformance to the require-ments of this standard for the above items.
22.8 Section 7.3.1.1.4 of the FSAR indicates that certain valves (7.3.1.1) in the containment isolation system are provided with air storage resevoirs.
Identify all valves in the isolation system which are of this type and prcvide a typical control diagram for these valves. Also, describe the degree of conformance of the instrumentation and controls for these valves to the safety criteria.
This should include a discussion of their testability, redundancy and power requirements.
dP ~ CW
22-7 22.9 The FSAR (Section 7.3.2.lp) indicates that the SFAS and (7. 3. 2.1 )
ESF systems are designed so that once a protective action is initiated, it continues until it is halted by deliberate operator action.
Section 4.16 of IEEE Std 279-1971 states that "The protection system shall be so dcsigned that,once initiated, a protective action at the system level shall go to completion.
Return to operation shall require subsequent deliberate operator action." Confirm that in your design a protective action at the system level will go to comoietion once initiated. Identify and justify any exceptions.
22.10 Incidents have occurred at a nuclear power plant that (None) indicate a deficiency in the control cir uit design.
These incidents involved the inadvertent disabling of a component by racking out the circuit breaker for a different component.
As a result of these occurrences, we request that you perform a review of the control circuits of all safety related equipment at the plant, so as to assure that disabling of one component does not, thcougn incorporation in other interlocking or sequencing controls, render other components inoperable.
All modes of test, separation and failure should be considered.
Also, your procedures should be reviewed to ensure the; provide that,whenever a part of a redundant system is removed fr;m service, the portion remaining in service is functionally tested immediately after the disabling of the affected portion and,if possible, before disabling of the affected portion.
2 2.11 State the extent of conformance of the safety systems (None) to Regulatory Guides 1.40, 1.41, 1.47, 1.68, 1.73, 1.75 and 1.80.
With regard to Regulatory Guide 1.47:
(1) The conditions of positions 3(b) and 3(c) are interpreted to include bypasses that result from manipulation of permanently installed electrical control devices located in any accessible area of the plant, and (2) The design criteria for the indication system should reflect the importance of botn providing accurate information for the operator and reducing the possibility.for the indicating equipment to affect adversely the monitored safety systems.
In discussing the Three Mile Island Unit 2 design criteria, the following should be considered:
49~C78 (a) The bypass indicators should be arranged to enable the operator to assess readily
22-8 the operating status of each safety system and detemine whether continued reactor operation is permissible.
(b) Means by which the operator can cancel erroneous bypass indications, if provided, should be justified by demonstrating that the postulated causes of erroneous indications cannot be eliminated by another practical design.
(c) Unless the indication system is designed in conformance with criteria established for safety systems, it should not be used to perform functions that are essential to the health and safety of the public.
Neither should administrative procedures require immediate operator actions based solely c-the bypass indications.
(d) The indication system should be designed and installed in a manner which precludes the postibility of adverse effects on the plant's safety systems.
Failure or bypass of a protection function should not be a credible consequence of failures occurring in the indication equipment and the bypass indication should not reduce the required independence between redundant safety systems.
(e) The indication syrtem should include a capability of assuring its operable status during normal plant operatior, to the extent that the indicating and/or annunciating function can be verified.
(f) Means provided in the control rocm for manually activating system level bypass indications, and the associated procedural or administrative controls, should be described.
22.12 Table 7.5-1 which lists the information readouts available (7.5.1.2) to the operator for monitoring conditions in the plant should be revised or supplemented to include the following:
(1)
Number of available channels.
(2)
Number of required channels.
(3)
Type, number and location of readouts (indicator / recorder).
(4)
The purpose or operator usage of each of the monitored parameters during normal and accicent or post-accident condi tions.
43-C79
22-9 22.13 Provide the criteria and design bases which established the (None) heat tracing requirement, temperature control, monitoring, and pcwer requirements for the boric acid tanks, borated water storage tanks, and spray additive tanks, and related piping of the chemical addition system.
Discuss the consequences of a single failure in the heat tracing temperature control or instrumentation of each of the above mentioned systems.
22.14 Provide the results of a review of your operating,
(:;ane) maintenance and testing procedures to determine the extent of usage of jumpers or other temporary forms of bypassing functions for operating, testing or maintenance of safety related systems.
Identify and justify any cases where the use of the above methods cannot be avoided.
Provide the criteria for any use of jumpers for testing.
22.15 (RSP, We have concluded from the infonnation presented in (6.0, 7.0 a nd the FSAR concerning the Auxiliary Feedwater System 15.0)
(AFS) that this. system is essential to plant safety and must be capable of satisfying its functional requirement after sustaining a break in its piping inside containment and a single electrical failure.
We will require that the instrumentation control, and electrical subsystems associated with the AFS be designed to conform to IEEE Std 279-1971 and IEEE Std 308-1971.
Therefore, (1) Modify your design of these subsystems to conform to these standards and criteria or justify the present design on scme other defined basis.
(2) ~If your design is modified, provide a sufficiently detailed description, including design bases and supporting analyses, to enable evaluation of the new design for confonnance with the stated criteria and standards.
W C80
22-10 22.16 (RCP)
The Staff has recently identified a concern with regard (None' to the apalication of tne single failure critcricn to manually-controlled electrically-operated vaives.
It hcs been cuncluded that where a single failure in an electric system can result in loss of capability to perfona a safety function, the effect on public safety nust be Ovaluated.
This is necessary regardless of whetner the loss of safety function is cau. sed by an active com;cnent failing to perform a requis'ite mechanical' motion, or by a passive component performing an undesirable nachanical mo ti on.
The following Staff position presents an acceptaLle means for neeting the single failure criterion with regard to this type of single f ailure.
(1)
Single failures of both active and passive conconents in the electric systems of valves and other fluid system ccaponents should be considered in designing against single failures, even though the fluid system component may not be called uten to function in a given safety systen operational sequence.
(2) Where it is determined that failure of a single active or passive component in an electric s:'s'-
c:
- ? :s-mechanical motion cf a p ::ive ccmpcncnt in : fluid :ystcm and this motion results in loss of capability to per#: en the system safety runction, it is acceptable, in lieu of design changes that als may be acceptacle. to disconnect oower to the electric systems of :ne ccm;onent.
The plant technical specifications should include a list of cil electrically-operated passive valves, and the required positions of these valves, to which the requirement for removal of electric power is applied in order to satisfy the single failure criterion.
(3)
Electrically-operated valves which are classified as active valves, but which are manually-controlled should be operated from the main controi rocm.
Such valves may not be included among those valves from which power is renoved in order to meet tne single failure criterion unless: ( a) electric power can be res crec to the valves from the main control room,( 5) valve operaticn is not necessary for at least 10 minutes following indication of a plant condition requiring such operation, and (c) it is demonstrated that there is reasonable assurance that all necessary operator actions will be performed within tne time shown to be acecuats by the analysis.
The plant technical specifica-i -s should inciude a list of the required cositions cf manually-controlled, electrically-c;erated valves and should identify those vcives to which the require en; 4-order for remcyal of e!:ctric pcuer is applied to satisfy the single failure criterion.
43 0S1
22-11 (4) When the single failu " criteri;n is satisficd by removal of electric t - er frcr passive valms or from active valves uaetir.g the recuirrena of (3).
above, the associated valvcs :hould have redun@ nt position indication in the rain control rec cnd :nc position indication syste:a sncuid itself meet the single failure criterion.
(5) The phrase " electrically-o gra:ed valves" in.cludes both valves operated directly by an electric device (e.g., a motor operated valve and a sciencic-operated valve) and those valves opcrated ir.directly by an electric device (e.o., an Tir opcrated valve whose air supply is controlled by an eiectric solenoid valve).
Therefore, please provide:
(a) An evaluation of all safety related fluid systems to identify all valves whose failure can result in the loss of capability to perform a system safety function.
(b) A description of the means orovided to meet the single failure criterien in safety related fluid systems where it is identitled t'lat a single f ail' re ' '.
result in tne loss of capability to ucrform :ne systr safety r_unc:1on.
22.17 Section 8.3.1.1.5 indicates that a synchronizing signal is (8.3.1.1) provided to the inverter frca the standby regulated ;c.er source.
Provide a discussion of the ccnsecuen:es of % e loss of this synchronizing signal and any indicators wnich may be available to indicate the signal loss.
22.18 Provide the following additional information with regards ::
(8.3.1.1) the Diesel Generator (D/G) controls and interlocks.
(1)
Section 8.3.1.1.8.3 states that several percissive interlocki, are to be satisfied for cceration in the fully automatic start mode.
Describe the incicaticns available to the operator to apprise him of tne s:atus of these permissive conditions.
(2) With regard to the information presented in Section 8.3.1.'.
8.2e, describe the capability for testing :be two out of three lube oil pressure (low) or two cu cf three crankcase pressure (high) coincident circuits used for tripping t':e D/Gs.
(3)
Provide the design basis for the shutdown reset push-button permissives (item a5 of Section 8.3.1.1.8.3) for operation in the fully automatic start mode.
(4)
Describa the permissive function of the stari. end sap 48'CS2
pushbutton (item b of Sectic
'.3.1.1.8.4) on the diesel generator contrcl '
.i in the control room.
It appears that this item is inconsistent with the opening sentence of this section.
The FSAR states that all essential motor circuits have 22.19 (8. 3.1.1 )
thermal overload devices and that small motors, below 5 HP are also equipped with a thermal switch built into the motor.
Discuss the effect of low bus voltage (e.g., during diesel operation) on motor torque and the resultant possibility for 0/L trip of motor operated valves prior to completion of their stroke.
Also, discuss any provisions for bypassing these 0/L's especially during emergency conditions.
22.20 (RSP)
The standby diesel generators (2000 kW continuous)
(8.3.1.2) described in Section 8.3.1.2.3b are in a size range that has not teen previously qualified for use in nuclear poi er plants.(It is noted that the diesel generators provided for Three Mile Island Unit 1 are rated 2500 kW continuous.)
We will require qualification testing for these units similar to that performed on the Zion 4000 kW diesel generators.
An. acceptable test program would include the following requirements.
(1) At least two tests acceptable to the Staff shall be perfonned.on each diesel to demonstrate the start and load capability of these units with some margin in excess of the design requirements.
(2) Prior to initial criticality, performance of at least 300 valid start and load tests, with no more than three failures allowed. This would include all valid tests performed offsite.
(A valid start and load test shall be defined as a start from design cold ambient conditions with kading to at least 50's of the continuous rating within the required time interval, and continued operation until temperature equilibrium is attained.
(3) A failure rate in excess of one per hundred will require further testing as well as a review of the system design adequacy.
State your intent with regard to meeting these requirements and provide a detailed description of your test program.
22.21 Provide a description of the switchyard batteries (None) installation.
This descr'.ption should include a discussion of the independence of tnese power supplies.
43 CS3
22-13 22.22 Verify that the batteries serving the protection systems (8.3.2.2) loads are independent Seismic Class I installations housed in separate rooms.
Describe these installations in more detail including the battery room ventilation systems.
22.23 (RSP)
The information presented in the FSAR concerning the battery (8.3.2.2) test program is not ccmplete.
We require that your design criteria include:
(1)
The " Procedure for Battery Capacity Tests," as specified in Section 5 of IEEE Std 450-1972, and (2) The frequency of " Performance Discharge Test," as specified in Section 5.3.6 of IEEE Std 3C8-1971 unless a demonstratable technical basis can be established for a greater interval between performance tests.
Revise the FSAR to inc hde the above requirements and justify any exceptions taken. Also, specifically address the capability of the batteries (assuming no other available sources) to perform their intended safety functions for the indicated degraded condition of 203 volts across the battery.
22.24 Regulatory Position 4d of Regulatory Guide 1.6 (Safety (None)
Guide 6) requires that at least one interlock be provided to orevent an operator error that would parallel redundant power sources.
Describe the interlocks which prevent an operator from paralleling the emergency diesel generators by manually closing the following breakers.
(1) Manually closing breakers TlE-2E-2 and T2E-lE-2 or breaker i.k-E-2 and T4E-3E-2 to parallel buses 2 'c and 2-2E or buses 2-3E and 2-4E respectively (reference Figure 8.3-3).
(2) Manually closing breakers T11E-21E2 and T{lE-11E2 or breakers T12E-22E2 and T22E-12E2 to parallel buses 2-llE and 2-21E or buses 2-12E and 2-22E respectively (reference Figure 8.3-5).
(3) Manually closing breakers T31E-41E-2 and T41E-31E-2 to parallel buses 2-31E and 2-41E (reference Figure 8.3-7).
It is noted that for most of tnese pairs of tie breakers, one of each pair is shown on the respective Figures as being norually closed (NC).
With these tie breakers normally closed, the scheme does not meet single failure.
Confirm that these breakers are all nomally open or justify your design on some other defined basis.
or 084
22-14 22.25 Your design provides alternate feeds from redundant buses (None) for make-up pump 1B (reference Figure 8.3-3), reactor building air cooling fan "C" (reference Figure 8.3-5),
nuclear service closed cooling pump "NS-P-lC" and make-up pump "18" auxiliaries (reference Figure 8.3-13), the mechanical trash rack SW-S-lC, the traveling screen SW-S-2C and the screen wash pump discharge strainer SW-S-3 respectively (reference Figure 8.3-7).
Describe the interlocks and administrative controls required for these circuits in order to meet the single failure criterion and position 40 of Regulatory Guide 1.6.
43 CSS
23-1 23.0 CORE DERFORMMCE BRANCH 23.1 Assuming the formation of axial gaps in the fuel column due (0.2.1) to inpile densification of the fuel, what is the calculated time to collapse of the worst case TMI-2 fuel rod?
23.2 What steps are taken to monitor the loadings for shipping and
( 4. 2.1.1.1 )
hardling conditions specified in section 4.2.1.1. l? What are the consequences of exceeding these loads?
23.3 In sizing the fuel rod for fission gas release, is the increased (4.2.1) fission gas which might be generated during a transient taken into account? If so, what method is used to estimate this transient fission gas release?
23.4 Describe the calculations done to d etermine the adequacy of (4.2-2) the 10% margins of internal fuel rod gas pressure over systen design and operating pressures during DNB as discussed in Section 4.2-2.
In these calculations, describe how transient fission gas release is taken into account.
(4.2.1)
(1) Specify what is meant by " excessive cladding strain."
(4.2.1)
(2) What system pressure is assumed during the DNB transient?
(4.2.1)
(3) What cladding properties are assumed for these calculations?
(4.2.1)
(4)
Is pellet cladding mechanical interaction taken into account in these calculations?
23.5 Describe the calculations done to show that primary stresses do (1.2.1.1.2.2) not act for a time exceeding 75% of the time required to produce stress rupture of the cladding as stated in section 4.2.1.1.2.2.
23.6 What is the cold He pressure in the fuel rods for TMI-2 (4.2.1) before irradiation?
23.7 What value is used for irradiation growth of the Zir:aloy (4.2.1.3.2.1) used for fuel assembly components (section 4.2.1.3.2.1)?
23.3 What is the calculated hot pressure in the worst case fuel (4.2.1) rod for TMI-2?
(1) How was this number obtained? Describe both the calculational method and the assumptions used.
49 ' C S.3
23-2 (4.2.1)
(2) What power histories are assumed in these calculations?
If pellet clad contact occurs, how is this handled in the fission gas release calculations?
(4.2.1)
(3)
Is the method of <erforming these calculations described in a topical report? What is the basis of the 3300 psia EOL internal pressure limit? Discuss the possible consequences of the internal pressure of the fuel rod exceeding the system pressure during both nornal and postulated transient?
(4.2.1)
(4)
Discuss how this 3300 psia internal pressure limit is consistent with the discussion page 4.2-2 " Maximum Internal Gas pressure" in which it is stated that "An additional basis is to ensure that compressive cladding stresses exist at reactor temperature conditions for wnich hydride formation may occur."
(4.2.1)
(5) Also discuss how the limit is consistent with tne 105 margins over system design pressure for short time DNB conditions.
(4.2.1)
(6)
Describe in detail the stress rupture calculations done to show that the 3300 psia internal gas pressure is a sound design basis.
Did these calculations include transient effects?
Estimate the number of rods in the core for which their internal gas pressure is greater than the coolant pressure.
23.9 Describe in detail or reference the tests reported in section (4.2.1.3.4) 4.2.1.3.4 in which B&W test fuel rods were irradiated for burnups up to 75,000 Mwd /Tu.
Were some of the rods prototypical of TMI-2 fuel rods? Give the values of the p arameters varied and describe the results of the irradiation.
Was any analytical verification of these results attempted?
23.10 In section 4.2.1.3.5.2 a fuel assembly ficw test is describec (4.2.1.3.5.2) in which flow induced vibrations were measured.
It is stated that
" Vibratory amplitudes have been found to be very small, and, except for a few isolated instances (attributed to pretest spacer grid damage), no unacceptable wear has been observed."
(4.2.1)
(1)
Discuss why such damaged spacer grids could not te present in TMI-2.
49 CS7
(
31-1 P
31.0 ACCIDENT ANALYSIS BRANCH 31.1 Provide a list of the volumes of regions of the contain-(6.2.3)
=ent which are not directly covered by the spray, and which have restricted com=unications with the main sprayed region.
31.2 Describe the tests to be performed to verify that the (6. 2. 3) system, as installed, is capable of delivering the proper mixture of boric acid, thiosulfate, and sodium hydroxide within the concentration limits specified in Section 6.2.3.1.
31.3 The response to Question 12.4 was inadequate as it did not (6.2.3) address the request for Laformation relating to the air tiltration systems. List each engineered safety features (ESF) filtration system.
Tabulate each ESF filtration system with respect to each position in Regulatory Guide 1.52 (see Bellefonte Nuclear Plant PS AR Tables 6.2-2a, 9.4-2, 9.4-3).
Discuss each item of non-compliance in detail.
Any references to other parts of the docket must be specific and indicate page and paragraph.
31.4 Estimate the gamma dose rates to control room operators due (6.2.3) to iodine accu =ulation within the recirculation charcoal filters.
Express the dose rates on a per unit curie basis with respect to the iodines within the charcoal filters.
31.5 In the event of a chlorine release, the ti=e period between (6.2.3) chlorine detection and a buildup of chlorine concentration within the control room to dangerous levels can be relatively short.
Consequently, (see enclosure on chlorine protection) it is necessary to have automatic actuation of an appropriate prutection action for the control room.
As outlined in the enclosure, plants with a single fresh air inlet should have an automatic control room isolation capability in conjunction with a chlorine detector signal.
Co= pare your plant design against the enclosed design reco==endations and provide us with the results of your review and a description of any changes that may be necessary to achieve a degree of protection equivalent to that which is described in the enclosure.
4W CSS
PROVISIONS FOR ADEOUATE PROTECTION AGAINST A CHLORINE RELEASE Adequate protection of the control room against an on-site chlorine release will be achieved if provisions are included in the plant design to isolate the control room automatically to limit the potential build-up of chlorine within the control room, and if equipment and procedures are provided to assure i= mediate use of breathing apparatus by the control room operators.
Similar precautions would help =itigate consequences of =ost postulated toxic gas releases.
To accomplish the automatic isolation quick-response chlorine detectors should be located in the fresh air inlets to the control room. These detectors should be able to detect and signal a chlorine concentration of 5 ppa or less. The detectors should cause isolation of the control room within 10 seconds after arrival of the chlorine.
Detectors should be provided at the control room fresh air inlet for all plants that have storage facilities that might accidently release a total of 500 pounds of chlorine. Additional detectors should be provided at chlorine storage locations that are less than 100 ceters fro =
the control roon or that may release more than 3 tons of chlorine as a result of any postulated accident.
These detectors should be placed, and the detector trip point adjusted, so as to assure detection of a leak or a container rupture.
Detector trip signals should initiate automatic isolation of the control room and provide an audible alarm to the operators. The =eans used to initiate automatic isolation should meet single active failure and seismic criteria.
Adequate isolation requires all openings to the control room to have low leakage characteristics.
This would include doors, dampers, and penetrations.
Total infiltration into the isolated control room should be less than 100 cfm* assuming a 1/8" water gage pressure differential across all openings and the maxi =um operating differential across the isolation dampers upstream of recirculating fans. This leakage limit should be reduced to 25 cfm* if chlorine storage is within 100 meters of the control room of if = ore than 3 tons of chlorine can be released as a result of any postulated accident.
Nor=al fresh air make-up should be limited to no = ore than 1 air change per hou r.
An administrative procedure should provide all doors leading to the control room be kept closed when not in use.
- These leakage rates are based on a control room volu=e of 100,000 cubic feet and thus should be adjusted as directly proportional to actual control room volume.
- 45) 0S:9
Control room isolation should be followed i==ediately by the Start-up and operation of the emergency recirculating charcoal filter or equivalent equipment designed to remove or otherwise limit the accumulation of contamination within the control room.
Under certain meteorological conditions control room isolation may not be sufficient by itself to limit chlorine concentrations to levels below those which cause physical discomfort or disability.
Therefore, the use of self-contained breathing apparatus should be considered-when developing a chlorine release c=ergency plan.
Since calculations indicate that rapid increases in chlorine concentrations are possible, emergency plan provisions cad rehearsal of these provisions for it=ediate donning of breathing apparatus on detection of chlorine release are neccssary.
Storage provisions for breathing apparatus and procedures for use should be such that operators can begin using the apparatus uithin two cinutes after an alars.
Donning of breathing apparatus should bc =andatory prior to the determination of the cause of an alarm.
A toxic environment =ay be present for several days or longer if a chlorine leak cannot be fixed or the leaking container removed.
In any event, adequate bcttled air capacity (at least six hours) should be readily available on-site to assure that sufficient time is available to locate and transport bottled air froa off-site locations.
This off-sice supply should be capable of delivering several hundred hours of bottled air to the members of the emergency crew.
1 solation and air supply equipment relied on should accenmodate a single failure cf an actrva component and still perform the required function.
(In the case of self-contained breathing apparatus this may be accomplished by supplying one extra unit for every three units required.)
Protection requirements for plants located nearby other facilities that store significant quantities of chlorine or plants located nearby cajer chlorine transportation routes vill be determined on a case-by-case basis.
Similarly plants having storage facilities that might accidentally release a total of 500 pounds of chlorine or less will be reviewed on a case-by-case basis to deter-mine need for protection against accidental release.
49~090
h 32.1-1 32.1.0 SITE AN ALYS IS BRANCH - METEOROLOGY 32.1.01 Provide the design basis snow and ice load for the (2.3.1.3) roofs of all safety related structures.
32.1.02 Since severe meteorological conditions (e.g.
- icing, (2.3.3) lightning and hail), that can seriously affect the meteorological instrumentation, may occur at this site and Regulatory Guide 1.23 states that the meteorolo-gical program is to be maintained to minimize extended periods of instrument outage, provide a discussion of the amount of time required to rest re the data collection capability of the operational onsite meteorological system in the event that the system is partially or totally impaired and, if totally impaired, any pro-visions and estimates of time required to obtain data from an alternate source.
32.1.03 An inspection of the meteorological tower at the site by the staff indicates that the wind measuring equip-ment located at the 100-foot above ground level (AGL) should be relocated to conform with the recommendations of Regulatory Guide 1.23.
Provide information concer-ning plans to either relocate the existing wind measuring equipment to (or add additional wind measuring equipment at) a level on the tower 10 meters (33-feet) abcve the level of the dike or the trees on the north end of the island; whichever is highest.
32.1.04 Provide any revisions to the meteorological extremes (2.3) presented in Section 2.3 (including the Tables) made necessary by meteorological events occurring subse-quent to 1971.
32.1.05 As recommended in Regulatory Guide 1.21 - " Measuring, (16. 3)
Evaluating and Reporting Radioactivity la Solid Wastes and Releases of Radioactive Materials In Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," provide a revision to Section 3.9 of the Technical Specifications reflecting the necessit of monitoring meteorological conditions whenever gaseous effluents are being released.
32.1.06 Provide a technical specification describing the post-(16.4) operational, onsite meteorological monitoring program including the tower location, instrumentation, maintenanc-and calibration procedures and schedules, data recording, filing and disposition, and provisions for control roon monitoring of meteorological parameters.
49 091
32.2-1 32.2.0 SITE ANALYSIS BRANCH - RYDR0 LOGIC ENGINEERING 32.2.1 Substantiate the adequacy of site protection for the effects (2.4.2.3) of local intense precipitation, including the PST.
Provide assurances that the site drainage system, including the roofs of safety-related buildings, will prevent flooding of safety-related facilities.
Describe and provida details (sizes, dimensions, cocposition, grades) of site draiu c,e facilities such as culverts, ditches, canals, and drains.
Present applicable design bases to substantiate the adequacy of the designs.
Provide a core detailed site =ap, with detailed topography, shewing locations of levees, channels, and culverts, any other features involving hydrologic considerations.
32.2.2 Properly reference Hydrometeorological Report No. 40.
(2. 4. 3.1) 32.2.3 Furnish detailed river cross-sections used in deter =ination (2.4.3.5) of water surface profiles and division of flow thrcugh the various Susquehanna River channels at the site.
32.2.4 Discuss sedimentation and estimate sedi=ent loads in the (2.4.9) vicinity of the intake structure.
Provide assurances that material cannot be deposited during a flood or a flood recession in a manner which would cause the cooling water intakes to be segregated from the river cooling water.
32.2.5 (1) The riprap and the riprap toe should be designed in (2.4.10) accordance with EM 1110-2-1601, Hvdraulic Desien of Flood Control Channels, or si=ilar conservative criteria, for protection against channel velocities if they are the controlling design basis.
Define the controlling design basis and criteria (wind waves or channel velocities) used to establish riprap size and thickness.
Provide the bases for filter thickness and gradation.
Provide the estimated scot-depths, and basis therefore, anticipated a.t.the toe of the dike during.the design flood and. PMF. _ Provide detailed channel cross-sections that were used in riprap design.
(2) Docu=ent that the riprap will withstand the ef f ects of flood waves and channel velocities created by the f ailure of Stoney Creek reservoirs, assuming various levels of 48~CO2
32.2-2 ficw (up to 1/2 PMF) in the Susquehanna 2iver.
Include wave heights and velocities and describe the nature of the waves.
(3) Discuss any adverse ef fects on the access bridge for flows up to and including the PMF, due to accu =ulation of deoris, scouring of bridge piers and abut =ents, and sedi=entation.
Will the bridge provide access to T2E for all Susquehanna River flood levels up to and including the PMF?
Discuss the need for access during floods.
(4) Describe place =ent of the riprap.
Was, or vill any riprap be placed underwater?
(5) Provide the bases for design ef additional riprap protection that =ay be required in the vicinity of structures which protrude into the channel flew area or otherwise
=ay produce vortices and/or other endesirable flow patterns.
(6) Furnish results and bases of co=putations of wave action on the levee and saf ety-related facilities for the levee design flood and the PMF.
Provide the bases for ycur decisien not to place riprap on the south dike across the island.
(7) Provide assurances that all safety-related structures have been adequately designed to withstand flooding and the static and dyna =ic effects thereof, such as uave action. runup, and splash during the occurrence of the PMF.
32.2.7.
Present in detail the hydraulic design bases for the decer=inatic (2.4.11) of water surface elevations during lowflew conditions.
32.2.8.
Provide assurances that liquid effluents can be discharged (2.4.12) through the plant discharge canal, or otherwise saf ely disposed of, at any river level up to and including the level requiring plant shutdown.
Discuss any flooding affects related to this structure.
32.2.9 Provide assurances that all necessary s'teps taken to prevent (2.4.14) flooding of safety-related facilities during various stages of E=ergency Closure, Shutdcwn, or Shutdewn Alert can be accomplished in the available ti=e.
The discussion should be referenced to the PMF hydrograph, water surface elevations, rate of rise and tine required (incl 2 ding bases for t1=e required) to acconplish necessary flood-prevention nctions such as gate closures and ficod-barrier place =ents.49-093
32.3-1 32,3.0 SITE ANALYSIS BRANCH - GEOLOGY, SEISMOLOGY, AND FOUNDATION ENGINEERING 32.3.1 The rip-rapped dike section between the two intake structures (NONE) shows some surficial distress in the half next to the Unit 2 intake structure.
In this area, possible erosion and sub-sidance of the supporting dike =aterials may have caused several low spots and sags to the surface of the rip-rap.
The construe:1on sequence of the dike and cofferdam for the Unit 2 intake does not assure that rip-rapped slopes were supported by properly engineered fill.
Investigate the quality of this fill and perform appropriate stability analysis of these slopes to assure that they will remain stable under static and seismic loads.
Show that the dike section cannot slide into the water intake area and create a blockage.
49 094
33-1 33.0
_RADICLCGICAL ASSESSEiT BRECd 33.1 In Section 12.1.1 discuss plant design features with regard (12.1.1) to Regulatory Guide S.8 "Infor=acion Relevant to Maintaining Occupational Radiation Exposure As Lcw As Practicable."
In particul r show hcw the guidance in paragraph C.3:
Facility and Equipcent Design have been used in the desi,.,.
Section 12.1 of the 73AR states m y of the objectives but does not provide specific exa=cles which de=onstrate that occupational exposures will be kept as 1cw as practicable.
The follcwing ques tions illustrate considerations that have not been centioned or have not been described in sufficient detail.
33.2 What type of surf ace will be used on walls, ficors and ceilings (12.1.1) in areas where equip =ent which =ay contain radioactive
=sterial is located? Discuss this type of surf ace with regard to buildup of centa=ination and with regard to decont-i-ation.
33.3 What types of covable and te=porary shielding will be available (12.1.1) on-site? Discuss provisions that will be cade to allcw the utilization of te=;orary shielding if necessary.
I 33.4 b"nat design features have been incorporated to ensure that (12.1.1) equip =ent which may require servicing will be accessible so that service tire and thus potential exposure ti=es will be =ini=ized?
33.5 Section 12.1.2 is inec=plete.
Provide scaled layout and (12.1. 2) arrange =ent drawings of Unit 2 shcwing the location of all radioactive sources of significence.
Provide on the laycuts the radiatica ene designations including zone boundaries for both nor=al operational and refueling conditions.
The layouts should show shield wall thicknesses, centrol access areas,
persennel and equip =ent decent 4-ation areas, conta=ination control areas, traffic patterns, location of the health physics facili les, location of airborne radioactivity and area radiation =cnitors, location of control panal(s) for radwaste equipcent and compenents, loca on of the ensite laboratory for analysis of che=ical and radioactive sa=cles, and location of the counting roo=.
33.6 Provide figures in suf ficient detail to show whether all (12.1. 2) aanual valves in radiation areas have shielding between then and a radioactive source.
Give procedures used in operating any unshielded valve.
4T 025
33-2 33.7 Section 12.1.3 is inco=plete as estimates of dose rates and (12.1. 3) isotopic inventory have not been provided.
Describe the scurces of radiation that are the basis for the radiation protection design it. the = saner used as input to the shield design calculation. The description should tabulate sources by isotcpic co= position or ga=na ray energy groups, strength (curie content) and gec=etry as well as provide the basis
~
for the values.
Scurce description should be provided for equip =ent of the radioactive vas ta =anage=ent syst;a:s,
spen: fuel storage pool, various auxiliary sys tens, and the equip =ent containing A-41 scurces.
33.8 If solid waste storage drums have manual hock-up wires, esti=nte (12.1. 5) the yearly =an-res occupational exposure during their usage.
Give esti=ated dose level in the area and the a=ount of exposure ti=c.
Co= pare this =an-rem exposure with the =an-ren exposure that would occur if re=ote drum cappers were used instead of =anual hook-up wires.
33.9 The solid waste storage space and facilities appear inadequate, (12.1.5) especially if equipnent like a de=ineralirer is out of service and =ust be drained.
Give procedures to be used if capacity of waste storage space is exceeded.
Esti= ate the==wd radiation levels and occupational exposures from these was te storage areas.
33.10 No esti= ate is given of doses to ecnstruction workers on Unit (12.1.6) 2 from the operation of Unit 1.
Provide esti=ated annual doses to construction workers due to radiation from the amdM ary building, the reactor building, outside tanks and storage areas which =ay contain radioactivity, and from radioactive effluents (direct radiation fro = the gaseous radioactive effluent pluce) from Unit 1, Provide esti=ated =m-rem doses for the construction period.
Include models,
assu=ptions, and input data.
33.11 Section 12.1.6.1 is inco=plete.
Only the total =an-re= is (12.1.6.1) esti=ated.
Provide the estimated occupancy of the varicus plant radiation areas during nor=al operation and anticipated operational occurrences.
For areas with expected airborne radioactivity concentration provide esti=ated =an-hours of oc cupancy.
Provide an estizate of the annual can-rem doses associated with major functions such as operation, maintenance, radwaste handling, refueling, and inservice inspection. Provide the basis, =odels, and assu=ptiens for the above values. For an exa=ple of an acceptable approach see Section 12.1.6.2 of the PSAR for Allens Creek Nuclear Generating S tatica.
02 ~ ON
41-1 41.0 QUALITY ASSURANCE 41.1(RSP) Met-Ed is requested to amend section 17.2.7 of its description (17. 2. 7) of the QA Progra::t for Station Operation for Unit 2 to include a statement which co:raits to comply with, the " Gray I'co':", USH-12S3 Revision 1, " Guidance on Quality Assurance Rcquire:. 'nts During, Design and Procurement Phase of Nuclear Pcwer Pl.'nu - Revision 1,"
dated :hy 24, 1974, and the " Green Ecc F,
PLASH-130f "Guilance on Quality Assurance Requirements During the Constn.ction PN.se of Nuclear Power Plants," dated
ay 10,19~4, for those desi n procurement, and construction activities which may occur durine the operations phase and for which the guidance is applicable, or identif>
any exceptions to the guidance contained within " Gray Book andior the " Green Book" and described acceptable alternatcd.
41.2 You state in Section 13.1.1.4 that the technical support for the (13.1.1.4) operation of Tnree Mile Island 2 will be provided 5 the :'.anacer, Generation Engineering, Shnaga; Generatjan Operations, una..mm. er, Generation Maintenance and their staffs. In regard to these groups, descrioe their assigned engineering staffs including the number of personnel assigned, their qualifications, educational backgrouncs and experience.
41.3 Describe for those groups defined above in 41.2 :. hat other pc:.a (13.1.1.4) plants for which they provide technical support.
41.4 Describe the specific succession of responsibiliU for the overall 13.1.2.2) operation of Three Mile Island 2, in the absence of the Plant Superintendent.
11.5 Indicate your expected date of assignment of a person to the po :i'. -
()3.1.3.2) of Nuclear Engineer, Unit 2.
41.6 Your response in Amendment 14 to iten 7 - 4/14.1. 3..'. contained in (None) our request for additional information dated March 19, IF4, inc h
+
that you planned to submit test abstracts for prcape:atienal :.r;s prior to December 31, 1974. 'In order for our review schedule to be maintained it is necessary that abstracts for all pccoperatic.c.,..
power and power ascensions tests be submitted by October IS,19 Your test abstracts for transient or dynamic power ascension tests show discuss the status of mode of operations of major plant control sytems.
49-ON
~
41-2 41.7 Provide a listing of Regulatory Guides. applicable to initial test (14.1.1) programs, that will be incorporated into the initial test program for Three Mile Island-2.
Provide justification for those Regulatory Guides or portions of Regulatory Guidas, applicable to initial test programs, that are not planned to be incorporated or followed.
41.8 Expand the discussion pertaining to how design requirements and (14.1.1.1) test requirements.are translated into test procedures.
In your reply, address the extent that responsible design organi:ations will be used to provide design performance requirements and acceptance criteria and describe the interfaces between such design organizations and other participants in the initial test program.
41.9 Expand the discussion on how individual test procedures, generic (14.1.1.2) test procedures and special procedures will be assigned safety related classifications.
Identify the organization or individual responsible for assigning classifications and the organizational units or individuals responsible for review and appmval of procedures in each classification.
41.10 Expand the discussion on test planning meetings.
Describe the (14.1.1.6) functions of these meetings, identify the participants at these meetings and clarify how these meetings interface with the functions and responsibilities of the Test Working Group.
41.11 Describe che organizations involved and the general methods that (14.1.3.1) will be used to develop and approve constmction test procedures.
This discussion should identify the organizations or persons responsible for establishing perfomance requirements and acceptance criteria for construction tests.
41.12 (14.1.1. 3)
Expand the discussion on prerequisite lists.
Identify the organi:ations or job positions responsible for assuring required prerequsites are satisfied. Describe the controls established to assure the review and approval of test results from major phasas of the test program before proceeding with the next phase (construction test phase, preoperation test phase, initial fuel loading and initial criticality and low power tests). Discuss the plans pertaining to the review and approval of test results at power test plateaus, during the pcuer ascension phase, before increasing power level to the r. ext test plateau.
41.13 Provide a resme of the individual designated as the GPU Initial (14.1.1. 4)
Fuel loading Supervisor and describe the review and approval chain for i'litial fuel loading procedures and initial criticality test proce cures.49-098
41-3 41.14 Modify the test schedule provided in Figure 14.1-1 to indicate (11.1.1.4) the time period scheduled to conduct major test phases relative to the scheduled date for initial fuel loading.
Tests to be conducted in each major test phase should be specifcally identified to pemit cross indexing to test abstracts.
A discussion of the overall test schedule should be provided to establish the sche 'iled time for development of test procedures for each phase, the schedule for establishing the organizations thn will dexeloo and carricinate in the test program and the scheduled time for having approved test procedures available for review, prior to their >se, by Regulatory Inspectors.
41.15 Expand the discussion on the planned use of operating and emergency (14.1.5) procedures during the initial test program.
In your response, indicate the schedule for pmviding draft operating and emergency procedures and describe the extent that such procedures will be trial use tested during the initial test program.
41.16 Describe the specific provisions established for initiating, reviewing (None) and approving plant repairs and field chang ts or modifications that are determined to be regnired by the performance of construction tests and preoperational tests.
Include a discussion of the controls provided to assure appropriate changes to test procedures and appropriate retest Mg.
41.17 Cur review of abnomal occurrende reports from operating power reactors (None) discloses that the causes for a large fraction of these occurrences could have been detected and eliminated during initial test programs.
Discuss your program or methods for review of reactor plant operating experiences and describe how the results of this program will be appropriately factored into the initial test program for Three Mile Island-2.
41.18 Provide a su:miary description of the controls established to assure (None) all applicable technical specification requirements will be satisfied prior to initial fuel loading and for subsequent testing.
41.19 Provide the number of GPU Startup and GPU Cognizant engineers to be (Nc ')
provided during each major phase of the test program and the schedule for providing these personnel.
41.20 The list below identifies certain systems, components and plant (None) features that do not appear to be scheduled for testing based on our review of test titles pmvided in Table 14.1-1, 14.1-2 and 14.1-3.
If the performance of these systems, components or plant features will be verified during, or as a part of some other scheduled test, identify which tests will be be used to demonstrate their functional adequacy.
If 5.e performance of these systems components in plant features is not scheduled to be verified by testing during the initial test program provide your reasons for not w 099
41-4
~
testing. Also included in the list are power ascension tests reconr. ended by Regulatory Guide 1.68 that do not appear to be scheduled for performance. State your position relative to the conduct of such tests in your reply.
(1)
Control Rocm hVAC System (2)
Reactor Building Purge and Recirculation System (3) Ccmbustible Gas Control Purge System (4)
Containment Penetration Roca Fihaust Air System (5) Diesel Generator Building hVAC System (6) Fire Pump House Ventilation System (7) Auxiliary Building Ventilation System (8)
Cable Roca E/AC System (9)
Leak testing of Centalment Penetrations and Airlocks (10) Vital Electrical Power Buses and Controls (11) Seismic Instrumentation (12) Reactor Coolant Leak Detection System (13) Containment Spray Additive Systems (14) Internediate Closed Cooling Water System (15) Concentrated Boric Acid System (16) Secondary System Radiochemical Sampling System (17) Radicactive Liquid Gaseous Collection, Treatment and Effluent Systems (13) Fire Protection System (19) Containment Vacuum Relief Valves49-100
i 41-5 (20) Main Steam Isolation Valves (21)
Emergency Core Cooling System Pump Raom $xhaust Air System (22)
Emergency Diesel Generator Auxiliary Systems and Fuel Supply Systems (23)
Communications Systems (24)
Emergency Lighting System (25)
Fuel Transfer System (26) Condenser Circulating Water System (27) Reactor Building Hydrogen Analyzer System (28) Condensate Storage Tank and hhin Condenser Hotwell and Associated Controls (29) Spent Fuel Cooling System (30) Reactor internal's and Reactor Coolant System Equipment and Component Vibration Tests (31) System Expansion and Restraint Tests for the Power Conversion System and Emergency Core Cooling System
'(32)
Control.Rocm Computer (33) Fuel Handling Building Ventilation System
~ - -
(34) Main Steam Systems'and Pressuri::er Electrcmatic Relief Valves (351 Asynnetric Control Rod >bnitor (36) Axial Xenon Oscillation Control Test (37) Reactor Coolant System Natural Circulation Tests 49 101
i*
\\
42-1 42.0 INDUSTRIAL SECURITY AND EMERGENCY PLANNING BRANCH 42.1 Discuss provisions for coping with non-radiological incidents, (13.3.1) e.g. toxic gas release, bomb threats, sabotage.
22.2 (R5P)
Your proposed date of January 1976 by which copies of written (13.3.1) agreements will be submitted as an amendment to the Emergency Plan is unacceptable.
It is our position that all agreements be submitted by April 1975.
42.3 What is the status of the " Plan for Hospital Reception of (13.3.1.2)
Radiation Casualties", to be developed for the Hershey Medical Center? Provide the effective date of the Plan.
42.4 What is the relationship between the applicant and Radiation (13.3.1.2)
Management Corporation?
Is there a letter of agreement or a contract for emergency services?
. i.
42.5 What additional means are available for determining a General (13. 3.1.4)
Emergency? An offsite survey could consume a time pericd that (13 A. 6. 3.1 )
might be critical for initiating protective measures in the (13A.S.3.2)
Exclusion Area and in the Low Population Zone.
42.6 Discuss what actions will be required of Unit 1 personnel (13.3.1.5)
(and visitors to Unit 1), following announcement of a (13A.6)
Unit 2 energency.
. ~. ~ Describe the onsite first aid and pe.rsonnel deco'ntamination '
~
a2.7 (13A.7.2) facilities as required by 10CFR50, Appendix E, Part IV, at
_ Section F.
12.3 Describe the program for training and for drill participation (13A.10) of those offsite personnel who may be called upon to provide emergency support as required by ICCFR50, Appendix E, Part IV at Sections H and I.
12.9 Provide additional information concerning the evacuation of perscns from the exclusion area and from any potentially affected sector of the envirens out to a distance of 4 niles from the center of the exclusien area.
Include explicit information on the following:
(1)
The time recuired to make predictions of projected dose and to designate the areas containing the copulation-at-risk.
(2)
The maans that will be employed and an estimate of the time required to noti fy the population-a t-risk.
47 102 I
n'
~4 42-2 (3)
The times required to effect the evacuation proper, frcm the exclusion area and from each sector, or increments thereof, of the environs.
Address in particular the following:
(a) the recreation area planned for the southern part of Three fiile Island, (b) the towns of Middletown, Goldsboro, and York Haven, and (c) the nearby islands in the Susquehanna River.
(4)
The identity and character of egress routes and the means assumed for effecting the physical evacuation.
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e s e e w
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