ML20033D361

Forwards Details of 11 SER Issues Discussed at 811130 Meeting

ML20033D361
Person / Time
Site:	Clinton
Issue date:	12/03/1981
From:	Geier J ILLINOIS POWER CO.
To:	John Miller Office of Nuclear Reactor Regulation
References
L30-81(12-03)-6, L30-81(12-3)-6, U-0354, U-354, NUDOCS 8112070572
Download: ML20033D361 (29)

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/LLINDIS POWER 00MPANY p

L30-81 (l2-03)-6 500 SOUTH 27TH STREET, DECATUR, ILtJNolS 62525 December 3, 1981 Mr. James R. Miller, Chief

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Standardization & Special Projects Branch S

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Division of Licensing Eg (f y@h -@

Office of Nuclear Reactor Regulations S3 O U. S. Nuclear Regulatory Commission 2/

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/gg Dear Mr. Miller-

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/n/b / l l.l, Clinton Power Station Unit 1 Docket No.

50-461 Attached are details related to the following items which were discussed with N. Fioravante, Auxiliary Systems Branch, during a meeting of November 30, 1981 to resolve issues for the Clinton SER:

ISSUES Structure, Systems, and Components to be Section 3.5.2 protected from Externally Generated Missiles Section 9.1.1

- New Fuel Storage-Sectioli 9.1.2 - Spent Fuel Storage Section 9.1.3 - Spent Fuel Pool Cooling and Clean-up System Fuel Handling System Section 9.1.4 Function Design of Peactivity Control System Section 9.2.4

- Potable and Sanitary Water System Section 9.3.1

- Compressed Air System Section 9.4.2 - Spent Fuel Pool Area Ventilation System Auxiliary and Radwaste Area Ventilation Section - 9.4.3 Section 10.3

- Main Steam Supply System Section 10.4.7 - Condensate and Feedwater System sg YNEN $$f8Q PDR.

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.Page 2 Shared Systems Grand Gulf Question 211.13, Clinton Response The above items are considered by the NRC and IP to be closed for CPS Licensing purposes.

Sincerely, J-91 i -

Manager, Nuclea/

J.D. Geier r Station Engineering Attachments cc:

J.H. Williams, NRC Clinton Project Manager-H.H. Livermore, NRC Resident Inspector N. Fioravante, Auxiliary Systems Branch

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Issue Section 3.5.2 - Structure, Systems, and Components

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to be Protected from Externally Generated Missiles-Additional information concerning missil'e barriers for HVAC intakes and exhaust blowout panels in the auxiliary building roof if needed.

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Response

Plan and elevation drawings, listed below, which show the ventilation openings and missile barriers for all HVAC systems, were sent to the NRC on November 4, 1981.

M14-1018-2, Rev. N HVAC Equipment Room Control Room Bldg. Plan El. 825' M14-1018-8, Rev. H HVAC Equipment Room Control Building Sections M14-1019-3, Rev. G HVAC Equipment Room, Diesel Generator and HVAC Building Plan El. 762' There are no blow out panels on the auxiliary building roof and exterior to any other building housing safety-related eg'uipment.

Figure 3.5-3, Sheet 5, will be revised.

Figure 1.2-14' correctly shows no blow out panels on the auxiliary building roof.

Action Required-Revise FSAR Figure ~3.5-3,. Sheet 5.

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_ Issue Section--9.1.1 - New. Fuel Storage The applicant should verify that the new fuel vault is-seismic Category I..

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Response

The new fuel stordge vault is a seismic Category I structure.

This statement will be added to FSAR Subsection; 9.1.1.1.-2.

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' Action Required' 1.

.IP to: revise'FSARiSubsection 9.1.1'.1.2.

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CPS-FSAR AMENDMENT 6 AUGUST 1981 CHAPTER 9 - AUX'ILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.1.1 New Fuel Storage 9.1.1.1 Design Bases 9.1.1.1.1 Storage Design Bases New fuel' storage racks are provided for approximately a.

38% of the full core fuel load in each unit.

b.

New fuel storage racks are designed and arranged so

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that the fuel. assemblies can~ be handled efficiently during refueling operations.

9.1.1.1.2 Safety Desian Bas'es - Structural The'new fuel s'orage racks, fully loaded with fuel a.

t assemblies, are designed to withstand all credible static and dynamic loadings, to prevent damage to the structure of the racks, and therefore the contained fuel, and to minimize distortion of the racks arrangement.

(See Table 3.9-2[h].)

{'*. g b.

The modules are designed to protect the fuel assemblies from excessive physical damage under normal or abnormal conditions caused by impacting from either fuel assemblies,' bundles or other equipment.

The racks are constructed in accordance with the c.

Quality Assurance Requirements of 10 CFR 50, Appendix B.

Th'e new fuel storage racks are categorized as Safety d.

Class 2 and Seismic Category I.

'e.

The new fuel storage facility is designed in accordance with General Design Criteria 2, 3,

4, 5,

61, 62, and 63.

The new fuel storage facility is also designed in accordance with Regulatory Guides 1.13, 1.29, 1.102, 1.115, and 1.117.

Its design AJJ 2"d** ""Y #' "'"ri integrity due to phenomena such as earthquakes, a

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flel sfuge va d is a seism;c. Cologory h

new I sinufure.

9.1.1.1.3 Saf e y Desfon'

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The new fuel storage racks are designed and maintained with

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sufficient spacing between the new fuel assemblies to assure that

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the array, when racks are fully loaded, shall be subcritical, by at least 5% K eff including allowance for calculational biases and 9.1 -

o Issue

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Section 9.1.2 - Spent Fuel Storage The. applicant sould verify the spent fuel dropped conditi'ons

- for the upper containment pool.

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Response

The fuel storage racks in the upper containment pool will. maintain l-l keff < 0.95 under the abnormal conditions of a dropped fuel bundle resting horizontally.on the top of the storage rack.

This ab-normal condition is bounded by the case where a fuel bundle-is adjacent to the fuel rack (item 9.1.2.3.1.2.J) in which case the-fuel rods face to face and at a' smaller separation-distance than is-achievable by being on top of the stored fuel'(i.e. separated by the fuel plenum and upper ' tie p1dte regions).

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Issue

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Section g. 1.3 - Spen't Fuel Pool Cooling and Cleanup System In Section 9.1.3 of the FSAR, the design heat load is based on the decay heat load from storage of an amound of fuel ecual to approximately 2.17 reactor cores.

However, in Section 9.1.2, a storage capacity of 4.70 cores in the spent fuel pool is indicated.

The applicant should reconcile the apparent imbalance between the design heat load and the storage capacity.

1.

Verify that the calculation for decay heat loads are made in accordance with Brr.nch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling."

Provide a summary of the calculations for the case of the maximum (last) normal refueling and for the case of a full-core off-load 30 days after the last laonnal refueling.

Indicate the assun.ptions including decay time before off-load for both cases.

2.

For both of the above cases, provide the time before pool boiling and the required makeup-rate assuming a loss of all pool cooling.

Response

The system-design heat load referenced in Section 9.1.3 is the design basis for the FPC&C pumps and heat exchaggers.

The system design heat load (19.7 X 10 Btu /hr) represents the peak. heat load resulting from 8 normal refueling cycles.

The

. spent fuel assemblics~in the pool would be equiv-alent to approximately 196% of one' reactor.

Prior to the utilization of neutron absorbing (BORAL)-

fuel racks, 217% of a reactor core represented the i

maximum storage capacity of the spent fuel storage pool.

The neutron absorbing fuel racks have in-creased the spent fuel storage capacity to 400%=

of a reactor co93 (See Section '9.1.2).

CPS-FSAR

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& 3 The FPC&C. system has two trains 'of pumps and heat

exchangers.- Each tragn has a heat dissipation capacity of 19.7 x 10 Btu /hr under normal system operation.

Under. abnormal conditions, an alternate cooling water m'edian (Shutdown Service Water System) ' '

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-can be u- ! in the FPC&C heat exchangers ingreasing the heat dissipation capability to 33.2 x 10 Btu /hr.

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1.

Calculations for the decay heat loads were made in accordance with Branch Technical Position calculations we/Po values used in decay heat ASB 9-2.

The P re obtained from the fuel assembly supplier, General Electric.

The' comparison be-

-low indicates that the P/Po used are more con,LL servative and results in higher heat load va19es

,than Figures 1,2,3 of Branch Technical Position ASB 9-2.

,e P/Po Time after used in Clinton Shutdown From Figures 1,2,3 Heat Load Cal-(Seco.tds) of ASB 9-2 culations 4

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-2 10 1.16 x 10 1.25 x 10 5

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-3 10 5.7 x 10 6.6 x 10 6

10 2.5 x'10-3 1

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2.85 x 10

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10 not given 1.25 x 10 8

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10 not given 5.9 x 10 9

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~4.0 x 10 Table 1 provides a' summary of the calculations for ;the maximum (Last). normal refueling cycle and for the case of a full-core.off load 30 days after the last normal' refueling.

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load is 50.33 x 10 Btu /hr.

Operating:both FPC&C system trains (cooled by shutdown Service Water System); the FPC&C system would 6-

' have. a heat dissipation capacity' of 66.4 ' x 10 -

Btu /hr.

If one or both'of the FPC&C trains'were-notLavailable, the Residual Heat. Removal (RHR)'

systemLcan be connected to the'FPC&C system to-

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- maintain spent fuel pool w'ater temperatures'be-

.lowtl50*F.

2..' Assuming-loss ofiall spent fuel pool' cooling capacity and applying the maximum heat. loads calculated for the-fuel pool,ethe summary-belowf provides thef time 5before pool boiling-and make--

- up; rates.

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' Time for Popl Pool water Make-Temp.'to Raise _

up Rate

• ' Pool Inventory 120'F to 212*F Recuired Avail.

Last normal re-52'3.6 Min.

. 50 gal /nin 150 gal /

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fueling (13 re-(8.7 hrs. )

min fueling,. cycles).

Full-core off-h.29 Min.

08 gal /

150 gal /

' l'oad plus last'.

(3.8 hrs.)

min min normal refueling Cl3 refueling

' cycles).

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Action Required IP to revise FC&R Section 9.1 3.

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T ele ?? Fmt 92e1 Sterm Feel Put fo%

PJIL-CCRZ CFF-CAD; azactor XCP.lAL REPJELDiG

,4 Shutdem 134 hrs. after Inst CYCLE tiemi RefMeliner b

% of Heat had 5 cf Heat Icad 0

6 Rehtel Cycleg Ta hmre Re.eter ce,e nt.3/h. r 10 Refuel C.eler A hmrs Panete= ce m Bt.4. r ?O 13 110 20 7.88 14 224 100 30.42 12 8760 21 1.78 13 468 20 4.66 11 717520 21 1.42 32 9118 21 1.75 6

10 26280 21 1.24 11 17878 21.

1,41 9

35040 21 1.17 10 26638 21

'.1.24 8

43000 21 1.13 9

35398 7

52560 23 1.21 8

44158 21 1.17 21 1.13 6

61320 5

70080

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23 1.17 7

52918 23 1.21 24 1.20 6

61678 23 1.17 4

78640 24 1.16 5

70438 24 1.20 3

87600 23 1.10

'4 79198 24 1.16 2

96360-26 1.22

'3 87958 23 1.10 1

105120 32 1.49 2

96718 26 1.22 - '

1 105478 32 1.49 TOIAL =

300 23.17 TOTAL =

400 50.33

' Note;1;- The Spent fuel storace pool heat leads are mazi u:a heat loads vnich decrecces as(tiIe el asse=blies newM.orep increases.

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Heat had = (3.413 Etu }l2 l((% of reactor);2G941.f7,Themal Pouq)

' *4 2 Reactor

(

'E-hr ),Po Coro

);

J3 74 = the time after reactor shutdom for the applicable refueling cycle.

yf. 8760 hmrs = 1 year.

no hours is the 'hortest possible time for voracu&ng $ cf a reacter core to apent fuel etorage pool, naas 54 s

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"$ of Reactor Core" represents the plenning refuel' g scheille and amounts for the Clinton Station.

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74. Full-core off-load heat load calcuatica is not base en 30 dey operator after last nomal refueling, as rectested in*AEB(DSER)ItemNo.10 The heat load calculati is be. sed en shorter ti o period (134 hours0.00155 days <br />0.0372 hours <br />2.215608e-4 weeks <br />5.0987e-5 months <br />) between last

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nomal refuelfr.g pincement of fuel assemblies in pool cnd a reactor shutdom for full-core eft-lead.

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71. ' Decay heat Icads are based on full power cperation for 16,000 pours prier to removal of itel: assemblies from the * * ?

reactor, c

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CPS-FSAR

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d.

Monitor containment pool water level and maintain i n

shielding water level above the reactor vessel acad kl>

sufficient to hold the radiation level below acceptabla limits.

Maintain upper containment and fuel building pool c.

water temperatures at or below 120o F under normal operating conditions while removing the decay and drywell convective heat associated with the design heat load.

The temperature limit of 1200 F is set to establish an acceptable environment.for personnel working in the vicini~ty of the fuel pools and to allow a safety margin for the temperature sensitive

resins in the filter demineralizer.-

The design heat loads is based on the decay heat load from storage of an amount of spent fuel equal to approximately 217% of one reactor core and heat transferred from the operating. reactor to the upper containment pool water..

The decay heat load (19.4 x 106 Btu /hr) is defined as the decay heat assoo;ia e with the sum of:

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1.

21% of core, )llCs Hours af ter reactor shutdown with a reshlta'nt heat load of 8.0 x 10' Btu /hr; f--

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2.

21% of core stored in the spent fuel pool for 1

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year with a resultant heat load of 1.8 x 106 Btu /hr; 3.

23% of' core stored in the spent fuel pool for 2 years with a resultant heat load of 1.6 x 10' B&u/hr; 4.-

23% of core stored in the spent fuel pool for 3 years with a resultant heat load of 1.3 x 106 Btu /hr; 5.

24% of core stored in the spent fuel pool fot 4 years with a resultant heat load of 1.3 x 106 Btu /hr; 6.

24% of core stored in the spent fuel pool for 5 years with a resultant heat load of 1.3 x 10*

Btu /hr; 7..

23% of core stored in the spent fuel pool for 6 years with a resultant heat load of 1.2 x 10' Btu /hr; 8.-

26% of core stored in the spent fuel pool for 7

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years with a resultant heat load of 1.3.x 10'

, Btu /hr; and v.1-13

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i INSERT'a f.

Each FPC&C ' train (pump and h'ea't exchanger) has the heat dissipation c,apability of 19.7x106 BTU /hr under normal operating conditions.. The normal co'oling water median for the FPC&L heat exchangers is the component cooling system.

One FP&C train can remove the decay heat associated with 8 normal rdfueling cycles ~119'6% of a reactor core; 18.79 x 106 BTU /hr).

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Each FP&C train has the capability to employ the shittdown service wate system as a cooling water median in place of the component cooling water system.

Under-this abnormal system operation, each FPC&C train has the heat dissipation capability of 33.2 x 106 BTU /hr.

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32% of core stored -in_ th.e spent fuel pool for B years with a resultant. heat -load-o_f _1. 6 x 106 Btu /hr.

h, The decay heat loads are based on full power operation for 4'l16 00o houff

.yone. prior to removal of fuel assemblies from the reactor.

E The heat'lo'ad to. the containment upper pool due to heat conduction and convection from the reactor and through the drywell head is estimated to be 0.3 x 105 Btu /hr.

This-heat-1valdE-combinca witn une decay lieYt I~d:rdTtrsulLs Jez,ign4 cat 1cc.; "

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l-9.1.3.2 System Descriotion The integrated fuel pool cooling and cleanup system maintains the upper containment pool, the spent fuel storage pool, storage pool and the fuel transfer pools below a spe and the cask cified temperature, below a specified radioactivity concentration level, and at a degree of clarity necessary to transfer and service the reactor internals and fuel bundles.

Two 100% capacity system trains are employed for each unit.

One train supplies the design basis cooling capacity with cleanup at a rate of 1000 gpm.

The other train acts as a backup system and is employed when servici.ng of the first train is required,

($)

i.e.,

during backwashing, maintenance, etc.

Both trains can be run in parallel when a larger than normal heat load is produced in the

pool, e.g., from a nonequilibrium discharge batch.

The primary components in each train ars the heat exchanger,

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filter-demir.aralizer unit, pump, and the associated piping and valves.

A piping and_ instrumentation diagram of the system is provided in Figure 9.1-4.

The majority of the equipment is located in the fuel building.

The valvss, piping, and instrumentation associated with the containment pool are located c

in the containment building.

The valves, piping, instrumentation, and equipment associated with the filter-demineralizers are located in the radwaste building.

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The FPC&C system cools all of the pools'by transferring the heat through heat exchangers to the component cooling water (CCW) system.

Each train contains a heat exchanger designed to transfer the system design heat load of 19.7 x 106 Btu /hr The

-pool water temperature is maintained at or below 1200 F.

The decay heat released from the stored spent fuel.is transferred to the component cooling water system.

The shutdown service water '

system replaces the component cooling water system as a cooling water source when the latter is not available.

The residual heat removal system may be connected to the FPC&C system to remove

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decay heat from the pools under the follcwing conditions:

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p

CPS-FSAR AMENDMENT 6 AUGUST 1981 s

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, The filter-demineralizers are' controlled from a local panel.

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Differential pressure and conductivity instrumentation is oro-vided for each filter-demineralizer unit to indicate when 'back-l wash is required.

Suitable alarms for parameters such as differential pressure, conductivity and flow are used to monitor the condition of the filter-deminera,lizers.

FPC&C system' instrumentation is provided for both automatic and remote-manual operations.

In order to protect the pumps, a low-low level switch stops the FPC&C pumps when the fuel pool skimmer surge tank level is too low.

Level switches are also located

..,..... -in. both upper containment. and spent fuel storage pools.- When>m... c.

ever the water level is too low, alarm and indicator lights are I

. activated.

Temperature elements are provided and pool cooling and cleanup system instrumentation and. controls are described in Subsection 7.6.1.9.

The FPC&C pumps are controlled from the main control room.

Low l

pump suction pressure automatically turns off'the pumps.

A low pump discharge pressure alarm is indicated in the control room and on the local panel.

These pumps' receive power from a Class lE electrical system.

The water in the spent fuel storage pool is maintained at a level which is sufficient to provide shielding for normal building occupancy.

F.adioactive particulates and solubles are removed l ($)

from the fuel pool by the filter-demineralizer units which ara l

l located in shielded cubicles.

For these reasons, the exposure of station perscnnel to radiation from the spent fuel pool cooling and cleanup system is as low as is re'sonably achievable (ALARA).

a l

Details of radiological considerations for this system are described in Chapter 12.

Fuel storage pool water is circulated by means of overflow i

through skimmers and scuppers around 'the periphery of the pool.

l This overflow is collected in.the fuel pool ~ skimmer surge tanks, i-and the pool water is pumped through the filter-demineralizer and heat exchanger and back to the pool through diffusers.

The containment pool water is collected by skimmers and scuppers and flows either into the spent fuel pool or is sent directly to the surge tanks.

From there a closed loop circulation is attained through the same pump, heat exchanger.and filter-demineralizer train.

The circulation patterns within the containment pool and spent fuel storage, pool are established by locating the diffusers and skimmers so particles-dislodged during refueling operations are swept away from the work area and out of the pools.

In the event of pipe rupture, the return lines to the pools are prevented from siphoning the pool by siphon breakers located just j below the low water level of the pool. 9.1-16

CPS-FSAR Heat from pool evaporation is, handled by the building ventilation system. Makeup water from the cycle condensate system is (I provided the system through a remotely operated valve. l Suppression pool cleanup is r.ormally accomplished through the standby Unit 1 fuel pool filter-demineralizer. Under some conditions, three of the four FPC&C filter-demineralizers being installed with Unit 1 may be tied together for increased suppression pool cleanup. For further' details,~ refer to Subsection 9.3.6' .g pf - c. 9.1. 3..) Safety Evaluation '9 O;s y The uaximum possible heat load off.. Ax,206 Btu /hr is the decay heat of the full core load of fuel'a~~t the eseAof the fuel cycle L plus the remaining decay heat of the spent fuel discharged at 13 previous refuelings. The temperature of the fuel pool water may be permitted to rise to approximately 1500 F under these l conditions. If the above situation should arise, the operator l has the following modes of cooling operati'n available to satisfy o l the cooling requirements: l ~ a.' Each of the two 100% capacity trains of the FPC&C system are capable of, maintaining the temperature of both pools at or beloy 200 F w!(ile removing the decayheatassociatedlw'$; @ n heat load.. (]) These trains may be operated in parallel when a larger than design heat load is produced. Both-trains running in parallel have a heat dissipation # apability of 39.4 x 106 Btu /hr. ,j/ t{ f. _Lf-oMy-one-tca.Lfhe f uel pool water the-FPC-&G-syst-e.a is 5 v s i l i. b l e foc-operation-and l rises above 1200 the FPCsc heat exchanger cooling i . -media may be transferred by the operator from the component cooling water system to the shutdown l . service water system. The design inlet water temperature of the component cooling water system is l 1050 F while the shutdown s.ervice water system design temperature is 950 F. This' change in cooling media increases the heat dissipation capability of the one train from 19.7 x 106 Btu /hr to 33.2 x 10' Btu /hr. ZnSeff 0 h p[. If it appears that fuel pool water temperatube will-l exceed 1500 F when the reactor is in a cold shutdown condition, the operator may' connect the FPC&C system to the residual heat removal system. The reactor will not be started up should either the fuel pool water temperature be above 1500 F or whenever portions of the RHR system are needed to cool the fuel pool. The connecting piping m. from~the fuel storage pool to the RHR system.is designed Seismic W Category I. The RHR system is normally isolated from the FPC&C 9.1-17

-m. c, l INSERT B Cooling one FPC&C heat exchans;_.r with' shutdown service. water and the second FPC&C heat exchanger with component cooling water increases the~FPC&C system heat dissipation 6 BTU /hr. capability to'52.9 x 10 . t r 6 / e 8 g M - / / O f / i' / l, t f a. - - o 9 3 y w 5 e '1 e p J' ^ c f I f', f 4 i;, /p i o l-e ) 4 y ,~, \\ ~ , F i g y ,. i. N / _,./, i - / r / l \\ ~ r' / I f .,e f ' 179' \\ \\ -. '/ ~ r f \\ [f ,f / 4 /.- /', f, e '/ / J i y / ,/ (' d.

D i ):- h > yn.* l ~ Issue ~~ Section 9.1.4 - Fuel llandling System ~The.. applicant shoulil provide a commitment to implement the interim actions of NUREG-0612 and provide additional'information concerning s 1;o.ad. handling in coritainment and the. fuel handling building. 1., ~ Issue: 1. . Verify that the.atixiliary platform and the 5-ton fuel "3. building crane are designed to seismic Category I criteria. [ '.,

Response

~ 1. The transfer of new fuel assemblies between the un-4 crating area and the new fuel inspection stand and/or the new fuel storage'va61t is accomplished using a 10 ton hook on the 125 ton ~ crane which is seismic-a' Category I (FSAR Section 9.1.4.1' will be revised). The auxiliary platform is seismic Category I'(FSAR Section ~ Verify that the polar crane )is of singic failure proof 9.1.4.2.7.2 will be revised. Issue: 2. 5 designandconstruction as delineated in Branch Technical ^ position ASB 9-1.

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'y Response- .y.. . l, 2. The polar. crane is of single failure. proof. design'.as ~ required.by NUREG 0554 which superseded ASB 9-1.' 'cIP ? ' h,as completed.the final' response to NUREG 0612, " Control of Heavy Lo' ds in Nuclear Power Plant's," arid a submitted it to the NRC Sept. 25, 1981. This submittal - /" includes a detailed comparison of the requirements of NUREG 0554.to.the as-built condition of 3 the polar crane.- Issue: 3. Figure 9.1-22 does not show the same ~1ayout as facilitics that is shown in Figure 9.1-24.

Also, Figure 9.1-17 shows the.use of a 5-ton bui1 ding ' crane.

If thereais noLS-ton crane, this drawing is in error. . Correct these discrepancies.

Response

, _... ~ -3. Figure 9.1-22 -is incorrect l and.will.be ' deleted. Figure 9.1-17 willibe. revised to show-the 10 ton @, m hook on 125 ton _ crane.. y, .s ? .i

l l 4 l i .. +, Issue: 4. Concerning the cask drop over the railroad tracks, identify the safety-related systems serviced by the cabic-tray located in the compartment under the railroad tracks. Describe the affect on these systems resulting from a cask drop.

Response

4. The following systems are serviced by the cable-trays located in the compartment under the railroad tracks: 1. auxiliary power, 2. component cooling water, 3. fuel pool cooling and cicanup, f '^ 4. shutdown service water, 5. shutdown service water pump room HVAC. a Lelow are listed the components that would be lost for each of the above systems and the2nsult of this lost: System Component Result Auxiliary power Cable between ' Loss of the 480 V Substation 1B division 2 shut-(1AP12E)and shutdown down service water service water MCC IB system. (1AP30E). Component Cool-Valves: ICC075B Loss of flow-to ing ut?re ICC076B fue1~ pool cooling -and cleanup heat exchanger 1FC01AB. cue c<.ol cool-Valves: 1FC015B Loss of flow ^ to:: fuel' inn and cleanup 1FC023, IFC026B pool cooling and ~ 1FC016B, IFC024A cleanup heat ex-- IFC011B, IFC004B changer 1FC01AB. Instument: IPT-FC109 Shut dow.n-s ervice. Val.es:

1SX020B, Lossfof the division water 1SX012B, 1SX016B,_

2 shutdown service ISX004B, ISX003B, water system. .lSX014B, ISX011B, Pump: 1SX0lPB Instrument: 1PT-SX030' Shutdown service Cable connection floss of division water pump room-between SSWS pump 2 shutdown' service-RVAC room 1B panel 1(lPL53JB) water system. -and the : control room, cable connection between SSWS MCC lB ^ -(LAP 30E) and' control room

... o l I I In the event of the loss of the division 2 shutdown service water (SSWS) or shutdown service water pump room HVAC systems, the division 1 systems would still be available to perform their functions,,w ;wrar Twurus c.uwonoms w/o_ sc,wwww. The loss of the component cooling water system (CCWS), does not compromise any safety-related system or component, since the CCWS is not required to assure safe shutdown. The fuel pool cooling and cleanup (FPC&C) system would suffer the loss of the division 2 train. However, each train is capable of maintaining the temperature of both fuel po.as at or below 120 F while removing the decay heat ansociated with the design heat load. The FPC&C heat exchanger ccoling media may be transferred by the i operators from CCHS to SSUS, Div. 1. Therefore, the temperature of the fuel pool can be maintained within the specified limits by the division 1 train of the FPC&C system and the division 1 shutdown service water system. See Subsection 9.1.3.3. Issue 5: 1n regard to the handling of loads over spent fuel in containment and the fuel building, provide verification that the maximbm potential kinetic energy capable of being developed by all objects handled above the spent fuel racks, if dropped from the height at which it is normally handled above the storage rack, does not exceed the kinetic energy of one fuel assembly and its associated handling tool.

Response

There are no " heavy loads" handled over the fuel building spent fuel pool. Crane stops are used to keep the crane out of this area. Any loads handled over the containment fuel transfer pool are handled by either the polar crane, which is single failure proof, or the fuel handling platform, which in not considered to be a " heavy load" handling device. The fuel handling plat' forms in the fuel and containment buildink,s~are ~ - 6E** ^'"b_ iientified as single-failure proof in our response to NUREG-0612.- ~ 'MMi& 'C16 YN5%?@ Un?'?Ndi3CES'tiMn'."' ""* * **""' '*"N Issue 6:We require the applicant to implement the interim measures of to the December 22, 1980, generic letter prior to the final implementation of NUREG-0612 guidelines and prior to the receipt of their operating license.

Response

The interim measures of Enclosure 2 are being implemented where appropriataMeuc,utc - Action Required: 1. IP to revise FSAR EEctions 9.1.4.1 and 9.1.4.2.7.2 and Figure 9.1-17. 2. Fipure 9.1-22 will be d

m-.-. 't i Issue Section 4.6 - Function Design of Reactivity Control System i The applicant has not responded to our concerns relating to the CRD return line nozzle cracking (NUREG-0619) and the BWR scram discharge volume. ( i

Response

CPS complies with the requirements of UUREd-0619 and the required flow will be demonstrated by test. This response was given in the response to Question 410.5. CPS intends to modify-the scram discharge system to meet the criteria enumerated in the Generic Safety Evaluation Report BWR Scram Discharge System. It appears-that the additional equipment with appropriate qualifications probably will j not be available-prior to fuel load. t +. i O + Action Required 1. IP to confirm compliance with NUREG-0619'duringLpreoperational' test. 2. IP to' modify the scram.dischargeisystem to meet the: established--' criteria prior 1to the endJof the'first refueling. 4

? iI Issue ~ ~ l Section 9.2.4 - Potable and Sanitary Water Systems . Process and instrumentation diagrams of the potable and sanitary water system are needed. Provide process and instru-mentation diagrams for the potable and sanitary water systems which demonstrate that there are no connections.to systens hav-ing a potential.for containing radioactive material.

Response

The following'are the required diagrams: A21-1044 Rev. G. A21-1045 Rev. D. A21-1121 Rev. E. A21-1122 Rev. D. A03-1247 Rev. H. A03-1245 Rev. M. A03-1243 Rev. M. A03-1242 Rev. T. M05-1061 Rev. D. Copies of these drawings have been supplied to the NRC - during the week of November 9, 1981. Action Required None

+ Issue ~ Section 9.3.1 - Compressed Air System . Additional information concerning instrument air quality is needed. Also the applicant should correct certain inconsistencies in the FSAR. _ Response 1. There are 12-207. air rimplifiers per unit. Figure 9.3-2, sheet 5' and Table 9.3-2 wili be revised. 2. Table 9.3-2 will be. revised to state the 3 service air compressors are shared by the two units. 3. Compressed air leaving the service air compressor filter dryers will be tested periodically (at least yearly) for dew point and particulate contamination. The maximum acceptable dew point will be -40*F G115 psig (-72*F G Atmos)' Particulate contamination will be determined using a test procedure simulon to ANSI / ASTM D2009-65. -Maximum acceptable particle size will be 3 microns-(3 x'10-6in) in diameter. Branch lines will be checked upon failure to meet - sthndards at filter - discharge or u'pon indication of trouble in any of the lines,un. Amurce,wu ruo<cartt ocozecnvp Acrio/fjema ce tam #, Action Required IP revise FSAR Tabic 9.3-2. i 4 L

-c W.ng, G E a CPS-FSAR / A q

d. A A 2.L3 U tg1 ~.,

TABLE 9.3-2 i f( N-/d-fj /). COMPRESSED AIR SYSTEM FAILURE AtIALYSIS j COMPONENT FAILURE CCp1MENTS

- dic< red Service Air Loss of Three service air compressors Compressor Compressor a r efG r o. i d e d-for--sach u r. i t, Ly fh

the third acts as standby .~ for the first two. Additional protection is provided by utilizing air receiver tanks and equipping nuclear safety- 'related air component systems .with individual accumulators. The air accumulators are of sufficient capacity to safely perform their shutdown function r in the event that the air supply system is lost. ~ Q -207o Instrument Loss of F6 u r, M' capacity instrument Air Amplifier Compressor air amplifiers are provided 150 psig, for each unit. The second set ADS Valve acts as standby for the first M ' Operation set. 'Also, storage bottles are 4 provided and each ADS valve has.its own air accumulator with sufficient capacity to safely shutdown the unit. D y = ' .= a t 6 i e G 9.3-26

_ / of

• ~

~ [ s..> n Issue .Schtion9.4.2 Spent F6el Pool . Area Ventilation System -. ~ - ,e.11 . Verify that the spent fuel pool area ventilation'sys' tem w:li-

.g..,

serves the same areas of secondary containment that are - t s served by-the. standby gas treatment system. [;f' ,.,-l.j{I. :, List ali of.the areas served by'the spent fuel pool area -

ventilation system an4,.show them on the P & ids.

~. 3. We cannot determine from the drawings (Figs. 1.2-3 and ~ following) where the division between the fuel building e and auxiliary building is located. Provide drawings that clarify this division. 4. Describe the means provided to assrre that'the temperature in th.e rooms housing the spent fuel pool cooling pumps can be maintained at acceptable lqvels for pump operation' - under accident and emergency conditions when the normal fuel building IIVAC system is not in operation.- Respdnse. q,.},F(- l1&2;. The Fuel Building ~llVAC system serves.the same areas.o'f I the secondary containment that are :s'erved-byJthe stahd-by gas treatment system. The areas, served by.the Fuel-,. -lC : Building IIVAC are shown on. Figure -9.4-2. The"hreas f. served by the'SGTS'arci.shown on Figure 6.5-3.. 3 The division between1the fuel building and the auxiliary - building is column row AD. The secondary containment. boundary. is' shown 'on ' Figure' 6. 2-132.- 4. 'These pump' motors are classified ac Non-Class 11E;- they ' E .are. not required for safe shutdown o.f ' theTplant. lTherefore, no safety-related source'of cooling islprovided. ' Action Required " ~- .None~ s. - _ g

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'Section 9.4.3 - Auxiliary and Radwaste Area Ventilation System cThe, applicant should verify. that the auxiliary llVAC system does- .not. service any area considered' secondary containment., < j;; .:)). - + J q

e :,c....

i .s

Response

The Auxiliary. Building IIVAC system does not serve any area that would.be considered secot.dary containment, i e;, any area. served by the SGTS. The areas served'by the. Auxiliary Build ' ing IIVAC are shown on Figure 9.4.3. The areas served by the SGTS are shown in Figure 6.5-3. s .s. -:e r t ~> ~. s ..,,.1 Action Required- \\None; . g t + s s 'I / G " 6 e ..E '. 3.. s

i-I Issue Section.10.3-Main Steam Supply System Additional information concerning the design of the main steam shutoff (block) valves is needed. In the FSAR description of the-matn steam support system, no mention is made of the main steam-shutoff'(block) valves. Describe the functional design and operation of these valves during normal, accident and transient conditions, and identify their power sources. 'Itesponse-1: l The. main steam shutoff (block) valves are located down-1 stream of the outboard' main steam isolation valves in the auxi-liary building steam tunnel. The main steam shutoff valves are leak-tight, motor-operated gate valvec powered by separate class 1E sources. They are normally open and are uanually actuated from the main control room when additional assurance. of steam line isolation is required. I Action Required IP: revise'FSAR Artic1c.10 3.2 to. include above information.

I .s 4 Issue Section 10.4.7 - Condensate and Feedwater System Additi.onal information concerning the design of the main feedwater block valves is needed. f I

Response

i The feeduater block valves are located upstream of the feedwater isolation check valves in the auxiliary building steam tunnel. The feedwater block valves are leak tight motor operated gate valves powered by separate Class 1E They are open during normal operation and are sources. manually activated from the main control room when additional assurance of feedwater line isolation is required and prior to startup when the feedwater is being recycled to t-he main condenser. 8 i Action Required None t_

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Issue Shared Systems There is concern with the delay in construction of the second unit and the impact on. shared systems. The applicant should identify those portions of the shared systems which will not be completed prior to receipt of an operating license. ~

Response

All portions of systems which were to have been shared by Units #1 and 2 will be completed.for use by Unit #1, prior to fuel loading. I i Action Required 6 6

i i ev,, Grand Gulf Question 211.13 In Section 3.5.1.2.2, you state that pressurized components, (3.5.1.2) namelydpressure vessels and pressurized bottles containing noncondensible gases with an operating pressure at or above 100 psig have been evaluated as potential sources of missiles. Provide this evaluation which should include the CRD scram nitrogen bottles and the safety-relief valve air accumulators acting as potential missiles. Describe protective barriers available to prevent missiles from striking nearby safety systems or components. Clinton Response: The supports for the air accumulator tanks supported at EL 781'.0" from the drywell wall have been evaluated for dynamic loads plus a postulated jet load of 1000 pounds caused by rupture of the vessel. If the drain valve of an accumulator would become a missile, the small amount of energy would be absorbed by the structural grating below it. l The ADS backup air bottles located in the basement of the Auxiliary building and the control room emergency breathing i air bottles located at El 800 in the control building are l supported by interconnected structural framing systems designed to maintain structural integrity during dynamic events. In both areas, the bottles are isolated from other essential equipment. Each HCU's nitrogen bottle is secured to the HCU main frame, which is mounted to the floor and attached to the seismic l i beam support. A pressurized HCU was dynamically tested i with loads simulating the worst case faulted condition. The HCU function was demonstrated with no support system failure. The design burst pressure is more than 5 times the maximum operating gas pressure, which makes failure improbable. .}}