NSD-NRC-98-5630, Forwards W Responses to FSER Open Items on AP600

Forwards W Responses to FSER Open Items on AP600

ML20217D440
Person / Time
Site:	05200003
Issue date:	03/20/1998
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-98-5630, NUDOCS 9803300020
Download: ML20217D440 (27)

Text

.,

e

q -

s NU bing$ouse Energy Systems -

B5 355-Pittsburg1 Pennsylvania 15230-0355 Electric Corporation

. DCP/NRCl309 NSD-NRC-98-5630 Docket No.: 52-003 March 20,1998 j

~ Document Control Desk

- U.S. Nuclear Regulatory Commission

' Washington, DC 20555 ATTENTION: T. R. QUAY

[

SUBJECT:

AP600 RESPONSE TO FSER OPEN ITEMS

Dear Mr. Quay:

Enclosed with this letter are the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1. Included in the table is the FSER open item number, the associated OITS number, and the status that will be designated in the Westinghouse status column of OITS.

The NRC should review the enclosures and inform Westinghouse of the status to be designated in the "NRC Status" column of OITS.

. Please contact me on (412) 374-4334 when you receive this package to confirm receipt.

bg /f g/

13rian A.~ McIntyre, na Advanced Plant Safety and Licensing

~jml Enclosures cc:

W. C. Iluffman, NRC (Enclosures)

/

J. E. Lyons, NRC (Enclosures)

T. J. Kenyon. NRC (Enclosures) 6/

J. M. Sebrosky, NRC (Enclosures)

D. C. Scaletti, NRC (Enclosures).

N.lJ. Liparulo, Westinghouse (w/o Enclosures) 2:

N

>:9903300020 990320 iPDR.ADOCK 05200003 E.

PDRj m =

b' N

DCP/NRCl309 PfSD-N8C-98-5630 March 20,1998 i

Table 1 List of FSER Open Items Included in Letter DCP/NRC1309 FSER Open Item OITS Number Westinghouse status in OITS 480.1159F (RI) 6583 Confirm W 720.423F (R3) 6135 Confirm W 720.466F 6672 Action N 720.467F 6673 Action N n.

M71m wpf

l i

Enclosure to Westinghouse Letter DCP/NRCl309 March 20,1998 I

l i

i I

i.n..g i

1

\\

l e

)

NRC FSER OPEN ITEM Question: 480.lI59F (OITS #6583)

A technical specification equivalent to standard technical specification STS 3.9.4 specifying the status of containment penetration requirements during core alterations and movement ofirradiated fuel assemblies within containment should be provided (similar to current Ar ')0 TS 3.6.8 with applicability during core alterations).

Response (Revision 1):

WithouHaking-eredit-fer Occ::i.90 t cle:Ute the-AP400-site Ecund=j de:: !imits-aree::, fer events r

oeeurring-during cer: :!aer::! n: er =evement-of-trradiated-fu:!. Th:::fere, STS LCO 2.9 ' !: nc:

applicab!: : AP400-This-differ:::: !: 5:::d en imprevements-in-analy:i: = :hede!c;y.

l The Standard Technical Specifications include a technical specification for containment penetrations

} during core alterations. The BASES of this specification explains that the applicable safety analysis I that requires this specification is for fuel handling accidents. Many operating plants then assume that l the containment is closed when performing the offsite dose analyses for this event. However, the l AP600 fuel handling accident offsite dose analyses do not require containment closure to meet the I specified limits. Therefore, application of the accepted technical specification screening criteria does

! not capture containment closure during fuel movement, and no corresponding specification was

! incorporated.

i Nesertheless, the NRC staff has requested that Westinghouse include a technical speficication for containment during core alterations. Per this request, Westmghouse has added Technical Specification 3.9.5 that specifies the status of containment during core alterations and movement of irradiated fuel assemblies within containment. The technical specification is based on STS 3.9.4 with some modifications based on AP600 differences.

The proposed status of the containment openings during core alterations and movement ofirradiated i

i fuel is different from STS 3.9.4 as discussed below-Equipment Hatches l

STS 3 9.4 requires the containment equipment hatches to be closed and held in place by four bolts. In j

the AP600 technical specification, the equipment hatch must be closed, or if open, an alternative j

barner between the fuel and the outside atmosphere shall be in place or capable of being established withm 15 minutes. The alternate barrier may consist of the staging area in the auxiliary building outside the equipment hatches provided that the doors in the staging area are closed to the annex j

building or the outside atmosphere. The alternate barrier may also consist of a temporary equipment hatch cover. In addition, if the containment equipment hatch is open, the containment air filtration system (VFS) shall be operable to provide for monitoring of air-borne radioactivity and filtration of containment atmosphere..

i i

wg 480.1159F-1 Resision 1

==

i i

)

1 i

l NRC FSER OPEN ITEM I

Justification:

i i As discussed above, the AP600 offsite dose analysis does not assume the containment equipment l hatches are closed, and therefore the technical specification screening criteria are not met.

! Funhermore, unlike many current plants, the containment equipment hatches do not open directly to 1 the outside atmosphere, but open into a staging area in the auxiliary building. The alternate boundary i prevents the egress of airborne fission products from being released to the atmosphere if the

! equipment hatches were open during a fuel handling accident. The operation of the VFS provides for

! monitoring of radioactivity releases from containment following a fuel handling accident.

l l Other Penetrations i

STS 3.9.4 requires one door in each airlock closed. In the AP600 technical specification, one door in

! each air-lock is closed or,if open, the VFS is operable to provide a filtered exhaust of the containment

! atmosphere. AP600 technical specification 3.9.5 also stipulates that the VFS is operable if the i temporary penetrations are open.

I l Justification:

l l In addition to the justification abose, the operability of the VFS will reduce the consequences of a fuel

! handling accident by filtering the containment atmosphere. These penetrations open into the auxiliary buildicg. Upon detection of high radiation, these portions of the auxiliary building are serviced by the VFS which will provide containment air filtration of these areas. Operation of the VFS in combination with the auxiliary building enclosure provide a sufficient alternate barner to the release of radioactivity directly to the environment.

AP600 technical specification 3.9.5 is attached. In addition to the above justifications provided for the deviation from the Standard Technical Specifications. the AP600 has made many improvements in shutdown safety. The AP600 passive safety systems are operable during all modes of operation including shutdown. Unlike current and evolutionary plants, technical specifications for the AP600 l

safety systems have been included for all shutdown modes. Passive safety systems designed to operate during shutdown are major contributors to the large reduction in AP600 shutdown nsk. The AP600 shutdown probabilistic risk assessment calculates a core damage frequency from internal events at shutdown of 9E.8 which is orders of magnitude smaller than current plants.

SSAR Revision:

Technical Specification 3.9.5 480.1159F-2 Revision 1 3 Westinghouse

t-Containment Penetrations 3.9.5

-:- 3. 9 : REFUELING OPERATIONS 3;9.5 Containment Penetrations LCO 3.9.5 The containment penetrations shall be in the~following status:

The equipment hatches closed and held in place by (four]

bolts or, if open, an alternative barrier between the fuel and the outside atmosphere shall be inplace or capable of-being established within [15' minutes];

---

Notes --------------------------

1.

The alternate barrier may consist of the staging area in the auxiliary building outside the equipment hatches provided that the-doors in the staging area are closed to the annex building and the-outside atmosphere.

'The alternate barrier may also consist of a temporary equipment hatch cover.

2.

If a containment equipment hatch, both doors in a personnel air lock, or spare penetration is open, the containment air filtration system (VFS) shall be operating to provide for monitoring of air-borne radioactivity in the containment.

APPLICABILITY: During CORE ALTERATIONS During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION TIME i

A.

One or more A.1 Suspend CORE Immediately containment ALTERATIONS.

penetrations not in required status.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

'/80.Hs'9Ffi ) -3 tI AP600 3.9-1 DRAFT March 20, 1998 4

o.

Containment Penetrations 3.9.5

, SURVEILLANCE REQ IREMENTS SURVEILLANCE FREQUENCY' SR 3.9.5.1 Verify each required containment 7 days penetration is in the required status.

3

..)-'

?

e

i _

~

• i y80,lli9ffRI ~Y AP600'

3. 9-2 DRAFT March 20, 1998

);.

Containment Penetrations B 3.9.5 B 3.9 REFUELING OPERATIONS B'3.9.5 Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met.

In MODES 1, 2,

3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, " Containment."

In MODES 5 and 6, this is accomplished by providing closure capability of containment penetrations as described in LCO 3.6.8,

" Containment."

In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent.

The LCO requirements are referred to as " containment closure" rather than " containment OPERABILITY."

Containment closure means that all potential escape paths are closed or capable of being closed.

Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product I

radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained within the requirements of 10 CFR 50.34.

For a fuel handling accident, the AP600 dose analyses does not rely on containment closure to meet the offsite radiation exposure limits, This LCO is provided as an additional level of defense against the possibility of a fission product release from a fuel handling accident during CORE ALTERATIONS.

The containment equipment hatches, which are part of the containment pressure boundary, provide a means for moving large equipment and components into and out of containment.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, an equipment hatch is considered closed if the hatch cover is held in place by at least (four) bolts.

Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

]

If the equipment hatch is open, an alternative barrier between the fuel and the outside atmosphere shall be in place or capable of being established within (15 minutes).

The equipment hatches open into a staging area in the auxiliary building.

These staging areas contain doors that open to the outside atmosphere and the annex building.

The alternate barrier may consist of the (continued)

&O.ll 59F(KO -f AP600 B 3.9-1 DRAFT March 20, 1998

l Containment Penetrations B 3.9.5 BASES BACKGROUND staging area in the auxiliary building outside the (continued) equipment hatches provided that the doors in the staging area are closed to the annex building and the outside atmosphere.

The alternate barrier may also consist of a temporary equipment hatch cover that provides equivalent isolation capability.

The alternate boundary prevents the egress of airborne fission products from being readily released to the atmosphere if the equipment hatches were open during a fuel handling accident.

If an equipment hatch is open during CORE ALTERATIONS, the containment air filtration system (VFS) shall be operable to provide for monitoring of air-borne radioactivity.

This system services the containment, and upon detection of high radiation, also services the fuel area and the auxiliary building staging areas.

The operation of the VFS provides for monitoring of radioactivity releases from containment following a fuel handling accident and, if operating, will provide filtration of the containment atmosphere.

If two doors in a personnel air lock, or spare containment penetration are open during CORE ALTERATIONS, then the containment air filtration system (VFS) shall be operable to monitor for the release of radioactivity and to provide a filtration of the air inside containment to j

minimize the consequences of a fuel handling accident.

These penetrations open into the auxiliary building.

i Upon detection of high radiation, these portions of the auxiliary building are serviced by the VFS which will provide containment air filtration of these areas.

Considering that these penetrate open into the auxiliary building and not directly to the atmosphere, and the VFS j

is in operation, a sufficient alternate barrier to the release of radioactivity directly to the environment is provided.

1 (continued)

//$0o//5')FfNIT(o AP600 B 3.9-2 DRAFT March 20, 1998

Containment Penetrations B 3.9.5

. BASES, APPLICABLE For the AP600, there are no safety analyses that SAFETY ANALYSES require containment closure during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, other than those discussed in LCO 3.6.8.

Fuel handling accidents,. analyzed in Reference 1, include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated-fuel assemblies.

The requirements of LCO 3.9.4,'

" Refueling Cavity Water Level,"

release of fission product radi ensure that theoactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 50.34.

Standard Review Plan, Section 15.7.4, LRev. 1 (Reference-2), defines "well within" to be 25%

or less-of the limiting dose guideline values.

fLCO' This LCO.provides defense-in-depth against the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment.

This LCO requires the containment equipment hatches to be closed, or if open, requires an alternate barrier be established.

This LCO requires that if an equipment hatch, personnel air lock, or spare containment penetration is open during CORE ALTERATIONS, then the containment air filtration system (VFS) shall be in operation to monitor for the release of radioactivity and to provide a filtration of the air inside containment to minimize the consequences of a fuel handling accident.

(continued)~

NBO.IlTW(NO ~l

.AP600 B 3.9-3 DRAFT March 120, 1998

Containment Penetrations B 3.9.5 BASES APPLICABILITY The containment penetration requirements are applicable during COPE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident.

In MODES 1, 2, 3,

and 4, containment penetration requirements are addressed by LCO 3.6.1.

In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist.

Containment closure capability in Modes 5 and 6 are addressed by LCO 3.6.8.

ACTIONS A.1 and A.2 If the containment equipment hatch or air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, the unit must be placed in a condition where the isolation function is not needed.

This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.

Performance of these actions shall not preclude complet!.on of movement of a component to a safe position.

SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position.

The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing.

Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment purge and exhaust isolation signal.

The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment.

The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations.

A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO.

(continued)

V60 p S W {R b - 8 AP600 B 3.9-4 DRAFT March 20, 1998

Containment Penetration B 3.9.5 BASES (continued)

REEIRENCES 1.

AP600 SSAR, Section 15.7.4 2.

NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

I I

1 i

i j

i a

9'30. n S9F(RI) - 9 AP600 B 3.9-5 DRAFT March 20, 1998

W..

i NRC FSER OPEN ITEM T

Question 720.423F (OITS - 6135)

In meeting the RTNSS criteria, credit was taken for external reactor vessel cooling (ERVC) as a strategy for retaining molten core debris in-vessel. This results in the majority of core melt accidents

(~90 percent) being arrested in-vessel, thereby avoiding RPV failure and associated containment challenges from ex vessel phenomena. Successful RCS depressurization and reactor cavity flooding are prerequisites for ERVC, and credit for these aspects of ERVC in the focussed PRA is appropriate since both functions are fulfilled by safety-related systems. However, the nonsafety-related RPV thermal insulation system is also required for successful ERVC. The thermal insulation system limits thermal losses during normal operations, but provides an engineered pathway for supplying water cooling to the vessel and venting steam from the reactor cavity during severe accidents. Attributes of the system include specific RPV/ insulation clearances and water / steam flow areas based on scaled tests, integral ball-and-cage check valves and buoyant steam vent dampers which change position during flood-up of the reactor cavity, and insulation panel and support members designed to withstand the hydrodynamic loads associated with ERVC.

If credit for ERVC is reduced, large release frequency and CCFP would increase proportionally since all RPV breaches are assumed to lead to early containment failure in the PRA. Under the most limiting assumption of no credit for ERVC, the large release frequency would approach the core melt frequency and CCFP would approach 1.0. In view of the reliance on ERVC to meet the Commission's large release frequency goals, the stalT will require an appropriate level of regulatory oversight of the RPV thermal insulation system. This oversight should provide reasonable assurance that the as-built j

insulation system conforms with design specifications contained in Chapter 39 of the PRA, and that the operability of the system is confirmed through periodic surveillance.

The RPV insulation design description and functional requirements are not currently included in the SSAR, ITAAC, or reliability assurance program. The design description and functional requirements for the RPV insulation should be added to the SSAR, and important criteria associated with the insulation design should be incorporated into the ITAAC, including information related to the j

necessary clearances / flow areas, and the check valves and steam vent dampers. The system should be included as a risk significant SSC in the reliability assurance program, and reliability / availability controls and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained.

Westinghouse Response (Revision 3):

Functional requirements for the reactor vessel insulation was incorporated in section 5.3.5 of Revision 14 of the AP600 SSAR. Based on discussions with the NRC staff, more information was requested to be included in the SSAR. SSAR Section 5.3.5 is modified to include the design bases and design description for the reactor vessel insulation and is attached. In addition, the reactor vessel insulation is included as a risk-significant SSC in the reliability assurar.ce program as shown in the proposed resision to SSAR section 17.4. The AP600 Certified Design Mav. rial is also revised to include appropriate ITAACs for the reactor sessel insulation per the response to RAI 720.442F.

720.423F-1 g WGStingh0USS Revision 3

==

NRC FSER OPEN ITEM

! SSAR Section 53.5 is further revised based on discussions held with the staff at the January 22,1998

! chapter closcout meeting and subsequent discussions. Design detail of the doorway between the

! RCDT room and the reactor cavity has been provided to show that the minimum flow area between

! these compartments is 8 ft'. Scoping evaluations of the damper and opening indicate that the i resistance through this opening, assuming a typical damper design, is such that it provides an acceptable pressure drop through the opening during ex vessel cooling.

SSAR Revision:

Revised SSAR Sections 53.5, 17.4 attached. See the response to RAI 720.442F for changes to the Certified Design Material.

i l

720.423F-2 Revision 3 3 Westirighouse

\\

1 i

y l!

)

5. Rtactrr Calant Systim cnd Cannictrd Sys Ims j

1 i

$.3.5 Reactor Vessel Insulation 5.3.5.1 Reactor Vessel Insulation Design Bases i

Reactor vessel insulation is provided to minimize heat losses from the primary system.

Nonsafety re:ated reflective insulation similar to that in use in current pressurized water reactors is utilized. The AP600 reactor vessel insulation contains design features to promote in-vessel retention following severe accidents. In the unlikely event of a beyond design basis accident, the reactor cavity is flooded with water, and the reactor vessel insulation allows heat removal from core debris via boiling on the outside surface of the reactor vessel. The reactor vessel insulation permits a water layer next to the reactor vessel to promote heat transfer from the reactor vessel This is accomplished by providing:

A means of allowing water free access to the region between the reactor vessel and insulation.

A means to allow steam generated by water contact with the reactor vessel to escape from the region surrounding the reactor vessel.

A support frame to prevent the insulation panels from breaking free and blocking water from cooling the reactor vessel exterior surface.

The reactor vessel insulation and its supports are designed to withstand bounding pressure differentials across the reactor vessel insulation panels during the period that the reactor vessel

)

is externally flooded with water and the core retained in the reactor vessel through heat j

removal from the vessel wall accomplished by the water. This is accomplished by providing a minimum flow area of 7.5 ft in the portions of the flow paths required to vent steam. The j

flow path from the reactor loop compartment to the reactor cavity provides an open flow path for water to flood the reactor cavity The reactor vessel insulation inlet assembles are designed to minimize the pressure drop dunng ex-vessel cooling to permit water to cool the l

vessel.

5.3.5.2 Description of Insulation A schematic of the reactor vessel, the vessel insulation and the reactor cavity is shown in Figure 5.3 7. The insulation is mounted on a structural frame that is supported from the wall of the reactor cavity. The vertical insulation panels are designed to have a mmimum gap between the insulation and reactor vessel not less than 2 inches when subjected to the dynamic loads in the direction towards the sessel that result during ex vessel cooling. A nominal gap (with no deflection) of more than twice the minimum gap is provided, The conical design of the bottom portion of the vessel insulation is constructed of flat panels.

l This provides a single point of contact with the spherical portion of the vessel in the event j

that an insulation panel becomes dislodged. This prevents hot spots from developing on the reactor vessel and permits sufficient flow to maintain in vessel retention. The nominal gap 9 20.GH(k3)

,3 Draft Revm..on: 21 February 13,1998 5.3-20 W W85tingh0US8

1 s

.&W~O

?ee

~

5. Rsccesr Csol:nt Systrm exd Carnictzd Syst2ms between the conical portion of the insulation and the spherical portion of the reactor vessel is not less than 9 inches.

The structural frame supporting the insulation is designed to withstand the bounding severe accident loads without exceeding denection criteria. The fasteners holding the insulation panels to the frame are also designed for these loads.

At the bottom of the insulation are water inlet assemblies. Each water inlet assembly is normally closed to prevent an air circulation path through the vessel insulation. The inlet assemblies are self actuating passive devices. The inlet assemblies open when the cavity is filled with water. This permits ingress of water during a severe accident, while preventing excessive heat loss during normal operation.

The total Cow area of the water inlet assemblics have sufficient snargin to preclude significant pressure drop during ex-vessel cooling during a severe accident. The minimum total Dow area for the water inlets assemblies is 6 ft'. Due to the relatively low approach velocities in the Dow paths leading to the reactor cavity, and due to the relatively large minimum now area i

2 through each water inlet assembly, with an area of at least 7 in, the water inlet assemblies l

Fe not susceptible to clogging from debris inside containment. This 7 in' minimum area is l

also provided in the steam vent path to minimi:e the potentialfor clogging the steam flow l

path.

Near the top of the lower insulation segment are steam vent dampers. These dampers are norrnally closed to prevent reactor vessel heat hss, and a small buildup of steam pressure under the insulation will cause them to open to the vent position. The steam vent dampers are passive, self-actuated devices and will operate when steam is generated under the insulation with the cavity filled with water.

Extensive maintenance of the vesselinsulation is not normally required. Periodic venfication that the vessel insulation moving parts can be performed during refueling outages.

5.3.5.3 Description of External Vessel Cooling Flooded Compartments Exocssel cooling during a severe accident is provided by Gooding the reactor coolant system loop compartment including a s ertical access tunnel, the reactor coolant drain tank room, and the reactor cavity. Water from these compartments replenishes the water that comes in contact with the reactor vessel and is boiled and vented to containment. The opening between the vertical access tunnel and the reactor coolant drain tank room is approximately 100 ft:

l Removable steel gratmg us provided over the mlet to the vertical access tunnel to restrict I

access to the lower compartments.

This gratmg precludes large debris from bemg I

transported mio the reactor cavsty durmg ex-vessel coolmg scenarios. Figure 5.3-8 depicts the Gooded compartments that provide the water for ex sessel cooling. The ep=:ng bc: ween the :=c c cce!=: d::ir :=h :cc

=d 'he :=c::: c=t iS-f = r c eb:: ue:!ca = h 'h :

iHic= c: p:=!:d: : Hr.uc Oc e : = M-4-f4* :c pc. :: u:= :c Occd :h: :c=:c ca :,

l cc =q =* =. = : The doorway between the reactor cavity compartment and the reactor coolant i

Draft Revision: 21 W W85tingh00$e 5.3-21 February 13,1998 720,423F(R 3) - Y

T

.iin dSi I

5. Rrcetir Calait Systim md Cemetid Syst1ms i

e l

l drain tank room consists of a normally closed doer and a damper above the door. The door l

and damper arrangement, shown in Figure 5.3-9, maintains the proper airpow through the l

reactor cavity during normal operation. The damper prevents air from powing into the l

reactor coolant drain tank compartment, but opens to permitfooding of the reactor cavity I

from the reactor ccolant drain tank compartment. The damper opening has an minimumpow l

area of 8ft' and is not susceptible to cloggingfrom debris that can pass through the grating l

over the inlet to the vertical access tunnel. It is constructed oflight-weight material to l

minimi:e theforce necessary to open the damper andpermitfooding and continued water l

pow through the opening during ex-vessel cooling. The damper provides an acceptable l

pressure drop through the opening during ex-vessel cooling.

5.3.5.4 Determination of Forces on Insulation and Support System The expected forces that may be expected in the reactor cavity region of the AP600 plant during a core damage accident in which the core has relocated to the lower head and the reactor cavity is reflooded have been conservatively established based on data from the ULPU test program (Reference 5). The particular configuration (Configuration III) reviewed closely models the full-scale AP600 geometry of water in the region near the reactor vessel, between the reactor vessel and the reactor vessel insulation. The ULPU tests provide data on the pressure generated in the region between the reactor vessel and reactor vessel insulation.

These data, along with observations and conclusions from heat transfer studies, are used to develop the functional requirements with respect to in-vessel retention for the reactor vessel insulation and support system. Interpretation of data collected from ULPU Configuration 111 experiments in conjunction with the static head of water that would be present in the AP600 is used to estimate forces acting on the rigid sections ofinsulation. Further evaluation of the forces on the reactor vessel insulation and supports is provided in the AP600 Probabilistic Risk Assessment.

5.3.5.5 Design Evaluation A structural analysis of the AP600 reactor cavity insulation system demonstrates that it meets the functional requirements discussed above. The analysis encompassed the insulation and support system and included a determination of the stresses in support members, bolts, insulation panels and welds, as well as deflection of support members and insulation panels.

The results of the analyses show that the insulation is able to meet its functional requirements.

The reactor vessel insulation prosides an engineered pathway for water-cooling the vessel and for venting steam from the reactor cavity The reactor vessel insulation is purchased equipment. The purchase specification for the I

reactor vessel insulation will require confirmatory static load analyses.

7ac.

23FWJ)6 Draft Resision: 21 February 13.1998 5 3-22 W Westinghouse

=

i

1

5. RI:ctrr Culs Syst;m and Cann1ct;d Syst;ms 5.3.6 Combined License Informntion 5.3.6.1 Pressure-Temperature Limit Curves The pressure-temp. curves shown in Figures 53 2 and $3-3 are generic curves for AP600 reactor vessel design, and they are the limiting curves based on copper and nickel material composition. However, for a specific AP600, these curves will be plotted based on material composition of copper and nickel. Use of plant-specific curves will be addressed by the Combined License applicant during procurement of the reactor vessel.

5.3.6.2 Reactor Vessel Materials Surveillance Program The Combined License applicant will address a reactor vessel reactor material surveillance program based on subsection 53.2.6.

5.3.6.3 Reactor Vessel Materials Properties Verification The Combined License applicant will address verification of plant-specific belt line material properties consistent with the requirements in subsection 533.1 and Tables 5.3-1 and 53-3.

5.3.6.4 Reactor Vessel Insulation The Combined License applicant will address verification that the reactor vessel insulation is consistent with the design bases established for in vessel retention.

5.3.7 References 1.

ASTM E-185 82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."

2.

Soltesi, R G., et al.," Nuclear Rocket Shielding Methods, Modification, Updating, and input Data Preparation. Volume 5 -Two Dimensional Discrete Ordinates Techniques,"

WANL-PR-(LL).034, August 1970.

3.

S AILOR RSIC Data Library Collection DLC-76," Coupled, Self Shielded.47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors."

4.

NRC Policy issue "Pressunzed Thermal Shock " SECY-82-465, November 23.1982.

5 Theofanous, T.G., et al, "In-Vessel Coolability and Retention of a Core Melt."

DOE /ID-10460, July 1995 Draft Revision: 21 W Westinghouse

$3-23 February 13,1998 3.R 0 N.R 3F(R 3) -lo

.N

5. React:r Coolant System aid Connected Systems il.

't REACTOR C

VESSEL

(.d

$~EAM VENT U) j\\

i

/

i I

5hIELD WALL l

. CORE i

INSUL ATION (2)

I k

W ATER INLET ASSEMBLIES (3) c (1) Minimum staan veet flow area provided in subsecuon 5 3 5.1 (2). Minimum gap betwesa imulauon and sessel insulanon provided in subseccon 5.3.5.2 (3). Mintmum f'ow area provided in subsecuen 5.3.5.2 Figure 5.3 7 Schemade of Reactor Vessel Insulation g

i Revision: 19 December 31,1997 5.3 36 3 Westinghouse w j

e 3

e

• 7 0

i

/

\\

<< ron FLo00E0 ac1 Cow.Af4mf

,eb

\\

0?".r_

r,-

5

_g._

1 s-5,c ::w aar4 =r %\\

20,

,,20

. car:C.6.ecass j

i N

,,,0

,J

=ar -

,,,0, s staCron v(55(L Cavite

, /* C h dC I'}(,;.3 T..M \\

i o-f(-

i I'y es n.a-s v..e..s..A u.3,.w M 7 g.g.,

s I

E G

j

71 4

y/

s T

/

N 4

/ N 0

/

E 1

N/

1

.M

\\

1

=

T M

N W

O R

O I

I l

l l

c e

A.

R v

P o.

M no O

nr C

cn i

1 Y

no T

c I

.t.

V vo 5

n A

0 t

1 e

C 1

1 M

R O

O D

R T

CA E

R

/

D N

A M

O O

R T

D C

$cG R

N g

f I

H N

{

E g iE 3

W T

5 E

B m

.yY F

A V

t R

tA O

S O

g g,

+

S D

O

1 e

e iL

!S

17. Quility Assurne

\\

l Table 17.4-1 (Sheet 8 of 10) i I

RISK-SIGNIFICANT SSCs WITillN Tile SCOPE OF D RAP l

System, Structure, or Rationale

• Insights and Assumptions l

Component (SSC)"'

I l

Reactor Vessel Insulation EP These devices provide an engineered now path to promote l

Water Inlet and Steam Vent in vessel retention of the core in a severe accident.

l Devices

-l Reactor Cavity Doorway EP This device provides a flow path to promote in vessel l

Damper retention of the core in a severe accident.

l System: Normal Residual Heat Removal System (RNS)

~

~

l RNS Pumps EP These pumps provide shutdown cooling of the RCS. They I

also provide an alternate RCS lower pressure injection I

capability following actuation of the ADS.

I l

The operation of these pumps is RTNSS-important during I

shutdown reduced-inventory conditions. RNS valve I

realignment is not required for reduced inventory I

conditions.

I RNS Motor-Operated RRW/FVW These MOVs align a flowpath for nonsafety-related makeup l

Valves to the RCS followmg ADS operation.

I System: Spent Fuel Cooling System (SFS) l SFS Pumps EP These pumps provide flow to the heat exchangers for I

removal of the design basis heat load.

Revision: 19 D'

December 31,1997 17 22 W Westingh0USS

?

e NRC FSER OPEN ITEM i

=

k s

Question: 720.466F (OITS #6672)

In response to RAI 720.426F, which considered the likelihood of downward relocation resulting from crust failure.

Westinghouse offered qualitative arguments that any secondary explosion would involve a complex relocation process, and would not be conducive to producing an explosion load to threaten the containment. The acknowledgement that the relocation process may be complex makes the qualitative arguments less convincing in absence of some quantifications. For example, how much melt mass will likely be released from the crucible upon primary explosion induced failure, how much will the lower plenum coolant inventory be at the time, what would be the estimate of explosion load for that combination of melt and water. These are the types of quantifications that would add credibility to the qualitative arguments. In the absence of such information, the staff believes that the downward relocation issue remains an open item.

Response

The response provided in revision I of open item 720.426F is intended to describe the boundary conditions for the induced downward relocation scenario in terms of the mass of melt relocated, water in the reactor vessel lower head and explosion loads. The exercise is performed to conceptually understand the scenario in a manner consistent with other scenarios in the in-vessel steam explosion ROAAM. The conclusion of the analysis presented in revision i of 720.426F is that the secondary relocation would be a low flow of melt into a lower plenum that has been voided of water by two separate mechanisms. Although the interactions are characterized as complicated, it is a simple concept that an explosion large enough to move the large inertial mass associated with the lower supporting structures and debris bed will preferentially move the much lower inertial mass associated with the water pool. Therefore, the water is entrained out of the lower head with the venting steam fro n the piimary explosion. Despite the fact that the analysis is qualitative, it is clear without coolant in the lower head, there can be no fuel-coolant interaction.

Numerical analyses are not required to show that no explosion is predicted, and therefore, none are presented.

PRA Revision: None.

720.466F-1 W Westinghouse

l e

e NRC FSER OPEN ITEM n;=-um g~

7 Question: 720.467F (OITS #6673)

During a meeting with Westinghouse on January 22,1998, the staff communicated its concern with the values chosen for the number and diameter of the jets as input for the hinged failure case. Specifically, the hinged failure case was modeled as 236 coherent jets with a diameter of 0.068 m, and the model was chosen to represent the upper metalic layer spilling from the " unzipped" tower crucible. The concern was not necessarily that the values were inappropriate, but that sufficient information on how Westinghouse arrived at these particular input parameters has not been provided. It is the staff's contention that a process to trail and error must have taken place to arrive at these two values that would result in TEXAS calculating a significant explosion. De staff believes this process should be documented and that this issue is an open item.

Response

For the hinged lower head failure case, it is assumed that the lower head separates from the cylindrical portion of the vessel at RPV failure. The parameters needed to implement the TEXAS model are the number and size of coherent debris gets produced by the hinged lower head failure. He total number of jets and the jet diameter were not based on a process of trial and error. Rather, they were calculated as described below.

Input:

Melt Mass -

75900 kg UO 8500 kg ZrO2 70000 kg Stainless Steel 12900 kg Zr Release -

Metal

1. = 15100 kg/s Melt Superheat = 80*K Definitions:

Melt mass flow rate, kg/s

=

Melt density, kg/m' p

=

2 g

= Gravitational acceleration, m/s V

= Volume of lower head, m' im Volume of debris in the cylindrical part of the RPV, m' V,,,

=

H,,,

= Height of the debris in the cylindrical part of the RPV, m V

=

Melt velocity, m/s 2

Melt jet area, m A

=

d

= Melt jet diameter, m N

Number of melt jets

=

72 m m T westinghouse

\\

n 1

a o

NRC FSER OPEN ITEM h

~

Analysis:

Volume of Debris (V )in RPV:

o Material Mass (kg)

Density (kg/m')

Volume (m')

UO 75900 10000 7.59 2

ZrO 8500 5600 1.52 2

SS 70000 7800 8.97 Zr 12900 6800 1.98 3

V, = 20.06 m Volume of hemispherical lower head:

y,.1.$.n.r 3

23 2.n.(2)3 3

= 16.76 m3 Vessel radius (r) = 2 m.

Volume of debris in cylindrical part of RPV:

Vg = 20.06 m3 - 16.76 m3

= 3.3 m3 l

\\

720.4678 2 w wesunpouse 1

I i

o o

NRC FSER OPEN ITEM

)

\\

]

i lleight of debris in the cylindrical part of RPV:

i

)

V*

11 4 = n +r2 3

, 3.3 m n (2)2

= 0.26 m i

Driving head of debris in cylindrical part is used to estimate the debris velocity:

2*bP y,3

" E# 'M P

where V-pgH4

= /2(9.8 m/,2) (0.26 m)

V = 2.26 m/s 4

Total Jet Area (estimated by applying equation of continuity):

A..$.

pV 15100 kg/s 3

(7800 kg/m ) (2.26 m/s) l i

= 0.857 m2 720.467F-3 W Westinghouse

~

o C

NRC FSER OPEN ITEM ME=EE 1,*

i..

z, Characteristic Size of Failure Taken as Jet Diameter (estimated from total jet area and RPV circumference):

2 A

0.857 m d-2nr 2n *2 m d = 0.068 m Total number of jets located around circumference to given equivalent total jet area:

A 0.857 m' y,

x(d/2)2 x (0.068/2)2 N = 236 PRA Revision: None.

720.467F-4 W85tingh00S8