ML20052C965

Forwards Addl Info Re Comparison of Planned Controls for Handling Heavy Loads W/Guidelines of NUREG-0612,per NRC 801222 Request

ML20052C965
Person / Time
Site:	Grand Gulf
Issue date:	05/04/1982
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR AECM-82-149, NUDOCS 8205060144
Download: ML20052C965 (66)

Text

l MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi EdMdIAMidE P. O. B O X 1640, J A C K S O N. MIS SIS SIP PI 3 9 2 05 May 4, 1982 kf NuctE Aa PsooucTioN DEPARTMENT p Office of Nuclear Reactor Regulation 2 r Division of Licensing  : . , _

U.S. Nuclear Regulatory Commission i ,, .d2 " j0 Washington, D.C. 20555 a

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Attention: Mr. Darrel G. Eisenhut, Director 'h

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SUBJECT:

Grand Gulf Nuclear Station Units 1 and 2 Docket Nos. 50-416 and 50-417 File 0260/16360 Heavy Loads Final Report AECM-82/149 Your letter dated December 22, 1980, requested a review of the planned controls for handling heavy loads at Grand Gulf Nuclear Station, a comparison of these controls to the guidelines of NUREG-0612 and the submittal of information to demonstrate that the guidelines have been or will be met.

On November 23, 1981, Mississippi Power & Light (MP&L), on behalf of Middle South Energy, Inc. and South Mississippi Electric Power l Association, submitted a report on the first phase of the requested review. The information submitted was that requested in Section 2.1 of Enclosure 3 to your December 22, 1980, letter. This letter transmits the information developed in the second stage of our review. It is that information requested in Sections 2.2, 2.3, and 2.4 of Enclosure 3 to your December 22, 1980, letter.

Yours truly, L. F. Dale

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Manager of Nuclear Services PJR/JDR:lm Attachment g30 y y cc: (See Next Page) l[

8205060144 820504 PDR ADOCK 05000416 A PDR AE2J1 Member Middle South Utilities System

AECM-82/149 MISSISSIPPI POWER O LIGHT COMPANY Page 2 cc: Mr. N. L. Stampley (w/a)

Mr. G. B. Taylor (w/a)

Mr. R. B. McGehee (w/a)

Mr. T. B. Conner (w/a)

Mr. Richard C. DeYoung, Director (w/a)

Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. J. P. O'Reilly, Regional Administrator (w/a)

Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W., Suite 3100 Atlanta, Georgia 30303 AE2J2.

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RESPONSES TO REQUESTS FOR INFORMATION IN SECTIONS 2.2,2.3, Ato 2.4 OF ENCLOSURE 3 TO PRC DECEMBER 22,1980 LETTER 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOL NUREG-0612, Section 5.l.2, provides guidelines concerning the design and opera-tion of lood-handing systems in the vicinity of stored, spent fuel. Information provided in response to this section should demonstrate that adequate measures have been token to ensure that in this crea, either the likelihood of a load drop which might domage spent fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria of NUREG-0612, Section 5.1, Criterio I through lil.

ITEM 2.2.1 Identify by name, type, capacity, and equipment designator, any cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool.

RESPONSE: At Grand Gulf Unit I there are two spent fuel storage pools. One in the Auxiliary Building at the 208' el. capable of storing up to 158% of a full core and another in Containment capable of storing up to 21% of a full core.

(See Figure I, Regions I and 12). The spent fuel storage pools and racks are described in FSAR Section 9.l.2.

With regard to the Auxiliary Building, the New Fuel Bridge Crone is capable of carrying loads over the spent fuel pool. It is on overhead bridge crane and has a capacity of 5 tons. In the containment, the Polar Crane is capable of carrying leads over the spent fuel storage area. It is an overhead bridge crone mounted on a circular rail with a main and auxiliary hoist with capacities of 125 and 35 tons, respectively.

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ITEM 2.2.2 Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying heavy loads or are permanently prevented from movement of the hook centerline closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load can fall into the fuel-storage pool.

RESPONSE: With regard to the Auxiliary Building spent fuel storage area, the Spent Fuel Cask Crane is also located at the 208' el. However, the spent fuel pool is protected from drops from this crane by restriction of the limits of cask crane travel as discussed in subsections 9.l.4.1 and 9.1.4.2.2.2 of the Grand Gulf FSAR. The limits of cask travel are shown in FSAR Figure 1.2-7.

With regard to both the Auxiliary Building pool and the Containment spent fuel storage creo, there are 1/2 ton capacity jib cranes that can lif t loads over the spent fuel storage racks. As indicated in our initial response, however, these crones are only used to carry lighter loads, such as channels, control rods or fuel assemblies that do not qualify as heavy loads.

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ITEM 2.2.3 Identify any crones listed in 2.2.1, above which you have evaluated as having sufficient design features to mak,e the likelihood of a lood drop extremely small for all loads to be corried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.l.6 or partial compliance supplemented by suitable alternative or additional design features). For each crane so the food-handling-system (i.e., crane-lood-evaluated, combination) provide ir; Attachment 1.

Information specified RESPONSE: It has not been necessary to evoluote either of the two crones identified above against the criteria of NUREG-0612, Section 5.l.6.

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ITEM 2.2.4 For cranes identified in 2.2.1, obove, not categorized according to 2.2-3, demonstrate that the criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria i through lli, provide a discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance. This response should include the following information for each crane:

a. Which alternatives (e.g., 2, 3, or 4) from those identified in NUREG-0612, Section 5.l.2, have been selected.

b. If Alternative 2 or 3 is selected, discuss the crone motion limitation imposed by electrical interlocks or mechanical stops and Indicate the circumstances, if any, under w.nch these protective devices may be bypassed or removed. Discuss any administrative procedures invoked to ensure proper authorization of bypass or removal, and provide ony related or proposed technical specification (operational and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops.

c. Where reliance is placed on crone operational limitations with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decay times, provide present ond/or proposed technical specifications od discuss administrative or physical controls provided to ensure that these assumptions remain valid,

d. Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.

l l e. Analyses performed to demonstrate compliance with Criteria I through 111 should conform to the guidelines of NUREG-0612, Appendix A. Justify any exception taken to these guidelines, and provide the specific 1 information requested in Attachment 2, 3, or l l 4, as appropriate, for each analysis performed.

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I RESPONSE: The heavy loads that could be handled by the New Fuel Bridge Crane or the Polar Crone were identified in Tables 6 and 4 respectively in MP&L's initial submittal to NRC regarding the heavy loads issue (letter from L.

F. Dale to Eisenhut dated November 23,1981). 'As indicated in the response to item 3.0 of that submittal, both procedural restrictions and Technical Specifica-tions have been developed to prevent carrying heavy loads over spent fuel in the rocks in these two storage pools. Nonetheless, the pool gates (weight approxi-mately 3.5 tons) must be lif ted in the pools.

For this reason, structural analyses were performed to estimate the number of fuel rods that could be damaged as a result of a gate drop onto spent fuel in the storage rocks. Based on this number of fuel rods, dose calculations were performed for drops in both the Containment and Auxiliary Building using the conservative model and assumptions employed for Fuel Handling Accidents described in Sections 15.7.4 and 15.7.6 of the FSAR. These analyses resulted in a calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dose of 8.5 Rem at the site boundary. This slightly exceeds the 1/410 CFR Part 100 limit of 6.25 Rem imposed by Criterion I of NUREG-0612, Section 5.1. This calculated dose is still, however, well within the 10 CFR Port 100 limit of 25 Rem. The calculated inhalation doses were only a small fraction of the NUREG-0612 limit of 75 Rem.

With regard to criticality, it was determined that the possibility of a Keff of greater than 0.95 could not be precluded in the very unlikely event that a gate drop occurred and impacted spent fuel that contained substantial U-235, such as could be the case in a core off-load situation. This is because the Unit I rocks rely on spacing alone to nrevent criticality, i.e., there is no neutron poison in the rocks or the pool water. A description of the design of the rocks is provided in FSAR Section 9.l.2.

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it is evident from the many conservatisms employed in the calculations and the unique conditions that must exist at the time of a load drop, that the likelihood of a load drop occurring that actually exceeds the NRC criteria is very small. Nonetheless, MP&L has initiated actions with the objective , of demonstrating that the NUREG-0612 criteria are met.

Every transfer of a heavy load having 6e potential of a drop with damage in excess of the criteria will be carried out under administrative controls. Within two years of the date of this response, as required by the NRC's letter of December 22, 1980, an evaluation will be conducted to determine if any additional measures beyond administrative controls' are required.

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2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG-0612, Section S.I.3, provides guidelines concerning the design and operation of lood-handling systems in the vicinity of the reactor core.

Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in this crea, either the likelihood of a load drop which might domoge spent fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits set by the evoluotion criteria of NUREG-0612, Section 5.1, Criterio I through 111.

ITEM 2.3.1 Identify by name, type, capacity, and equipment designator any cranes physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel.

RESPONSE: The only handling system within containment physically capable of carrying heavy loads over the reactor vesse! is the Containment Polar Crone.

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ITEM 2.3.2 Justify the exclusion of any cranes in this crea from the above category by verifying that they are incapable of carrying heavy loads, or are permanently prevented from tho movement of any load .

either directly over the reactor vessel or to such a location where in I the event of any lood-handling system failure, the load may land in or on the reactor vessel.

RESPONSE: The only other handling system inside the containment capable of  ;

moving loads over the vessel is the Refueling Platform used for refueling operations, its' function is to handle single fuel assemblies and perform other I vessel servicing functions, i.e., no heavy loads as defined in NUREG-0612 are handled by this handling system.

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ITEM 2.3.3 Identify any cranes listed in 2.3.1 above which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by .

suitable alternative or additional design features). For each crane so evaluated, provide the load-handling-system (i.e., crane-lood-

. combination) information specified in Attachment 1. i RESPONSE: As indicated in the response to item 2.2.3 above, it has not been found to be necessary to evaluate the Polar Crane against the criteria of i NUREG-0612, Section 5.1.6.

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ITEM 2.3.4 For cranes identified in 2.3.1 above not categorized according to 2.3.3, demonstrate that the evaluation criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in your response to Section 2.4 of this request. With respect to Criteria I through Ill, provide o discussion of your evaluation of crane operation in the containment and your determination of compliance. This response should include the following information for each crane:

ITEM 2.3.4.o. Where reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices con be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such action. Discuss any related or proposed technical specifications concerning the bypassing of such interlocks.

RESPONSE: The crone load blocks have not been included in any of the heavy load drop evaluations described in subsequent responses for the reasons given below:

NUREG-0612 requires that the load block and hook be considered as a heavy load.

The load block is used for handling numerous loads, including the reactor vessel head, drywell head, steam dryer, and moisture separator. in moving these loads, the hook, load block, rope, drum, sheave assembly, motor shaf ts, gears, and other load bearing members are subjected to significant stresses approaching the load rating of the crane. By comparison, these components are subjected to a considerably smaller load when only the hook and load block are being moved.

Based on this, it is not considered feasible to postulate a random mechanical failure of the crane load bearing components when moving either the main hoist or auxiliary hoist load block without a load.

The only two feasible failure modes for dropping of the main hook and load block would be:

1) A control system or operator error resulting in hoisting of the block to a "two blocking" position with continued hoisting by the motor and subsequent parting of the rope (this situation can be prevented by operator action prior I to "two blocking" or by an upper limit switch to terminate l

hoisting prior to "two blocking"); and l

l 2) Uncontrolled lowering of the load block due to failure of l the holding broke to function (the likelihood of this can be l made small by use of redundant holding brakes).

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The Grand Gulf polar crane main hoist and auxiliary hoist are each provided with two redundant and diverse upper limit switches to interrupt power to the hoist motor prior to "two blocking". .When power is removed, holding brakes are automatically applied. One of the two limit switches is a geared limit switch driven off the drum shaf t. The other is a counter weight switch that is released when the load block comes up against a trip bar; the trip bar will stop power to the main hoist at more than 40" and the auxiliary hoist at more than 26" below the low point of the sheave assembly.

The holding brakes are solenoid released, and spring applied on loss of power to the solenoid. Two holding brakes are provided for each hoist on the polar crane; each holding broke has sufficient capacity to hold the rated load and is rated 150% of full motor torque. Additionally, inspection procedures anure that the  :

limit switches and holding brakes are functional and properly adjusted.

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l With the provisions described above, the redundant limit switches will reduce the likelihood for "two blocking" and the redundant holding brakes will reduce the likelihood of uncontrolled lowering of the load block. Based on these features, it is concluded that a drop of the load block and hook is of sufficiently low

likelihood that it does not require load drop analyses.

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ITEM 2.3.4.b. Where reliance is placed on other, site-specific considerations (e.g., refueling sequencing), provide present or proposed technical specifications and discuss administrative or physical controls provided to ensure the continued validitiy of such considerations. _

RESPONSE: Loads only lifted over the vessel when the reactor vessel head or moisture separator is in place were not considered as loads that could potentially drop into the core. These are: the drywell head and the steam dryer. No administrative controls are required to enforce this situation, because it is physically impossible to disassemble or reassemble the reactor such that these loads would be carried over on open vessel.

In addition, the portable radiation shield is installed in the reactor well after the head has been removed, but before the dryer or separator has been removed. It is removed from the reactor well af ter the dryer and separator have been installed. This sequencing is enforced by written procedures governing the installation and removal of the portable radiation shield and will be strictly enforced by individuals in charge of lif ts by the Polar Crane.

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ITEM 2.3.4.c Analyses performed to demonstrate compliance with Criteria i through 111 should conform to the guidelines of NUREG-0612, Appendix A. Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.

RESPONSE: There are three potential consequences of interest when considering load drops onto the open reactor vessel. They are: 1) loss of reactor vessel integrity,2) fuel cladding damage and the resultant radiological dose, and

3) fuel crushing and the possibility of a resulting criticality condition. Criteria i through 111 in Section S.I of NUREG-0612 address each of these potential consequences. The evaluations below have been performed to address these issues.

Analyses were performed to determine the structural consequences of dropping the vessel head or the shroud head assembly during maintenance operations.

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Reference:

NEDC-23566, Structural Analysis of Reactor Pressure Vessel and Internals for Vessel Head Drop, Shroud Head Assembly Drop and Steam Dryer Assembly Drop Conditions). The consequences of dropping the steam dryer assembly can be extrapolated from the analysis of the shroud head assembly drop, since o steam dryer drop would generate less kinetic energy than the shroud head assembly drop and the impacted structure would be the same for both cases. It was postulated that the vessel head would be dropped from o height of approximately 40 ft. above the vessel-head flange, and that at impact the head would be rotated 900 from the inplace orientation causing a point impact on the vessel. This height corresponds to the maximum possible carry height as limited by physical restrictions. The shroud head and dryer were assumed to be dropped from a height sufficient to generate the steady state velocity of the two assemblies as they move through water and are under the action of the fluid drag forces. An axisymmetric impact of the shroud head and dryer assembly on the main body of the shroud is assumed. In addition, nonoxisymmetric impact of the shroud head on the shroud was also considered.

The vessel loads due to the postulated impacts were determined by dynamic, elastic-perfectly plastic finite element analyses. In the vessel head impact l analysis the vessel was characterized by one-dimensional oxial and shear springs.

Implicit in this characterization are the assumptions that the extensional and I

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inextensional cylindrical shell motions are uncoupled, and that the lateral and transverse strains are negligible compared with the longitudinal strains. In the shroud head impact analysis the impacted shroud body is chorocterized by a series of one dimensional springs. Lateral strains are assumed negligible and transverse strains are precluded by the assumption of an axisymmetric impact.

The finite element program used for these analyses, DYANI, is a direct stiffness formulation using direct integration techniques in the time domain. The time integrator is based on on incremental formulation of the linear acceleration operator. The program has the Wilson-Theta operator as an option, and this option was employed with a value of 1.4 for the Wilson-Theto constant.

The results of the finite element analyses can be summarized as follows:

Drop of: Consequences Vessel Head Local yielding of the vessel top flange Damage to the vessel and vessel head will be considerable which could render these components irreparable.

Reactor skirt does not yield and will remain stable.

Vessel maintains its normal position.

Deflections of connected piping are within acceptable  !

limits.

No damage to fuel rods and hence no release of roCoative materials.

Shroud Head / Steam Local yielding of the upper shroud and shroud support Dryer struts.

No instability of the shroud support structure.

Damage to the internal components is considerable and could render them irreparable.

No damage to the fuel rods and hence no release of radioactive materials.

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in addition drops of the drywell head onto the vessel head and the portable radiation shield onto the separator (dryer conservatively ignored) were evoluoted by comparing the available drop energies to those in the GE onalyses described obove. Based on this comparison, these drops were found to be bounded by the GE head drop and dryer drop onalyses.

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On the basis of the analyses described above, it is concluded that NUREG-0612 ,

Criteria 1-ill are met for all postulated drops into the reactor well.

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2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT I REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING NUREG-0612, Section S.I.5, provides guidelines concerning the design and operation of lood-handling systems in the vicinity of equipment or components required for safe reactor shutdown and decay heat removal. Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small or that damage to such equipment from load drops will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1 and their loads in your response to 2.l.3.3.

ITEM 2.4.1 Identify any cranes listed in 2.1.1 above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by suitable alterantive or additional design features). For each crane so evaluated, provide the lood-handling-system (i.e., crane-load-combination)information specified in Attachment 1.

RESPONSE: The handling systems of interest listed in response to item 2.1.1 are the Containment Polar Crane, the New Fuel Bridge Crane and the LPCS and RHR "C" Equipment and Hatch Holst. It has not been necessary to evaluate any of these handling systems aginst the criteria of NUREG-0612, Section 5.1.6.

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ITEM 2.4.2 For any crones identified in 2.1-1 not designated as single-failure-proof in 2.4-1, o comprehensive hazard evoluotion should be provided which ,

includes the following information: ,

c. The presentation in a matrix format of all heavy foods and potential impact areas where domoge might occur to i safety-related equipment. Heavy loads identification should include designation and weight or cross-reference to information provided in 2.1-3-c. Impact areas should be identified by construction zones and elevations or by some other method such that the impact crea can be located on the plant general arrangement drawings. Figure I pro-vides a typical matrix. j

b. For each interaction identified, indicate which of the load and impact area combinations con be eliminated because of separation and redundancy of safety-related equip- i ment, mechanical stops and/or electrical interlocks, or j other site-specific considerations. Elimination on the basis of the aforementioned considerations should be i supplemented by the following specific information:

(1) For load /torget combinations eliminated because of separation and redundancy of safety-related equip-ment, discuss the basis for determining that load i drops will not offect continued system operation (i.e., the ability of the system to perform its safety-related function).

(2) Where mechanical stops or electrical interlocks are to be provided, present details showing the areas where crane travel will be prohibited. Additionally, provide a discussion concerning the procedures that ,

are to be used for authorizing the bypassing of interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability offer operations which require bypassing have been completed.

(3) Where load /torget combinations are eliminated on the basis of other, site-specific considerations (e.g.,

maintenance sequencing), provide present and/or proposed technical specifications and discuss admin-istrative procedures or physical constraints invoked to ensure the continued validity of such considero-tions.

c. For interactions not eliminated by the analysis 2.4-2-b, above, identify any handling systems for specific loads which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely  ;

small and the basis for this evoluotion (i.e., complete  ;

l compliance with NUREG-0612, Section 5.l.6, or partial I

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compliance supplemented by suitable alternative or cdditional design features). For each crane so evaluated, provide the food-handling-system (i.e., crane-lood-combination)information specified in Attachment 1.

. d. For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate using appropriate analysis that damage would not preclude operation of sufficient equipment to allow the system to perform its safety function following a load drop (NUREG-0612, Section 5.1, Criterion IV). For each analysis so conducted, the following information should be provided.

(1) An indication of whether or not, for the specific load being investigated, the overhead crane . handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).

(2) The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A.

(3) The information requested in Attachment 4.

RESPONSE: For the 3 handling systems listed in 2.4.1 above, a combination of systems and structural evaluations was utilized to determine if Criteria 111 and IV of NUREG-0612 are met for all postulated load drop scenarios. To assist these evaluations, o set of safety functions were identified corresponding to these criteria. The goal of these evaluations then became to demonstrate that the applicable safety functions could be accomplished for all load drop scenarios.

The safety functions identified and the corresponding NUREG-0612 criteria are indicated in Toble 1. The load impact regions evaluated for each region are also indicated in Table I and are displayed in Figures I, 2, and 3.

The systems and structural evoluotion methodologies are described below.

Tables 2 through 15 have been prepared to provide a region-by-region presenta-tion of the loads of interest, the evaluations undertaken, the results of the evaluations and conclusions based on these results.

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As indicated in the tables, Criterio ill and IV have been satisfied for all postulated load drops within each region. The basis for acceptance was, in most cases, that sufficient redundancy and separation of equipment required to accomplish the applicable safety functions exists or that equipment was not impacted based on structural analyses. The basis for acceptance in two specific situations is not equipment separation or redundancy. For this reason they have been discussed here in more detail than provided in the tables. The first relates to cooling fuel in the in-containment storage pool with the Fuel Pool Cooling and Cleanup System (FPCCU) and the second relates to RHR piping used for extended core cooling in Region 8.

Spent Fuel Cooling - The results of the systems evaluations conservatively performed in accordance with the methodology described below indicate that in every region except Regions 13 and 14, FPCCU cooling to the pool could potentially be lost as a result of pipe breaks. In addition, in a number of these regions, supplemental or backup pool cooling via the RHR system could also be lost.

The overall resu t of these system losses was judged to be acceptable, however, based on the following rationale. If both FPCCU and RHR cooling to the pool were lost,* hoses could be routed to the pool to provide makeup from any available water source. Therefore, spent fuel in the pool would always remain covered with water. Makeup would only be necessary if boiloff from the pool were to occur. Whether or not boiloff would occur and when would be highly dependent on the amount, power history, and decay history of spent fuel in the storage rocks. In any event, as long as makeup could be provided to the pool, there should be no spent fuel damage and, therefore, no off-site dose con-sequences.

Calculations were performed to determine how much time would be available to effect repairs and restore cooling in the event all cooling to the pool was lost.

The results for several decay times and numbers of assemblies in the pool are

• This assumes the worst case condition, i.e., (1) communication with the FPCCU or RHR system vio on open reactor vessel and/or water in the reactor well is not possible, and (2) backup water sources such as a supply from the condensate transfer system are not available (nonsafety system).

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presented below. The 170 assembly case corresponds to all rocks having spent fuel in them. ,

Number of Approximate Approximate Time (Hrs)

Assemblies in Time (Hrs) Until To Begin Uncovering Decay Time Pool Boiloff Begins Fuel (No Makeup) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 170 6 67 7 days 170 14 149 30 days 170 24 251 l 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 85 13 134 [

t 7 days 85 29 29 8 7 30 days 85 49 502 As con be seen, even in the most conservatively postulated situation, there are at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> available before boiling would occur and almost 3 days before j fuel would be uncovered assuming no makeup could be provided, which would not be the case. In the unlikely event that a load drop were to occur that resulted in a complete loss of cooling to the pool, the spent fuel situation in the pool is likely to be much more favorable than the most conservative situation.

Accordingly, there would likely be many hours available to restore cooling to the pool. For this reason, it is judged that there is reasonable assurance that cooling of fuel in the in-containment rocks con be accomplished following any load drop scenario in the containment.

RHR Piping - Region 8 r

Region 8 was found to contain equipment whose loss could result in a complete loss of the RHR shutdown cooling mode. An evaluation was performed to determine if in fact the RHR piping in the region could be lost.

The piping of interest in the region are two feedwater lines (24"-DBA-13) that l deliver flow from the RHR system to the reactor vessel and the RHR suction l line (20"-DBA-64) from Recirculation Loop B to the RHR system. The load j drops of interest would be the possible drop of a Regenerative or Nonregenera-tive Heat Exchanger or Hatch Cover into the heat exchanger compartment 20

impacting the compartment floor obove the steam tunnel. Figure 4 depicts the postulated drop and the location of floors ond. piping in relation to the piping of interest.

The following process was utilized to determine that it is highly unlikely that any of the lines listed above would be lost as a result of a drop into Region 8 from above the 208' elevation. Initiaily, structural analyses of drops to the heat exchanger compartment floor (170' elevation) were conducted to determine whether the load would perforate (i.e., break all the way through the floor),

cause overall collapse of the floor, or cause scabbing from the underside of the floor. The analyses indicated that scabbing only is possible. This would mean ,

that pieces of concrete on the order of a couple of inches thick (thickness up to the first rebar) could potentially impact piping below.

Next, a review of the piping arrangement in the steam tunnel indicated that the feedwater lines of interest are shielded from direct impact of these scabbed concrete sections by the larger main steamlines above. In addition, the I

mainsteam lines and feedwater lines are enclosed in guard pipes. Accordingly, it was judged that it is inconceivable that damage to the feedwater lines would result from this load drop scenario.

With regard to the RHR suction line, it is further protected by a 4 ft. thick concrete slab that forms the steam tunnel floor at elevation 133' and a guard pipe of its own. Again, it is inconceivable that this pipe could be damaged as a result of the postulated load drop scenario.

Based on-this evaluation, it was concluded that there is reasonable assurance that RHR shutdown cooling could be maintained following any postulated heavy load drop in Region 8.

SYSTEMS EVALUATION METHODOLOGY As part of the evaluation of heavy load handling operations at Grand Gulf, o number of potential load drop regions in both the Containment and the Auxiliary Building were addressed by performing a " systems evaluation." For the regions 21

involved, refer to Tables 2 through IS. The impact regions are displayed in Figures I,2, and 3 and extend down through o,Il elevations of the building unless limited on the basis of structural evaluations. The objectives of the " systems evoluotions" were to demonstrate that safe shutdown, long-term cooling, and fuel pool cooling could be achieved and/or maintained assuming that certain equipment was lost as a result of postulated load drops.

In order to demonstrate the ability to accomplish these objectives, it was necessary to (1) identify the safety functions required to be accomplished for each region, (2) identify plant systems required to accomplish the identified safety functions, (3) identify the equipment associated with these systems or their support systems that could potentially be lost if a load drop were to occur in the region, and (4) determine the resultant effects of the loss of this equipment on the ability to accomplish the identified safety functions.

PLANT CONDITIONS AND SAFETY FUNCTIONS ,

t To determine the safety functions that must be performed to accomplish the f objectives listed above, the foilowing plant conditions were assumed during load handling movements in the Containment and the Auxiliary Building.

CONTAINMENT BUILDING - PLANT CONDITIONS Region 12 in the containment contains spent fuel rocks capable of storing up to 21% of a full core. Accordingly, Safety Function I in Table I is applicable. For the purpose of evaluating whether or not this safety function could be accom-plished, heavy load drops inside containment were postulated to occur when there was recently discharged irrcdiated fuel in the in-contoinment spent fuel storage racks and in the Auxiliary Building spent fuel storage rocks. This could be in either a refueling core discharge (1/3 core) situation (or some fraction ,

thereof) or a full core discharge situation. As indicated below, the systems required to accomplish Safety Function I, spent fuel cooling, are different for these two situations.

It is not anticipated that any of the heavy loads handled by the Containment .

l Polar Crone would be lif ted until the plant has been shut down for some time.

22 l

Accordingly, all heavy load drops inside containment were postulated to occur when the reactor was shut down and cooled down. This condition corresponds to either Operational Condition 4 (Cold Shutdown: mode switch in Shutdown, overage reactor coolont temperature less than 200 F) or Operational Condition S (Refueling: mode switch in Shutdown or Refuel, overage reactor coolant temperature less than !40 F) as defi .ed in Table 1.2 of the GGNS Unit l Technical Specifications.

The overall safety function that must be accomplished during either of these operational modes is extended core cooling. Therefore, Safety Functions 2 and 3 in Table I are applicable. Two safety functions have been included related to maintaining extended core cooling to address two different plant configurations:

(I) reactor vessel head in place, and (2) reactor vessel head removed. Whether the plant is in one configuration or the other offects the available methods and systems for cooling the core.

Auxiliary Building - Plant Conditions Safety Function I, spent fuel cooling, is also aplicable to load drop evaluations in the Auxiliary Building. Accordingly, heavy load drops were postulated to occur under any of the some spent fuel offload conditions described above for the containment.

Load handling operations in the Auxiliary Building could take place during any operational condition. For this reason, the functions that could have to be accomplished following a postulated load drop to assure that the reactor con be safely shutdown and cooled are Safety Functions 2, 3, or 4 in Table 1. The applicable operational conditions defined in the GGNS Technical Specifications are:

1 Operational Condition 1: Power Operation; Mode Switch in

Run, coolant at any temperature l

Operational Condition 2: Startup; Mode Switch in Startup/ Hot Standby, coolant at any temperature Operational Condition 3: Hot Shutdown; Mode Switch in l

' Shutdown, coolant temperature greater than 2000F l

23

Operational Conditions 4 and S: defined above in Section 1.1.1

^

Systems Selected to Accomplish Safety Functions For the purpose of performing the systems evoluotions described in this appendix, certain plant " safety systems" were selected to accomplish each of the safety functions of interest. " Safety systems" only were relied on. This means that many of the systems used to accomplish normal. plant shutdown, such as offsite power and condensate, feedwater and circulating water systems, were not relied on to perform the systems evoluotions.

Further, although there are many combinations of safety systems that may be used for safe shutdown and core cooling, initially only certain systems were selected for evaluation. The reason for this was to minimize the effort required to trace out system piping and cables within the plant. Additional systems, or portions of systems, were included in the review only if it was determined to be necessary or prudent to do so.

Safety Function I - Spent Fuel Cooling As indicated in Figure S, cooling of spent fuel in the Auxiliary Building and/or the in-containment pools can be accomplished by the Fuel Pool Cooling and Cleanup System (FPCCU) for on overage core discharge (% core), and the FPCCU supplemented by the RHR system for a full core discharge. The explicit design basis heat loads for these two situations, including assumptions regarding previous core discharges residing in the pools, are described in FSAR Sec-tion 9.l.3.

Figure 6 identifies two redundant groups of systems selected for evaluating extended core cooling with the head in place. Each group in turn includes alternative equipment paths.

Figure 7 identifies two redundant groups of systems selected for evaluating [

24

7 l

extended core cooling with the head removed. Each group in turn includes alternative equipment paths.

Figure 8 indicates the functions that must be accomplished to achieve shutdown and core cooling from power operation. Systems selected for evoluotion purposes to accomplish each of the functions are otso identified. For some functions, alternative systems are included.

Load Drop Matrices For each region evaluated, o matrix was developed that outlined each of the regions to be evaluated and the critical system components relied on to perform the system function. The critical components include required support functions such as cooling water, power supplies, and electrical cabling.

Each region of concern was evaluated to determine which of the system components of interest could be lost as a result of a load drop. The evoluotion was performed by reviewing plant drawings and system descriptions including small and large piping layout drawings and plant walkdowns. Information from ,

the fire exposure zone evaluations was used to locate cabling of importance within the regions. This information was then entered into the matrices and a determination made as to the effect of its loss on the safety function being considered. This determination considered both the effect of losing individual components and of losing all components identified within a region.

Steps in the Systems Approach The following summarizes the steps that were performed in the systems evaluations for each safety function evaluated:

Completion of Matrix (1) Identify the system (including any support systems) components selected for accomplishing the safety func-tion of interest, organize into component groups for purposes of evaluation, and enter in matrix columns.

(2) For each region of concern, evoluote the potential for 25

damage / loss of system components based on a detailed-review of the equipment piping and cabling layout. Enter the components assumed lost in to the appropriate box on the matrix.

(3) Compare system equipment required (item 1), with equip-ment lost (Item 2), and determine if the safety function for which the system is relied on could be lost.

(4) Review for other potential system interactions based on equipment damaged / lost and determine if safety function could be lost.

Conclusions (5) If the system evaluation reveals that the system could accomplish its safety function following a load drop into the region of interest, then no further evaluation is necessary.

(6) If the system evaluation reveals that the system function could potentially be lost, then evoluote the possibility of relying on alternative safety systems to accomplish the same function following a postulated load drop into the region.

(7) The overall safe shutdown conclusion regarding a partico-for region is the composite for that region of the conclusion for all the systems required to accomplish the safe shutdown functions.

STRUCTURAL EVALUATION METHODOLOGY Structural load drop analyses performed to support evaluations related to safety functions 1-4 in Table i typically involved determination of structural response of ,

concrete floor slabs to dynamic impact loadings. The heavy loads which could potentially be dropped onto various floor slabs were evaluated to identify loads which control local response (e.g. penetration, scabbing, spalling, perforation, etc.); loads that control overall structural response (e.g., large inelastic deformations or abrupt failures of principal structural members, etc.); and/or loads that may induce behavior that exhibits combined response such that either overall or local failure modes would control. The results of this evoluotion are tabulated in Table 16.

l l

26

Where the controlling mode of response to postulated load drops is listed as

" local", these loads were evaluated to determine the potential for slab penetra-tion or perforation. Scabbing of the concrete deck backface was evaluated for all loads. Where it was found that postulated drops are capable of producing this scabbing effect, it was decided that the consequences of scabbing would be considered in the systems evaluations, i.e., its potential for damaging equipment below the floor was evaluated.

Postulated drops of the spent fuel pool gate, new fuel shipping container, portable radiation shield (cattle chute), and various hatch covers fall in this category of loads which control loca: response, and bound other load drops that potentially lead to local effects. A discussion of the local effects evaluation methodology is provided below.

Where the controlling mode of response is listed as "overall structural", these load drops were evaluated to determine the potential for producing gross and intolerable distortions of primary structural members and possibly propagating failures. Postulated drops of the spent fuel pool gate, the new fuel shipping containers, the portable radiation shield, various hatch covers, and the steam dryer and separator, fall in this category and bound other load drops that potentially lead to "overall structural" effects.

Overall Structural Response Evaluations A model of each floor location was developed with the objective of evaluating structural behavior for postulated flat and oblique drops of these loads.

A load drop methodology was developed to investigate the important modes of structural behavior. The objective of this methodology is to characterize structural behavior in terms of the available strain energy up to prescribed performance limits. These limits are dictated by either ductile or brittle modes of failure. The ductile mode is characterized by large inelastic deflections without complete collapse, while the brittle mode may result in partial failure or total collapse. The available internal strain energy that con be absorbed by the floor system without reaching those limits of unacceptable behavior is balanced 27

against the externally applied energy resulting from a heavy load drop. It has been assumed that momentum is conserved and the kinetic energy of the drop drives the mass of the floor and induces strain. As on additional conservatism, no credit was taken for potential sources of energy dissipation such as c.oncrete crushing and penetration.

An iterative step-wise linear static analysis was performed for the postulated load drops whose controlling mode of response was determined to be "overall structural." The objective was to determine force-deflection for important points in the structural model. The computation procedure of the analysis is based on a network interpretation of the governing equations, the principal feature of which is the segmentation in processing of the geometrical, mechani-col and topological relationship: of the structure. This allows a concise and ,

systematic computation algorithm that is applicable for different structural 1ypes.

For each impacted structural system (floor slab or slob-beam composite), o model was developed and the response of the system to the dynamic impact loading was determined. The model was loaded in the direct vicinity of the drop location. This is considered to be conservative in view of the fact that the slab will help transfer load away from the drop vicinity and result in a more favorable redistribution of the load.

The stiffness properties of the supporting beam grid, where opplicable, were represented assuming on effective reinforcement for the beam / slab composite section consistent with ACl 318-77 (Reference 1).

The model was loaded until the moment capacity of any section or the allowable deflection was reached. This moment capacity is defined by Chapter 10 of ACI 318-77 (Reference 1).

Generally, the ultimate load of a stab / grid system is reached prior to exceeding the hinge rotational capacity of particular sections provided that on unstable l mechanism has not formed. This was found to be the case in this analysis. The hinge rotational capacity was used as a criteria to set the maximum allowable l 28 1

I

level of deflection for the stab / grid system. The hinge rotational capacity for concrete structures was devaloped in References 2 and 3 based on test results given in References 4 and 5 and is given as:

r u = 0.0065 (d/c) 6. 0.07 (1)

where, r u = rotational capacity of plastic hinge (radions) d = distance from the compression face to the tensile reinforcement, c = distance from the compression face to the neutral oxis at ultimate strength.

The maximum deflection for a beam with a plastic hinge at its center, is then given by:

Xm = (ru L)/4 (2)

where, e

Xm = maximum deflection, L = span of beam Rotations of the magnitude governed by equation C.I result in cracking which is confined to o region below (above) the tensile reinforcement. Generally speaking the section will remain intact with no crushing, spalling or scabbing due to flexure; however, scabbing may occur os a result of shock wave motion associated with the reflection of tensile waves from the rear surface or shear plug formation. The potential for scabbing was evoluoted for all load drops.

The load / deflection history up to the point of the ultimate loading, coupled with the maximum allowable deflection, defines the maximum level of strain energy obsorption, provided that a shear failure has not occurred. The shear stress at limiting sections was checked and compared to allowables os specified in Chapter 11 of ACI 318-77 (Reference 1).

29

For each area where the potential for overall structural response modes was considered possible, an assessment of the bounding drop was made. The criteria for selection was impact energy of the postulated drop.

In addition to the conservatisms previously mentioned, the following conserva-tisms are also inherent in the methodology used in the evaluation:

1) Static material strengths for concrete and steel were used. Test data shows that this property increases with the increased strain rates associated with dynamic load-ings. For example, References 6 and 7 recommend dynamic increase factors of 1.25 for the compressive strength of concrete and 1.20 for the flexural, ter.sile and compressive strength of structural steel.

2) Design (minimum) material properties for concrete and steel were used. No increase was taken for the aging of '

concrete which con amount to a factor of up to 1.35 (Reference 8) of increased strength. Also, the average strength for structural steel is nearly a factor of 1.25 (Reference 9) higher than the minimum yield requirement specified by ASTM. While these factors above minimum code strength exist and contribute to structural margins, they were not used in the evaluation.

3) Equation (l) for hinge rotational capacity was used. This corresponds to support rotations of the order of 2 degrees with minimum cracking and no crushing or scabbing. To  ;

meet necessary performance requirements (i.e. halting propgating failures), larger rotations in the range of 5 to 12 degrees could be tolerated. Such rotations would lead to crushing, spalling and scabbing of the section Reference 7); however overall load carying capability is expected to remain intact. Experimental observations (Reference 10) suggest even further capability for well designed and well anchored slabs. Failure modes at such levels initially appear to be controlled by yielding in shear and flexture followed by membrane stretching until failure occurs, normally at the support edge of the stab.

Use of these larger rotational capabilities would have resulted in~ greater energy absorbing capabilities of the grid system.

4) The analysis used ACI 318-77 allowable shear stresses. A significant body of data suggests the existence of higher shear capabilities on the order of 10 Vf'c to 20 Vf'c (References 11-19). It is expected that the shear capabil- ,

! ities for these beams would tend to be in the higher end of I

l 30 l

the range since the majority of the beams are " deep".

Deep beams behave os tied-orches with significant re-serve capacity. .

5) in many cases, the analysis neglected the two-way resis-tance capability of the sicb. It is expected that the slab would contribute increased strength particularly at larger deformations.

6) The load was distributed directly under the dropped shield plug. In reality a more favorable food distribution would exist due to the load distribution capability of the slob.

7) No credit was taken for local energy dissipation associ-ated with any crushing of the shield plug or the immediate surface of the floor.

Local Structural Response Evoluotions Selected loads such as the portable radiation shield (cattle chute), various hatch covers and equipment were evoluoted to ossess the acceptability and potential consequences of postulated drops. The acceptance criteria were based on the capability of the concrete slabs to resist perforation, penetration, and underside scabbing.

Procedures recommended in References 20 and 21 were followed. Th3 modified National Defense Research Committee (NDRC) formula (Reference 22) was chosen because it has been shown to give the best fit with available experimental dato (References 23 and 24). The NDP.C formula for the depth of penetration, x (inches), of a solid cylindrical missile is given by:

x=( 4 KNWd (V)l.8 /(l000 d) )h for x/d62.0 (2) or x = (KNW (V)l.8 / (1000 d) + d for x/d22.0 (3) 31

where W = weight of the missile (pounds) d = diameter of missile (inches) .

V = impact velocity of missile (feet /second)

N = missile shape factor -

= 0.72 flat-nosed missiles

= 0.84 blunt-nosed missiles

= 1.00 spherical-nosed missiles

= 1.14 sharp-nosed missiles K = concrete penetrability factor

= 180/f (f'c = concrete compressive strength pounds / square inch The thickness of reinforced concrete needed to resist impact without perforation and scabbing are given by the following Army Corps of Engineers formulae which 3

can be used in conjunction with equations (2) and (3)(Reference 25).

ts/d = 2.12 + 1.36 (x/d) for 0.65dx/d dit.75 (4) tp/d = 1.32 + 1.24 (x/d) for 1.35 5 x/d 6.13.5 (5) where is = concrete thickness required to prevent scabbing tp = concrete thickness required to prevent perforation Equations (4) and (5) were later extrapolated for small values of x/d (Reference

39) giving, ts/d = 7.91 (x/d) - 5.06 (x/d)2 for x/d 50.65 (6) 32

tp/d = 3.19 (x/d) - 0.718 (x/d)2 for x/d$1.35 (7) .

i A 10 percent margin on thickness has been applied in the use of equations (6) thru (7) as recommended in Reference 20.

P Limited penetration and scabbing was predicted for the set of bounding heavy load drops considered; however, in no case were the concrete slabs predicted to be perforated.

i P

e

)

33

REFERENCES

1. Building Code Requirements for Reinforced Concrete, ACI 318-77, American Concrete Institute, December 1977.

2. ACl 349-76, Code Requirements for Nuclear Safety-Related Concrete Structures, Appendix C "Special Provisions for Impulse and Impactive Effects", American Concrete Institute,1976.

3. Kennedy, R. P., "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile impact Effects", Journal of Nuclear ,

Engineering and Design, Vol. 37, No. 2, May 1976.

4. Mottock, A. H., "Rotationni Capacity of Hinging Region in Reinforced Concrete Beams", Flexural Mechanics of Reinforced Concrete, ASCE 1965-50 (ACI SP-12), American Society of Civil Engineers,1965.

5. Corley, W. G., " Rotational Capacity of Reinforced Concrete Beams",

~

Journal of Structural Division, ASCE, Vol. 92, No. STS, Proc. Paper 4939, Oct.1976, pp.121-146.

6. " Design of Structures for Missile impact", Topical Report BC-TOP-9A, Bechtel Power Corporation, September 1974.

7. Structures to Resist the Effects of Accidental Explosions, TM5-1300, Department of the Army, Washington, D.C., July 1965.

8. Neville, A. M., Properties of Concrete, J. Wiley & Sons, New York,1975.

9. Design of Structures to Resist the Effects of Atomic Weapons - Strength of Materials and Structural Elements, TM5-856-2, Department of the Army, Washington, D. C., August 1965.

10. Personal communication between Professor William J. Hall and Howard A. .

Levin, October 5,1981.

34

ll. Wang, C. K. and Salmon, C. G., Reinforced Concrete Design, Intext Educational Publishers, New York,1973. ,

12. Ferguson, P. M., Reinforced Concrete Fundamentals, J. Wiley, New York,1973.

t I

13. Untrauer, R. E. and C. P. Siess, " Strength and Behavior in Flexure of Deep ,

Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 230, University i

of Illinois, Urbana, October 1961. l

14. Austin, W. J., et ci, "An investigation of the Behavior of Deep Members of Reinforced Concrete and Steel," Civil Engineering Studies Structural Research Series No.187, University of Illinois, Urbono, January 1960.

15. de Paiva, H.A.R., and C. P. Siess, " Strength and Behavior in Shear of Deep Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 231, University of Illinois, Urbana, Oct.1961.

16. de Paiva, H.A.R., and W. J. Austin, " Behavior and Design of Deep -

Structural Members -- Port 3 - Tests of Reinforced Concrete Deep Beams," Civil Engineering Studies Structural Research Series No.1974, University of Illinois, Urbana, March 1960. ,

17. Winemiller, J. R. and' W. J. Austin, " Behavior and Design of Deep Structural Members - Part 2 - Tests of Reinforced Concrete Deep Members with Web and Compression Reinforcement," Civil Engineering Studies Structural Research Series Report No.193, University of Illinois, Urbana, August 1960.

18. Newmark, N. M. and J. D. Haltiwonger, " Air Force Design Manual -

Principles and Practices for Design of Hardened Structures," AFSWC-TDR-62-138, December 1962.

3S

19. Crawford, R. E., et al, "The Air Force Manual for Design and Analysis of Hardened Structures," AFWL-TR-74-102, October 1974. j

20. Civil Engineering and Nuclear Power, Report of the ASCE Committeo on impactive and impulsive Loads, Vol. V,' American Society of Civil Engineers, September 1980.

21. Structural Analysis and Design of Nuclear Plant Facilities, American Society of Civil Engineers,1980.

22. Vossollo, F. A., Missile impact Testing of Reinforced Concrete Panels, HC-5609-D-1, Colspan Corporation, January 1975.

i 23. Stephenson, A. E., " Full Scale Tornado Missile impact Tests," Electric Power Research Institute, Final Report NP-440, July 1977.

24. Beth, R. A. and Stipe, J. G., " Penetration and Explosion Tests on Concrete Slobs", CPPAB Interim Report No. 20, January 1943.

25. Beth, R. A., " Concrete Penetration" OSRD-4856, National Defense Research Committee Report A-319, March 1945.

4 I

36

TABLEI

. RELATED APPLICABLE LOAD NRC NO. SAFETY FUNCTIONS IMPACT REGIONS CRITERIA l Accomplish Spent Fuel Cooling All Regions 111 in the Spent Fuel Pool and the Upper Containment Pool 2 Maintain Extended Core Cooling - All Regions IV Reactor Vessel Head Bolted in Place 3 Maintain Extended Core Cooling - All Regions IV Reactor Vessel Head Removed 4 Accomplish Reactor Shutdown, Regions 1-4 IV Depressurization and Core Cooling 5 Limit Radiological Dose at Regions I,12 & 13 i Site Boundary from impact of Spent Fuel to 1/4 of 10 CFR Part 100 l 6 Limit Kef t to Less than 0.95 Regions I,12 & 13 11 Crushing of Fuel

TABLE 2 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REClON FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. I #5. Dose New Fuel e Fuel Pool A. Structrual A. Damage to NUREG-0612 2.2.4 As described in Spent Fuel Bridge Crane Cote Analysis of rods in as guidelines the response Pool #6. Criticality Drop Onto many as 10 can be met to item 2.2.4, (see Table 1) e FPCCU Rocks (Gate) assemblies acceptable See Figure i Filter predicted solutions are Denim B. Dose Assess- available. The Hatch ment B. Calculated possible solu-Covers Whole Body tions will be C. Criticality Dose of 8.5 investigated e Miscel- Evaluation rem is slight. and an accept-laneous ly greater able solution loads as than NUREG- implemented.

much as 0612 limit of Stons 6.25 rem, but wcll within 10 CFR, Part 100, limit of 25 rem.

C. A Keff greater than 0.95 can-not be pre- .

cluded under worst case conditions.

TAILE 2 i

(continued)

FINAL APPLICAELC. LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTAPiE ACCEPTANCE ITEM NO. REMARKS

1. l. Fuel Pool D. Structural D. Liner pene- NUREG-0612 2.4.2 Cooling Analysis of tration--minor guidelines Drop to Pool leakage only, are met.

Floor (Gote) Therefore, no significant loss of pool inventory predicted.

1 e

e

. - - - . 7. . ---%< -,a-re- --- , _-r.-ec- --e.n., .r ,e, y e,,r ,-e v.-,e -. - r-.- ., .,-,-,--:--.*-..me- m i sm-m - 3 w- we i 3-- m - 2r-, -c-.

TABLE 2 (continued)

FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAtOLING OF ANALYSES RESULTS BAS:S FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. l #2, #3, and #4. New Fuel e Fuel Pool E. Systems E. Minor leakoge NUREG-0612 2.4.2 Shutdown ed Bridge Crane Gate Evoluotion from pool would guidelines Core Cooling have no ef f ect are met.

eFPCCU on Safety Func-l-iller tions 2,3, ed 4, Demin because (1) there Hatch is no equipment Covers related to these functions below o Miscellaneous the pool that S-ton load could be domoged and (2) minor weepoge would not result in flooding that could of fect brooder areas.

e A

TAELE 3 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RE*VLTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEFTABLE ACCEPTANCE ITEM NO. REMARKS

1. 2 #1. Fuel Pool N w Fuel e Fuel Pool Structural Yes, spent fuel NUREG-0612 2.4.2 Aux. Bldg. Cooling Bridge Crane Gate analysis of cooling ed guidelines 20e el. 5-ton load core cooling are rnet; no floor e FPCCU drop. Analyzed can be occom- equipment

1. 2, #3, and #4. Demin for local and plished. Impacted at See Figure 1 Shutdown and Hatch overoll floors below.

Core Cooling Covers response.

e Miscel-laneous 5-ton load

_ ~_ . _ , . -

TABLE 4 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. 3 #l. Fuel Pool New Fuel e New Fuel Systems . Yes, spent NUREG-06I2 2.4.2 Aux. Bldg. Cooling Bridge Crane Containers Evaluations fuel cooling guidelines Miscellaneous and core cool- are met Hatch #7, #3, and #4. e Miscellaneous ing con be Shutdown md foods os accomplished

. See Figure I Core Cooling much as 5 tons 1

TAELE 5 FINAL APPLICABLE LOADS ARE CONCLUSION tilC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE RECION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. 4 #1. Spent Fuel LPCS & o Hatch Systems Yes, spent NUREG-0612 2.4.2 Aux. Bldg. Pool Cooling RHR C Covers Evaluations fuel cooling guidelines LPCS & Equipment ed core cool- ore met RHR C ond Hatch e Miscel- ing can be Hotches #2, #3, and #4. Monorail laneous occomplished Shutdown ed loods os 97 el. Core Cooling heavy os i 14' el. 10 tons 139' el

. See Figure 2 t

- , - - - -- ., , , - a -. p ,-

TABLE 6 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAtOLING - OF ANALYSES RESULTS BASIS FOR RESPONSE REClON FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. 5 #1. Spent Fuel Containment Miscel- A. Systems A. Yes, Spent The intent of 2.4.2 A. Even though Containment Cooling Polar Crane laneous Evaluation fuel cooling NUREG-0612 FPCCU could Equipment Equipment con be oc- guidelines are potentially be Hotch Area Being Removed complished. met. lost to in-or installed contoinment See Figure I of Contain. fuel storage ment 208' el. pool, there is reasonable

1. 2 and #3. A. Systems A. Yes, RHR NUREG-0612 assurance that Core Cooling Evoluotion shutdown guidelines spent fuel cool-cooling are met- ing can be can be RHR shutdown provided os accomplished. cooling con be described in maintained. the response to NRC ltem 2.4.2.

s

~m  % w .e ---

9 , , e--" n <-a --

g -w , , - -, .cn .

l l

TABLE 7 FINAL LOADS ARE CONCLUSION NRC APPLICABLE BASIS FOR RESPONSE SAFETY HAtOLING OF ANALYSES RESULTS ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS REGION FUNCTIONS SYSTEM INTEREST PERFORMED A. Yes, spent The intent of 2.4.2 A. See REMARKS

1. 1. Spent Fuel Contoinment Drywell A. Systems

1. 6 fuel and RHR NUREG-0612 for Table 6 Drywell Cooling Polar Crone Head Evoluotion shutdown guidelines Head cooling con are met- B. Assumes that Storoge #2. Extended be occom- spent fuel and structural Area Core Cool- 8. Structural analysis will Anolyses plished RHR shutdown ing, RV cooling con be verif y no domoge See Figure i Head in maintoined, to equipment in Place the drywell.

B. Structural analysis of a drop of the drywell heod onto its storage loco-tion of 208'el will be per-formed to verify the assumption that domoge to equipment in the drywell

• within this region would not occur.

i TABLE 8 FINAL APPLICABLE LOADS ARE CONCLUSION NRC .

HANDLING OF ANALYSES RESULTS BASIS FOR RESPONSE SAFETY REMARKS.

FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO.

REGION A. Systems A. Yes, spent The intent of 2.4.2 A. See REMARKS

1. 7 #l. Spent Fuel Contoinment o RWCU for Table 6 Groting Cooling Polar Crone HX's Evoluotion fuel and RHR NUREG-0612 shutdown guidelines Area - NE quadrant of #2 and #3. o Miscellaneous cooling are met -

Equipment con be spent fuel containment Extended Core Cooling occomplished and RHR shut-See Fip re I cooling con be maintained.

4 o

e

" W #

' ' *~ 1-- -

e -

,--y e.

,=9v , , , , , , _ , , , _ _ , , . , _

TAELE 9 FINAL LOADS ARE CONCLUSION NRC APPLICABLE RESPONSE HAPOLING OF ANALYSES RESULTS. BASIS FOR SAFETY- ITEM NO. REMARKS FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE REClON A. Systems A.I Yes, Spent The intent of 2.4.2 A.I. See REMARKS

1. 8 #1. Spent Fuel Containment e Hotch for Table 6 Cooling Polar Crane Covers Evoluotion Feel Cooling NUREG-0612 RWCU con be oc- guidelines Heat Exchangers A.2. See Response

1. 2 and #3. e Heat complished are met -

Areo Spent fuel to item 2.4.2 Extended Core Exchangers B. Structural for basis for Anotyses A.2 Yes, RHR cooling con See Figure 1 Cooling occeptable result.

suction and be maintoined.

injectinn (FW) RHR shutdown lines are in cooling is not this region likely to be below 208' el. lost.

If piping integrity is lost, RHR shut-down cooling could be lost.

However,it was demonstrated based on structural onalyses and protection offorded by inter-vening structures

• and m..pererits that loss of '

these lines was inconceivable.

B. Structural analyses performed to support con-clusion in A.2 obove. Only scabbing pos- ,

sible.

TAE.E 10 FINAL APPLICABLE LOADS ARE CONCLUSION NRC HAtOLING OF ANALYSES RESULTS BASIS FOR RESPONSE SAFETY REMARKS FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO.

REGION A. Systems A. Yes, spe.it The intent of 2.4.2 A. See REMARKS

1. 9 #1. Spen' Fuel Containment e RV Head for Table 6.

Groting Cooling Polar Crone Insulatica Evoluotion fuel and RHR NUREG-0612 shutdown guidelines

. In SE quadrant #2. Extended cooling con are met -

Core Cool- be accom- spent fuel See Figure I ing, RV plished, and RHR shut-Heod in down cooling Place con be maintoined.

e

= m ---- r r - -v- " - - =- ----~+M r - - -' - ~+ -- ' -- - v - v

ens TAILE ll FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HANDLING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS The intent of

1. 10 #l. Spent Fuel Containment e RV Head A. Systems A. Yes, spent NUREG-0612 2.4.2 A. See REMARKS Cooling Polar Crone Evoluotion fuel and RHR guidelines for Table 6.

RV Head Storoge e Portable shutdown are met-Areo #3. Extended Rodiation cooling Spent fuel Core Cooling- Shield con be and RHR shut-See Figure i Heod Removed accomplished down cooling

• e RV Head con be maintained Insulation B. Structural B. Structural B. Assumes that eRWCU Analyses analysis of structural anoty-Filter o RPV head sis will verif y Demineral- drop onto its no domoge to izer storage loco- equipment in Hatches tion at 208' the drywell.

el will be performed to verify the assumption that damage to equipment in the dry- ,

well within this region ,

would not occur.

TABLE 12 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS The intent of

1. ll #1. Spent Fuel Containment e Portable A. Systems A. Yes, spent NUREG-0612 2.4.2 A. Sx REMARKS Groting Cooling Polar Crane Radiation Evoluotion fuel and RHR guidelines for Table 6.

- Area in SW Shield shutdown are met -

quadrant of #2 and #3. cooling can Spent fuel contoinment Extended Core e RWCU Filter be wcom- pool and RHR 20ff el. Cooling Demineralizer plished. shutdown cool-Hatch Covers ing can be See Figure I maintained.

1

-_ -.._y _ _ - - _ . . , _ _ _ _ _ my,, , , _ . _ , , ,_._ry.__,, , , , , . _ - , _ - _ _ , _ _ , s ,w ,. y ..__. ,_ ,

TABLE 13 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAMX.ING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. 12 #5. Dose Containment e Steam A. Structrual A. Domoge to NUREG-0612 2.2.4 As described in Dryer & Polar Crone Dryer Analysis of rods in as guidelines the response Fuel Storoge #6.Criticolity Drop Onto mcny as 10 can be met to item 2.2.4, Pool (see Toble 1) e Portable Rocks (Gote) assemblies occeptable Rodiation predicted solutions are See Figure i Shield B. Dose Assess- ovoitable. The ment B. Calculated possible solu-e Pool Whole Body tions will be Gates C. Criticality Dose of 8.5 investigated

< Evoluotion rem is sli.f t- and an occept-e Miscel- ly greater cble solution laneous than NUREG- implemented.

Equipment 0612 limit of 6.25 rem, but well within 10 CFR, Part

.100, limit of 25 rem.

C. A Keff greater than 0.95 con-not be p.e-cluded under .

worst cose conditions. .

TAE.E 13 (continued FINAL APPLICABLE LOADS ARE COrCLUSION PRC SAFETY HANDLING OF ANALYSES RESULTS BASIS FOR RESPOtGE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS fl. Spent fuel , D. Structural D. Liner pene- NUREG-0612 2.4.2 Cooling molyses of tration and guidelines are drop to pool scobbing pos- met-spent fuel floor (Dryer) sible, minor pool water (Appendix C) leokoge; no inventory and significant cooling con be loss of maintained.

pool inventory.

Rocks are in lower end of pool - would not be uncovered.

E. System E. Yes, spent The intent of Evoluotion fuel cooling iAJREG-0612 (Appendix B) con be g>idelines are occornplished. met - Spent fuel 4 cooling con be mainteined

.-- m. . - -, ,, , , - , - - ,, -, ~-

TAfLE 14 FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REGION FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

1. 13 #5. Dose Containment e RV Head A. Structural A. Yes, no NUREG-0612 2.3.4 Reactor Polar Crone Anolyses fuel domoge guidelines are Cell #6.Criticolity e Drywell (CE Analyses, predicted met-no fuel (See Table 1) Head see FSAR, domoge or See Figure i Table 9.l-7, crushing e Dryer NEDC-23566 predicted.

and GE letter, o Seporotor Smith to Dole dated 2/5/82).

e Portable TERA performed Radiation evoluotion to Shield assure that GE analyses bounded other postulated drops.

1. 2 and #3. A. Structural A. Yes, vessel NUREG-0612 2A.2 Extended Core Analyses integrity are met-Cooling (GE Analyses, maintained vessel see above) integrity maintained. ,

a e

h

---p ,-e------. , - _ . - , , - + - - , - . - + - - .+ y , ,,,, 7-. .-c -

3, gr , -- - -

• TAELE IS FINAL APPLICABLE LOADS ARE CONCLUSION NRC SAFETY HAPOLING OF ANALYSES RESULTS BASIS FOR RESPONSE REClON FUNCTIONS SYSTEM INTEREST PERFORMED ACCEPTABLE ACCEPTANCE ITEM NO. REMARKS

- # 14 fl. Spent Fuel Containment e Shroud A. Structural A. Yes, Liner pene- NUREG-0612 2.4.2 Separator Cooling Polor Crone Head / Analyses tration ed guidelines Storoge Separator scabbing are are rnet -

Pool #2 ed #3. possible; Spent fuel Extended Core e RWCU HX minor leakage edRHR See Figure I Cooling Hotch into drywell shutdown Covers would not of- cooling cm fect Spent be maintained.

fuel or RHR shutdown cool-ing. No electrical equip-ment in drywell required to operate to occomplish Sofety Functions 1, 2 or 3.

TABLE 16

SUMMARY

OF CONTROLLING STRUCTURAL BEHAVIOR RESULTING FROM POSTULATED HEAVY LOAD DROPS CONTROLLING MODE OF RESPONSE APPROXIMATE WEIGHT HANDLING OVERALL LOAD TONS SYSTEM STRUCTURAL LOCAL

1. Reactor Pressure 11 7 Polar Crane X Vessel Head (RPV)

2. Steam Dryer 40 Polar Crane X

3. Shroud Head / Steam 68 Polar Crane X Separator

4. Drywell Head 61.5 Polar Crone X

5. Portable Refueling 12 Polar Crane X X Shield

6. RPV Head insula- 10.5 Polar Crone N/A tion w/ Support Structure

7. Reactor Well/ Steam 3.5 Polar Crane X X Dryer Storage Area Gote

8. Upper Containment 3.5 Polar Crone X X Fuel Pool / Transfer Pool

9. Load Block 5.5 (M) Polar Crane N/A 1.0 (AUX)

10. RWCU Regenerative 15 Polar Crone X X HX Hatches (2)

11. RWCU Non- 15-17 Polar Crane X X Regenerative HX Hatches (3)

12. RWCU Filter Demin- 20 Polar Crane X X erolizer Hotebes (3)

r TABLE 16 (continued)

CONTROLLING MODE

. OF RESPONSE APPROXIMATE WEIGHT HANDLING OVERALL LOAD TONS SYSTEM STRUCTURAL LOCAL

13. New Fuel Shipping 1.5 New Fuel X Containers Bridge Crane

14. Fuel Pool and 3 New Fuel X Cleon Up Filter Bridge Crane Demineralization Hatch (2) ,

15. Spent Fuel Pool 3.5 New Fuel X X Gate Bridge Crone i

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J FIGURE 4 FEAVY LOAD DROP INTO REGION 8

O UPPER CONTAINMENT POOL COOLING SPENT FUEL POOL COOLING AVERAGE FULL CORE CORE DISCHARGE / HEAT DISCHARGE LOAD FUEL POOL FUEL POOL ,

COOLING AND COOLING AND CLEANUP SYSTEM CLEANUP SYSTEM RHR SYSTEM (BOTH LOOPS)

FIGURE S SAFETY FUNCTION NO. I SPENT FUEL COOLING f

EXTENDED CORE COOLING DIVISION I I 2 DIVISION 2 Ir4 DEPENDENT PATHS ALTERNATlVE ALTERNATIVE PATHS PATHS RHR "A" RHR "B" SHUTDOWtJ LPCS SHUTDOWN LPCI"C" COOLitJG MODE COOLING MODE I

MANUAL MANUAL RELIEF VALVE ADS RELIEF VALVE ADS OPERATION OPERATION RHRS RHRS

< SUPPRESSION POOL SUPPRESSION POOL COOLING MODE COOLING MODE FIGURE 6 SAFETY FUNCTION NO. 2 EXTEtOED CORE COOLING-HEAD IN-PLACE

. SYSTEMS SELECTED FOR PURPOSES OF EVALUATION *

• The systems shown in the figure are only those used to evoluote whether the safety function con be accomplished following a postulated load drop, i.e., oil systems available to occomplish the safety function ore not necessarily shown in the figure.

EXTEt4DED CORE COOLING DIVISION l DIVISION 2 p

PATHS ALTERNATIVE ALTERNATIVE PATHS PATHS RHR "A" LPCS RHR "B" LPCI"C" RHRS RHR5 CONTAINMENT CONTAINMENT COOLING MODE COOLING MODE FIGURE 7 SAFETY FUNCTION NO. 3 EXTEtOED CORE COOLING-HEAD REMOVED SYSTEMS SELECTED FOR PURPOSES OF EVALUATION *

• The systems shown in the figure are only those used to evoluote whether the safety function con be accomplished following a postulated lood drop, i.e., all systems available to occomplish the safety function are not necessarily shown in the figure.

l f-

SHUTDOWN AND CORE COOLING FROM POWER REQUIRED FUNCTIONS MONITORING INITIAL CORE EXTENDED DEPRESSURI-SCRAM $ TEM ZATION COOLING / MAKEUP PA AME ERS C ING I

RV ADS RPS LEVEL INDEPENDENT INDEPENDENT PATHS PATHS RCS CRDMS PRESSURE LPCS LPCl RHR "A" RHR "B" B

FIGURE 8 SAFETY FUNCTION NO. 4 SHUTDOWN POWER /COOLDOWN SYSTEMS SELECTED FOR PURPOSES OF EVALUATION *

• The systems shown in the figure are only those used to evoluote whether the safety function con be accomplished following o postulated load drop, i.e., all systems avoitable to accomplis;i the safety function are not necessarily shown in the figure.

... _ . -