ML18046A941

Forwards Addl Responses to NRC 801222 & 810203 Ltrs Re Control of Heavy Loads,Per Generic Ltr 81-07.Mods Required to Fully Satisfy NUREG-0612 Guidelines Have Not Been Completely Identified

ML18046A941
Person / Time
Site:	Palisades
Issue date:	09/23/1981
From:	Johnson B CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield 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 GL-81-07, GL-81-7, NUDOCS 8109300287
Download: ML18046A941 (20)

Text

consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Ml 49201 * (517) 788-0550 September 23, 1981 Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 US Nuclear Regulatory Commission Washington, DC 20555 1\ DOCKEI' DPR-20 -PALISADES PLANT -CONTROL OF HEAVY LOADS (GENERIC LETTER 81-07) NRC letters .dated December 22, 1980 and February 3, 1981 requested Consumers Power Company review of the Pa.lisades Plant controls associated with the ling of heavy loads. These documents requested a determination of the extent to which the guidelines of Enclosure 1 (:NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants") are presently satisfied at the Palisades Plant and the identifi.cation of what. c*hanges and modifications would be required in order to fully satisfy these guidelines.

Specifically, they requested the tion of interim actions provided by Enclosure 2 (Staff Position -Interim tions for Control of Heavy Loads) and additional information identified by Enclosure 3 (Request for Additional Information on Control of Heavy Loads). In addition, they requested confirmation that implementation of those changes and modifications determined necessary to comply with the guidelines commence as soon as possible without waiting on staff so that all such changes; beyond the above interim action will be completed within two years of submittal of Sec+,ion 2.3 of the above required report. If any changes or modifications that would be required to fully satisfy the guidelines of Enclosure 1 are sidered not necessary, justification for such determinations was requested.

Consumers*'

Power Company letter dated May 15, 1981 provided our confirmatH)D:

that the intent* of the five interim action items was.presently being met at the sades Plant and that total implementation of these required actions would te achieved by the start of our 1981 refueling outage. Until that time, Palisades would administratively control the use of all overhead handling systems to pre-clude movement of "Heavy Loads", as defined in Section 1. 2 of NPREG-0612, less the interim actions for the particular case in question are met. Consumers Power Company letter dated July 6, 1981 provided Part I to the NRC I requested report wh:i.ch ad.dressed Section 2.1 of NRC Generic Letter 81-07 dated / * '(D December 22, 1980. The evaluation provided in our <July 6, 1981 submitta.l did not identify any changes to be necessary but did indicate that Consumers Power Company would await completion of the integrated assessment of SEP modifications before Il!aking a final commitment to changes, modifications or implementation scliedule.**

1 ,

• -,* **. iPDR ' '-..........__J . .J ..... The attachment to this submittal provides Part II to the NRC requested report which addresses Sections 2.2, 2.3 and 2.4 of NRC Generic Letter 81-07. Consumers Power Company has identified the extent to which the guidelines of NUREG-0612 are met at the Palisades Plant. The changes and modifications that would be quired in order to fully satisfy the guidelines of NUREG-0612 have not been pletely identified at this time. In the responses to Sections 2.2, 2.3 and 2.4 where discussions concerning posed Technical Specification changes were suggested, it is Consumers Power Company's position that permanent Palisades Plant procedures provide the required restrictions or constraints to meet the requirements under consideration in the applicable area of discussion.

Brian D Johnson Senior Licensing Engineer CC Director, Region III, USNRC . NRC Resident Inspector

-Palisades CONSUMERS POWER COMPANY Palisades Plant Additional response to NRC letters dated December 22, 1980 and February 3, 1981 Control of Heavy Loads (Generic Letter Docket No 50-255 License No DPR-20 At the request of the Commission and pursuant to the Atomic Energy Act of 1954, and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations ther.eunder, Consumers Power Company submits .our response to NRC letters dated December 22, 1980 and February 3, 1981 entitled "C_ontrol of Heavy Loads (Generic Letter 81-07)". Consumers Power Company's response is dated September 23, 1981. R B DeWitt,* Vice President Nuclear Operations Sworn and subscribed to before me this 23rd day of September 1981 ..

Helen I Jackson County, Michigan My commission expires December 14, 1983 .... *": -.,,. .. , . . /,-: .:.. '* , '( ;: .. ,. 1*,_ .J-. l '

PALIS.ADES PLANT 2 .2 SPECIFIC REQUIREMEN'l'S FOR OVERHEAD HANDLING SYSTEMS. OPERATING IN THE VICINITY OF FUEL STORAGE POOLS 2.2-1 Identify by name, tYJJe, capacity and equipment designator, any cranes physically capable (ie, ignoring interlocks, mov.eable mechanical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool.

• RESPONSE:

1) Name: Fuel Building Crane, With Main Hoist and Auxiliary Hoist TYJle: Indoor Electric Overhead Traveling Bridge, Single Trolley Crane, Pendant Controlled Capacity:

Main Hoist Rating -100 Tons Auxiliary Hoist Rating -15 Tons Equip Designator:

L3, Serial No 31957 2) Name: Service Platform, With Fuel Handling Hoist and. Auxiliary Fuel Handling Hoist TYJle: Indoor Electric Traveling Bridge Capacity:

Auxiliary Hoist -1 Ton (With Load Cell at 1,750 lb) Equip Designator:

Service Platform 1 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 manently prevented from movement of the hook centerline closer than 15 feet to the pool.boundary, or by providing a suitable analysis ting that for any failure mode, no heavy load can fall into the fuel storage pool. RESPONSE:

The service platform and associated hoists are excluded in this area from the above catagory because they are used only for handling single fuel assemblies and reactor components that are not, by definition, classified as heavy loads. 2.2-3 Identify any crane listed in 2.2-1 above, which you have evaluated as having sufficient design features to make the liklihood of a*load drop extremely small for all loads to be carried and the basis for this ation (ie, complete compliance with NUREG 0612, Section 5.1.6 or partial compliance supplemented by suitable alternative or additional design features).

For each crane so evaluated, provide the load handling tem (ie, crane-load combination) information specified in Attachment

l. RESPONSE:

We have not performed a single faili.ireproof evaluation to substantiate the compliance of the fuel building crane .to NUREG 0612, Section 5.1.6. 2 2.2-4 For cranes identified in 2.2-1 above, not categorized according to 2.2-3, demonstrate that the criteria of NUREG 0612, Section 5.l, are satisfied.

Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria I through III, 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 (eg, 2, 3 or 4) from those identified in NUREG 0612, Section 5.1.2 have been selected.

RESPONSE:

Concerning the Fuel Building Crane, we have chosen option No 4 in Section 5.1.2 as our response.

The effects 9f drops of heavy loads in the spent fuel pool area was previously analized and submitted under.Appendix J of the Palisades Plant.Final Analysis ReJ;>ort Amendment

29. This analysis entitled "Evaluation of Postulated Cask Drop Accidents" is our basis for addressing NUREG 0612, Section 5.1 recommended Guidelines I, II, III and IV. In conjunction with this analysis, we have defined our heavy load. paths in the spent fuel pool area and have developed a plant procedure entitled "Movement of Heavy Loads in The Spent Fuel Area," Procedure No FHS-M-23.

Our analysis and this plant procedure satisfies the intent of NUREG 0612, Section 5.1.2 recommended guidelines.

b. If alternative.

2 or 3 is selected, discuss the crane motion tion imposed by electrical interlocks or mechanical stops and cate the circumstB.rices, if any, under which these protective devices may be bypassedor removed. Discuss any administrative procedures invoked to ensure proper authorization by bypass or.removal and pro"'.'* vide any related or proposed technical specification (operational and surveillance).provided to ensure the operability of such electrical interlocks or mechanical stops. RESPONSE:

This question doesn '*t apply to our fuel pool Gantry Crane response due to our not choosing either alternative 2 or 3 from NUREG 0612, Section 5.1.2. c. Where reliance is placed on crane operational limitations with spect to the time of the storage of c.ertain quantities of spent fuel at apecific post-irradiation decay times, provide present and/or proposed technical specifications and discuss tive or physical controls provided to ensure that these tions remain valid. 3 RESPONSE:

The only area in the spent fuel pool where reliance is placed on crane operation limitation with respect to time of storage is the north tilt pit. According to our procedure No FHS-M-23 entitled "Movement of Heavy Loads in the Spent Fuel Pool Area," Section 4.4, states that heavy load movement is limited in the following way: If fuel is stored in the north tilt pit, movement of the heavy load over the tilt pit shall not be allowed unless the fuel has decayed for a.minimum of 22 days with the charcoal Hiter.of the spent fuel pool area ventilation in service, or 77 days with the charcoal filter bypassed.

In addition, Technical Specification No 3.21 entitled "Movement of Shielded Shipping Cask in Fuel Handling Areas" specifically limits movement of the ten (10) ton shipping cask through the fuel handling areas in the same manner as stated in Section 4.4 of Procedure No FHS-M-23.

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 discus.s administrative or physical controls provided to ensure that these assumptions remain valid. RESPONSE:

Since Procedure NO FHS-M-23, Section 4.4 and Technical Specification No 3.21 include the limitations of heavy load movement in conjunction with the physical location of specific fuel modules at certain post-irradiation decay times, the response in Section 2.2-4c above, is also applicablefor this section. e. Analyses performed to demonstrate compliance with criteria I through III 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:

We have met the intent of NUREG 0612, Appendix A for fuel building load drop analyses.

Our analyses results are listed in the Palisades FSAR Appendix J. *. This topic is also addressed in the following Consumers Power Company letters submitted relative to the addition of new fuel racks: 1) CP Co letter dated 11/1/76 to Albert Schwencer, USNRC. 2) CP. .. Co .letter dated 2/8/77 to Albert Schwencer, USNRC. rp0881-0309b-46 4 PALISADES PLANT 2.3 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT 2.3-1 2.3-2 *2. 3-3 Identify by name, type, capacity and equipment designator, any cranes physically capable taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel. RESPONSE:

1) Name: Reactor Building Crane 2) Type: Polar Crane (Indoor Electric Overhead Circular Traveling Bridge, Single Trolley Crane) Capacity:

Main Hoist Rating -125 Tons Auxiliary Hoist Rating -15 Tons Equipment Designator:

Ll, Serial No 31909 Name: Reactor Building Jib Crane Type: Jib Crane Capacity:

1 Ton Equipment Designator:

LlA Justi-fy 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 the movement of any load, either directly over the reactor vessel or to such a location where, in the event of any load*-handling system failure, the load may land fn or on the reactor vessel. RESPONSE:

...

• The Reactor Building Jib Crane is exCluded because it will be limited to loads less than the 1,300 lb heavy load threshold liinit and their respective load paths will be restricted to prevent movement of loads over irradiated fuel or equipment required for safe shutdown or decay heat removal. I.dentify 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 I this evaluation (ie, complete compliance with NUREG 0612, Sec-tion 5.1.6 or partial compliance supplemented by suitable alternative.

rp0881-0359a-46 2.3-4 5 or additional design features).

For each crane so evaluated, provide the load-handling system (ie, crane-load combination) information specified in Attachment

1. RESPONSE:

We have not performed a single failure-proof evaluation to ate the compliance of the fuel building crane to NUREG-0612, Section 5. 1. 6. 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 III, provide a 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: a. Where reliance is placed on the installation and use of cal interlocks or mecl_ianical stops,* indicate the cirumstances under which these protective devices can 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:

Reliance has not been placed on electrical interlocks or mechanical stops. Plant procedures govern the.load path restrictions in the containment

.. b. Where reliance is placed on other site-specific considerations (eg, refueling sequencing), provide present or proposed Technical Specifications and discuss administrative or physical controls provided to.ensure the continued validity of such considerations.

RESPONSE:

According to our plant procedure number FHS:..M-24 titled, "Movement of Heavy Loads in the Containment Building Area (649 ELV)," a heavy load shall never be moved within 19 feet from the . center line of the reactor vessel unless covered by a specific handling procedure.

Our plant procedure number FHS-M-25 titled "Specific Requirements for Moving Heavy Loads Ins.ide Containment Near Reactor Vessel," identifies site-specific requirements for movement of the reactor missile shields and other heavy loads. over and near the reactor vessel. c. Analysis performed to demonstrate compliance with Criteria I through III should conform with the guidelines of NUREG 0612, rp0881-0359a-46 6 Appendix A. Justify any exception taken to these guidelines and provide the specific information requested in Attachments 2, 3 or 4 as appropriate for each analysis performed.

RESPONSE:

We feel the Palisades Plant meets the objectives of NUREG. 0612, Section 5.1 recommended guidelines.

The potentiai for a drop on any safety-related equipment which would hamper either safe shutdown of the reactor or cooling of the reactor is extremely small. According to Palisades FSAR, "the containment polar crane shall not be used to transport loads over the primary coolant *system if the temperature of the coolant-or steam in the izer exceeds 225°F." This guideline assures no load will be carri.ed in the containment building by the polar crane while the plant is on line. During refueling, the only reactor cooling system running is the shutdown cooling system. If a heavy load did drop on the shutdown cooling system while the plant is off line, plant procedure OMP 17 covers alternative systems available to the operator.

Any significant radiation that may be released by a heavy load drop will actuate the radiation detectors and containment isolation will occur as covered in the Palisades Plant Specifications.

rp0881-0359a-46 2.4 2.4-1 2.4-2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL OR SPENT FUEL POOL COOLING . . 7 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 (ie, complete compliance with NUREG 0612, Section 5 .1. 6 or partial compliance supplemented by suitable alternative or additional design features).

For each crane so evaluated, provide the load-handling system (ie, crane-load combination) information specified in Attachment

1. RESPONSE:

We have not performed a single failure-proof evaluation to substantiate the compliance of the reactor building polar crane to NUREG 0612, Section 5.1.6 For any cranes identified in 2.1-1 not designated as single-failure proof in 2.4-1, a comprehensive hazard evaluation should be provided which the following information:

a. The presentation in a matrix format of all heavy loads and potential impact areas where damage might occur to safety-re!l.ated equipment.

Heavy loads identification should include designation and weight or cross-reference to information provided in 2.1-3c. Impact areas should be identified by construction zdnes and elevations or by some other method such that the impact area can be located on the plant general arrangement drawings.

Figure 1 provides a typical matrix. RESPONSE:

A matrix for each crane identified in 2.1-1 and not excluded in 2.1-2 or 2.4-1 is contained in Table 1. Each matrix presents typical heavy loads and potential impact areas where damage might occur to safe shutdown related systems. The hazard eliminations and justifications shown for each load/impact area combinations refer to one of the five (5) categories listed below: 1. Crane travel for this area/load comb.ination prohibited by electrical interlocks or mechanical stops. and is identified by the letter "A." 2. System redundancy and separation precludes loss of capability of system to perform its safe shutdown related function following this load drop in this area and is identified by the letter "B." 3. Site specific considerations eliminate the need to consider load/equipment combination and is identified by the letter "c *II .rp0881-0359b-46 8 4. Likelihood .of handling system failure for this load is extremely small (ie, Section 5.1.6, NUREG 0612 satisified and is identified by the letter "D." 5. Analysis demonstrates that crane failure and load drop will not damage safe shutdown related systems and this category is identified by the letter "E." b. For each interaction identified, indicate which of load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, mechanical stops and/or electrical interlocks or other site-specific considerations.

Elimination on the basis of the aforementioned considerations should be supplemented by the following specific information:

(1) For load/target combinations eliminated because of separation and redundancy of safety-related equipment, discuss the basis for determining that load drops will not affect continued system operation (ie, 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 sion 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 fo:r verify:i.ng that interlocks are restored to operability after operations which require bypassing have been completed.

(3) Where load/target combinations are eliminated.on the basis of other, site-specific considerations (eg, maintenance sequencing), provide present and/or proposed Technical Specifications and discuss administrative procedures or physical constraints invoked to ensure the continued vafidity of such considerations.

RESPONSE:

The following is a discussion addressing the related equipment and/or systems identified in the load/impact area matrix. (1) Spent Fuel Pool Pumps rp0881-0359b-46 The hazard elimination category chosen was "b.11 If the spent fuel pool pumps are impacted, sufficient makeup would be available from the safety injection and refueling water storage tank or the fire protection system to preclude uncovering of the fuel.

9 (2) Cooling Water Return Line (From Spent Fuel Pool Cooling System) . The hazard elimination category chosen was "b." If the cooling water return line to the spent fuel pool was impacted, sufficient makeup would be available from the safety injection and refueling water storage tank or the fire protection system to preclude uncovering of the fuel. (3) Spent Fuel in North Fuel Tilt Pit The hazard elimination category chosen was "c." According to our plant procedure number FHS-M-23 titled "Movement of Heavy Loads in the Spent Fuel Pool Area," Section 4.4, heavy load movement over the spent fuel pit is limited. Section 4.4 states that if fuel is stored in the north tilt* pit, movement of the heavy load over the tilt pit shall not be allowed unless the fuel has decayed for a minimum of 22 days with the charcoal filter of the spent fuel pool area ventilation in service or 77 days with the charcoal filter bypassed.

(4) Two Fuel Bundles in South Fuel Tilt Pit The hazard elimination category chosen is "c" because ther.e are administrative procedures regulating heavy load travel over the mechanical tilt pit while fuel is present. (5) Spent Fuel Pool The hazard elimination categories chosen were "a" and *11 II e. A system of trolley and bridge electrical interlocks normally prevent the trolley from moving over the storage pool. This is listed in our FSAR Amendment No 17, Item 13. Plant procedure numbers FHS-M-23 defines when these interlocks can be bypassed.

Appendix J of the Final Safety Analysis Report, Amendment.29 titled, "Evaluation of Postulated Cask Drop Accidents" covers the analysis of dropping a heavy Toad in the spent fuel pool. (6) Low-Pressure Injection Line High-Pressure Injection Line rp0881-0359b-46 Redundant High-Pressure Line According to the Palisades FSAR, "the containment polar crane shall not be used to transport loads over the primary coolant system if the temperature of the coolant or steam in the pressurizer exceeds 225°F." This guideline assures no 10 loads will be carried in the containment by the polar crane while the plant is on line. The low-pressure injection line is keeping the reactor cool during shutdown.

Both the pressure and redundant high-pressure injec_tion lines are immediately available to keep the shutdown reactor cool if the low-pressure line were crushed by a heavy load. Our analysis noted that all three lines run fairly close to each other; ie, 48 inches center-to-center.

In the unlikely event a heavy load should drop on all three lines, our plant procedure OMP 17 provides other alternate shutdown cooling systems (ie, the boric acid flow path which approaches the reactor from the opposite side of containment).

(7) Reactor: No heavy load drop analysis has been performed to date for the reactor vessel. c. For interactions not eliminated by the analysis of 2.4-2b 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 evaluation (ie, complete compliance with NUREG 0612, Section 5 .1. 6 or partial compliance supplemented by suitable al ternatiye or additional design features).

For each crane so provide the load-handling system (ie, crane-load combination) information specified in Attachment

1. RESPONSE:

Neither the polar crane nor the fuel pool gantry crane comply with NUREG .0612, Section 5.1.6; therefore, thi.s section is not applicable.

d. For interactions not eliminated in 2.4-2b or 2.4-2c above, demonstrate using appropriate analysis that damage would no.t 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 analys.is 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 seism,ic accelerations equivalent to those of a safe shutdown earthquake (SSE). rp0881-Q359b-46 RESPONSE:

A seismic analysis has not been performed for the fuel building loaded crane. For the crane, the stresses resulting from the following seismic loads combined with normal operating stresses in no case exceed the yield point of the component materials.

The seismic load was calculated as 60 percent of the dead load applied in any 11 horizontal direction and 15 percent of the dead load applied in either direction vertically.

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

We have met the intent of NUREG 0612, Appendix A, for fuel building analyses.

Our analysis results are listed in the Palisades FSAR, Appendix J. (3) The Information Requested in Attachment 4 rp0881-0359b-46 RESPONSE: (a) Initial Conditions/Assumptions Discuss the assumptions used in the analysis including:

-Weight of Heavy Load: 25 Tons -Impact Area of Load: 11.67 Ft 2 -Drop Height and Thicknesses:

Air Water Total Location (Ft) Q!L Q!L Thickness Fuel Pool 2.5 39.75 42.25 30", Plus 12 11 Wall Railroad Track 25.5 25.5 18 11 Washdown Pit 16.5 . 16 .5 18" . Elev 649 1.5 1.5 12", Near 54 11 Wall -Drop Locations:

Cask Storage in Fuel Pool On Railroad Track Cask Washdown Pit At Elevation 649 to the south of Hatch in Fuel Pool Area -Assumptions regarding credit taken in the analysis for the action of impact limiters:

None -Assumptions regarding drag forces caused by the environment:

Drag in water -0.886

-Load combinations considered:

25-ton cask with 46.25-inch diameter impact section. -Material properties of steel and concrete:

Concrete:

F = 3000 psi c Steel: F = 40 KS! y (b) Method of Analysis 12 Provide the method of analysis used to demonstrate that sufficient load-carrying capability exis.ts within the wall(s) or floor slab(s). Identify any computer codes employed and provide a description of their capabilities.

If test data was employed, provide it and describe its applicability.

RESPONSE:

The method described in Bechtel Topical Report BC-TOP-9 was used, without computer codes; (c) Conclusion Provide an evaluation comparing the results of this analysis with . Criteria III and IV of NUREG 0612,.-Section 5 .1. Where safe shutdown equipment has a ceiling or wall separating it from an overhead handling system, provide an evaluation to demonstrate that postulated load drops do not penetrate the ceiling or cause secondary missiles that could prevent a safe shutdown system from performing its safety function.

RESPONSE:

There is no damage sufficient to cause structural damage. When the cask drop analysis was calculated in the cask storage area, a 2 .5-foot a*ir drop and 40. 5-foot drop through water resulted in a penetration to the bottom of .495 inches; not enough to cause spalling.

As another example, when the 25-ton cask was analyzed to drop in the cask wash down pit, a 16.5-foot air drop resulted in a penetration of .33 inches which.won't cause spalling.

Redundant systems in the containment and the spent fuel pool were analyzed in Sections 2.2 and 2.3. rp0881-0359b-46

(/) ro (]) c+ t--' 0 1-1) \JI LOAIB CRANE LOCATION IMPACT AREA TABLE 1 FUEL BUILDING CRANE -MAIN HOIST FUEL BUILDING AREA NO 1 .,_., ______ ..,.. _____

Hazard Safety-Related Elimination Elevation Equipment

__ y. ______________

.............

""V"' .... ....,,..,.

__ __, Spent Fuel Shipping Cask (50,000 Lb) Pool Divider Gates (2,300 Lb) 1----.-Fuel Transfer Canal Valve ( 2, 300 Lb) 590 612 1. 590'. 590' Spent Fuel Pool Pumps Cooling Water Return Line (From Spent Fuel Pool Cooling System) Spent Fuel in Fu4ure Fuel Tilt Mech Pit Two (2) Fuel Bundles -South Tilt Pit b b c c* !-ShieldB-1-oc_k_s

__ _.-! .....

(5,300 Lb) Spent Fuel -*-------------------. 649' Pool e ' a . Fuel Transfer Carriage (2,000 Lb) --.,... __. . ..,.. .. ....... --Irradiated Specimen Shipping Cask (20,500 Lb) NOTE: All Loads Can Pass Over All Systems

.. mcrrnsu""-""*

Ott£,..a,m_..._,.,,......,,""'*"'""""""....,,_""'"':iwoe . .,.. * ...,_,.._.,,..,www_w.._

___ i;,,,..cwww_,_.._..,.,._.,., ....,.,,L.,.

______ _,_mn::_,,.,.,.,....,...._.,._,_.,j ro (J) c+ [\) 0 ...., V1 LOAIB CRANE LOCATION IMPACT AREA t WWWUWW . WWW UWWWD l Plant Equip I PCP Rotor L'::5: --I lt'ailed Fuel I . Container (Unknown)

In-Core Cask Spent Resin Cask ( 20 , 500 Lb) Elevation 590' 612 1 590' 590' TABLE 1 (Contd) FUEL BUILDING CRANE -MAIN HOIST FUEL BUILDING AREA NO 1 Safety-Related Equipment

• spent Fuel Pool Pumps Cooling Water Return Line (From Spent Fuel Pool Cooling System) Spent.Fuel in Future Fuel Tilt Mech Pit Two (2) Fuel Bundles -South Tilt Pit Hazard Elimination b b c c ------..... ----11-----------

..... ..y..--...................

__ .._ ____ ..,_ ________ ....d I In-Core Cask Spent Filter Cask (20,500 Lb) ---*---New Yuel Shipping Container (5,000 Lb) *-------------------

Crane Load Block (Unb10wn) 649' Spent Fuel Pool NOTE: All Loads Can Pass Over All Systems e, a Ul ::r ro ro c+ w 0 1-1) \Jl CRANE LOCATION ... IMPACT AREA LOAIB Pool Divider Gates (2,300 Lb) -*-----Fuel* Transfer Canal Valve (2,300 Lb) ----.--I Shield Blocks (5,300 Lb) ------Irradiated Specimen Shipping Cask (20,500 Lb) ----*--In-Core Cask Spent Resin Cask (20,500 Lb) t. Elevation

"' 590' : -612 1 590' -590' -649' -NOTE: All --i . ! ' . I ' ; i i : I I ' ' I ' . I Corttd} i : . ' i i. ' : FUEL .. ;BUILD ING CRANE--I ' BUILDING AREA NO 1 ! ; ! Safety-Related Hazard : Elimination

'.Equipment*

• r.,,.+,.,..Q-orv

., i ; ' i b Spent: Fuel ' I Pool Pumps : ' : Cooling Water I Return (From! : ; b Spenti Fuel Pool Svstem) l CoolihP.'.

Spent* Fuel in Fuel c Tilt Mecp Pit ; Two (2) Fuel Bundles '-P.outh c Tilt Pit i . Spent Fuel e, a Pool Loads Can Pass Over All Systen;is

: i ... i I ' ' ' ' I AuXILIARY

• HOIST : : ' ' ' : ; ; : ' ' ' i ' : . -I ; I : i ; ' ! *-* ' : : i : i I ' ' ' ' ' i i ' : ' ' ' ., I ' ' .. , .. *** ! j .. ***r ' : I ; : *' '---' * ' ! ' ; ,. ; ' .. .. .. I ; *, ' ... ; ' ' " ; ' ' i ... I Ul <D <D c+ V1 LOADS CRANE LOCATION IMPACT AREA In-Core Cask Spent Filter Cask (20,500 Lb) New Fuel Shipping Container (5

Lb) J ---.-.. ----Fuel Transfer Carriage (2,000 Lb) Elevation 590' 612 1 590' 590' TABLE l (Contd) FUEL BUILDING CRANE *..,. AU'.AILIARY HOIST FUEL BUILDING AREA NO l Safety-Related Equipment Spent Fuel Pool Pumps Cooling Water Return Line (From Spent Fuel Pool Cooling System) Spent Fuel in Fnture Tilt Mech Pit .Two ( 2) Fuel Bundles -Soutli Tilt Pit. Hazard Elimina'eion( . f'a+<>uorv b b c c ------... ----

...... 649' Spent Fuel Pool e, a ----*---NOTE:* All Loads Can Pass Over All Systems Cfl ::r Cl) Cl) c+ Vl 0 H) LOADS . CRANE LOCATION IMPACT AREA Reactor Vessel Head (133 Tons) Core Supj?ort Barrel (1)!1,000 Lb) **uiJiJer-rnter:na:rs**--( 43, 000 Lb) ----.. ---.. *TABLE 1 (Contd) REACTOR BUILDING POLAR CRANE CONTAINMENT BUILDING ALL AREAS OUTSIDE THE 19-FOOT RADIUS AROUND THE REACTOR VESSEL Hazard Elevation Safety-Related Elimination Equipment 590'* 590 I 590' Low-Pressure Injection Line High....;.

Pressure Injection Line Redundant Pressure Injection Line Categorv b b b CONTAINMENT BUILDING ALL AREAS WITHIN THE 19-FOOT RADIUS AROUND THE REACTOR VESSEL Safety-Related Elevation Eliminati:

Equipment Category 649' Reactor None Vessel Inservice Inspection Tool (20,000 Lb) ------..... ----

________ .._. Reactor Coolant Pump; Rotor and Impeller (33,500 Lb) ---------Missile Shielus

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_ Crane Loau Bloch (Unknown)

NOTE: All Loads Can Pass Over All Systems . '