ML20081D272

Control of Heavy Loads at Nuclear Power Plants - Catawba Nuclear Station Units 1 & 2 (Phase Ii), Draft Technical Evaluation Rept

ML20081D272
Person / Time
Site:	Catawba
Issue date:	10/31/1983
From:	Shaber C, Stickley T EG&G, INC.
To:	NRC
Shared Package
ML20081D276	List: ML20081D272
References
CON-FIN-A-6457, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8311010109
Download: ML20081D272 (27)

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CONTROL OF HEAVY LOADS AT NUCLEAR PCWER PLANTS CATAWSA NUCLEAR STATICN UNITS 1 AND 2 (PHASE II)

Docket No. [50-413]

[50-414]

Author C. R. Shaber Principal Technical Investigator T. H. Stickley Published October 1983 EG&G Idano, Inc.

Idaho Falls, Icaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6457 XA Copy Has Been Sent.to.fDR

'@y$1(6/6/64

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ABSTRACT The Nuclear Regulatory Cem. mission (NRC) has requested that all nuclear plants, either operating or under construction, submit a response of consistency with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." EG&G Idaho, Inc., has contracted with-the NRC to evaluate the responses of those plants presently under construction. This report contains EG&G's evaluation and recommendations for Catawba Units 1 and 2 for the requirements of Sections 5.1.2, 5.1.3, 5.1.5, and 5.1.6 of NUREG-0612 (Phase II). Section 5.1.1 (Phase I) was covered in a separate report [1].

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EXECUTIVE

SUMMARY

Catawba Nuclear Station Units 1 ano 2 is not :ctally consistent with the guidelines of NUREG-0612. In general, inconsistencies exist in the following areas:

o Submittal of promised analyses for 5 hoists in the Auxiliary area have not been received.

o Two postulated load drops stated, "to envelope all others" are under analysis for the Reactor Buildings. These were promised for September but have not been received for review at present, o A list of 24 hoists in'the Reactor Building are identified as handling heavy loads over vital equipment. Action, or risk control for these heavy .lcads are not acdressed.

The main report contains recommendations which will aid in making the above items consistent with the appropriate guidelines.

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CONTENTS AESTRACT ....... ............... ............. ... . ........ ........ 11 EXECUTIVESUMMARIY................................................... iii

1. INTR 00VCTION'...................... .................. ... ..... 1 PurposeofReview..............................)..........

1.1 1 4

l 1.2 Generic Background ........................................ 1  ;

1.3 Plant-Specific Background ................. . ... ... ..... 3

2. ' EVALUATION ' AND RECOMMENCATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.1 Overview ............. ........ ... ....................... 4 2.2 Heavy Load Overhead Handling Systems ...................... 4 2.3 Guidelines ............................ . ....... ......... 4

3. CCNCLUDING

SUMMARY

....................... .... ................. 21 3.1 Guideline Recommendations'........... ... . ............... 21 3.2 Additional Recommendations .............. ....... ......... 21 3.3 Summary ............................... ................... 22

4. REFERENCES ... .................... ........... ....... ... ..... 23 TABLES 2.1 Nonexempt Heavy Load-Handling Systems ........................... 5 I

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CCNTROL OF HEAVY LCA05 AT NUCLEAR POWER PLANTS CATAWEA NUCLEAR STATION UNITS 1 AND 2 (PHASE II)

1. INTRCOUCTION 1.1 Purpose of Review This technical evaluation report cocuments the EG&G Icabo, Inc.,

review of general load-handling policy and procedures at Catawba Nuclear Station Units 1 and 2 (Catawba 1 and 2). This evaluation was performed with the objective of assessing conformance to the general load-handling guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" [2], Sections 5.1.2, 5.1.3, 5.1.5, and 5.1.6.

This constitutes Phase II of a two phase evaluation. Phase I assesses conformance to Section 5.1.1 of NUREG-0612 anc was documented in a separate report [1].

1.2 Generic Backcround Generic Technical Activity Task A-36 was establishec by the U.S.

Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants o assure.the safe nandling of heavy loads and to recommend necessary changes to these measures. This activity was initiated.by a letter issued by the NRC staff on May 17, 1978 [3], to all power reactor applicants, requesting information concerning the control of heavy loads near spent fuel.

The results of Task A-36 were reported in NUREG-0612, " Control of Heavy Loads at Nuclear Power. Plants." The staff's conclusion from this evaluation was that existing measures to control the handling cf neavy loacs at coerating plants, althougn provicing protection fecm certain potential problems, de not adeauately cover the major causes of load-hancling accidents and should be upgraded.

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In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two phase objective using an accepted aoproach or prctec-ion chilosophy. The first portion of the objective, achieved through a set of general guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load-handling systems at nuclear power plants are designed and op.erated such that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. The second portion of the staff's objective, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ensure that, for load-handling systems in areas where their failure might result'in significant consequences, either (a) features are provided, in addition to those requiced for all load-handling systems, to ensure that the potential for a load drop is extremely small (e.g., a single-failure proof crane) or (b) conservative evaluations of load-handling accidents indicate that the potential consequences of any load drop are acceptably small. Acceptability of accicent

. consequences is quan ified in NUREG-0612 into four accident analysis evaluation criteria as follows:

o " Releases of radioactive material that may result from camage to spent fuel based on calculations involving I

accicental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses shoulc show that coses are equal to or less than 1/4 of Part 100 limits);

o " Damage to fuel and fuel storage racks based on calculations involving accicental dropping of a postulated heavy load does not result in a configuration of the fuel such that k,ff is larger than 0.95; o " Damage to the reactor vessel or the scent-fuel pool based on calculations of damage following accicental dropping cf a postulated heavy load is limited so as nct to result in 2

water leakage that.could uncover the fuel. (makeup water provided to overcome leskage should be from a borated source of adequate concentration if the water being lost.is borated); and o " Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental droppinglof a postulated heavy load, will be limited so as

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not to result in loss of recuired safe shutdown functions."

The approach used to cevelop the staff guidelines for minimi:ing the potential for a load' crop was based on defense in depth. This plan includes proper operator training, equipment design, and maintenance coupled with safe load paths and crane interlock cevices restricting movement over ciritical areas.

Staff guidelines resulting from the foregoing are tabulated in Section 5 of NUREG-0612.

1.3 plant-Soecific Backcround On Decemcer 22, 1980, the NRC issued a letter [4] to Duke Power Company, the applicant for Catawba 1 and 2 requesting that the applicant review provisions for handling and control of heavy loads at Catawba 1 and 2, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guidelines. Duke Power Company proviced responses to this request on September 24, 1981 [5], July 1, 1982 [6], October 21, 1982 [7].

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water leakage that.could uncover the fuel, (makeup water provided to overcome-leakage should be from a borated source of adequate concentration if the water being lost.is borated); and o " Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of recuired safe shutdown functions."

The approach used to develop the staff guidelines for minimi:ing the potential for a load ' crop was based on defense in depth. This plan includes proper operator training, equipment design, and maintenance coupled with safe lead paths and crane interlock cevices restricting movement over ciritical areas.

Staff guidelines resulting frcm the foregoing are tabulated in Section 5 of NUREG-0612.

1.3 Plant-Soecific Backcround On December 22, 1980, the NRC issued a letter [a] to Duke Power Company, the applicant for Catawba 1 and 2 requesting that the applicant review provisions for handling and control of heavy leads at Catawba 1 and 2, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guidelines. Duke Power Company proviced responses to this request on September 24, 1981 [5], July 1, 1932 [6], October 21, 1982 [7].

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2. EVALUATION AND RECOMMENCATIONS 2.1 Overview The following sections summart:e Duke Power Company's review of heavy

load handling at Catawba 1 and 2 accompanied by EG&G's evaluation, i

conclusions, and recommendations to the applicant for making the facilities more consistent with the intent of NUREG-0612. -

2.2 Heavy Load Overhead Handlina Systems Table 2.1 presents tne applicant's list of overhead hancling systems which are subject to the criteria of NUREG-0612. The applicant has indicated that the weight of a heavy load for the facilities as 1500 pounds per the NUREG-0612 definition.

2.3 Guidelines The basic guidelines of NUREG 0612 for Phase II evaluati'ns a are quoted and followed with: A, the oplicants statements, B, EG&G evaluations, and C, recommendations. The criteria includes guideline 5.1.4 for Soiling Water Reactors only and 5.1.6 an alternative choice to other guidelines. Catawba 1 and 2 a PWR plant needs to show consistency with guidelines 5.1.2, 5.1.3, 5.1.5, and if the alternative is chosen 5.1.6, in the Phase II.

2.3.1 Scent-Fuel Pool Area [NUREG-0612, Article 5.1.2]

(1) "The overheac crane and associatec lifting devices used for handling heavy loads in the spent-fuel pool area should satisfy the single-failure proof guidelines of Section 5.1.6 of this recort.

OR (2) "Each of the following is provided:

(a) Mechanical stops or electrical interlocks snould be providea that prevent movement of the overhead crane load block over or within 15 feet hori:ontal (a.5 meters) of the spent-fuel pool. These mechanical 4

TABLE 2.1. OVERHEAD HANDLING SYSTEMS SUBJECT TO NUREG 0612 (Handles Heavy Loads >1500 Pounds)

Auxiliary 5uildine and Scent Fuel Pool l Has Vital Equipment Has Spent in Orop Fuel in GA No. Crane or Hoist (monorail or Jib) Size Area Oroo Area l

A001A Fuel Handling Bridge 125 Ton . / l A0018 Fuel Handling Auxiliary Hoist 10 Ton /

A002A Fuel Handling Bridge 125 Ton / l A0029 Fuel Handling Auxiliary Hoist 10 Ton /

A003 Containment Spray Pumo Monorail 5 Ton /

• A004 Containment Spray Pump Monorail 5 Ton /

A005 Residual Heat Removal Pump 5 Ton /

Monorail A006 Containment Spray Pump Monorail 5 Ton /

A007 Containment Spray Pumo Monorail 5 Ton /

A008 Residual Heat Removal Pump 5 Ton /

Monorail A009 Motor Driven Auxiliary Feedwater 5 Ton / (Study in Progress)

Pump Hoist f A010 Auxiliary Feedwater Pump Hof st 5 Ton / (Study in Progress)

A011 Miscellaneous Equipment Manoratl 3 Ton /

A012 Centrifugal Charging Pump 5 Ton /

Monorail A013 Centrifugal Charging Pump 5 Ton /

Monorail A014 Miscellaneous Equipmen: Manorafi 5 Ton /

A015 Reciprocating Pumo Monorail 6 Ton /

l A016 Safety :njection Pump Monorail 5 Ton /

A017 Safety Injection Pump Monorail 5 Ton /

A018 Centrifugal Charging Pump 5 Ton /

Monorail A019 Centrifugal Charging Pump 5 Ton /

Monorail A020 Reciprocating Charging Pump 6 Ton /

Monorail A021 Safety Infection Pump Monorail 5 Ton /

A022 . Safety Injection Pump Monorail 5 Ton /

A023 Miscellaneous Equipment Monorail 5 Ton /

A024 Miscellaneous Equipmen: Monorail 5 Ton /

A025 Seal Water Heat Exchanger 3 Ton /

Monorail 5

TABLE 2.1. (continued)

Auxiliary Buildino and Soer.t Fuel Pool Has Vital Equipment Has Spent in Orop Fuel in

.GA No.- - Crane or Hof st (Monorail or Jib) Size Area Oroo Area A026 Seal Water Heat Exchanger 3 Ton /

Monorail-A027 Turbine Driven Auxiliary 10 Ton Study in Progress Feedwater Pump Monorails A028 Turbine Driven Auxiliary 10 Ton Study in Progress Feedwater Pump Menorail A036A Fuel Manipulator Bridge Crane 10 Ton /

A0368 Fuel Manipulator Crane Auxiliary 2 Ton /

A037A Fuel Manipulator Bridge Crane 10 Ton /

A037B Fuel Manipulator Crane Auxiliary 2 Ton /

A038 Component Cooling Pump Monorail 3 Ton /

.A039 Component Cooling Pump Monorail 3 Ton /

-A040 Component Cooling Pump Monorail 3 Ton /

A041 Component Cooling Puma Monorail 3 Ton /

A042 Componen: Cooling Pump Nonorail 3 Ton / -

A043 Component Cooling Pump Monorail 3 Ton /

A044 Component Cooling Pump Monorail 3 Ton /

A045 Component Cooling Pump Monorail 3 Ton- /

-A046 Component Cooling Puma Monorail 3 Ton / (Study in Progress)

A047 Component Cooling. Pump Monorail 3 Ton /

A048 Comconent Cooling Pump Monorail 3 Ton /

A0d9 Component Cooling Pump Monorail 3 Ton /

A050 Component Cooling Pumo Manorail 3 Ton /

A051 Componen: Cooling Pump Monorail 3 Ton /

A052 Fuel Pool Cooling Pump Monorail 10 Ton /

A053 Fuel Pool Cooling Pump Manorail 10 Ton /

A056 Fuel Pool Cooling Pump Monorail 2 Ton /

A057 Fuel Pool Cooling Pump Monorail 2 Ton /

A059- Fuel Pool Cooling Pump Monorail 2 Ton /

-A075 Diesel Generator Bridge 2 Ton /

A076 Diesel Generator Bridge 2 Ton /

A077 Diesel Generator Bridge 2 Ton /

A078 01esel Generator Bridge 2 Ton /

ACS3 Main Steam Isolation Valve Heist 7-1/2 Ton /-

A034 Main Steam Isolation Valve Hoist 2 Ton /

A035 Main Steam Isolation Valve Hois: 2 Ton /

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TABLE 2.1. (continued)

Auxiliary Suildinc and Scent Fuel Pcol Has Vital Equipment Has Spent in Orop Fuel in GA No. Crane or Hof st (Monorail or Jib) Size Area Orop Area A086 Main Steam Isolation Valve Hoist 7-1/2 Ton / ,

A087 Main Steam Isolation Valve Hoist 2 Ton /

A088 Main Steam Safety Valve Hoist 2 Ton /

A089 Main Steam Safety Valve Hoist 7-1/2 Ten /

A090 Main Steam Safety Valve Hoist 2 Ton- /

A091 Main Steam Safety Valve Heist 2 Ten /

A092 Main Steam Isolation Valve Holst 7-1/2 Ton /

A093 Main Steam Safety Valve Hoist 2 Ton /

A094 Main Steam Safety Valve Hoist 2 Ten /

Reactor Building Has Vital Reactor Equicment Vessel is in Orop in GA No. Crane or Hoist (Monorail or Jib) Size Area Droo Area R005 ~ Miscellaneous Equipment Jib- 5 Ton /

- R006 Miscelaneous Equipment Jib 5 Ten /

R007 Fuel Manipulater 10 Ton / /

R008 Fuel Maniculator 10 Ten / /

RC09 Equipment Hatch Access Cover 15 Ton /

- R010 Equipment Hatch Access Cover 15 Ton /

R011 Miscellaneous Equipment- 3 Ten /

R012 Miscellaneous Equipment 3 Ton /

R013 Polar 175 Ton / /

R014 Polar 175 Ton / /

ROIS Polar Auxiliary 25 Ton / /

R016 Polar Auxiliary 25 Ten /

R017 Pressure Hatch Jib 2 Ten /-

ROIS Pressure Hatch Jib 2 Ten /

. R019 Reactor Coolant Pump Pressure 3 Ten /

Hatch Jib R020 Reactor Coolant Pump Pressure 3 Ten /

Hatch Jib R021 Reactor Cociant-Pumo Pressure 3 Ten /

Hatch Jib 7

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TABLE 2.1. (continued)

Reactor Buildine Has Vital Reactor Equipment Vessel is in Orop in GA No. Crane or Hoist (Monorail or Jib) Size Drop Area Oroo Area R022 Reactor Coolant Pump Pressure 3 Ton /

Hatch Jib R023 Reactor Coolant Pump Pressure 3 Ton /

Hatch Jib R025 Reacter Coolant Pump Pressure 3 Ton /

Hatch Jib R026 Reactor Coolant Pump Pressure 3 Ton /

Hatch Jib R027 Fuel Manipulator Auxiliary 2 Ton / /

c uel Manipulator Auxiliary RC28 2 Ton / /

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-stops or electrical interlocks should not be bycassed when the pool contains." hot" spent fuel, and should not be byoassed without aoproval from the shift supervisor

-(or other designated plant management personnel). The mechanical stops and electrical interlocks should be verified to be in place and operational prior to

_ placing " hot" spent fuel in the pool.

(b) The mechanical stops or electrical interlocks of 5.1.2(2)(a) above should also not be bypassed unless an analysis has demonstrated that damage due to postulated load drops would not result in criticality or cause

! leakage that could uncover the fuel.

' (c) To preclude rolling if dropped, the cask should not be carried at a height higner than necessary and in no 4.

case more than six (6) inches (15 cm) above the operating floor level of the refueling building or '

other components and structures along the path of t

travel.

(d) Mechanical stops or electrical interlocks should be provided to preclude crane travel from areas where a postulated load drop could damage ecuipment from redundant or alternate safe shutdown paths.

-(e) Analyses should conform to the guidelines of Appendix A.

9.3 (3) "Each of the following are provided (Note

This alternative is similar to (1) above, except it allows movement of a r

heavy load, such as a cask, into the-pool while it contains

" hot" spent fuel if the pool is large enough to maintain wide separation between the load and the " hot" spent fuel.):

(a) " Hot" spent fuel should be concentrated in one location in the spent-fuel pool that is separated as much as possible from load paths.

(b) Mechanical stops or electrical interlocks should be provided to prevent movement of the overhead crane load block over or within 25 feet horizontal (7.5 m) of the

" hot" spent fuel. To the extent practical, loads should be moved over load paths that avoid the spent-fuel pool and kept at least 25 feet (7.5 m) from the " hot"_ spent fuel unless necessary. When it is necessary to oring loads within 25 feet of the restricted region, these mechanical stops er electrical  !

interlocks should not be byoassec unless the spent fuel has decayed sufficiently as shcwn in Table 2.1-1 and 2.1-2, or unless the total inventory of gap 4

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activity for fuel within the protected area would result in off-site doses less than 1/4 of 10 CFR Part 100 if released, anc such bypassing shou!c recuire the approval from :Se shif t supervisor (or c:her designated plant management individual). The mechanical stops or electrical interlocks should be verified to be in place and operational prior to placing " hot" spent fuel in the pool.

(c) Metnanical stops or electrical interlocks should be provided to restrict crane travel from areas where a postulated load drop could damage equipment from redundant or alternate safe shutdown paths. Analyses have demonstratec tnat'a postulated load drop in any location not restricted by electrical interlocks or mechanical stops would not cause damage that could result in criticality, cause leakage that could uncover tne fuel, or cause loss of safe shutcown equipment.

(d) To preclude rolling, if dropped, tne cask should not be carried at a height higher than necessary and in no case more than six-(6) inches (15 cm) acove the operating floor level of the refueling building or other components and structures along the path of travel.

(e) Analyses should conform to the guidelines of Appendix A.

OB (4) "The effects of drops of heavy loads should be analyzed and shown to satisfy the evaluation criteria of Section 5.1 of this report. These analyses should conform to the guidelines of Appendix A."

A. Summary of Aeolicant's Statements Using NUREG criteria on requirements, the determination of cranes to be considered are, by General Arrangement Number; A001a A002a A036a A037a A00lb A002b A036b A037b Two load drops were evaluated, wnich encompass all otner possible load drops at tne Spen: Fuel Pool area. These two loac crops are: 1) cask crop evaluation and 2) fuel assembly crop analysis.

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In 1) crane stops are positioned to prevent the cask from being moved into the fuel pool area. The cask area is separated from the scent fuel pool by a 3 foot tnick reinforcea concrete wall . Analyses made using maximum crane speed impacts at the stops and a vertical cask lifted position to give maximum swing and horizontal displacement show that the cask center of gravity remains on the cask side of the wall. Tipping studies, assuming that the cask catches the edge of the wall also show :nat the cask center of gravity remains on the cask side of the wall. The evaluation conclusions are tnat the cask would not enter tne spent fuel pool due to dropping or tipoing of the cask.

2) Fuel Assembly Orco is based on a spent fuel assembly being the heavisst object moved over the spent fuel pool.

The highest level abvove the fuel storage racks from which

, it coulc be dropped is 2 feet 2 inches through water.

Assumption made is that the fuel assembly striking the pool floor results in rupture of the cladding of all the fuel rods in the assembly, in spite of the many controls and limitations to prevent this.

An analysis of the postulated accident was performed using a conservative approach based on Regulatory Guice 1.25 and a realistic approach. The basis from Regulatory Guide 1.25 l

l listing 14. parameters used for the analysis model is given.

It shows the whole body dose and thyroid dose at the EAS are conservatively calculated to be 0.66 rem and 10 rem

respectively. The doses from this accident are well witnin the 10 CFR 100 limits. The conclusion is that Catawba Nuclear Station is in full compliance to Section 2.2 Enclosure 3, of the generaic letter. e.g., NUREG 0612 Article 5.1.2.

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l S. EG&G Evaluation The information given in submittals indicate many hoists that do not handle heavy leacs, that do not have vital equipment in the drop area, and co not have spent fuel in the crop area. Table 2.1, therefore, lists only those handling systems justifying the special concern over heavy loads. The units listed by the applicant are those identified as cause for concern in the spent fuel area.

EG&G has no basis to dispute this.

The evaluaiion given on the cask drop supplements the designed in crane stops and barrier wall to justify their conclusion that the fuel cask cannot be drcpped or tumbled into the spent fuel pool.

The information on fuel assemoly drop anc breakup are stated to be conservative and the prameters listed confirm that the analysis model is evaluated conservatively. There is no basis for EG&G to discute Duke Power Company conclusions presented to show that they are in ecmpliance with Section 2.2 Enclosure 3, of tne generic letter.

C. EG&G Cenclusions and Reccmmendations Within the premise used and results stated Duke Power Company has shown that Catawba Units Nos. I and 2 are consistent with the guidelines for NUREG 0612 Article 5.1.2 for heavy load handling at the spent fuel pool area.

2.3.2 Reactor Buildinc [NUREG-0612. Article 5.1.3]

(1) "The crane.and asscciatec lifting cevices used for hancling heavy loads in the containment building shculd satisfy the single-failure-proof guicelines of Section 5.1.5 of :nis report.

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9.3 (2). " Rapid containmect isolation is provided with prompt automatic actuation on high radiation so that postulated releases are within limits of evaluation Criterion I of Section 5.1 taking into account delay times in detection and actuation; and analyses have been performed to show that evaluation criteria II, III, and .IV of Section 5.1 are satisfied for postulated load drops in this area. These

, analyses should conform to the guidelines of Appendix A.

93 (3) "The effects of drops of heavy loads should be analy:ed and shown to satisfy the evaluation criteria of Section 5.1.

Loads analyzed should include the following: reactor vessel head; upper vessel internals; vessel inspection platform; cask for damaged fuel; irradiated samole cask; reactor coolant pump; crane load block; and any other heavy loads brought over or near the reactor vessel or otner equipment required for continued decay heat removal and maintaining shutdown. In this analysis, credit may be taken for containment isolation if such is provided; however, analyses should establish acequate detection and isolation time.

Additionally,- the analysis should conform to the guidelines of Appencix A."

A. Summary of Acolicant's Statements For the purcose of our review we broke the operating conditions in the Reactor Buildings cown into four conditions.

Condition 1 (Full or Reduced ?ower)

During this condition the reactor is in operation. All vital systems are required.

Condition 2 (Hot Shutdown)

In this condition the reactor has been shut down out the reactor coolant system is cressuri:ed above 300 psig and all vital systems are recuired.

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Condition 3 (Cold-Shutdown, Fuel Loaded)

In this condition the reactor is completely shut down and the reactor coolant system depressurized allowing removal of the reactor head. Fuel is present in the l core. During this concition only cecay heat removal systems are needed.

Condition 4 (Cold Shutdown, Fuel Unloaded)

This condition is identical to Concition 3 except no fuel is present in the vessel. No vital systems are required.

Review of Load Handling Systems Used During Different Conditions of Operation Condition 1 & Condition 2 During these concitions miscellaneous hoists are used for minor repair work cr for obtaining cold-shutcown but do'not handle heavy loads. Uncer these conditions a heavy load drop cannot occur precluding the need to favestigate further the load hancling systems or coerations during these conditions.

Condition 3 During this condition all load hancling systems are subject to use. All loac handling systems will be reviewed.

Condition 4 A heavy load drop during this condition would have no impact on the station since the Unit is down with the fuel unloaded and out of the building. For this reason it will not be necessary to review icad handling systems and operations during this condition.

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Based on this review, only load handling cperations curing Condition 3 require further investigation.

The latter half of Table 2.1 summari:es the fincings from the initial review of load nandling systems in the Reactor Buildings. The Cranes and hoists identified are tnose meeting the requirements in Category 3 above. The loads handled by these units and additional information are -

presented on individual information sheets submitted in response to the guidelines of the Phase I report requirements.

Only two (2) of the crane / load comoinations possible while the Unit is at cold shutdown will be considered because tney envelope the others.

1. The first accident consists of dropping the reactor head onto the vessel flange. This accicent could cause structural camage to the vessel and attached piping resulting in loss of cooling water. (Criteria III)

2. The second accident consists of an oblique drop of the reactor head onto the uoper internals. This accident could damage fuel, releasing radicactive gases (Criteria I) and result in increased gao activity above acceptable limits.

(Criteria II)

These analyses are scheduled to be completed in September of this year. The results will be forwarcec to you at that time.

B. EG&G Evaluation The tabulation of cranes and noists to be consicered, ne operational evaluation and loads to be handled nave preoared 15

the way for the analyses. The actual work necessary to show censistency with guidelines remains. This requires that in addition to NUREG 0612 Article 5.1.1 (Phase I) ene cf three options of Article 5.1.3 be satisfied. The analyses tnat are reportec to be schedulec for completion in September (1982) have not been forwarced for review. The scope of information being limited to two crane / load combinations needs a clear presentation to verify that they in fact, envelope the others.

C. EG&G Conclusiens anc Recommendations No valid review can be made on the incomplete suomittal.

Please expedite forwarding of the ccmpleted analyses.

Include information to show that the two loads do, in fact, encompass the others.

2.3.3 Other Areas [NUREG-0612. Article 5.1.51 (1) "If safe shutcown equicment are beneath or directly adjacent to a potential travel load patn of overhead handling systems, (i .e. , a cath not restricted by limits of crane travel or by mecnanical 3: ops or electrical interlocks) one of the followf9g shculd be satisfied in acdition to satisfying ne general guidelines of Secti:n 5.1.1:

(a) The crane and associated lifting devices should conferm to the single-failure pecof guidelines of Section 5.1.6 of this report; C3 (b) If the load drop could imcair the oceration of equipment or cabling associated with reduncant or dual safe shutdown paths, mechanical stops or electrical interlocks should be provided to prevent movement of loads in proxicity to :nese redundant er dual safe shutdown equiprent. (In this case, credit should no be taken for intervening floors unless justified by analysis.)

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S3 (c) Tne effects of load drops have been analy:ed and tne results Indicate tnat damage to safe shutdown equioment wculd not prtclude coeration of sufficient ecuipment to.

achieve safe shutdown. Analyses shculd conform to the guicelines of Appendix A, as applicable.

(2) "Where the safe shutdown equipment has a ceiling separating it from an overhead handling system, an alternative to Section 5.1.5(1) above would be to show by analysis that the largest postulated load-handled by the handling system would not penetrate the ceiling or cause spalling that could cause failure of the safe shutdown equipment."

A. Summary of.Apolicant's Statements The charted information presented in Table 2.1, in the column, "Has Vital Equipment in Orco Areas" represents the heavy load risk and requires additional analysis. These deco area considerations include an allowance for load shape and ara orojected to the basement.

Tne additional analysis for the Auxiliary Building hoists are tabulated in matrix sheets, wnere the analysis for each load is resolved to a "ha: arc eliminatien category." With a few specified exceptions it has been concluded that heavy load hanc!ing operations in tne Auxiliary Buildings will have no effects en vital systems. The exceotions are uncer further study.

The hoists handling heavy loads over vital equipment in the Reactor Building have been tabulated, but at oresent no aralysis has been submitted.

B. EG&G Evaluation The submittal cata shows that hoists icentifisc as A009, A010, A027, A028 and ACdo are uncergoing a detailed ir.restigation and that the evaluation will se presentec at the earliest possible date. All, "otner area" noists in the 17

Auxiliary Building are shown in the load / impact area matrix sheets with Ha:ard Elimination Categories censistent with NUREG guidelines.

In the Reactor Building there are 24 hoists identified as handling heavy loads over vital equipment. in Table 2.1 eight of these are shown to handle heavy loads over the

( reactor vessel and are discussed in 2.2.3 above. However,

[ all 2A are recorted to have vital ecuipment in their load drop areas. There are no matrix analyses or other discussions to show how the heavy load drop risks are controlled, and it is not clear how the two load drops discussed in 2.3.2 are to, " encompass all other possible lead drop accidents" in the reactor area.

C. EG&G Conclusions and Recommendations The detailed investigation under way on the five hosits snould be completed and facts submitted to confirm censistency with the guideline of Article 5.1.3. Other hoists' in the Auxiliary Suilding have suitable presentations made to show they are consistent with the guidelines.

The areas of risk identifiec in the Reactor Building load drop :enes need better presentation of facts to show in what manner the guideline requirements are met. Those risks ccvered in the analysis in progress for guidelines of Article 5.1.3 snould be completed and reported. The other Reactor Area hoists with heavy loads being handled over vital ecuicment should be covered herein to show consisterey with guidelines of Article 5.1.5.

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2.3.4 Single-Failure-Prcof Handling Systems [NUREG-0612, Article 5.1.6] I (1) " Lifting Devices:

(a) Scecial liftine devices that are used for heavy leads in tne area where the crane is to be upgraded should meet ANSI N14.61978, " Standard For Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More For Nuclear Materials," as specified in Section 5.1.1(4) of :nis report except that the handling device should also comply with Section 6 of ANSI N14.5-1978. If only a single lifting device is provided instead of dual devices, the special lifting device shculd have twice the design safety factor as required to satisfy the guidelines of Section 5.1.1(4). However, loads that have been evaluated and shown to satisfy the evaluation criteria

, of Section 5.1 need not have lifting devices that also comply with Section 6 of ANSI N14.6.

(b) Lifting devices that are not soecially designed and that are used for handling heavy loads in the area where the crane is to be upgraded should meet ANSI 830.9-1971, " Slings" as specified in Section 5.1.1(5) of this report, except that one of the following should also oe satisfied unless the effects of a drop of tne particular loac have been analyzed and shown to satisfy the evaluation criteria of Section 5.1:

(1) Provide dual or redundant slings er lifting devices such that a single compcnent failure or malfunction in the sling will not result in uncontrolled icwering of the load; 01 (11) In selecting the proper sling, the lead used should be twice what is called for in meeting Section 5.1.1(5) of this recort.

(2) "New cranes should be designed to eeet NUREG-0554,

" Single-Failure-Proof Cranes for Nuclear Power Plants."  :

.or coerating plants or plants under construction, the crane snould be upgraded in accoroance with the implementaticn guidelines of Appendix C of this reocrt.

(3) " Interfacing lift points such as lifting lugs or cask trunions shoulc alsa meet ene of the following for heavy loads handled in the area wnere the crane is to be utgraded 19

unless the effects of a drop of the particular load have been evaluated and shown to satisfy the evaluation criteria of Section 5.1:

(a) Provide redundancy or duality such that a single lift point failure will not result f. uncontrolled lowering of the load; lift points should have a design safety factor with respect to ultimate stren;th of five (5) times the maximum combined concurrent static and dynamic load after taking the single lift point failure.

E (b) A non redundant or non-dual lift point system should have a design safety factor of ten (10) times the maximum combined concurrent static and dynamic load."

A. Summary of Acolicant's Statements The Applicant has not indicated that Catawba 1 or 2 has any single failure proof handling systems. No alternative upgrading to show improved reliability acpears to have been chosen and no statements are presented concerning this alternative guideline, Article 5.1.6 of NUREG 0612.

9. EG&G Evaluation The information submitted at this time provides no basis for review to NUREG 0612 Article 5.1.6 guideline.

C. EG&G Conclusions and Recommendations There is presently no basis for recommencations on Submittals for this guideline. However, if in the several unresolved pending hoist analyses it is found that the alternative method of upgrading is necessary the guidelines of NUREG 0612 Article 5.1.6 shocid ce used.

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+.  :

3. CCNCLUDING

SUMMARY

3.1 Guideline Recommendatiens Spent Fuel Pool Area, NUREG-0612 Article 5.1.2.

Within the premises used the Analyses show that Catawba 1 and 2 are consistent with this guideline.

Reactor Building, NUREG-0612 Article 5.1.3.

The framework for an analysis has been scoped but the analysis submittal has not been made. It should be excedited and include information to show how the two lead drops scoped envelop all others.

Otner Areas, NUREG-0612 Article 5.1.5.

The detailed investigations on Auxiliary Building hoists No. A009, A010, A027, A023 and A046 should be completed and sent for review. The Auxiliary 3uilding risks, involving other hoists with analysis given in matrix sheets show consistency with guideline requirements.

Single Failure-Proof Handling Systems NUREG-0612 Article 5.1.6. This alternative of upgrading apparently was not chosen. It is not addressed. No Single Failure-Proof Systems have been identified.

3.2 Additional Recommendations If, in the process of completing the studies to develop facts that show consistency with guidelines, it becomes necessary to up grade, the best option of guidelines in NUREG 0612 Article 5.1.6 shoulc be identified and followed.

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N 3.3 Summary Duke Power Company has made a commendable presentation shcwing consistency to the guideline for heavy load handling in the Auxiliary Builcing. An exception, involving 5 hoists :till under investigation,

needs resolution.

In the Reactor Building and for other hoists and areas having the analyses promised for September (1932) has not been received for review. Follow-up action is needed.

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e -

4 REFERENCES

1. Phase I Final Report is.being held up, awaiting promised suomittal of information. Ucon issue it will show: C. R. Shaber and T. H.

Stickley, EG&G Idaho. Control of Heavy Loads at Nuclear Power Plants. Duke Power Company, Catawba Nuclear Station Units 1 and 2.

[date]

2. ' NUREG-0612, Control of Heavy loads at Nuclear Power Plants, NRC.

3. V. Stello, Jr. (NRC), Letter to all applicants.

Subject:

Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 17 May 1978.

4 USNRC, Letter to Duke Power Company.

Subject:

NRC Recuest for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 22 December 1980.

5. William O. Parker, Jr., Letter to Mr. Harold R. Denton, Director

, 0ffice of Nuclear Reactor--Regulation,' U.S. Nuclear Regulatory Commission, Washington D.C.

Subject:

Catawba Nuclear Station Docket Nos. 50-413, 50-414, Control of Heavy Loads. NUREG-0612.

Sectember 24, 1981.

6. William 0. Parker, 'Jr., Letter to Mr. Harold R. Denton, Director

~ Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington D. C.

Subject:

Catawba Nuclear Station Occket Nos. 50-413, 50-414,- Contre! of Heavy Loads. NUREG-0612.

July 1, 1982. <

4

7. Hal B. Tucker, Letter to Mr. Harold R. Denton, Direc:ce Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington D.C. Suoject: Catawba Nutlear Station Cocket Nos. 50-413, i 50-414, Control of Heavy Loads. NUREG-0612. October 21, 1952.

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