ML20076D212

Control of Heavy Loads (C-10),Prairie Island Nuclear Generating Plant,Units 1 & 2, Technical Evaluation Rept

ML20076D212
Person / Time
Site:	Prairie Island
Issue date:	05/17/1983
From:	Vosbury F FRANKLIN INSTITUTE
To:	NRC
Shared Package
ML20076D213	List: ML20076D212
References
CON-NRC-03-81-130, CON-NRC-3-81-130, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-08075, TAC-8075, TER-C5506-384-3, TER-C5506-384-385, NUDOCS 8305230021
Download: ML20076D212 (31)

Text

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& Preparedby 1

tj , Frank!!n Research Center Author: F. W. Vosbury

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Philadelphia, PA 19103 FRC Group Leader: I. H. Sargent k '

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Washington, D.C. 20555 Lead NRC Engineer: F. Clemenson 2

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'This report was prepared as an account of Iwcrk ~

sponsored b'y an agencv of the United States j Government. Neither the United States Government nor any agency thereof, or any of their emcloyees. makes any warranty, expressed or implied, or assumes any legal Ifaci!ity or respons!bility for any inird party's use, or the results of such use, cf any Informatien, apca-ratus, product or process disciosed in this repert or represents that its use by such third party woutd not infringe privately owned rights.

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j TECHNICAL EVALUATION REPORT l CONTROL OF HEAVY LOADS (C-10)

NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 NRC DOCKET NO. 50-282, 50-306 FRC PROJECT C5506 NRC TAC NO. 08075, 08076 FRC ASSIGNMENT 13 N RC CONTRACT NO. NPC-03-81-130 FRC TASKS 384, 385 Prepared by Franklin Research Center Author: F. W. Vosbury 20th and Race Streets .

Philadelphia, PA 19103 FRC Group Leader: I. R. Sargent Preparedfor Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: F. Clemenson May 17, 1983 I m.

,, .his repcrt was precared as an account of work sponsored by an agency of the Unite # 5 ates Government. Neither the United States Government nor any agency thereof, or any c!. their employees, makes any warranty, expressed or implied, or assumes any 'egai liability or

+

q resocnsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

Prepared by- Reviewed by: Approved by:

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/ S Sf Principal Aut orQ Droup dader bepartment Dir/ctorf/

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.. 0. Franklin Research Center A Division of The Franklin Institute The Benen Franun Parhey. PMa.. Pa 19103 (215) 448 100o l

TER-C5506-384/385 CONTENTS Section Titig Pags 1 INTRODUCTION. . . . . . . . . . . . . 1 1.1 Purpose of Review . . . . . . . . . . . 1 1.2 Generic Background . . . . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . . 2 2 EVALUATION . . . . . . . . . . . . . . 4 2.1 General Guidelines . . . . . . . . . . . 4 2.2 Interim Protection Measures. . . . . . . . . 21 3 CONCLUSION . . . . .

. . . . . . . . . 24 3.1 Genera'1" Provisions for Load Handling . . . . . . 24 3.2 Interim Protection Measures. . . . . . . . . 24 4 REFERENCES . . . . . . . . . . . . . . 26

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TER-C5506-384/385 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.

Mr. I. L Sargent and Mr. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.

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1. INTRODUCTION 1.1 PURPOSE OF REVIEW This technical evaluation report documents an independent review of general load handling policy and procedures at Northern States Power Company's (NSP) Prairie Island Nuclear Plant Units 1 and 2. This evaluation was performed with the following objectives:

o to assess conformance to the general load handling guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" [1] ,

Section 5.1.1 o to assess conformance to the interim protection measures of NUREG-0612, Section 5.3.

1.2 GENERIC BACKGROUND Generic Technical Activity Task A-36 was established by the U.S. Nuclear Regulatory Commi'ssion (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to ensure the safe handling of heavy loads and to recommend necessary changes in these measures. This activity was initiated by a letter issued by the NRC staff on May 17, 1978 [2] to all power reactor licensees, requesting information concerning the control of heavy loads near spent fuel.

i 1 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 of heavy loads at operating plants, althcugh providing protection from certain potential problems, do not adequately cover the major causes of load handling accidents and should be upgraded.

In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed t.r. achieve a two-phase objective using an accepted approach or protection philosophy. The first portion of the objective, achieved through a set of general guidelines identified in NUREG-0612, Section 5.1.1, is to ensure that all load handling systems at 3 J.JJ Franklin Research Center a o, a w n. rm w -

TER-C5506-384/385 nuclear power plants are designed and operated such that their probability of failure is uniformly small and appropriate for the critical tasks in which thay are employed. The second portion of the staff's objective, uchieved

, through guidelines identified in NUREG-0612, Sections 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 (1) features are provided, in addition to those required 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 (2) conservative evaluations of load handling accidents indicate that the potential consequences of any load drop are acceptably small.

Acceptability of accident consequences is quantified in NUREG-0612 into four accident analysis evaluation criteria.

A defense-in-depth approach was used to develop the staff guidelines in order to ensure that all load handling systems are designed and operated so that their probability of failure is appropriately small. The intent of the guidelines is to ensure that licensees at all operating nuclear power plants

_ _ , perform the following:

o define safe load travel paths through procedures and operator training

, so that, to the extent practical, heavy loads are not carried over or near irradiated fuel or safe shutdown equipment o provide sufficient operator training, handling system design, load l handling instructions, and equipment inspection co ensure raliab~ a .

operation of the handling system.

Staff guidelines resulting from the foregoing are tabulated in Section 5 of NURSO-0612. Section 6 of NUREG-0612 recommended that a program be initiated to ensure that these guidelines are implemented at operating plancs.

' l.3 PLANT-SPECIFIC BACKGROUND On December 22, 1980, the NRC issued a letter [3] to NSP, the Licensee for the Prairie Island Nuclear Plant, requesting that the Licensee review provisions for handling and control of heavy loads, 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 O dh) Franklin Research Center

TER-C5506-384/385 these guidelines. On April 2,1981, NSP provided a partial response [4] to this request. Additional information was provided in subsequent reports on July 15, 1981 [5], August 31, 1981 [6], December 9, 1981 [7], and February 3, 1982 [8]. A draft technical evaluation report was prepared based on these submittals and was informally transmitted to the Licensee for review and comment. On August 30, 1982, a telephone conference call was conducted with representatives of NRC, FRC, and NSP to discuss unresolved issues. As a result of this call, additional information was forwarded by NSP on November 8, 1982 [9]. This information was incorporated into the draf t TER, which was then reissued as a draft final TER in which several issues remained unresolved. In order to resolve these issues, a meeting was held at NRC headquarters on March 18, 1983. Subsequent to that meeting, NSP submitted additional information in response to these issues in a letter dated April 8, 1983 [10]. This final report is based on the information provided in References 4 through 10.

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2. EVALUATION This section presents a point-by-point evaluation of load-handling provisions at Prairie Island Nuclear Generating Plant Units 1 and 2 with respect to NRC staff guidelines provided in NUREG-0612. Separate subsections are provided for both the general guidelines of NUREG-0612, Section 5.1.1 and the interim measures of NUREG-0612, Section 5.3. In each case, the guideline or interim measure is presented, Licensee-provided information is summarized and evaluated, and a conclusion as to the extent of compliance, including recommended additional action where appropriate, is presented. These conclusions are summarized in Table 2.1.

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2.1 GENERAL GUIDELINES The NRC has established seven general guidelines which must be met in order to provide the defense in-depth approach for the handling of heavy loads. These guidelines consist of the following criteria from Section 5.1.1 of NUREG-0612:

( o Guideline 1 - Safe Load Paths o Guideline 2 - Load Handling Procedures o Guideline 3 - Crane Operator Training o Guideline 4 - Special Lifting Devices o Guideline 5 - Lifting Devices (Ndt Specially Designed) o Guideline 6 - Cranes (Inspection, Testing, and Maintenance) l o Guideline 7 - Crane Design.

These seven guidelines should be satisfied for all overhead handling systems and programs in order to handle heavy loads in the vicinity of the reactor vessel, near spent fuel in the spent fuel pool, or in other areas where a load drop may damage safe shutdown systems. The Licensee's evaluation of the extent to which these guidelines have been satisfied and an independent assessment of this evaluation are contained in the succeeding paragraphs.

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a. o Talate 2.1. Prairie Island tintts 1 and 2/NUREG-0612 Compliance Mattia m t.,, -

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• weight or Interie Interim Guideline I Guideline 2 Guideline 3 Guideline 4 Guideline 5 Guideline 6 Guideline 7 Measure 1 Capacity Safe Imad Measure 6 Crane Operator special Lifting Crane - Test Teche.ical special 3Q Heavy Imade (tonal Pathe gocedures Training Devices Sidnee and Inapection Crane peelen Specificatione Attention R *

1. Containment Polar .

Cranosl23 230/20 -- --

C -- --

C s

C -- --

a. Reactor 28.1 C C --

C - -- - -- --

Missile Shields

b. PER 20.25 C C -

C - - - -- --

Minelle sis t elde

c. Vessel 40.5 C C --

C - -

e g need #

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' d. Upper 25.0 C C --

C -- -- -- --

C Internals

e. tower 85.7 C C --

C -- -- -

Internals C

t. Weeao1 3.6 C R -- --

C -- -- -- --

Studs 9 IS! Tool 2.0 C A -- --

C -- -- -- --

h. RCP Motor 39.7 R R -- --

C -- -- -- -

!. RCP Pump 27.6 R R -- --

C -- -- -- --

J. RcP 6.6 R m -- -

C -- -- - --

Flywheel g

19 23 A

t, tn C = Licensee action complies with titlREG-0612 Guideline. o m

P = I.lcensee action par tially comp!!ss willa HistM-4,63 2 G.eldeline, m - Licensee has proposed revisions /moJilicattuns wlatete a:Wly with NUREG-0612 Guideline. CD

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( or Guideline 1 Culdeline 2 Culdeline 3 Guideline 4 Culdeline 5 Cuideline & Culdeline 7 19easure 1 Measure 6 8r Capacity Saf e 104 0 Csane Operator Special Lif ting Crane - Test Technical Special Ih

,: a pleavy inails (tonal Pithe Ps ocedur es _ Ts e tning Devices Slings and Inspection Crane Deelqn Specificatione Attention j Q 2. Aua lliar y Building Crane 125/25 -- -- C -- -- C C -- --

e. taw ruet 3.3 C C -- -- C -- -- -- C Shilting Containere 1

b. heat e.1 C y -- -- C -- - -- --

Enchanger j kesoval l Matchee ,

c. Heat 0.55-0.95 C R -- -- C -- -- -- --

Eachanger e j m Buredlee .

3. Turbine BullJing Cranes (21 120/25 -- -- C -- -- C C -- C i

a. HP Cover 42.8 C R -- C -- -- -- == --

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b. LP 11 61.2 C R -- C -- -- -- -- --

Outer Casing

c. LP 82 61.2 C R -- C -- -- -- -- --

Outer Casing

d. LP 11 25.0 C M -- C -- - -- -- --

Inner Cyl . 11 h

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-- 1 l.P B 2 25.0 C H -- C -- -- -- --

e. O Inner Ut Cyl. 81 Ut o

f. LP 81 45.0 C H -- C -- -- -- -- --

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fy We19ht Interla Interin n or Guideline 1 Guldeline 2 Guideline 3 Guldeline 4 Guideline 5 cuideline 6 culdeline 7 reasure 1 Measure 6 g Capacity Bate toad Crane operator special Lifting , Crane - het Technical Special

-* Heavy loads (tone) Pathe Procedures Tsetnine Devices , , Elines and inspection Crane Deelen speelficatione attention it

g. LP 82 45.0 C R --

C C -- - -- -

Inner Cyt. 83

h. IIP Rotor 35.0 C R -- C C -- -- -- --

1. LP Rotor 80.0 C R -- C C -- -- -- --

J. Condensate g.3 C R -- -- C - -- -- --

Pump

k. Condensate 6.0 C R -- == C# -- -- - --

Pump Motor 4

3 1. vertical 3.2 C R -- --

C -- - -- --

Cooling Water Pump Motor

m. vertical 1.3 C R -- -- C -- -- -- --

Cooling  !

Water Pua.p  ;

n. Spare Rotor 6.25 C R -- -- C -- -- -- --

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o. Generator 123.0 C 2 -- C -- -- -- -- --

potor '

4. Spent l'uet 3.0/3.0 -- -- C -- -- C C C --

Cranes [

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a. Olvider 1.3 C C -- -- C -- -- - --

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b. 1901 Covers 1.0 C R -- -- -- -- -- -- -- Ut un C2 m

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TER-C5506-384/385 2.1.1 NUREG-0612, Heavy Loads overhead Handling Systems

a. Summary of Licensee Statements and Conclusions The Licensee has evaluated the load handling systems at the Prairie Island plant and concluded that the following systems should be subject to the general guidelines of NUREG-0612:

o containment polar cranes (2) i o auxiliary building crane ,

1 o turbine building cranes (2) o spent fuel pool crane.

Other load handling systems were eliminated from further consideration under NUREG-0612 for the following reasons:

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1. Physical separation from safety-related equipment. It was determined

, by inspection that a load drop could not damage any system or component required for plant shutdown or decay heat removal for the following load-handling system': s

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o filter room crane o 3-ton new fuel handling crane o manipulator cranes o 1-ton trolley above auxiliary building general exhaust fan. l

2. Single-purpose system. Each of the following load handling systems  ;

is used for maintenance of a single piece of safety-related equipment; consequently, thes~e systems' carry ' heavy leads over safety-relared equipmenc I

only when plant conditions have been established to allow such equipment to be 1

l removed from service: '

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I o Nos. 11/12 diesel cooling water pump trolleys '

o D-1/D-2 diesel trolleys o 1-ton trolley above main steam isolation valve (MSIV) (2) o trolley between main steam and feedwater lines o 1-ton trolley inch main steam relief header (MSRH)

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o portable 5-ton trolleys above residual heat removal (RHR) heat exchanger removal hatch (2) o 1-ton trolley above relief header in fuel handling and vent fan room o 6-ton trolley above RHR pit covers (2) .

i b. Evaluation i

The Licensee's determination that NUREG-0612 is not applicable to those lif ting devices identified is consistent with NUREG-0612 guidance for the following reasons: (1) adequate separation from irradiated fuel and safety-related equipment exists or (2) the lif ting device is used only when a safety-related component or system that might be damaged by a load drop is placed out of commission (presumably following the establishment of appropriate plant conditions) prior to the lif t.

c. Conclusion NSP's identification of load handling systems subject to compliance with the guidelines of NUREG-0612 is consistent with NRC requirements.

2.1.2 Safe Load Paths (Guideline 1, NUREG-0612, Section 5.1.l(1)]

" Safe load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to impact irradiated fuel in the reactor vessel and in the spent fuel pool, or to impact safe

, shutdown equipment. The path should follow, to the extent practical, structural floor =cmbers, beams, etc., such that if tne load is dropped, the structure is core likely to withstand the impact. These load paths should be defined in procedures, shown on equipment layout drawings, and clearly markad on the floor in the area where the load is to be handled.

Deviations from defined load paths should require written alternative procedures approved by the plant safety review committee."

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a. Summary of Licensee Statements and Conclusions The Licensee states that the primary equipment of concern in the turbine j 1

building is the 4.16-kV switchgear. Adequate protection is af forded this j l

switchgear in the following manner:  ;

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The switchgear is protected by an 18-inch-thick concrete floor, o

! The switchgear is redundant and a single load drop will not cause loss of safeguard switchgear.

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The "two crane has been modified with redundant limit switches to prevent blocking." '

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The area covering the switchgear has been marked as an exclusion area for moving heavy loads.

o Any load movement procedure, in this area would require preparation of a special f

Safeguard equipment located in the auxiliary building is limited to the boric acid storage tanks (BAST) and the spent fuel pool. An exclusion area i

has been established above the BASTS similar to that established for the 4.1 kV switchgear in the turbine building. Control of load movements over the spent fuel pool is governed by administrative procedures which prohibit movement of heavy loads in the area as per the technical specifications.

All load handling operations in containment are controlled by procedure D58, " Control of Heavy Inads."

This procedure contains general load handling l

precautions, including the prohibition of carrying heavy loads in the vicinity of spent fuel and certain safety-related equipment. This procedure requires i

load-specific procedures to provide detailed requirements for handling particular components.

Where specific procedures have been prepared (e.g.,

reactor head, internals, and missile shields) , they are referenced. For additional heavy loads that may be identified in the future, D58 provides guidance for the preparation of a specific load handling procedure including the requirements for defining a safe lead path consistent with NUREG-0612.

The Licensee has provided drawings showing the safe load paths to be used for the reactor vessel head and missile shields.

Any new load handling procedure developed pursuant to the requirements of D58, or any revision to an existing procedure, will be approved by the plant operations committee.

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In lieu of marking safe load paths on the containment floor, NSP will assign at least one member of each load handling crew, in addition to the crane operator, the responsibility of ensuring that safe load paths are followed.

In addition, the Licensee has provided the following comments regarding load movement in the containment:

I o Dimensions of heavy londs nearly span the distance between floor beams.

o Movements of heavy loads are restricted by where the load is designed to rest when not in use; for example, the upper and lower internals

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l are stored underwater in the refueling pool, and movement of the I reactor vessel head, stored on the permanent head stand, is constrained by the refueling pool walls and the steam generator vault walls.

o The physical dimensions of heavy loads and the space available for

. their laydown on floors to which the crane has access do not allow deviation from procedural load paths.

b. Evaluation The establishment of exclusion areas for the 4.16-kV switchgear in the turbine building and for the BASTS in the auxiliary building is consistent with Guideline 1 because the exclusion area is relatively small and well defined, and the establishment of individual safe load paths would unnecessarily restrict the handling of loads in the remainder of the building.

The containment load paths currently employed by NSP are consistent with l

this guideline and indicate that paths to be prepared for additional lif ts will also be consistent with this guideline.

The Licensee's commitment to require plant operations co=mittee review / approval for procedural changes affecting safe load paths, deviations from safe load paths, or proposed new safe load paths is also consistent with

! this guideline.

The Licensee's plan to provide assistance to the crane operator to ensure that load paths are followed through the use of a designated member of the load handling party is also consistent with this guideline.

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c. Conclusion and Recommendation 1

Actions taken at Prairie Island Units 1 and 2 are consistent with this guideline. It is further concluded that this guideline will be satisfied I

during future operations at Prairie Island Units 1 and 2 based on the

. Licensee's commitment to provide and implement safe load paths as outlined in 1

Section 2.1.2.a above.

2.1.3 Load Handling Procedures [ Guideline 2, NUREG-0612, Section 5.1.l(2} }

" Procedures should be developed to cover load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment. At a minimum, procedures should cover handling of those loads listed in Table 3-1 of NUREG-0612.

These procedures should include identification of required equipment; inspections and acceptance criteria required before movement of load; the steps and proper sequerice to be followed in handling the load; defining the safe load path; and other special precautions."

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a. Summary of Licensee Statements and Conclusions The Licensee has indicated that no rpecific procedures are in place for handling turbine generator components. However, procedures will be generated if it becomes necessary to carry these heavy loads over the 4.16-kV switchgear.

I In the auxiliary building, proceduras exist for the handling af the new l

fuel shipping containers. Prccedures to handle heat exchanger removal hatches I and heat exchanger bundles will be written prior to handling these loads and i

will take into consideration circumstances present at that time. Detailed f

procedures exist for the handling of spent fuel pool divider gates and pool covers.

Loads handled by the containment polar cranes are procedurally controlled as follows:

Load Procedure Reactor vessel (RV) Missile Shield D3 Section 4.28/D7 Section 4.23 Removal / Replacement Pressurizer Missile Shield D5 8.3 .1 Removal / Replacement i

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RV Head Removal / Replacement D3 Section 4.26, 4.27/D7 Section l

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Upper Internals Removal / Replacement D4.1 Section 2.0/D6.1 Section 2.0 i Lower Internals Removal / Replacement D4.2 Section 2/ Procedure to be l written before use Vessel Studs (in handling box) Procedure to be written before use l

In-Service Inspection Tool Procedure to be written before use Reactor Coolant Pump Notor Procedure to be written before use i

Reactor Coolant Pump Internals Procedure to be written before use In addition, the Licensee has indicated that existing procedures and those to be prepared will provide the information specified in NUREG-0612, Section 5.1.1(2) . '

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[ b. Evaluation The procedures which have been implemented and the Licensee's commitment to develop certain load handling procedures prior to the handling of the specific loads noted satisfy the requirements of this guideline.

c. Conclusion The Prairie Island Plant ccmplies with Guideline 2 of NURIG-0612.

2 .1.4 Crane coerator Traininc ! Guideline 3, NUREG-0612, Section 5.1.l(311

" Crane operators should be trained, qualified and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976 ' Overhead and Gantry Cranes * [11]."

a. Summary of Licensee Statements and Conclusions The Licensee has stated that no exceptions are taken to the guidance in ANSI B30.2-1976 with respect to operator training, qualification, and conduct.

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b. Evaluation and Conclusion Crane operator training, qualification, and conduct at Prairie Island

'9 Units 1 and 2 are consistent with Guideline 3 of NUREG-0612.

2.1.5 Special Lifting Devices [ Guideline 4, NUREG-0612, Section 5.1.l(4)1 "Special lifting devices should satisfy the guidelines of ANSI N14.6-1978, ' Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials' [12].

This standard should apply to all special lifting devices which carry heavy loads in areas as defined above. For operating plants certain

inspections and load tests may be accepted in lieu of certain material requirements in the standard. In addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined '

maximum static and dynamic loads that could be imparted on the handling device based on characteristics of the crane which will be used. This is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static load) of the load and of '

the intervening components of the special handling device."

a. Summary of Licensee Statements and Conclusions 1

The special lifting devices used at the Prairie Island plant include:

o turbine spreader assembly o internals lifting rig L

o reactor vessel head lifting rig. ~

The Licensee,has stated that.these devices were designed by Westinghouce pric:

to the existence of ANSI N14.6-1978.

The reactor vessel head lifting rig, internals lifting rig, and turbine spreader assembly were designed and built during 1970-1971. No formal stress report was prepared and no design specifications were written. The devices were designed, fabricated, assembled, and inspected in accordance with Westinghouse requirements specified on detailed manufacturing drawings and purchase order documents. In general, Westinghouse requirements meet the intent of ANSI N14.6-1978, but do not comply with all specific requirements of the standard. The following is a tabulation of those areas not in strict compliance and the associated Licensee commentary:

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ANSI N14.6 Paragraph \.3.2.1.1, requires \ ik wailh strength materials are

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the design, when usl~nig materials with , @ ' use'd in these devices and yield strength's above 804 'tAlt. heir 'N N ithe frac.ture toughness was

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ultimate air,pneM to b'eu.Ot.S on the . - ' not determhed. However, materials trattiNe,ec[uchrmdd;5ad not the stress Me' sign factors the lists'd he'sQn factory. % Z' listed were\bsed in the'

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analysis and\the resulting stresses'are within those

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2. ANSI N14'.b. Paragraph 5.1, . lists ownti ]2) Westinghouse Quality, Release responsibilities, and 5.1.1' and 5.1 2i _ is considered to be an '

s require the owner to verify that the L - acce' p table alternate to

special lif ting' devices meet the pe'e ~ ~ 9 ; ve'

ify that the criteria for

• formance criteria of the de' sign specia u, S, sert'ified material testing m fication by records and witness of

, 9- reports, nondestructive m'\ '

testing.

4 _ exam'ination (NDE), and C..

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J l 15 doc'umentation required by '

Westingt,ouse drawings and l s

g purchasing documents were Y)

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3. ANSI N14.6,k Sections 3.2-and 3 3 4 .

3M The devices were originally ' ^

resuIres thht t$e rigI'iie inifiellyh.

t9.". wit at 150t maxistm load ' **

,che load tested 1 cad. A I c to a $.only 100% of[ N -

u ,a fd,@.93 by NDE of criticak lo&gi- -  ?.50n of the aaxhu.. T m loadtest  ; ofi

, bearitig' parts and welds and a'Iro \ ' s is ' imp (.ssible to perform '

i d i antwal 1504 load test or antwal'4D2fY, S at tTe' Prairie Island plant'.

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The following. infor eution is

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• t' provic!ej releventhto','.'the benefit of an overload test s of each specisl l}Iting device O \ .y \

,g iy at the Prairie'Islan'ci,' plant.

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3 k Internals Lif tirG Rig his

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device is deiigned to lif t

-1 ' either the upper, or lower

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internals assembly. The lower f internals assembly,-which is gi3 $

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C' s greaterkan twice the weight L 3

,t 'N 'ys\ l of the, {.@sy intezrfals g assembly, hac.been lif ted.

l This lift'~constitur.cs an l ,. overload test for this device l , 'N  %' _

with respect to its use in a lif t for which NUREG-0612 is

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N) Requiremen Remarks

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applicable. (The lower

-(3 r .I . 1 internals lif t is not

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_ subject to NUREG-0612 since it is accomplished only when i

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all fuel is removed from the

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3 C-reactor vessel).

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Reactor Head Lifting l

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Device - This device is a s 1_. fairly simple mechanical s3 - L '~

. . N 1; * ' system assembled from

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, , structural members, i,

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clevises, lugs, and pins. A W W

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stress analysis has been conducted which demonstrates

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that all stresses are below those allowed by ANSI l ,

ar N14. 6-1978.

l Turbine Component Lifting Assembly - This device is essentially an I-beam

, spreader assembly which

, transfers the weight from slings connected to the lifted component via yokes tg and beam hangers to slings connected to the crane hook.

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A stress analysis has been i d ,'" X conducted and demonstrates l d (! -

k that all stresses are below l

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- j those allowed by ANSI N14.6-i

' 1978.

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4,.3 puREG 0612, Paragraph 5.1.l(4) recuires 4. The applicable cranes have 4 [ht the stress design factor stated been investigated to i g t \M Section 3.2.1.1 of ANSI N14.6 be

' determine the dynamic impact based on the combined maximum static force imparted on a load due

\ and dynamic loads that could be imparted to the sudden app 11 cations oy the handling device based on i

of the brakes of the hoists.

characteristics of the crane which will The impact factors were

-belused. This is in lieu of the guide- computed by first establish-sline in Section 3.2.1.1 of ANSI N14.6, ing the minimum stopping which' bases the stress design factor distance of the load under

' only on the weight (static load) of the the maximum braking. A

_ Coad and of the intervening components formula given in " Whiting of 'the special handling device. Crane Handbook," 4 th Edition, I

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TER-C5506-384/385 Requirement Remarks l

p. 166, was used. The braking distance according

! to the formula is a function of icwering speed of the load, horsepower for lifting

' the load, rotational moment of intertia of the motor, gears, drums, etc. , the braking torque, and the efficiency of the system.

The impact factor does not consider flexibility of ropes and cranes. The maximum impact factor calculated is 34.

The Licensee has also indicated [6] that these special lifting devices are inspected prior to use in accordance with the requirements of ANSI N14.6-1978.

Such inspection will include the nondestructive examination of welds and other critical components [10].

b. Evaluation Information provided by the Licensee indicates that the design and fabrication of special lifting devices will provide a degree of load handling reliability equivalent to that expected from an initial design in accordance j with ANSI N14.6-1978. Although not part of the original design, a complete stress analysis has been completed for these devices. This stress report demonstrates that actual factors of safety on material yield and ultimate stress substantially exceed those specified in ANSI N14.6-1978. Furthermore, the manufacturing controls implemented by Westinghouse are expected to provide a degree of quality assurance equivalent to that inherent in ANSI N14.6-1978.

Although no specific 150% overload tests have been conducted on the three special lifting devices subject to NUREG-0612, it can be concluded that the proof of workmanship expected to be demonstrated by such testing can be otherwise determined.

In the case of the internals lift rig, the past use of this rig for i

lifting the lower internals more than adequately demonstrates its capacity for UNilF nklin Research Center A Opveman of The Frerman areasuee

TER-C5506-384/385 handling the upper internals. It should be noted in this case that only the lif t of the upper internals is of interest with respect to NUREG-0612 since the plant conditions required prior to lif ting the lower intervals eliminates this lif t from the jurisdiction of NUREG-0612.

In the case of the reactor vecsel head lif t rig, a review of the rig design and stress analysis indicates that an additional test is not necessary to provide the high degree of assurance of freedom from errors in fabrication or inadequate material properties expected to be demonstrated by such a test.

The device is of fairly simple design. There is little use of welded connections. Almost all load bearing connections are lugs and clevises with large diameter pins. All material used was procured to AISI or ASTM specifications and provides yield strengths in excess of five times calculated stress with margins to ultimate proportionately higher. This substantially exceeds the requirements of ANSI N14.6-1978.

Similarly, the turbine ccaponent lifting assembly is made of material with large safety marg' ins. Further, this device is assembled almost entirely with mechanical connections in an extremely simple design consisting essentially of a wide flange beam with two welded beam hangers near the center and yoke assemblies at either end. At five times rated load, the stress calculated in the beam hanger to beam weld is approximately 1/4 of the weld design capacity.

In summary, both the reactor vessel head and turbina component lif t rigs are of such simple design and large material safety margins that it is highly unlikely that errors of fabrication or inadequate material properties will render them incapable of lif ting 150% of their design load.

The Licensee's commitment to a continuing inspection and examination program is consistent with ANSI N14.6-1978.

c. Conclusion and Recommendations The special lif ting cevices subject to NUREG-0612 at Prairie Island Units 1 and 2 will provide a degree of mechanical reliability consistent with that inherent in Guideline 4.

4 dudd Franidin Research Center 4 on .# n. r,.nen w l

) . .

TER-C550 6-384/385 2.1.6 Lifting Devices (Not Specially Designed) (Guideline 5, NUREG-0612, i  : Section 5.1.l(5))

"Lif ting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9-1971, ' Slings'

[13]. However, in selecting the proper sling, the load used should be the sum of the static and maximum dynamic load. The rating identified on the sling should be 'in terms of the ' static load' which produces the maximum static and dynamic load. Where~ this restricts slings to use on only certain cranes, the slings should be clearly marked as to the cranes with which they may be used."

a. Summary of Licensee Statements and conclusions The Licensee has stated that slings used with cranes at the Prairie Island plant comply with the design and inspection requirements of ANSI B30.9-1971. Plant Procedure D58, " Control of Heavy Loads," specifies a minimum factor of safety of 5, and the rated capacities of the slings shall be

. taken as those listed in Tables 3 through 14 of ANSI B30.9-1971.

Analyses performed show that the design load rating of the slings is based upon the maximum static and dynamic loads. Analysis has shown that the loading due to dynamic loads is very small (approximately 3t) . Assuming that the design load for a particular sling is based solely on static loads, the combined dynamic and static load for the cranes at the Prairie Island plant closely approximates the static design load of the slings.

Since the dynamic loads have been shown to be very small, no penalty is required to be assigned to the slings. Therefore, the actual sling load rating is the design ~1oad, and no further marking or restrictions are

necessary.

l Slings shall be visually inspected each day they 4t6 used. The condition of their replacement and/or repair will comply with requirements of Section 9.2.8 of ANSI B30.9. Further, the operation of the cranes is controlled by administrative procedures which include inspection for safe operating practices with wire slings compatible with ANSI B30.9-1971. .

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3. F anidin Research Center A Cmmen of The Franum insname

^

TER-C550 6-384/385

b. Evaluation and Conclusion Prairie Island Units 1 and 2 comply with Guideline 5 of NUREG-0612.

2.1.7 Cranes (Inspection, Testing, and Maintenance) [ Guideline 6, NUREG-0612, Section 5.1.1(6))

"The cra,ne should be inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2-1976, ' Overhead and Gantry Cranes,' with the ,

. exception that tests and inspections should be performed prior to use '

where it is not practical to meet the frequencies of ANSI B30.2 for

, periodic inspection and test, or where frequency of crane use is less

than the specified inspection and test frequency (e.g., the polar crane l inside a PWR containment may only be used every 12 to 18 months during refueling operations, and is generally not accessible during power operation. ANSI B30.2, however, calls for certain inspections to be performed daily or monthly. For such cranes having limited usage, the inspections, test, and maintenance should be performed prior to their use) ."

a. Summary of Licensee Statements and Conclusions The Licensee has stated that procedures for inspection, testing, and maintenance of cranes are in effect that satisfy the criteria of ANSI B3 0.9-1976, Chapter 2-2.

Evaluation and Conclusion b.

Prairie Island Units 1 and 2 comply with Guideline 6 of NUEEG-0712.

2.1.8 Crane Design [ Guideline 7, NUREG-0612, Section 5.1.l(711 l

"The crane should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976, ' Overhead and Gantry Cranes, ' and of CMAA-70, ' Specifications for Electric Overhead Traveling Cranes' [14]. An alternative to a specification in ANSI B30.2 or CMAA-70 ,

may be accepted in lieu of specific compliance if the intent of the specification is satisfied."

l

a. Summary of Licensee Statements and Conclusions The Licensee has stated that the major cranes and hoists at Prairie Island Units 1 and 2 (i.e., 120-ton turbine building cranes, 125-ton auxiliary build-i

' M) 00! Frank 5n Research Center 1

4 ow n. n.,en

. , - - - ,.,s,

TER-C5506-384/385 ing crane, 230-ton polar cranes, and 3-ton spent fuel pool cranes) were manufactured in accordance with ECCI-61 [15] and USAS B30.2-1967. The Licensee performed a point-by-point review of the specifications to which their cranes were built and those required by CMAA-70. The Licensee's review vorified that the cranes substantially meet the requirements of CMAA-70.

b. Evaluation The Licensee has provided verification that CMAA-70 requirements have bcen satisfied for cranes subject to review or adequate justification that the esquirements of CMAA-70 have been satisfied by equivalent means.

c. Conclusion Prairie Island Units 1 and 2 comply with Guideline 7 of NUREG-0612.

2.2 INTERIM PROTECTION MEASURES The NaC has established six interim protection measures to be implemented I'

ot operating nuclear power plants to provide reasonable assurance that no heavy loads will be handled over the spent fuel pool and that measures exist to reduce the potential for accidental load drops to impact on fuel in the core or spent fuel pool. Four of the six interim measures of the report

!. consist of Guideline 1, Safe Load Paths; Guideline 2, Load Handling Procedures; Guideline 3, Crane Operator Training; and Guideline 6, Cranes (Inspection, Testing, and Maintenance) . The two remaining interim measures cover the following criteria:

1. Heavy load technical specifications

'2. Special review for heavy loads handled over the core.

The status of the Licensee's implementation and the evaluation of these interim protection measures are summarized in the succeeding paragraphs of this Section.

& 3 Franklin Research Center s o~ n a m vren,o m mne

~.,

TER-C5506-384/385 2.2.1 Technical Specifications (Interim Protection Measure 1, NUREG-0612, Section 5.3(1)]

I

" Licenses for all operating reactors not having a single-failure-proof overhead crane in the fuel storage pool area should be revised to include a specification comparable to Standard Technical Specification 3.9.7, l

' Crane Travel - Spent Fuel Storage Pool Building,' for PWR's and Standard Technical Specification 3.9.6.2, ' Crane Travel,' for BWR's, to prohibit handling of heavy loads over fuel in the storage pool until implementation of measures which satisfy the guidelines of Section 5.1."

8. Summary of Licensee Statements and Conclusions l In a previous submittal to the NRC, the Licensee revised the Prairie l

Island Technical Specifications to meet the requirements of Interim Protection <

Macsure 1 of NUREG-0612. This submittal added Section 3.8 (B.1) to the Prairie Island Technical Specifications prohibiting the handling of heavy loads over spent fuel in the fuel storage ool area.

b. Evaluation and Conclusion .

Prairie Island Units 1 and 2 comply with Interim Protection Measure 1.

l 2.2.2 Administrative Controls (Interim Protection Measures 2, 3, 4, and 5, NUREG-0612, Sections 5. 3 (2)-5.3 ( 5) ]

" Procedural or administrative measures (including safe load paths, load l

handling procedures, crane operator training, and crane inspection] ...

l can be accomplished in a short time period and need not be delayed for completion of evaluations and modifications to satisfy the guidelines of Section 5.1 of [NUREG-0612) ."

c. Summary of Licensee Statements and Conclusions Summaries of Licensee statements and conclusions are contained in discus-sions of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.

b. Evaluations, Conclusions, and Recommendations Evaluations, conclusions, and recommendations are contained in discussions of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.

l O l E

' d Franklin Research Center A Dramen of The FrannM meende

- . - - . - - - - . . . =. __- ._-

l TER-C5506-384/385 2.2.3 Special Reviews for Heavy Loads Over the Core (Interim Protection Measure 6, NUREG-0612, Section 5.3 (1)]

"Special attention should be given to procedures, equipment, and personnel for the handling of heavy loads over the core, such as vessel internals or vessel inspection tools. This special review should include the following for these loads: (1) review of procedures for installation

, of rigging or lifting devices and movement of the load to assure that sufficient detail is provided and that instructions are clear and concise; (2) visual inspections of load bearing components of cranes, slings, and special lifting devices to identify flaws or deficiencies j that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g. , hand signals, conduct of operations, and content of procedures."

4

a. Summary of Licensee Statements and Conclusions The Licensee has stated in Reference 5 that the interim actions described in Reference 3 were implemented.

b. Evaluation and Conclusion Prairie Island Units 1 and 2 comply with Interim Protection Measure 6.

l i

~ ~

d000 Franklin Research Center A Dnessen of The FrerenInsetute

TER-C5506-384/385

3. CONCLUSION This summary is provided to consolidate the results of the evaluation contained in Section 2 concerning individual NRC staff guidelines into an cverall evaluation of heavy load handling at Prairie Island Nuclear Generating Plant Units 1 and 2. Overall conclusions and recommended Licensee actions, l where appropriate, are provided with respect to both general provisions for l

load handling (NUREG-0612, Section 5.1.1) and completion of the staff recommendations for interim protection (NUREG-0612, Section 5.3).

3.1 GENERAL PROVISIONS FOR LOAD HANDLING The NRC staff has established seven guidelines concerning provisions for hendling heavy loads in the area of the reactor vessel, near stored spent fusi, or in other areas where an accidental load drop could damage equipment required for safe shutdown or decay heat removal. The intent of these guidelines is twofol'd. A plant conforming to these guidelines will have d veloped and implemented, through procedures and operator training, safe load l trevel paths such that, to the maximum extent practical, heavy loads are not entried over or near irradiated fuel or safe shutdown equipment. A plant conforming to these guidelines will also have provided sufficient operator I

training, handling system design, load handling instructions, and equipment inrpection to ensure reliable operation of the handling system. As detallad in Section 2, it has been found that load handling operations at the Prairie Island plant can be expected to be conducted in a highly reliable manner consistent with the staff's objectives as expressed in these guidelines.

3.2 INTERIM PROTECTION MEASURES The NRC staff has established (NUREG-0612, Section 5.3) that certain msasures should be initiated to provide reasonable assurance that handling of hsavy loads will be performed in a safe manner until final implementation of tha general guidelines of NUREG-0612, Section 5.1 is complete. Specified mrasures include the implementation of a technical specification to prohibit nMn Research Center ~4~

i A Dmmon af The Franann heemme

TER-C5506-384/385 the handling of heavy loads over fuel in the storage pool; compliance with Guidelines 1, 2, 3, and 6 of NUREG-0612, Section 5.1.1; a review of load handling procedures and operator training; and a visual inspection program,

~ including component repair or replacement as necessary of cranes, slings, and special lifting devices to eliminate deficiencies that could lead to component '

failure. Evaluation of information provided by the Licensee indicates that Euch measures have been implemented at Prairie Island Units 1 and 2.

e l

A dbh0 Franklin Research Center A !>= mon of The Fremde heenme

TER-C5506-384/385

4. REFERENCES

1. NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" NRC, July 1980

2. V. Stello, Jr. (NRC)

Letter to all Licensees

Subject:

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

3. NRC Letter to Northern States Power Company

Subject:

Request for Review of Heavy Load Handling at Prairie Island Nuclear Generating Plant December 22, 1980

4. L. O. Mayer (NSP)

Letter to the D. G. Eisenhut (NRC)

Subject:

Control of Heavy Loads ,

April 2, 1981

5. L. O. Mayer (NSP)

Letter to D. G. Eisenhut (NRC)

Subject:

Control of Heavy Loads July 15, 1981

6. L. O. Mayer (NSP)

Letter to D. G. Eisenhut (NRC) l

Subject:

Control of Heavy Icads August 31, 1981 7.- L. O. Mayer (NS' P)

Letter to D. G. Eisenhut (NRC)

Subject:

Control of Heavy Loads December 9, 1981 l

8. L. O. Mayer (NSP)

Letter to D. G. Eisenhut (NRC)

Subject:

Control of Heavy Loads February 3, 1982

9. D. M. Musolf (NSP)

Letter to pirector, Office of Nuclear Reactor Regulation

Subject:

Control of Heavy Loads November 8, 1982 O dNU Franklin Research Center A Ns.on as The Fransen mesame

e-TER-C550 6-384/385

10. D. M. Musolf (NSP)

Letter to Director, Office of Nuclear Reactor Regulation

Subject:

Control of Heavy Loads April 8,1983

11. American National Standards Institute

" Overhead and Gantry Cranes" ANSI B30.2-1976

12. American National Standards Institute

" Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 KG) or More for Nuclear Materials" ANSI N14.6-1978

13. American National Standards Institute

" Slings" ANSI B30.9-1971

14. Crane Manufacturers Association of America

" Specifications for Electric Overhead Traveling Cranes" CMAA-70, 1975

15. Electric Overhead. Crane Institute

" Specifications for Electric Overhead Traveling Cranes" EOCI-61, 1961 1

l I

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