ML20126F966
| ML20126F966 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/13/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Groce R Maine Yankee |
| References | |
| NUDOCS 8103240568 | |
| Download: ML20126F966 (25) | |
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o, UNITED STATES -
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NUCLEAR REGULATORY COMMISSION wAsMiscToN, o. c. 20sss j
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March 13, 1981
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Docket ha, 50-309 Mr. Robert H. Groce I
Senior Engineer - Licensing Maine Yankee Atomic Power Company 1671 Worcester Road Framingham, Massachusetts 0170i i
Dear Mr. Groce:
i During recent telephone comunications, you indicated that you may have received a BWR enclosure in lieu of a PWR enclosure to our December 22, 1980 letter regarding Control of Heavy Loads.
This letter transmits a PWR enclosure for your use. We apologize for any h onvenience the error may have caused you.
Sincerely, k
obert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Enclosure:
As stated cc: See next page 9
7 t,.s 2 319BI"1.,
& BED u.s.Tj,#
c THIS DOCUMENT CONTAINS g g40g4B P00R QUAUTY PAGES
Maine Yankee Atomic Power Conpany 1
1 cc:
E. W. Thurlow, President Mrs. L. Patricia Doyle, President Maine Yankee Atomic Power Conpany SAFE POWER FOR MAINE i
Edison Drive Post Office Box 774 Augusta, Maine 04336 Camden, Maine 04843.
Mr. Donald E. Vandenburgh First Selectman of Wiscasset Vice President - Engineering Municipal Building Yankee Atomic Electric Company.
V. S. Route 1 20 Turnpike Road Wiscasset, Maine 04578 Westboro, Massachusetts 01581 Stanley R. Tupper, Esq.
John A. Ritsher, Esquire Tupper and Bradley Ropes & Gray 102 Townsend Avenue.
225 Franklin Street Boothbay Harbor. Maine - 04538 i
Boston, Massachusetts 02110 David Santee Miller, Esq.
Mr. John M. R. Paterson 213 Morgan Street, N. W.
Assistant Attorney General Washington, D. C.
20001 State of Maine Augusta, Maine 04330 Mr. Paul Swetland Resident Inspector / Maine Yankee Mr. Nicholas Barth c/o U.S.N.R.C.
Executive Director P. O. Box E Sheepscot Valley Conservation Wiscasset. Maine 04578 i
Association, Inc.
P. O. Box 125 Mr. Charles B. Brinkman Alan, Maine 04535 Manager -~ Washington Nuclear-Operations Wiscassett Public Library Association C-E Power Systems High Street Conbustion Engineering, Inc.
Wiscasset, Maine 04578 4853 Cordell Avenue, Suite A-1 Bethesda, Maryland 20014 Mr. Torbet H. Macdonald, Jr.
Office of Energy Resources Director, Criteria and Standards Division State house Station #53 Office of Radiation Prograns (ANR-460)
Augusta, Maine 04333 U.S. Environmental Protection Agency Washington, D. C.
20460 Robert M. Lazo, Esq., Chairman Atomic Safety and Licensing Board U.S. Envi. onmental Protection Agency U.S. Nuclear Regulatory Commission Region I Office Washington, D. C.
20555 ATTP: EIS C0ORDINATOR JFK Federal Building Dr. Cadet H. Hand, Jr., Director Boston, Massachusetts 02203 Bodega Marine Laboratory Univer 'ty of California Bodega cJy, California 94923 Mr. Gustave A. Linenberger Atomic Safety and Licensing Board State Planning Officer U.S. Nuclear Regulatory Commission Executive Department i
Washington, D. C.
20555 189 State Street
. Augusta, Maine 04330
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ENCLOSURE 2 j !
STAFF POSITION -
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!NTERIM ACTIONS FOR I'
ij CONTROL OF HEAVY LOADS i!
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,3 (1) ' Safe load paths should be defined per che guidelines of Section
.t 5.1.1.(1) (See Enclosure 1);
.(2) Procedures should be developed and implemented per the guidelines l
of Section 5.1.1(2) (See Enclosure 1);
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(3) Crane operators should be trained, qualified and conduct themselves per the guidelines of Section 5.1.1(3) (See En:losure 1);
(a)
Cranes should be inspected,' tested, and maintained in accordance with the guidelines of Se: tion 5.1.1(6) (See Enclosure 1); and (5)
In addition to the above, special attention should be given 'to procedures, equipment, and personnel for the handling of heavy leads over the core, such as vessel internals or vessel inspection tools.
This special review should include the following for these -
1 cads:
(1) review of procedures for installation of rigging or lif ting devices and movement of the load to assure that sufficient cetail 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 that
- culd lead to failure of the. component; (3) apprcpriate repair and replacement of defective components; and (a) verify that the crane c erators have been properly trained and are familiar with specific procedures used in handling these loads, e.g., hana signals, conduct of operations, and content of procedures.
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' ENCLOSURE 3 REQUEST FOR ADDITIONAL INFORtiATION ON CONTROL OF HEAVY LOADS 1.
INTRODUCTION i
e Verification by the licensee that.the risk associated with load-handling failures at nuclear power plants is extremely low will require a systematic evalua-The folleving specific information tion of all lead-handling systems at each site.
requests have been organized to support such a systematic approach, and provide a basis for the staf f's review of the licensee's evaluation. Additionally, they have been crganized to address separately the two hazards requiring investigation (i.e.,
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I' radiological consequences of damage to fuel and unavailability consequences of
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f' da: age to certain syste=s).
The following general inforcation-is provided to assist in this evaluation and reduce the need for clarification as to the intent and expect-ed results of this inquiry, i
Risk reduction can be demonstrated by either of two approaches:
1.
f likelihood of f tilure is cade extre=ely lov throuch enhanced The j
a.
handling-syste= design fea*ures (NUREG 0612, Section t
5.1.6).
t The consequences of a f ailure can be shown to be b.
acceptable (NUREG 0612, Section 5.1, Criteria I-IV).
i Regardless of the approach selected, the general guidelines of NUREG 0612, Section 5.1.1, should me satisfied to provide maximu=
practical defense-in-depth.
Evaluations concerning radiological consequences or criticality 2.
saf ety, where used, can rely on either the adoption of generic analyses reported in NUREG 0612, requiring only verfication that these generic assumptions are valid fer a specific site, or employ a site-specific analysis.
I Systens required for safe shutdown and continued decay heat removal 3.
therefore, identified in this request.
are site-specific and are not, Individual plants should censider syste=s and componen.s identified those systems or in Regulatory Guide 1.29, Fosition C.1 (except j
portions of syste=s that are required solely for (a) emergency core cooline, containment' heat removal, or (c) post-accident l
(b) post-accident at=esphere cleanup), for evaluation and recognize that containment is similar to that f>3entified in the approach taken in this respect The fact that a load-handling Regulatory Guide 1.29, ?osition'C.2.
syste: may be prevented fro: eserating during plant conditions re-quiring the actual or potentia *. use of sone of these systems,'is rec-(
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ornized in this request for information..
4, The scope of this systematic review should include all heavy loads carried in areas where the potential for non-
. compliance with the acceptance criteria (NUREG 0612, i
Section 5.1) exists. A summary of typical loads to be considered has been provided in NURIG 0612, Table 3.1-1.
It is recognized that some cranes will-carry additional r
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miscellaneous loads, some of which are not identifiable' in detail in advance. In such cases an evaluation or analysis demonstrating the acceptability of the handling
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.of a range of loads should.be provided.
3.
.At some sites loads which must be1 evaluated will include i
lic'ensed shipping casks provided for the transportation of irradiated fuel, solidified radioactive vaste, spent' resins, or other byproduct material. Licensing under 10CTR71.is not evidence that lif ting devices for these shipping casks meet l
the criteria'spe~cified in'NUREG 0612, Sections 5.1.1(4), 5.1 1(5), 5.1.6 (1), or 5.'. 6(3), as appropriate,'and thus does
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eliminate. the need to provide appropriate information 1
not concerning these devices. A tabulation (Attachment 5) is provided.to indicate multiple-site use of these shipping casks.
licersee's evaluation, as reported in response to this :
The results of the request, should provide information suf ficient fo'r the staff to conduct an in-the intent of 'this ef f ort (i.e... the uniform a
dependent review to determine that reduction of the potential hazard from load-handling-system failures) has been satisfied.
2.
IllFORMATION REQUESTED FROM THE LICENSEE GENERAL REQUIREMENTS FOR ~0VERHEAD HANDLING SYSTEMS 2-.1 T~.*AEG 0612, Section 5.1.1, identifies scveral general guidelines related to I
h the design and operation of overhead load-handling systems in the areas w ere l
fuel is stored, in the vicinity of the reactor core, and in other areas of l
spent in damage to equipment required for safe l
the plant where a load drop could result shutdova or decay heat remov'a1.
Infornation provided in response to this section of potentially hazardous load-handling operations at a sheuld identd#- the extent site and the extent of conformance to appropriate load-handling. guidance.
che results cf your review of plant arrangements to Teport 1.
identify all overhead handling systems from which a load drop may resuir 4.n da age to any' system required for plant for any shutdodn or ;
ay heat remo;tal(takingnocredit
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interlocks, technical specifications, operating procedures, or detailed structural analysis).-
2.
Justify the exclusion of any overhead handling system from the above category by verifying that there is sufficient physical separation from any load-impact point and any saf ety-related co=ponent to permit a determination by inspec-tion that no heavy load drop can result in damage to any systes or component required for plant shutdown or decay heat removal.
3.
With respect to the design and operation of heavy-load-handling systems in the containment and. the spent-fuel-pool area mad those load-handling systene identified in 2.1-1, above, provide your evaluation concerning co=pliance with the guidelines of NUREG 0612, Section 5.1.1.
The following specific information should be included in your reply:
Drawings or sketches sufficient to clearly a.
identify the location of safe load paths, spent fuel, and safety-related equipment.
b.
A discussion of measures taken to ensure that load-handling operations re=ain within safe load paths, including procedures, if any, for deviation from these paths.
A tabulation of heavy loads to be handled by each c.
crane which includes the load identification, Icad weight, its designated lif ting device, and verifi-cation that the handling of such load is governed
.by a written procedure containing, as a minimuc:,
the information identified in NURIG 0612, Section 5.1.1 (2).
d.
Verification that lif ting devices identified in 2.1.
3-c, above, comply with the requirements of ANSI N14 6-1978, or ANSI B30.9-1971 as appropriate. yor lif ting devices where these standards, as supplemented by NUREG 0612, Section 5.1.1(4) or 5.1.1(5), are not met, describe any proposed alternatives and demon-strate their equivalency in terms of load-handling reliability.
Verification that ANSI B30.2-1976, Chapter 2-2, has e.
been invoked with respect to crane inspection, testing, and maintenance. Where any exception is taken to this standard, sufficient information should be provided to de:enstrate the equivalency of proposed alternatives.
f.
Verification that crane design co= plies with the guide-lines of CMAA Specification 70 and Chapter 2-1 of ANSI 330.2-1976, including the demonstration of equivalency of actual design requirements for instances where spe-cific complia'nce with these standards is not provided.
R l
Exceptions, if any, taken to ANSI E30.2-1976 with g.
respect to operator training, qualification, and conduCC.
.}
- 2. 2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS NUREG 0612, Section 5.1.2, provides guidelines concerning the design and operation of load-handling systems in the vicinity of stored, spent fuel.
Infor=ation provided in response to this section should demonstrate that ade-qutte ceasures have been taken to ensure that it. this area, either the likeli-hood of a load drop which might da= age spent fuel is extremely small, or that the estimated consequences of such a drop vill not exceed the limits set by the evaluation criteria of NUREG 0612, Section 5.1, Criteria I through III.
1.
Identify by name, type, capacity, and equipment designator, any cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carry-ing loads which could, if dropped, land or f all into ti e spent fuel pool.
Justify the exclusion of any craner in this area from he 2.
above category by verifying that they a're incapable of carrying heavy loads or are permanently prevented fro:
ove-to the col ment of the hook centerline closer than 15 feet boundary, or by providing a suitable analysis demonstr.,ing for any failure nod _, no heavy load can fall into the that fuel-storage pool.
3.
Identify any cranes listed in 2.2-1, above, which you have evaluated as having suf ficient design features to nahe the likelihood of a load drop extremely small f or all loads to be carried and the basis f or this evaluation (i.e., complete compliance with NURIG 0612, Section 5.1.6 or pertial com-pliance supplemented by suitable alternative or additional design features).
For each crane so evaluated, provide the load-handling-system (i.e., crane-load-ccmbination) inf orma-tion specified in Attachment 1.
For cranes identified in 2.2-1, above, not categorized accord-4.
ing to 2.2-3, demonstrate that the criteria of NUREG 0612, Section 5.1, are satisfied.
Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this Vith respect to Criteria I through III, provide request.
a discussion of your evaluation of crane operation in the fuel area and your determi:stion of compliance.
This spcnt response should include the f ollowing information for each crane:
k'hich alternatives (e.g., 2, 3 or 4) from those a.
identified in NrREG 0612. Section 5.1.2. have been selected.,, - -,.
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b.
If Alternative 2 cr 3 is selected, discuss the crane motion limitation imposed by. electrical r
interlov:s or mechanical stops and indicate the circumstances, if any, under which these protective r
devices may be bypassed or removed. Discuss any
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administrative procedures invoked to ensure proper
- i authorization of bypass or removal, and provide.
j any related or proposed technical specificatien (operational and surveillance) provided to ensure
. the' operability of such electrical interlocke or mechanical steps.
c.
k'here reliance is placed on crane operational limitations with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to' ensure that these assumptions remain valid.
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d.
Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed. techni-cal specifications and discuss administrative or physical controls provided to ensure that these assueptions remain valid, e.
Analyses performed to demonstrate co:pliance viti Criteria I through III should confor= to the gui lines of NUREG 0612,-Appendix A.
Justify any ex-ception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis pe-formed.
i 2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG 0612, Section 5.1.3, provides guidelines concerning the design and cperation of load-hansling systems in the vicinity of the reactor core.
Infor-ma:icn provided in response to this section should be suf ficent to, demonstrate j
that adequate measures have been taken to' ensure that in this area, either the likelihood of a load drop vhich might damage spent fuel is extremely small, or that the estimated ceasequentec of such a drop will not exceed the limits set
.bv the evaluation c-iteria of NUREG 0612, Section 5.1, criteria I through III.
1.
Identify by name, type, capacity, and equipment designator, any cranes physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel.
j 1.
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Justify the exclusion of any cranes in this area from the 2.
above category by verifying that they are incapable of carrying heavy loads, or are perranently prevented from the movement of any load either directly over the reactor vessel or to such a location where in the event of any,
load-handling-system f ailure, the load may land in or on the reactor vessel.
Identify any cranes listed in 2.3-1, above, which you 3.
have evaluated as having sufficient design features to mak.e the likelihood of a load drop extremely small for all loads l
to be carried and 'the basis for this evaluation (i.e., com-plete compliance with NUREG 0612, Section 5.1.6, or partial cocpliance supplemented by suitable alternative or additional For each crane so evaluated, provide the
, design features).
load-handling-system (i.e., crane-load-combination) informa-tion specified in Attachment 1.
For cranes identified in 2.3-1, above, not categor1 zed accord-4.
ing to 2.3-3, demonstrate thet the evaluation criteria of NU?lG 0612, Section 5.1, are satisfied. Compliance with Criterion IV vill be demonstrated in your response to Sec-With respect to Criteria I through tion 2.4 of this request.
III, provide a discussion of your evaluation of crane opera-tion in' the containment and your determination of compliance.
This response should include the following information for each crane:
Where reliance is placed on the installation and use a.
of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or bypassed and the administrative pro-cedures invoked to ensure proper authorization of such action.
Discuss any related or proposed technical specification concerning the bypassing of such interlocks.
Where reliance is placed on other, site-specific con-b.
siderations (e.g., refueling sequencing), provide f
or proposed technical specifications and dis-present cuss administrative or physical controls provided to i
ensure the continued validity of such considerations.
Analyses performed to demonstrate compliance with c.
Criteria I through III should conform with the guide-lines of NUREG 0612, Appendix A.
Justify any ex-ception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
l 5PECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OP l
l 2.4 A.EAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HE REM 3 VAL, OR SPENT FUEL POOL COOLING 5JPIC 0612, Secticn 5.1.5, preyides guidelines concerning the design and er components
- pers: ion cf load-handling systems in the vicinity of equipnent l l
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t required for safe reactor shutdown and decay heat removal.
Information pro-vided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small, or that damage to such equip-ment from load drops will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evaluated. in this section have been previously identified in your response to 2.1-1, and their loads in your response to 2.1-3-c.
1.
Identify any cranes listed in'2.1-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for. all leads to be carried and the basis.for this evaluation (i.e., complete compliance with NUREC 0612,.Section 5.1.6, or partial com-liance supplemented by suitable alternative or additional design features).
For each crane so evaluated, provide the load-handling-systet (i.e., crane-load-combination) informa-tion specified in Attachment 1.
2.
For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, a comprehensive hazard evaluation should be provided which includes the following information:
The presentation in a matrix format of all heavy a.
loads and potential impact areas where damage might occur to safety-related equip ent.
Heavy loads identification should include designation and weight or cross-reference to infor=ation pro-vided in 2.1-3-c.
Iepact areas should be identi-fied by construction zones and elevations or by some other method such that the impact area can be located on the plant general arrangement drawings.
l Figure 1 provides a typical matrix, b.
For each interaction identified, indicate which of the load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, mechanical stops and/or electrical interlocks, or other site-specific considerations.
Elimination on the basis of the aforementioned considerations should be supplemented by the following specific information:
(1) For lead / target combinations eliminated because of separation and redundancy of safety-related equipment, discuss the basis for determining that load drops will not affect continued system operation (1.e.,
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the ability of the systen to perform its safety-related function). __._,.
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(2) k"nere mechanical stops.or electrical' inter- '
locks are to be provided. present. details showing the areas where' crane travel vill be.
prohibited. - Additionally, provide a discus-sion concerning the procedures that are to be used for authorizing the bypassing of interlocks. or removable stops, for. verifying
-that incerlocks are functional prior to crane use, and f or verifying that interlocks are restored to' operability.after operations which require bypassing have been completed.
. I (3) Where load / target combinations are eliminated I
on the basis of other, site-specific consid-erat.ons (e.g., maintenance sequencing), pro-.
i vide present and/or proposed technical. speci-fications and discuss administrative procedures or physical constraints invoked.to ensure the continued validity of such considerations.
c.
For interactions not eliminated by the analysis.of. 2.4-2-b, above, identify any handling systems for specific loads which you have evaluated as having sufficient design fea-l tures to make the likelihood cf a load drop extremely small-and the basis for this evaluation (i.e., complete compliance with NUREG 0612, Section 5.1.6, or partial compliance sup-piemented by. suitable alternative or additional design f ea- ~
tures).
For each crane so evaluated, provide the. load-handling-syste=, (i. e., crane-lead-combination) information specified in Attach =ent 1.
d Tor interactions not eliminated in 2.4-2-b or 2.4-2-c..
above, demonstrate using appropriate analysis that damage
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vould not preclude operatien of suf ficient equipment to i
allow the syste to perf orm its saf ety function following p
a load drop (Ni.:F.EG 0612, Section. 5.1, criterion IV).
For
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' 8.S each analysis so conducted, the following information k ' /. f,.r should be provideu oGs oh /
(1) An indication of whether or not,. for the d /.
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specffic load being investigated, the over-
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head crane -handling syst e is designed and
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constructed such that the hoisting system (g
vill retain its load 'in the event of~ seismic e
accelerations aquisalent to those of a saf e
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shutdova. earthquake (SSE).
L Tne basis for any exceptions taken to the (2) analytical guidelines of KG.EG 0612, Ap-pendix A.
4 (3)
The information requested in Attachment
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NOTES TO TIGURE 1 Note 1:
Indicate by sy=bols the safety-related eq11pment.
The Ifcensee should provide a list consistent with the clarification provided in 1.2-3.
Note 2: Hazard Elimination Categories Crane travel f or this area / load combination prohibited a.
by electrical interlocks or mechanical stops.
b.
System redundancy and separation precludes loss of capability of system to perform its safety-related function following this load drop in this area.
Site-specific considerations eliminate the need to con-c.
sider load / equipment combination.
d.
Likelihood of handling syste= failure for this load is extremely s=all (i.e. section 5.1.6 N'UREG 0612 satis-fied).
Analysis demonstrates that crane failure and load drop e.
will not damage safety-related equipment.
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i Attachment (1)
SINGLE-FAILURE-PROOF HANDLING SYSTEMS Provide the name of the manufacturer and the design-rated load (DRL). If; 1.
the e.aximum critical load (MCL), as defined in hTREG 0554, is not the same as the DRL, provide this capacity.
2.
Provide a detailed evaluation of the overhead handling system with respect to the f eatures of design, f abrication, inspection, testing, and operation as delineated in hTRIG 0554 and supplemented by the identified alteinstives specified in hTREG 0612 Appendix C, This evaluation mu6t include a point-
.by-point co=parison for each.section of hTREG 0554.
2f the alternatives of h*UREG 0612, Appendix C, are 'used f or certain applications in lieu of
- i co= plying with the reco==endation of hTREG 0554, this should be explicitly stated.
If an alternative to any of those contained in hTREG 0554 or hTREG 0612, Appendix C, is proposed, details must be provided on the proposed alternat'ivetodemonstrateitsequivalency.b!
to the seismic analysis employed to demons'trate that the ever-3.
Vith respect head handling syste= can retain the load during a seismic event equal to a safe shutdown earthquake, provide a description of the method of analysis, the assu=ptions used, and the cathematical model evaluated in the analysis.
~
Tne description of assu:ptions should include the basis f or selection of trolley and load position.
/
Provide an evaluation of the lifting devices for each single-failure-proof 4
handling system with respect to the guidelines of b'UREG 0612, Section 5.1.6.
Provide an evaluation of the interfacing lif t points with respect to the 5.
guidelines of hTREG 0612, Section 5.1.6.
1/ If the crane in question has previously been approved by the staff as satisfying NUREG 0554, Reg. Guide 1.104, or Part B to.BTP-ASB9-1, please reference the date of the staff's safety evaluation report or approval letter in lieu of providing the information requested by item 2.
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Attachment (2)
ANALYSIS OF RADIOLOGICAL RELEASES The following information should be provided for an analysis conducted to demonstrate compliance with Criterion 1 of KUREG 0612, Section 5.1.
1.
INITIAL CONDITIONS / ASSUMPTIONS Identify the time af ter shutdown, the number of fuel a.
assemblies damaged, and the assumed duration of radio-logical release associated with each accident analyzed.
NUREG 0612, Table 2.1-2, provides the assumptions used b.
to arrive at generic conclusions concerning radiological dose consequences. To rely on the radiological dose analysis of NURIG 0612, the licensee shculd verify that these assu:ptions are conservative with regard to the plant / site evaluated.
If the assumptions are not con-servative.for the specific plant, or if a more cite-specific analysis is required, the licensee should identify plant-specific assumptions.used in place of those tabulated.
Identify and provide the basis (e.g., USNRC Regulatory c.
Guide 1.25) for any assumptions employed in site-specific identified in NUREG 0612, Table 2.1-2.
analyses not Dose calculations based on the termination or mitigation
- d. - of radiolegical releases should be supported by informa-tion suf ficient to demonstrate both that the time delay assumed is conservative and that the system provided to accomplish such termination or mitigation vill perform its saf ety functicn upon demand (i.e., the syste= meets the criteria for an Engineered Safety Feature).
Specific information so provided should include the following:
(1) Details concerning the location of accident sensors, parameters monitored and the values of these parameters at which a safety signal vill be initiated, system response time (including valve-operation time), and the total tice required to automatically shif t fro: nor al operation to isolation or filtra-tion f ollowing an accident.
(2) A description of the instrumentation and con-trols associated with the Engineered Safety Feature which includes inf ormation suf ficient to denenstrate that the requirements (Section 4) of IEEE 2 7 9-1971,." Crit eria f er Protection Systers f or Suelear Fever Generating Staticns "
are satisfied.
2-1
-.~. - -.~
-.~
4
_ 3) A description of any Engineered Safety
~.
(
l Feature filter system which includes infot-mation suffic'ient tol demonstrate compliance with' the guidelines of USNRC Regulatory Guide 1.52, " Design Testing, and Maintenance Criteria for Engineered Safety Feature Atmos-phere Cleanup System Air Filtration and.
Absorption Units of Light-Water-Cooled I
Nuclear Power' Plants."
(4) A discussion of any initial conditions (e.g., manual valves locked shut,' containment airlocks or equipment hatches shut) necessary I
to ensure that releases will be terminated or mitigated upon' Engineered Safety Feature actuation and the. measures employed (i.e., Tech-nical Specification and administrative controls) to ensure that these initial conditions are satisfied snd that Engineered Saf ety Teature systems are operable prior to.the load lif t.
2.
.5 THOD OF ANALYSIS Discuss the method of analysis used to demonstrate that post-accident dose vill be vell within 10CFR100 limits.
In presenting methodology used in determining the radiological consequences, the following information should l
bt provided.
A description of the mathematical or physical model a.
employed.
b.
An identification and su==ary of any co:puter program used in this analysis.
The consideration of uncertainties in calculational c.
methods, equipment perf ormance, instrumentation response, characteristics, or other indeterminate eff ects taken into account in the evaluation of the results.
3.
CONCL"SION Provide an evaluation comparing the results of the analysis to Criterion 1 cf Nr72C 0612, Section 5.1.
If the postulated heavy-load-drop accident analyzed bounds other postulated heavy-load drops, a lirt of.these bounded heavy loads should be provided.
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O Attachment (3)
CRITICALITY ANALYSIS The following information should be provided for analysis conducted to demon-strate compliance with Criterion II of NURIG 0612 Section 5.1 1.
INITIAL CONDITIONS / ASSUMPTIONS i
The conclusions of NUREG 0612, Section 2.2, are based on a particular' model fuel assembly. If a licensee uses the results of Section 2.2 1
. rather than performing an independent neutronics analysis, the assump--
t tions should be verified to be' compatible with plant-specific design.
I For any analysis conducted, the following assu=ptions should be provided as a minimum:
l a.
k'at er/UO2" "** #"*i b.
The boron concentration for the refueling water and spent-fuel pool l
c.
The amount of neutron poison in the fuel d.
Fuel enrich =ent e.
The reactivity insertion value due to crushing of the core f.
The k value allowed by technical specifications ff fort $ecoreduringrefueling 2.
METHOD OF ANALYSIS Provide the method of analysis used to demonstrate that accidental dropping of a hcavy load does not result in a configuration of the fuel such that k,ff is larger than 0.95.
The discussion of the method of analysis should include the following information:
a.
Identification of the computer codes ecployed b.
A discussion of allowances or compensation for calculation and physical uncertainties 3.
CONCLUSION provide an evaluation comparing the results of the analysis to Criterion II 1
of NUREG 0612. Section 5.1.
If the postulated heavy 1oad-drop accident 3 l
- ~ -..
bounds other postulated heavy-load drops, a list of these bounded he.avy loads should be provided.
e i
I
)
l l
I i
l i
J I
1 1
3-2 i
l l
9
0 Attachment (4) i AKALYSIS OF PLANT STRUCTURES The following infor=ation should be provided fc:: analyses conducted to demon-strate co:pliance with Criteria III and IV of NUREG 0612 Section 5.1.
~
INITIE CONDITIONS / ASSUMPTIONS 1.
J
, Discuss the assu=ptions used in the analysis, including:
l
- a. %"eight of heavy load 2
b.
I= pact area of load c.
Drop height d.
Drop locarion Assu=ptions regarding credit taken in the analysis for e.
the action of impact limiters
+
f.
Thickness 'of valls or floor riabs i=pacted g.
Assu=ptions regarding drag forces caused by the i
enviro = ment h.
Load combinations considered
- i. Material properties of steel and concrete 2.
?2TE03 07 AJU.I.YSIS Provide the method of analysis used to de=enstrate that sufficient load-carrying capability exists within the vall(s) or floor slab (s).
Identify any co:puter codes e: ployed, and provide a descriptiu of their capabilities.
If test data was e= ployed, provide it and describe its applicabi2ity.
3.
CONCLUSION Provide an evaluation co: paring the results cf this analysis with Criteria III and IV of hT?IG 0612, Section 5.1.
Enere safe-shutdown equipment has a ceiling or vall separating it from an overhead handling systes, provide an evaluation to de=onstrate that postulated load drops do not penetrate the ceiling or cause secondary =issiles that could prevent a safe-shutdown syste: fro: perferring its safety function.
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Atta' hment (5) c 1 of 6 l SHIELDED SHIPPING CASKS CERTIFICATED FOR NUCLEAR POWER PLANTS I - Fuel (New and Seent)
CROSS LOT IN CERT.
N0 DEL PRIMARY L:CENSEE L35. (APPROX;)
SECONDARY LICENSEE 4956 FA-1,' 2, 3',
J Ceneral Electric Co.
TVA 5450 RCC, 1, 2, 3 Westinghouse Electric VEr, DLC 5805 Vandenburgh
' Chem-Nuclear Systems, 70,000 APC, CPL, DLP, DPC, TPL, FPC, JCP, b7P, Inc.
VEP 5901 b75 Model 100 Nuclear Puel Services 126,200
' CPC, FCE 593B Eh7 P 48,000
- 'EC 6078 907Al Co
- bustion Engineer-t'? 00 APL 927C1 ing, Co.
7000 62:6 3
Babcock & Wilcox Co.
6940 DPC, FPC 6273 48 (Series) 4500 VEP 6373 PB-1 Che:-Nuclear Systems, 67,050 APC. BEC, CPL, DPC FPL, TPC, CPC, JCP, Inc.
MYA, MIC, NNE, NSP, PNY, TVA, VEP 6400 Super Tiger Vestinghouse Electric 45,000 API., CPC, DLP, DLC, MEC, h7P, SMU, VEP Co.
669G b75-4 Nuclear Puel Services, 50,000 BCE, BEC, CWE, DLP, DP C, ITL, TP C, J CP,
Inc.
MYA, RCE, SCE, M.7, 9001 IP 300 Ceneral Ile.ctic Co.
140,000 CPL, CW 9010 b11-1/2 KL Industries, Inc.
47,500 3::C, TPL, VYC 9 01.4 CE-1600 General Electric Co.
23,000
- 5e. attached ifst of abtreviaticns.
1.
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rem **--vg rr?-+e-g----wgp eg**w-m+q-
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- y=f-y w w qey-w-wgy yw-=*ePer '
P1rwg Tr y-1
e Attachment (5) '
2 of 6 SHIELDED SRIPP]NG CASKS CERTIFICATED FOR NUCLEAR POWER PLANTS II - l'aste GROSS LOT IN CERT.
MODEL PR KGY LICENSEE LES. ( AFPROX. )_
SECONDARY LICENSEE
k'P S 6058 13-1 Nuclear Engineering Co.
30,000 APL; CPC, DLP, IEL, MIC, NPP, NSP, PGE, SMU, TEC, VEP 6144 6144 Nuclear Engineering Co.
42,000 APC, APL, CPL, CEC, CPC, DLP, DPC, FPL, FP C, UP C, IEL, J CP,
MEC, NPP, NSP, PGE, PNY, RGE, SMU, VIP 6244 62't che -Suelear Syste s, 46,000 APC, CPL, Cb'I, DPC,
~
Inc.
'6272 Poly Panther Nuclear Engineering Co.
6100 APL, CPC, DLP, MEC NPP, SMU, VE?
l 6568 LL-60-150
- ennessee Valley Auth.
73,000 6574 F0; 200 Eitt an Nuclear and 47,000 APL, BCE, CL'E. CEC, t
DLP, DLC, IMI, J CP,
l Development Corp.
MYA, MEC, NFP, PEC, PNY, VYC, YAC 6601 L*_-50-100 Che=-Suelear Systers, 70,000 APC, BEC, CPL, CYA, CEC, CPC, DLP, DPC, Inc.
.NSE, PEG, RGE, TVA, VIP Kuclear Ingineering Co.
45,000 APL, CPC, DLP, MIC, NTP, SML' VEP 6679 1/2 Super i
Yiger 6722 E5-33-150 Tennessee valley u:5.
51,000 l
- See at:a:hed list of sbtreviatiens.
..... w...c. 3 s, 3 of 6
-)
t SHIELDED SHIPPING, CAsi3 CERTIFICATED FOR NUCLEAR P7rlER PLANTS -
II - Weste d
GROSS LOT IN CERT.
M07EL PRIMGY LICENSEE LES. (/JPROI.)
SECONDARY LICD;SEE*
6744 Poly Tiger Nuclear Engineering Co.
35,000 APL, BEC,,CPC, DLP, 12C, NPP,
- SMC, TIP h;ue; _ ar Engineerdag Co.
60,000 APL, CPC, DLP, NPF,
6771 SN-1 SMU, VEP 9074 J2-100 28,000 DLC i
9079 EN-100 Ser. 2 Eitt:an Nuclear and 98,000 APL, BCE, CIC, CWE, Develop ent Corp.
DLP, Ie., JCP, 117A, HEC, NFP, PIC 9050 E';- 600 Eit::an Nuclear and 42,000 BGE, CWE, CIC, 3:y, Develop ent Corp.
U 2, I D., J CP,.P.'A,
~
9056
' EN-100 Ser. 1 Eitt=an Nuclear and 46,000 APL, EGE, CWE, D12, Develop:ent Corp.
IME, JCP, ')2A, }ZC,
N?P, NNE, ?EC, RGE,.
VYC
@C59 E';-1005 Eit:=an Nuclear and 36,300 3GE, CVE, CEC, IME, Develop ent Corp.
J CP, iflA N??, PEC
@092 D;-300 Rit::an Nuelear and 43,000 MYA Develop:ent Corp.
B093 EN-400 Eitte.an Nuclear and 43,000 MYA Develep:ent Corp.
3094 CNSI-14-193-E Cher-Nuclear Syste=s, 56,500 APC, APL, EIC, CPL, Inc.
J CP,.". C, NM?, NNE,
'NSP, OPP, ?0E, ?EC, PGC, PNY, FIG, TVA, VEP l
$096 CSSI-21-300 Che=-Nuclear Syst e:s, 57,450 APC, APL, CPL, CE,
Inc.
DPC, TPL, ??C, GPC, JCP, MEC, 109, IC;E, PSY, PEG, VEP
- Set at t ad.e d li st c f :.lbrevin d:ns.
Attaen.ent (5).
4 of 6 SHlELDED SHIPPING CASKS CARTIFICATED FOR NUC!. EAR POWER PLANTS 11 - Weste CROSS LOT IN CIR;.
MO It PRIVARY LICENSEE -
LES. (AFPROX.)
SECONDARY LICENSEE
TPC, CPC, JCP, XEC, NHP, VEP 9105 A1.-33-90 Che=-Nuclear Systems, 41,300 AFC, CPL, CkT, CEC, Inc.
DPC. TPL, TPC, JCP, NFP, NMP, NNE, PGC, VIP, k1P 9111 CN6-80A Che:-Nuclear Syste.=s,
'Inc.
DPC. TPL, TPC, CPC, MEC, NNE, PCC, SM;;.
VEP 9123 7-100 Che:-Nuclear Syste.=s,
7000 A? C, BEC, CPL, CK'I, Inc.
CYA, DPC. TPL, TPC, GPC, JCP, MEC, NMP, NNE, NSP,* VIP 9122 16-450 Che=-Nuclear Systems, 61.000 EEC Inc.
- See attached list of abbreviatiens.
5 of 6 SHIELDED SHIPPIfG CASKS CERTIFICATED FOR NUCLEAR P0h'ER PLANTS III - Byproducts GROSS LOT IN RERT.
M3??_
PRD%RY LICD;SEE LES. (AFPROX.)
SEC0h*DARY LICIliSEE 5971' GE-200 10,000 PEC 5980 CI-600 18,500 NNE ESP 6275 LL-25-4 Che=-Nuclear Syste=s, 30,000 APC, CPL, D?C TPL, Inc.
FPC, NPP, VIP 90E1 C);S-1600 Che=-Nu clear Sy st e=s,
.26,000 APC, 3GE, CPL DPC, Inc.
FPL, FPC, GPC, NSP, TVA, VIP f
4 e
- See attached If st of abbr evia t ier.s.
N t
' Attachment (5) t.q:E
' LICENSEE' ABBREV!AT:0NS 6 of 6 '
C t'
s a
APC Alaba=1 Power Company.
' AFL' Arkansas-Power and Light' Company BIC 3eston Edison' Company' BCE 3altimore cas and I3ectric Co=pany
[
CIC Consolidated Edison Company p
CPC Consu:ers Power Company
- i:
CPL Carolina' Power and Light Company; CEI Cc== nvealth Edi, son Co:pany CYA Connecticu: Yankee Ate:ic Power Company DLC Duquesne Light Company DLP Dairyland Power Cooperative DPC Duke ?:ver'Coupany l
- (
YPL Tiorida Power and Ligh: Co=pany
!9
.t CPC Ga'orgia Power Co:pany IIL Iowa Electric Light and Power Co=pany l
IMI Indiar.a and Michigan Electric Company
[
JCP Jersey Central Power and Light Ccapa:7 MIC Me:ropolitan Idison Ce:pany MYA Maine Yankee Ato:ic Power Co:pany i
NMP Niagara Mohawk Power Corpora: ion l
NNI Nor:heas: Nuclear Emergy Co=pany NPP Nebraska Public Power Corporation NSP Nerthern 5:stes Power Co=pany 0?P 0=aha Public Pever Distrie:
PIC Philadelphia Electric Cc:pany PIG Public Service Electric and Gas Company PGC Por: land General Electric Company PNY Pcver Au:hori:y of the 5:ste of N'ev York RGE Roches:e Gas and Electric Corporatics SMU Sacra:en:o Municipal Utilities Corpora: ion TIC Toledo Idisen Co:pany i
' Tennessee Valley Authority VI?
Virginia Elec:rie and ?ever Co=pany VYC Ver=::: Yankee Nucles: ?:ver Corpora:icn i
YAC Yankee A:ecic I!ec:ri: Cc:pany
'JMp Vis::nsin-M1:hipar. Feve r C:=;any V77 Vis: rpir, Publir servi:e Cer;e::: fen 2
i f
e
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