ML20095A163
| ML20095A163 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/17/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR GL-81-07, GL-81-7, NUDOCS 8408210396 | |
| Download: ML20095A163 (36) | |
Text
.
- f eo DUKE POWER GOMPANY P.O. HO x 33180 CHARLOTTE. N.C. 28242 HALH. TUCKER Tztmenown vara raremswr (704) 073-4331 WUCLEAR PRODUCTION August 17, 1984 Mr. Harold ' R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission
. Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4
Subject:
.McGuire Nuclear Station Docket Nos. 50-369 and 50-370 NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants"
Dear Mr.-Denton:
On December 22 1980, Mr. D.~. G. Eisenhut (NRC/0NRR) issued a letter. requesting
-that Duke Power Company review its controls-for.the handling of heavy loads to determine the extent to which the guidelines of NUREG-0612 were satisfied at.
McGuire Nuclear Station, identify the changes and modifications that would be required in-order to fully satisfy _those guideline's, and provide information documenting the results of.our. review and implementation of the required changes and modifications. NRC Generic Letter _81-07, " Control of Heavy Loads",
was issued on February 3, 1981 correcting several minor errors.in the December 22, 1980 letter.
NUREG-0612, Section 5.1, gives recommended guidelines for the control of-heavy -loads which assure that either (1) the potential for a load ' drop is extremely _small, or (2) for each area addressed certain specified evaluation criteria are satisfied.- Toward this end, the NRC developed a defensc-in-depth capproach-for controlling the handling of heavy loads which encompasses an lintent to~ prevent a- well as= mitigate the consequences of postulated accidential
-load drops. The subsections provide guidelines on-how the defense-in-depth approach may be satisfied for various plant areas. 'Section 5.1.1 ~ identifies several general guidelines related to the design and operation of overhead load-handling systems in the areas where spent fuel is stored, in the vicinity of the reactor core, and in other areas of the plant where a load drop could result.
in damage to equipment required for safe shutdown-or decay heat removal.
Sections 5.1.2 through 5.1.5 provide specific guidelines concerning the design-and: operation of load-handling systems in the spent fuel ~ pool area-PWR
'(Section 5.1.2), containment building-PWR (Section 5.1.3), reactor building-BWR
-(Section 5.1.4), and:other areas containing safe shutdown equipment (Section-5.1.5). In addition, in order to assure safe handling of heavy loads in the interim period until measures at operating plants are upgraded to satisfy.the A-guidelines lof Section. 5.1, . Section 5.3 gives various interim protection qg'7
. measures to be implemented. .c D O \
P
r e
O Mr. Harold R. Denton, Director August 17, 1984 Page Two Duke Power Company has submitted responses to the NRC's December 22, 1980 letter via Mr. W. O. Parker Jr.'s letters to Mr. H. R. Denton (NRC/0NRR) dated June 2, 1981; August 5, 1981; October 8, 1981; November 23, 1981; January 15, 1982; March 3, 1982; June 4, 1982; and July 26, 1982. Ms.
E. G. Adensam's letter of September 9, 1982 transmitted a draft technical evaluation report (dated August 31, 1982) which was developed by the Franklin Research Center (under a technical assistance contract to the NRC) to assess McGuire's conformance to the general load-handling guidelines of NUREG-0612 Section 5.1.1 (corresponding to Section 2.1 of enclosure 3 of the December 22, 1980 letter), and to the interim protection measures of NUREG-0612 Section 5.3 (corresponding to enclosure 2 of the December 22, 1980 letter). My letter of November 1, 1982 provided a response to this draf t TER, as well as submitting further informaton requested by the December 22, 1980 letter. As a result of continued evaluation of McGuire Nuclear Station a revised dradt TER (dated January 12, 1983) was transmitted by Ms. Adensam's June 10, 1983 letter. This draft TER found that load-handling operations at McGuire can be expected to be conducted in a reliable manner generally consistent with the staff's objectives as expressed in the general load-handling guidelines, and that (with certain exceptions) the interim protection measures have been satisfactory implemented. However, the draf t TER identified several unresolved items for which additional information was requested. My letter of July 15, 1983 supplied a partial response to these items; additional information is provided below.
Guideline No.1 - Safe Load Paths (NUREG-0612, Section 5.1.l(1))
Safe load paths for all major equipment handled in the reactor building (including the reactor building polar cranes) were developed and were in use in April, 1983. Load paths for the ice condenser bridge cranes were developed and in use in April, 1983. However, it has subsequently been determined (as shown in attached FSAR Figures 1.2.2-14, 1.2.2-15, and 6.2.2-1) that the ice condenser bridge cranes do not handle heavy loads in the areas where spent fuel is stored, in the vicinity of the reactor core, or in other areas of the plant where a load drop could result in damage to equipment required for safe shutdown or decay heat removal (load drop would be onto a three foot thick concrete floor (bottom deck) and could only damage ice condenser equipment), and therefore it is concluded that the requirements of NUREG-0612 are not applicable. My July 15, 1983 response indicated that visual aids would be provided to assist the crane operator in ensuring that the designated safe load paths developed in the auxiliary building which were not painted on the floor (i.e. for fuel handling area cranes, A108A and A111A) are actually followed. Procedures to implement the use of these visual aids were to be in effect by September 1, 1983. However, lines designating the safe load paths for the fuel handling area cranes have since been painted on the floor (work was completed in March 1984). These lines are only partially covered by plastic during
r _ - - .
P Mr. Harold R. Denton, Director August 17, 1984 Page Three fuel cask movement. The lines, along with having the designated travel path shown in an enclosure to the procedure and having a handling supervisor verify the load path, are adequate.to meet the intent of guideline No. 1.
All designated load paths which have been-developed in other buildings, including the reactor building polar cranes (and ice condenser bridge cranes although not required as indicated above), are either marked on the floor, roped off, designated in procedures, or are clearly marked on drawings in the lift supervisors handbook. With the submittal of this information full compliance with guideline No.1 (and consequently Interim Protection Measure No. 2 of NUREG-0612 Section 5.3) has been demonstrated.
Guideline No. 2 - Load Handling Procedures (NUREG-0612, Section 5.1.l(2))
Procedures for handling heavy loads and using safe load paths for cranes located in the reactor building which comply with the requirements of guideline No. 2 were developed and approved / implemented April 29, 1983.
With the submittal of this information full compliance with guideline No.2
_ (und consequently Interim Protection Measure No. 3 of NUREG-0612 Section 5.3) has been demonstrated. Since compliance with the remaining Interim Protection Measures (Nos. 1, 4, 5 and 6) has previously been acknowledged in the January-12, 1983 TER, McGuire has fully implemented all Interim Protection Measures of NUREG-0612 Section 5.3. (Note that although compliance with Interim Protection Measure No. 1 of NUREG-0612 Section 5.3 was not .
requested by the December 22, 1980 letter, Duke Power Company demonstrated compliance with this measure in response to the August 31, 1982 TER).
Guideline No. 4 - Special Lifting Devices (NUREG-0612, Section 5.1.1(4))
As indicated in my July 15, 1983 response, the following are the special lifting devices at McGuire which must be evaluated for compliance with the requirements of ANSI N14.6-1978 (As supplemented by NUREG-0612, Section 5.1. -
1 (4)):
- a. Reactor vessel head lifting rig and load cell
- b. Reactor internals lifting rig
- c. Reactor Coolant Pump motor lifting rig
- d. Control Rod Drive Mechanism (CRDM) missile shield lifting rig The August 31, 1982 draft TER supplied a list of the specific sections of ANSI N14.6-1978 which must be addressed to determine compliance with guideline No. 4. Attachment No. 1 provides an assessment of each of these sections for items e and d above which were designed and constructed by Duke Power Company. Items a and b were manufactured by Westinghouse, and Duke Power is currently pursuing additional information from them necessary for the evaluation. Once this information is received an assessment will be made and submitted to the NRC. NRC/0IE Information Notice 83-71, i
b
o
' Mr. Harold R. Denton, Director August 17, 1984 Page Four
" Defects in load-bearing welds on lifting devices for vessel head and internals", was reviewed with respect to the McGuire reactor vessel head and internals lifting rigs. An inspection procedure meeting the requirements of ANSI N14.6-1978 has been written for these lifting devices.
The lifting 'evices d are visually inspected annually. All load bearing parts and welds are inspected and any part or weld that appears suspicious or for which a defect is detected is subjected to an appropriate non-destructive examination by QA personnel. Duke Power Company believes the
-preventative maintenance program in place at McGuire is adequate to insure the structural integrity of the reactor head and internals lifting devices.
Further, these devices were load tested by Westinghouse to 125% of maximum load (NUREG-0612 requires that the lifting device be tested in accordance with ANSI N14.6, i.e. a load test of 150% of maximum load prior to initial.
use). In addition, the questions raised in the information notice will be addressed in the above mentioned NUREG-0612 lifting rig analysis for these rigs which has been requested from Westinghouse. Since compliance with the remaining general load-handling guidelines (Nos. 3,5,6, and 7) has previously been acknowledged in the January 12, 1983 TER, McGuire has demonstrated full compliance with all guidelines of NUREG-0612 Section 5.1.1 (except as noted for guideline No._4).
Duke Power Company's July 26, 1982 submittal contained information in accordance with the specific requirements of Section 2.3 of enclosure 3 of the December 22, 1980 letter (corresponding to NUREG-0612 Section 5.1.3) and indicated that further information concerning reactor vessel head drop analyses would be forwarded later. Based on the evaluation of Safe Shutdown and Decay Heat Removal equipment, it was determined that the following approach would satisfy the intent of NUREG-0612. The polar crane is assumed to be capable of a drop at any point within the crane wall.
Since the head is the largest item that can be dropped, all smaller loads would be covered by the reactor vessel head. This simplifies the analysis since we assume any/all equipment can be damaged by a load drop. The July 26, 1982 submittal stated that only two of the crane / load combinations possible were to be considered because they envelope the others: (1) dropping of the reactor vessel head onto the vessel flange, and (2) oblique drop of the reactor vessel head onto the upper internals. These analyses were performed by Westinghouse with the following results. Duke Power supplied Westinghouse with drawings of the interior concrete layout and also provided information concerning the height of drop and medium through which the head would fall, and a sketch showing the relative elevations of structures and components in the safe load path of the vessel head from the reactor vessel to the head storage stand. The information available within Westinghouse included: the masses of the objects involved in the impact, the stiffnesses of the reactor vessels, nozzles, supports, and loop pipiag, the length of the reactor vessel guide studs; and the details of the vessel heads,-reactor vessels, and vessel nozzles. The information was used to develop the conditions and scenarios for which the postulated drop accident was evaluated. Attachment 2A is the report documenting this analysis. The
l e
i Mr. Harold R. Denton, Director August 17, 1984 Page Five results of the analysis shows that the reactor vessel nozzle stresses caused by the head drop are within allowable limits. The reactor vessel support impact loads, however, exceed the faulted condition vertical allowable loads. Consequently, the reactor vessel supports had to be evaluated to determine the effects of these high loads on the ability of the system to maintain adequate residual heat removal and, thereby, prevent excessive radiation releases. Maintenance of core cooling capability in this case is dependent upon the loop piping and essential auxiliary piping to remain intact following the vertical displacement of the reactor vessel. The purpose of this support / piping investigation was to determine the maximum vessel displacement and the subsequent effects on the essential piping.
This additional analysis was performed by Duke Power and demonstrated that primary loop piping, auxiliary piping and vessel supports are adequate to maintain core cooling capabilities. A third load drop of the reactor vessel head was considered. This case is a drop on the operating floor which causes concrete spalling and subsequent possible damage to equipment on lower levels of the Reactor Building. This case was not mentioned in 1
earlier submittals covering the Reactor Building. Attachment 2B provides a summary of this case's analysis which concludes that the postulated reactor vessel head drop does not penetrate the operating floor; some scabbing does occur on the underside, however, the structural stabili,ty and functional requirements are maintained. Additionally, attachment 2C provides summary information on analyses performed by Duke Power Company for reactor vessel head drop on vessel internals primary shield wall, load swing into refueling canal wall, and drop onto the reactor internals.
Attachment 2D provides the radiological dose consequence for a reactor vessel head drop, and attachment 2E provides a criticality analysis.
One means of complying with the guidelines of NUREG-0612 sections 5.1.2, 5.1.3, 5.1.4, or 5.1.5 is upgrading of the crane and lifting devices to conform to the single-failure-proof guidelines of section 5.1.6, one of which was NUREG-0554, " single-failure-proof cranes for nuclear power plants". Mr. D. G. Eisenhut's letter dated December 19, 1983 (NRC/0NRR Generic letter 83-42, " Clarification to generic letter 81-07 regarding response to NUREG-0612") indicated that in the course of reviewing crane designs against NUREG-0554, concerns of a generic nature were identified which indicate that NUREG-0554, until revised, may be deficient in assuring single failure proof cranes. It was stated that this aspect of single failure proof crancs would be part of the NRC's review of Duke's submittals if we take credit for a single failure proof crane to satisfy NUREG-0612. In regard to this, McGuire has not and will not use single failure proof cranes as compliance to NUREG-0612.
Should there be any questions in this matter, please advise.
e s
Mr. Harold R.'Denton, Director August- 17 1984 Page Six' Very truly yours, d & k' H'a l B. Tucker PBN/mjf-Attachments
.cc: Mr.' J. P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region 11 101 Marietta Street, NW, Suite 2900 n
Atla'ta, Georgia 30323 Mr. Ralph Birkel Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
-Mr. W. T.. Orders Senior Resident inspector McGuire Nuclear Station Mr. Amarjit Singh Auxiliary Systems Branch .
Division of Systems Integration Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
p e-
.t bec: (w/ attachments)
R. C. Futrell J. S. Warren W. H. McDowell R. B. Priory G. W. Hallman M. D. McIntosh (MNS)
J. F. Streetman R. E. Hunning R. W. Rider (MNS) R. E. Pratt R. L. Dick _ C. J. Sylie J. W. Cox (CNS) S. B. Hager R. O. Sharpe J. B. Swords P. M. Abraham R. P. Ruth (MNS)
D. Mendezoff (MNS) C. F. York (MNS)
R. L. Gill A. V. Carr S. A. Gewehr J. C. Rogers P. D. Stephenson A. L. Snow P. T. Farish S. K. Blackley J. McGarry J. L. Elliott John Snyder (CNS) W. H. Taylor P. R. Herran J. N. Underwood D. E. Colson R. C. Giles W.-0. Henry E. D. Lindsay E. M. Geddie Section File MC-815.07 (81-07)
Donna Hendrix Section File MC-801.01 M. S. Tully
6 t
ATTACHMENT 1 ASSESSMENT OF SPECIAL LIFTING DEVICES FOR COMPLIANCE WITH GUIDELINE NO. 4 The following is an assessment of the Reactor Coolant Pump Motor Lifting Rig and the Control Rod Drive Mechanism (CRDM)
Missile Shield Lifting Rig for Compliance with the requirements of ANSI N14.6-1978 (as supplemented by NUREG-0612, Section 5.1.1 (4)). Strict interpretation of compliance of existing special lifting device design with the criteria of ANSI N14.6-1978 cannot be made. Accordingly, only those sections directly related to load-handling reliability of the lifting devices need be addressed. Several sections of ANSI N14.6-1978 do not contain requirements concerning load-handling reliability: Scope (Section.1),
Definitions (2), Design Considerations to Minimize Decontamination Efforts (3.4), Coatings (3.5), Lubrication (3.6), Inspector's Responsibilities (4.2), and Fabrication Considerations (4.3) .
Evaluation of compliance with Section 6 (Special Lifting Devices for Critical Loads) need not be included since no load has been determined to be a " critical load." The specific sections of ANSI N14.6-1978 referenced below are those stated in the August 31, 1982 TER for which compliance or equivalence must be demonstrated in order to determine compliance with guideline no. 4. (Note that the TER referenced the applicable ANSI sections both by number and brief description. However, several of the numbers did not correspond to the sections indicated by the description. For these-cases Duke Power assumed the. descriptions referenced the section for which compliance or equivalence with was intended).
Section 3.1.1: Design calculations and drawings (and notes on drawings) cover specification criteria defined in Section 3.1.1.
Section 3.1.2: The design specifications, calculations,and drawings identify all load bearing (i.e. critical) components, define their critical characteristics, and specify material identification, qualification,and control for these components.
There are general statements about material requirements for other (non-critical) components. Fabrication practices are discussed in the specification and on the drawings (i.e. notes on drawings).
In process testing and inspection with acceptance criteria is also specified, as well as requirements for final product testing and inspection to applicable acceptance criteria. Duke's QA program was followed on the design of the special lifting devices and is in accordance with 10CFR 50, Appendix B.
Section 3.1.3: Signed stress analysis exist which demonstrates the adequacy of the special lifting devices and their components with respect to any loads that may be imposed upon them during the performance of their functions. These analysis also demonstrate appropriate margins of safety.
Page Two (continued)
Section 3.1.4: All special lifting devices are unique to the item they were designed to lift. Any repairs made would also be unique to each lifting device; therefore, all repairs would be handled on an individual basis and would depend on the severity of the problem. Any major repair would be subject to 125% load test.
Section 3.2.1: Duke Power performed an analysis of the McGuire Reactor Coolant Puap Motor Lifting Rig and the CRDM Missile Shield Lifting Rig per ANSI N14.6-1978 Section 3.2.1 as modified by NUREG-0612 Section 5.1.1(4) . Both of the rigs are constructed of A-36 steel having a minimum yield strength of 36 KSI and a minimum ultimate strength of 58 KSI. The rigs are required to meet stress deaign factors of 3 for minimum yield strength and 5 for ultimate strength. Since the ratios of 58 KSI/36KSI and 5/3 are approx-imately equal (1.61 and 1.67 respectively), checking one of the stress design factors essentially checks the other stress design factor (if one fails both will fail, if one passe both will pass).
Therefore, the analysis of each rig consisted of checking the following case: The rig must be capable of lifting 300% (stress design factor of 3) of the maximum static plus dynamic (15% of static) load without exceeding the yield strength of the materials of construction. For the McGuire RCP Motor Lifting Rig (static load = 88.3 kips, static + dynamic' loa'd = 102 kips) analysis determined t' hat'the rig would be able to meet this case (300% =
306 kips), an'd therefore is also capable of lifting 500% (stress design factor at 3), of this load without exceeding the ultimate strength of the mat'erials. For the McGuire CRDM Missile Shield Lifting Rig (static load = 145. kips, static + dynamic load =
167 kips) analysis determined chat the rig would n'ot be able to meet this case. The ris'was determined to be capable of lifting 120% (201 kips) the maximum static plus dynamic load or 139%
(201 kips) the maximum static load without exceeding the yield strength of the materials of construction. Consequently, the rig would also not be capable of meeting the ultimate strength stress design factor. The McGuire CRDM Missile Shield Lifting Rig (and RCP Motor Lif ting Rig) was designed and built prior to the publication of ANSI E14.6-1978, and therefore is not designed in accordance with that standard. In addition, these rigs do not carry shipping. containers of nuclear material. However, the NRC has taken the position (Ref. " Synopsis of issues associated with NUREG-0612", dated May 4, 1983 which was transmitted with the January 12, 1983 TER) that for special lifting devices subject to NUREG-0612, it should be able to be dembnstrated that, from a design standpoint, they are as reliable as a device for which ANSI N14.6 was developed. Although not originally specified to be designed in accordance with ANSI N14.6, the special lifting devices were provided by Duke Power Company in accordance with appropriate quality assurance and quality control procedures,for a specific application. These lifting rigs meet their applicable design
Page Three (continued) specifications and therefore it is Duke Power Company's opinion that these rigs are designed to an acceptable factor of safety.
This position is consistent with exception No. 1 to guideline No. 4 as given in the above referenced synopsis of issues.
l .
Section 3.2.4: Not applicable to the Reactor Coolant Pump Motor Litting Rig and the CRDM Missile Shield Lifting Rig because these rigs do not have any load bearing pins, extension links or adapters.
f However, all members are designed to meet section 3.2.1.
l Section 3.2.5: All slings at McGuire Nuclear Station are inspected and tagged with a color coded I.D. tag annually. The slings comply with ANSI B30.9 (1971).
t i
Section 3.2.6: Temperatures inside the Reactor building will be l maintained at all times at a level that will ensure ductile conditions (i.e. brittle fractures won't occur) for the material used in these applications (i.e. carbon steel) in accordance with paragraph AM 218 of the ASME boiler and pressure vessel code,Section VIII Division 2, and therefore impact testing is not required. All temperature requirements of the design specification are met.
Section 3.3.1: Not applicable to the Reactor Coolant pump motor lif ting rig and the CEDM Missile Shield Lif ting Rig because the lifting rigs are protected from the environment and from galling.
Lamellar tearing is not a problem although it is considered in the design process.
Section 3.3.4: The rig's design assures even distribution of load.
Section 3.3.5: All load-carrying components have been fitted with cotter pins or lock pins and/or lockwire.
Section 3.3.6: The spent fuel cask lifting yoke is the only lifting device with a remote actuating mechanism. The yoke is in full view of the operator at all times and, therefore, does not require a position indicator.
Section 4.1.3: Design drawings indicate materials to be used, and the rigs were fabricated to those requirements.
Section 4.1.4: The rigs were fabricated in accordance with design requirements and using generally accepted good practices.
Section 4.1.5: All welders in the Construction Departrent at McGuire were qualified in accordance with ASME Section IX and their qualifications are documented and controlled in accordance with Quality Assurance procedure (QAP) L-100, entitled the " Welding Program", and QAP I-1, entitled " Qualification of Welders and Operators". Likewise, all procedures are qualified in accordance with the ASME Boiler and Pressure Vessel Code as directed by QAP L-100.
) ,
,, ' ' n. ; . '
y ,
, 3 Page Four continued s
.s,,
Section 4.1.6: As directed by,the Code of Federal Regulations, s Duke Power Company has operated,the Construction Department Nuclear Station sites,in accordance with a strict Quality Assurance program which meets all the requirements of this section. Quality Assurance h
I
-procedure F-l~ dictates the control of field Construction Procedures (CP).4 These lifting rigs were fabricated and welded in accordance with CP 57, " Welding of Structural Steel Miscellaneous Steel and Other Steel Construction for Nuclear Structures and Certain Non-
' Nuclear Structures" and CP 859, " Structural Steel fabrication, Erection and Inspection".
Section 4.1.7: All structural steel tor McGuire were ordered, received, controlled and issued in accordance with the Quality Assurance procedures for_ Nuclear Safety Related materials. The specific materials for these lifting rigs is designated to be A-36.
This material was color coded in accordance with CP 395, " Identification Marking and issuing of Miscellaneous Steel Stock For QA Condition Uses" and was issued ~accordingly, although these lifting rigs were not Nucicar Safety Related.
4f7.1reme enouqh, the fuel could be crushed to the point where it fragments. This fragmentation changes the heterogeneous nature of a
- li v
- pKit f uel assenhly and approaches a honogeneous mixture. A homogeneous 1 ( mi.xture is by nature a less reactive syst.em than a het.erogeneous j., .
4 geometry, all ele.e remaining the same (water /UO2 volume ratio, quantity of U-215,etc.).
- This is mainly due to the resonance escape probability ubich, for a given ratio of fuel to moderator, can be increased by I using a heterogeneous lattice system (such as fuel rods in an assently).
for example, in a system consisting of natural uranium (0.7 percent t F uranium-235) and graphite (acting as a moderator) the value of K.o is increased f roni a maxinnam of 0.85 in a homogeneous mixture to about 1.08 i
in an optimized heterogeneous (or lattice) system. The lattice spacing J
ni a pWR fuel assenhly is designed close to optinal to maximize neutron cronomy, llomogenization of a fuel assenhly would not necessarily pro-dote the same magnit.ude of change but the trend would be the same, llowever, with light water as moderator, it is apparently impossible to p! achieve a critical system with natural uranium as fuel under any circum--
stances. Spent f uel generally has an enrichment less the.n 0.9 weight i
pm cent (1-235. very close to natural uranium. Therefore, there appears tn he no potential for crititality of totally spent fuel that has been 4
i
[ o ushed to the point of breaking up and fragmenting, regardless of the mannitude of f ragmentation and the final configurat ion. This cannot he catagorically stated for fuct with U-235 content above approximately 150% of natural, such as fresh or partially spent fuel. The drop in the multiplication factor due to ti c !loss of lat tice geometry (homogeniz'ation)
F may not be enough to overcome the higher fuel enrichment Damage to fuel Stored in Spent fuel Pool
~
4 p'IR assenblies are stored in racks that depend on fuci separation to-
'ene. ore subcriticality. This separation produces a large water /ll02.
volume ratio that keeps the multiplication factor well below its desfqn basis value of 0.95. 'If the fragmented fuel spreads out into a soup lite. mixture, the resulting system is even more subtritical than before. Spent fuel storage pool chemistry.is typically'2000 ppm *
(minimum) bnron as horic acid. Huron concentrations have an ef fect ni approximately 103 Ak /k per 1000 ppm. In Unke power testimony bef ore lho Atomic Safety and I.irensing Board -for Oronec spent fuel t ransportat irn and storage at ikGuire, it was stated thal, assuming ma>.iraum enrichment of any assenhly. then in the Oconee pool '(1.2%)
and 2000 prni boron, massive damage to 226 assenblics produced Eef f =
3
'0.45. If the damaged fuel breaks up and falls in on itself, the
- i .
4
?
.' 2 l '
~ __ _ _ _ _ _._.___.__ _ _. __
water /UO2 volinne ratio may ( crease to a more optimal value. If this thange is large enough to ov;rcomo the loss of lattice geometry, the multiplication f actor would increase, in this unlikely st.cnario, the -
' 2000 ppm of boron would be more than adequate to ensure subcriticality assuming all but the freshest of fuel. If the fuel storage racks con-tained a solid poison (usually a boron composite), subcriticality would be maintained by its distribution through the damaged system along with the dissolved bnron. The crushed stainless steel racks also act as a poison by absorbing neutrons. lhe stainless steel, act ing in conjunc-tion with the 2000 ppm boron dissolved in the water, will prevent criti-c.ality in the spent fuel pool of fresh or spent fuel regardless of the nonnitude of fragmentation and the final configuration. If fuel and racks were crushed so compactly that all' water, borated or not, was forced out. criticality would not be. a concern. Light water reactor f uel needs water as a moderator of neutron energy to sustain a fission chain; lack of a moderator prevents criticality.
Damage to Fuci While in Reactor Vessel
.. In a reactor core, approximately 10I gli'/ movable k and all control rods have fixed burnable a totalhave poisons wortha of total unrth of approximately 4.5% ^ k/k. 1 heir presence has a significant elfcct. f or liqht water reactor fuel to achieve criticality after massive damage, the fuel must not have seen significant burnup (fairly high enricFn.cnt) and must he moderated by the proper amount of water, i.e. the Water /UO2 volume ratio must be near optiimon. Even if this highly unlikely scenario occurred, the effect would he only to boil water, not cause a steam explosion. The concern of potential criti-
' ca.ity of massively damaged fuel can he eliminated by ensuring that a large amount of low burnup fuel cannot be affected by a dropped load and that the boron concentration is maintained at its required level.
B. Reconfiguration of Fuel Assembly and Pin lattice Theory .
The more limiting case of a dropped load accident occurs when the fuel pins are not crushed but are pushed closer together so that the spacing of the fuel lattice is changed. in a spent fuel pool, this highly unlikely scenario would require extensive damage to the fuel storage racks while maintaining the integrity of the fuel. The assemblies must be pushed -
closer and the pins in each assembly close in on each other. This configuration of, fuel in the storage pool resembles a reactor core, i.e.
an infinite array of fuel rods. Although the pins in an assembly are already at a near optimum lattice spacing for pure water, squeezing out borated water is actually removing an absorber of neutrons so that packing pins tighter in some instances makes criticality more likely.
', 3
,c .
Damage to Fuel Stored in Spent Fuel Pool tlew fuel storage pools contain nn water to act as a moderator so -
criticality is not a problem. Optimum moderation from hydrogenous fire fighting substances are no longer considered credible scenarios in new fuel storage pools.
In testimony by Charles R. Marotta of the NRC before the Atomic Safety and Licensing Board on Oconee spent fuel transportation and storage at McGuire, he states that: a) subcriticality will be assured for spent Oconee fuel with 2000 ppm boron in the pool water and b) subtriticality will be assured for fresh McGuire fuel with at 1 cast 2000 ppm boron, if the average fuel enrichment is 2.6% U-235 or less. Duke Power testimony on the same occasion states that, assuming an enrichment of 1.2% U-235 for Oconee spent fuel and 2000 ppm boron, the pushing toget-her of 226 assemblies produces a Keff = 0.95. Both the NPC and Duke analyses assumed an infinite number of fuel assenblies. This is a standard calculational technique that introduces conservatism into the
, analysis since it eliminates neutron leakage in the horizontal direc-tion.
The smaller the number of assentlies actually involved, the less chance of realistically attaining criticality because of increased leakage in a smaller system. Also, the less fuel (U-235) involved, the smaller r the chance of attaining a critical mass. NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", hypothesizes this scenario in Section 2.2, Criticality Considerations. This worst case configuration where assemblies and pins are pushed together into optimal reactivity volume ratios without damage to the fuel seems extremely unlikely. The fuel storage racks would prevent this configuration from occurring in a spent fuel pool.
Damage to Fuel WhilE in Reactor Vessel In a reactor, the boron concentration in the coolant is decreased as the fuel is burned up during the life of a cycle. This has no effect on our criticality consideration because this accident can only occur .
during shutdown with the vessel head off. The vessel head is not re-moved until the boron concentration is brought up to a prescribed level.
During refueling at McGuire and. Catawba, boron concentration will be
?.2000 ppm; whereas, at Oconee the concentration will never be less than 1835 ppm. Near the end of the refueling operation, the average ,
fuel enrichment is approximately 2.6 weight percent U-235 while at the start the average enrichment is closer to 2.0. As can be seen from the attach'ed graphs, criticality cannot he absolutely ruled out with the prescribed scenario. This goes for bnth a core in the reactor vessel and a core that has been discharged into the spent fuel storage rool early in its cycle. Once again, fuel enrichment, boron concentra-tion and water /UO2 volume play important parts.
~
... ~
If this scenario is deemed nossible in the reactor vessel, the possi-bility of the core becoming critical cannot be ruled out. As stated in fil1 REG-0612, observing the attached graphs indicates subcriticality
- can be maintained by raising the baron concentration above 2500 ppm.
If the worst case load drop is judged feasible, raising the boron concentration could be the solution.
tems changes to accommodate the higher concentration.This, however, ma C. Applicability of Criticality Curves to Duke Nuclear Plants lhe attached graphs plot the neutron multiplication factor as a function of the Water /tJ02 volume ratio for Westinghouse 15 x 15 fuel. Oconee uses Babcock and Wilcox 15 x 15 fuel while McGuire and Catawba use
. Westinghouse 17 x 17 fuel (Catawba is scheduled to use an " optimized" design variation of the Westinghouse 17 x 17). fiumerous comparisons were made reac tivi ties of
. the three dif ferent fuels to determine their relative The computer code OCELOT was used to model infinite arrays of the various fuel rods in order to calculate the respective neut ron multiplication factors. The analysis showed the three fuels to behave virtually the same from a nuclear criticality standpoint; no signi ficant di f ference exists. The very slight variations in calculated neutron multiplication factors show fuel used by Duke to be slightly more convervative than Westinghouse 15 x 15 fuel and thereby bounded by the plotted curves. Ilowever, any and all di f fer-ences are slight enough that the three *ypes of fuel can be considered the same.
V. C0!!CttJS10tlS Recriticality of damaged fuel is dependent on a number of factors but the following statements can be made:
1.
In general, recriticality in a spent fuel pool is not a problem if dissolved boron is present in the required amount and the storage racks are included.
- 2. Subcriticality is ensured if the damaged fuel is totally spent (close -
fn natural enrichment).
- 3. Damage to fuel in a reactor core must be studied closely to determine the potential for criticality.
4 '
Increasing the boron concentration during shutdown from 2000 ppm to above 2500 ppm would ensure subcriticality in the reactor vessel.
VI. REFEREllCES
- 1. "fluclear Reactor Engineering" by Samuel Glasstone and Alexander Sesonske.
- 2. Duke Power Company, Amendment to Materials License SNft-1773 for Oconee fluclear Station Spent fuel Transportation and Storage at McGuire fluclear Station, Docket fio. 70-2623, Af fidavit of S. B. llager
. ~
VI. REFERENCES (Continued)
. 3.
Duke Power. Company, Aroendment to Materials License SUM-1773 for Oconee Station, Station fluclear Docket Spent fuel Transportation and Storage at McGuire Nuclear No. 70-2623, Testimony of Charles R. Marotta.
4 "Recriticality Potential of TF11-2 Core", by C. R. Marotta.
5 fl0 REG /CR-2155, "A Review of the Applicability of Core Petention Concepts to Light Water Reactor Containments", Sandia National Labs, September 1981.
6 ITUREG-0612 " Control of Heavy loads at Nuclear Power Plants".
7 Memo to file, " Criticality Analysis", October 2, 1979, by Norman T. Simms.
Tile: MC-1201.28, CN-1201.28, P81-1201.28
- 8. McGuire fluclear Station, Technical Specifications.
- 9. Catawba Nuclear Station, Technical Specifications.
- 10. Oconee fluclear Station, Technical Specifications.
Attachments 5
i 4
e e
6 0
--- . -.m . , _ . - t .,, -_ , _ . , . -- ~ . . - - . , . - , - _ . , = - , - - ,m.
, . .i
. Parameters tre si folfsws. (The dim:nsions used are those of Westingbouse 15 x 15 fuel)
Fuel Pellet Diameter . . . . . . . . . . . . . . . . 0.3659" Zire Clod inside Diameter . . . . . . . . . . . . . 0.3734" Ziec Clad Outside Diemeter ........... 0.4220" A s - B u it W/U R at io . . . . . . . . . . . . . . . . 1.647 T e m pe r a t u r e . . . . . . . . . . . . . . . . . . . . . 20 0EGC Fuel M eterial . . . . . . . . . . . . . . . . . . . . 0.9 w/o U235 1.0 - Oppm _
9 0.9 -
)
C OB{ ~
[ -
1000 ppm ii 0.7 [ ~
0.6 - -
2000 ppm O.5 - ~ '
3000 ppm
.~
. ~
~
k )
0.4 - ~
i 3
0.3 ! '
O5 1.0 1.5 2D 2.5 3.0 3.5 4.0 Water /UO2 Volume ristio (W/U Retial NEUTRON MULTIPLICATION FACTOR FOR INFINITE ARRAY OF FUEL RODS IN BORATED VIATER
. . . , - - - . . . , , . , . - - - . - - .- ,,e . - . , , . , , . . . . . - . , . , - - - . . - - - - - -
. Parameters are as follows. (Tt+ dimensions,used are thsse
. cf Westinghouse 15 x 15 fuell Fuel Petiet Diameter . . . . . . ......... 0 3659
Zire Cted inside Diameter . . . . . . . . . . . . . 0.3734
. Zire Cfed Outside Diameter ........... 0.4220" As-Built W/U Ratio ................ 1.647 Temper at ur e . . . . . . . . . . . . . . . . . . . . . 20DEGC F uel M a t e rial . . . . . . . . . . . . . . . . . . . . 2.0 w/o U 2 35 1.3 I I I I I I Oppm l
1.2 -
}
l 1.1 -
A b 1000 ppm 1.0 /
s 4
r 4
' . , 0.9 -
2000 ppm _
=
.w" i
i 0.8 - -
3000 ppm l 0.7 -
l -
7 a
l l
~
0.6
! ! I I I i 1 b 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 l Water /UO2 Volume Ratio (W/U Ratio) :
! NEUTRON MULTIPLICATION FACTOR FOR INFINITE ARRAY OF FUEL RODS IN BORATED WATER 1 .
e.
A Parameters are as follows. (The dimensions used are those of Westinghouse 15 x 15 fuel)
Fuel Pellet Diameter . . . . . . . . . . . . . . . . (13659" Zire Clad inside Diameter . . . . . . . . . . . . . 0.37M
- Zire Clad Outside Diameter ........... 0.4220**
As Built W/U Ratio ................ 1.647 T e m pe r a t u r e . . . . . . . . . . . . . . . . . . . . . 20 D EGC F uel Material . . . . . . . . . . . . . . . . . . . . 3 5 w/o U235 I I I I l l 1.4
)
O ppm 1.3 -
1000 ppm 1.2 -
1 2000 ppm I*I 3000 ppm ~
l 2 3 5000 ppm 2 ^
0.9 -
7000 ppm ~
7 n
5 d.
OE -
05 1
1.0 I
1.5
\l 2.0 l
2.5 I l M 3.0 3.5 4.0 Water /UO2 Volume Ratio (W/U Ratiol :
NEUTRON MULTIPLICATION FACTOR FOR INFINITE ARRAY OF FUEL RODS IN BORATED WATER 9
..,,...--._,_,,,,---.,--.v...--,..--.-.,- -. .
y .
SUPPLEMENTAL EVALUATION OF IN-VESSEL CRITICALITY FOR NUREG-0612, " CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS" McGUIRE NUCLEAR STATION According to Section 4.2.2 of Appendix A, the licensee can demonstrate that crushing the ' core will not drive it critical by using the core refueling neutronics analysis for uncrushed fuel and showing that keff for an uncrushed core is no greater than 0.90 Then, using the estimated 0.05 maximum reacti-vity insertion due to crushing, the maximum achievable keff is still less than 0.95. We will show compliance with the 0.95 limit using McGuire specific fuel pa rameters.
From Table A.1 of WCAP-9323, "The Nuclear Design and Core Physics Characteristics of the W. B. McGuire Unit 1 Nuclear Power Plant Cycle 1," we see that during fuel loading (ambient temperature and pressure) with all control rods in, a boron concentration of 1367 ppm gives Keff=0.95 McGuire Technical Specifications states that the baron concentration will be maintained at a minimum value of 2000 ppm during refueling. To account for the difference in reactivity between 1367 ppm boron and 2000 ppm boron, we refer to Figure A.10 of WCAP-9323. For BOL, 680F an average value for the dif ferential boron worth is -11.95 pcm/ ppm (this assumes linear interpolation of the curve out past the 1500 ppm end point). Thus; (2000 ppm-1367 ppm)(-11.95pcm/ ppm)(10-5Ak/k/pcm)= .0756ak/k Keff = 0.95 0756 = .8744 The calculated value of keff is well below the 0.90 limit given.
f ff g
- 'g f ;. 'f ' 'E
. (' ._. .
- g. -~
g3 j I
}. .
a g,, I
- j. g j.
t.
Tg j.
- y , . _m .p ,
[t. v i l', , H ,5 l , 'ih
~ ~
! 1, , -
r :
m i. .- - .,t i, .
e m- _
- .z
,! _a y ly lpg y &=p,f...f' p.
e l 1: C s.
n,g h;,,{.
sa , .-
e g .
,7., .
- 5. ~ ' l'!.%~/<l s g ._ .. .. .._ - ! .ml _. .. hlcir li ! NWM*. ! p .*E , N 4' ..
,)
.',r
, it M !
- l l I $:
M _ = =
[! l- -' }- Ik~
'.' bh , .
% ~
7 J .: -
4 - ,' r ,, dj g- t' ('~y N)..
d ..
'. t I i (.p%!
B ',"
N I I f- '"gl Hl.b J ' .i . ~ .
Fjily>i
.. y. . .
Yh t
ltk.3hE::
i 4! h '
I Ik -
I iN .e {
fr' e lq'a.ajiii 1!
l I
H .
e- f 8 v
1 Nkfhr lh 3
fafl 9124;ibdb*! / -
- -T *- '-
j'd [M~ 'Y' l l* 6 e' ( *i !! I' L ~~Jhj!' " " " ~ ~
I]IE!j. %ad{' . d,!^"-~ 1 ;-ll{~I~ ' ~ ~
t p lg ML _I --~ qi.;' gf , *f
' ~IlI
~
!' >m ~~- a- ' -
17s .
j' j M ,d.h k '
I p-
~
,,. l -[ l a.-[nd wlj i, a t . t. . .-
e, I4 ,. ., li5 e2,- a. .
- 14 cp g
- f.rt j g-
- h, i f, ,
- g a ~.
$g)ith ';ti
. }i J L,.,_ p 2 ,7
! 4khT~f 1- (fT' lI ' I ! !!
e D '.A kt
.. ; . .m . s i i na I
. h-. (-)~@t y *y- ' ' ,.
j j^i r'-
'_.dj .
t i
.. U!
_, c gF-
~
j f.
1' .
.. k .L' :. . ..
!, a y 2 ,4
! 1 13 '
1 tiet! ai 2 e d r2 1
.3 ,
L,,,,,,,,, GENERAL ARRAEDtENT CONTA!!WENT
~I l '
AND REACTOR BUILDIE SECTION McGUIRE NUCLEAR STATION Figure 1.2.2-14
- Revision S
- AL __ , _ x. --
- w. _
e
- 1. &._ _._ im---__--_--------------------- - - - - - ----- - "- ~ -~ ~ '---' - - " ~ ' - - ' - - ' ~ ^ - ~ ~ - ' ~ ~ ~ - - - ~ ~ ~ ~ ~ ~ ~ ~~~
C C T
.=.-
7 h -.
t '
-..~,,.o ,-_ .. . . . .
s,- +s ,.. - - _
(""* - F g- - - ' . % \
y-r , , _ ' . - ]"
n .
e
. . :X-_ ,__:
e . - - - .
N .r.,
ve e._ 3 .. . - - <mw, e i
e ; == '.
1
., .- (.
- r----
- i u j.
L. _. ,.
j
- s. (.Ib ; . u: =.
d~ I
=~ =.,== -.
-a -..
gy .- 2 a e:. 33 n. _ g-
. . g y - , _.g..,._,_.3w_.._-- n -
i -7,i e '" -
V;; -
47 G,h=w-i A-. 7 .r l
p
- ;g.
_ %m ,
x-i
., j L
j...
=
_.<4 m3
- -f.__ "-a y_
1 10.;1111 g._;I;'! InQ}-
]r T t '" _ '
4 ,
1 ,,
,2 et <
j \ dI ,m>
(mn 6,.- u (L..
, . - ,.e, f
,I - m-.3 O, - m. l'i ,
a*='
4 =. .
4 .
I
,rm - ~12%
/ 3 '
,,g,
!A =
,,=s . w.
- .-- 5u . m' ,.
'j
~~ 1.,
l1 h-* _h1 *
..,"-- w- .~
'. ,O. , ,O, r .
mr 1
. i c-n_;w, , =-
'~
j
- O,
.O . -
- .h f -~- M o. w .. 3 . p T P
[
",* /
R_ JL :.- ..' . 4 swl. m . ' ll(dl.i l
A<
.n4 . t .. -
- 7 7+ -
s; i ; -
s, a4 .i y:., +.E
- 1. i, .r ry} -
, 'l*-
~
p g 4
i ne I. _
g, .
j
-y. n
..- 3 --r
'u w-.m. y r. ,-4 m...;, .,
1 i -
' p, -
n l-ID'..[ -
m.i -. ;
7T1J -"':"6 ; '
g (n ,f[ --:_ " _ --
UnA IO
, . ,y . , - 'N' r7O w,. .sa
- ..u_ s n2, :~-
O y ., - . + '
..._b.
l :. .-
Y I * * "
- *;a==
CI "rty::c_ - % 3 ., * - - -
t . _ , , ,
L--
- (
.'Til2-2.----. y gl i fo.
.a:-.. ,
- g. s*i m, Li .
s.; g. r- m.,p~ u m i -
l- %
Li sp .
_. i m.- . .
M'.v.ts=m.-a f rurr cn '
$r- .-
..a. i.,+ ,- -F ~~ .-? _.
en ;i
.: _ . .. . ,. ~o,.h+1, ~ ,
~%.
~~ t n. x -O.-
r&;j [g w
n("fn-
- ] 9 i..
q i -
a
[ ' M "- a
--- g, -- jq t
l +,1,. , mql .,
.1- -
- .n_4 f - 1>n 7 - ll 4
l :
- E j
- p. ,, ,* f . ..J.. ir-- r
.. . . <? '
L l
[
-4 O b. "'
. --=n , ._g
, ._q l - ',
6
. .i
\ * . * '
,. .+ ,.
. g i ,
MM N , .
e e a a i 4 ,
rd. .d-. 6. d .- ,ld
. := = 4 e .6 en r f
- k. WN I
4 .
unw .
- L 8z ,
, ,M 6 - . . wu s a
.: 5m 9%
gc i.
t--
4 at pB 8E sg a
~ m-t 4M N ) r-
.' 3 S9 ao
- 5~ n
> (,n Q "Ee 4>
w 2A f
s
w I
c lC8 COndanggp (J ,^- '~Q-
- U ,,
3 -
M ns M.
D' w
- _ d ,q -
h y n AIR HAnotggg - "
UNIT g-NTE RMEDI ATE t Q & .
CRANE I '
AIR DISTRisurgon N ,y DCE BASKET
- Ii ;j, l*;,',"f CRANE Watt
~..,I" m
s-ll, : 1 ~
c " ~~
B f44; b
' Ib f f[.;
- ** I
~~ i, . *
( STt Au LATTICE FRAME
' ~~**"~~~~
l'D?$g'- p q! )::]l )N ~N GENERATOR ENCLOSURE
?%q m T] l' s t',*
N~ ,: $1 }. *:.s: , >;*.
~
",;:sL.
' a, y
'I e
' ? .,*r . .
E% o t
.V -
s % :
t'~ /
% I q
- 'i SHELL ,
I W. ql
x i
'l ~j Co '
m h.*
- i WALL PANELS
~
i
_v C'% ,x g!, a t.o::-), [3
- ,< w t*. , . .* ,
- 'h , jjl l f; (.)
, Il GENERATOR Ql .f,{ l TURNgng i ifk,. * ., '
VANES k R%s .l l (* ,1 I
],o
, f .I, "?' . ,
. Q, , - [ A .l1's W;
S$bJ4 %
.- f tW
/
y// .g> [ !l, 'lrA}\ ,.
Y l Nr%L
% 4:
'%' .] lYIDER DECK :
WW
/ ..
- e f % a-
',,, INLET 000Rs
.,q w. , /.
e,' ,
L0wER T 'M* *N. '
. *. SUPPORT STRUCTURE FLOOR s ,
SOMETRIC OF icg C0tIDEllSEP
) McGUIRE NUCLEAR STATION Figure 5,g,g_)
~