ML20097G377
| ML20097G377 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/12/1996 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9602210190 | |
| Download: ML20097G377 (15) | |
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I Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Montce!:o. Minnesota 55362-9637 February 12,1996 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Notification of a Deviation from Licensing Basis information Concemino Tornado Effects on the Reactor Buildina Superstructure i
This submittal provides notification to the NRC of an identified condition which is a deviation in implementation of the Monticello licensing basis. Evaluations performed as part of the Monticello design basis reconstitution project have determined that information, incorporated into the plant licensing basis by reference, is inconsistent with design basis information for the Monticello plant. In addition, this submittal provides clarification of licensing basis information submitted in response to an NRC generic letter dated May 17,1978, conceming the movement of heavy loads.
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The Monticello staff has identified that information incorporated by reference into the Updated Safety Analysis Report (USAR), conceming the capability of the Reactor Building superstructure to withstand tomado wind loading conditions, is inconsistent with the design basis for the structure as well as specific licensing basis statements contained in the USAR.
Attachment A of this submittal provides a discussion of the Monticello Reactor Building design
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basis pertaining to wind and tomado phenomena, the licensing basis pertaining to wind and tomado phenomena, identification of the deviation from the facility licensing basis, a safety significance evaluation, and our pianned corrective actions to correct this deviation. The planned corrective actions restore compliance with the current licensing basis. We have confirmed that the intent of the licensing basis has been satisfied with the existence of the identified deviation in that the guidelines of 10 CFR Part 100 are not exceeded as a result of postulated events. The Monticello staff has evaluated this concem for reportability in accordance with regulatory requirements. We have determined that the issue does not meet the criteria for reportability as contained in 10 CFR 50.73,10 CFR 50.72,10 CFR Part 21,10 CFR 50.9, or the plant technical specifications. We have evaluated the potential for an adverse impact on the public health and safety. We have determined that a substantial safety hazard does not result from the postulated event and that a reasonable assurance of safety is maintained.
Attachment B provides clarification of information provided to the staff in response to an NRC generic letter dated May 17,1978. The generic letter requested information conceming the 2496 NsP H:\\ DATA \\NRCCoRR4A78047A. ooc 9602210190 960212 DR ADOCK0500g3
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movement of heavy loads in the vicinity of spent fuel and the protection of stored spent fuel.
This clarification is subject to resolution of the licensing basis deviation described in Attachment A of this submittal.
This submittal is provided to inform the NRC staff of deviations from the licensing basis as described herein. The issue identified in Attachment A is to be resolved in accordance with the provisions established by 10 CFR 50.59. Information conceming the licensing basis issue described in Attachment A has been provided to the nuclear industry via the Institute of Nuclear Power Operations (INPO) Nuclear Network due to potential industry interest. This submittal provides the following new commitment to the NRC to correct the deviation descrited in Attachment A:
Modifications are to be implemented to resolve the deviation from the current licensing basis. The modification design will ensure that a failure of the structural members of the Reactor Building superstructure will not occur when exposed to tomado winds.
This submittal clarifies licensing basis information as described in Attachment B. This notification is provided in accordance with established guidance conceming modification of docketed non-legally binding commitments. Specifically, the information supersedes the Monticello response to item 9 submitted by letter dated July 21,1978," Control of Heavy Loads Near Spent Fuel"in response to NRC Generic Letter, dated May 17,1978, conceming movement of heavy loads in the vicinity of spent fuel.
Please contact Mary Engen, Sr Licensing Engineer, at (612) 295-1291 if you require further information.
h$ni*? h William J Hill Plant Manager Monticello Nuclear Generating Plant c:
Regional Administrator-til, NRC NRR Project Manager, NRC Sr Resident inspector, NRC State of Minnesota, Attn: Kris Sanda Attachments: (A) Evaluation of Reactor Building Superstructure Licensing Basis Pertaining to Tomado Events (B) Clarification of Commitments Related to Protection of Spent Fuel from Potential Missiles
T Attachment A Evaluation of Reactor Building Superstructure Licensina Basis Portainina to Tomado Events i
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Monticello Reactor Buildino Desian Basis l
l Design basis is defined in 10 CFR 50.2. The definition states:
Design basis means that information which identi6es the speci6c functions to be pertoimed by a structure, system, or component of a facility, and specific values or ranges of values chosen for controlling parameters as reference bounds for design.
The values may be (1) restraints derived from generally accepted " state of the art" practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident Ibr 7
which a structure, system, or component must meet its functional goals.
1 The Monticello Reactor Building is a reinforced concrete structure from the foundation at elevation 888'-3" to the refueling floor at elevation 1027'-8". The Reactor Building superstructurn consists of I-beam framing covered with an outer layer of aluminum siding and
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j sn inner layer of steel siding. The siding extends from the refueling floor level,1027'-8", to the top of the building,1074'-2". The inner layer of steel siding is screwed to horizontal wind girts
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(members spanning the vertical I-beams). The outer layer of aluminum siding is screwed to horizontal sub-girts which are clipped to the inner siding. See Figure 1, page 9, this attachment, " Typical Reactor Building Panel Cross Section." The seams of the outer and inner layers of siding are caulked to minimize building leakage. The reactor building superstructure above elevation 1027'-8"is a non-Class I structure, in accordance with original plant design and any modification thereto. The structure was designed to meet design basis loads pertaining to a Class 11 structure while ensuring that the structure will not collapse during a Class I seismic event. The structure is part of secondary containment; however, it is not required to perform any secondary containment function during a tomado event.
s Class il structures are defined as those structures which are not essential for the safe shutdown of the plant or for the removal of decay heat, but are required for power generation.
The steel superstructure above elevation 1027'-8" provides secondary containment during
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i normal plant operations and during postulated accidents such as a Loss of Coolant Accident (LOCA) or refueling accident. These accidents are not postulated to occur as a result of j
l tomado events nor are tomado events postulated to occur during accident mitigation l
[ Reference Monticello Updated Safety Analysis Report (USAR), Sections 12.2.1.2 and I
i 12.2.2.1.1; and license application amendment 4, dated January 10,1967). The reactor I
building superstructure below elevation 1027'-8" is a Class I structure, in accordance with original plant design and any modification thereto. Class I structures are defined as those structures whose failure could cause significant release of radioactivity or which are vital to j
i safe shutdown of the plant under normal or accident conditions and the removal of decay and
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i sensible heat from the reactor, l
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.l Attachment A February 12,1996
- Page 2 The Reactor Building, including the superstructure, was designed to withstand wind forces based on a maximum wind velocity of 100 mph,30 feet above ground with a gust factor of 1.1.
Increases in wind velocity due to elevation are to be considered in accordance with Table 1(a) of ASCE Paper 3269 [ Reference Monticello USAR Section 12.2.1.6]. The Reactor Building, excluding the steel superstructure, was designed to withstand tomado loading conditions resulting from a rotational wind having a tangential velocity of 300 mph, a differential pressure l
between inside and outside enclosed areas of 2 psi, and a torsional moment resulting from applying the rotational wind on one-half of the structure [ Reference Monticello USAR Sections 12.2.1.4 and 12.2.1.8]. Additional design requirements pertaining to seismic loads, dead and live loads, etc., are provided in section 12.0 of the Monticello USAR.
Reactor Buildina Tornado Event Licensina Basis History i
. The current licensing basis is defined in NRC Inspection Manual, Part 9900, Technical Guidance as:
i Current licensing basis (CLB) is the set of NRC requirements applicable to a speciRc plant, and a licensee's wntten commitments for assuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modification and additions to such commitments over the life of the license) that are j'
docketed andin effect. The CLB includes the NRC regulations containedin 10 CFR Part 2,19, 20, 21, 30, 40, 50, 51, 55, 72, 73,100 and appendices thereto; orders; Iicense conditions; exemptions, and Technical Specifications (TS). It also includes the plant-specific design basis information definedin 10 CFR 50.2 as documentedin the most recent Final Safety Analysis Report (FSAR) as required by 10 CFR 50.71 and the i
licensee's commitments remaining in effect that were made in docketed licensing i
correspondence such as licensee response to NRC bulletins, generic letters, and l
enforcement actions, as weII as licensee commitments documented in NRC safety evaluations orlicensee event reports.
l Northem States Power Company (NSP) submitted an application dated August 1,1966, with the Atomic Energy Commission (AEC), predecessor of the NRC, for all necessary licenses for construction and operation of the Monticello nuclear facility. The facility construction permit was issued on June 19,1967. The information contained in the application, including the Preliminary Safety Analysis Report, and in Amendments 1 through 8 to the application, was evaluated by the AEC staff as well as the Advisory Committee on Reactor Safeguards (ACRS).
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During this phase of the licensing for the Monticello Nuclear Generating Plant, a question was raised by the AEC staff regarding the plant design basis pertaining to tomado events.
Amendment 4 to the Monticello application, submitted January 10,1967, provided a response to an AEC question which stated:
Describe the design basis for the plant to withstand the wind and pressure effects of tomadoes (AEC question 3.9 contained in Amendment 4).
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i Attachment A February 12,1996 Page 3 d
i The Monticello response to this question restated the tomado wind loading design criteria.
Further, the response to this AEC question clearly established the criteria as applicable to structures necessary to protect equipment and systems to permit safe shutdown of the reactor during tomado conditions and established the criteria as applicable specifically to that portion of the reactor building below the refueling floor. This information was subsequently incorporated into the Monticello Final Safety Analysis Report (FSAR) for AEC provisional 3
operating license review and has been incorporated into section 12.2.1.8 of the Monticello USAR.
By amendment 9, dated November 7,1968, to the Monticello license application, NSP filed the Monticello Final Safety Analysis Report (FSAR) in connection with application for a provisional j
operating license. Amendment 15, dated July 2,1969, to the Monticello license application submitted revised information for the Monticello FSAR. Information submitted with this revision j
of the FSAR specifically excluded the reactor building superstructure from the listing of Class I structures and from the consideration of tomado loads presented in section 12.0 of the FSAR.
This information is consistent with the Reactor Building des!gn basis and was subsequently incorporated into the Monticello USAR.
On October 15,1969, NSP submitted amendment 21 to the Monticello license application.
This amendment provided response to AEC staff concems regarding structural and seismic i
matters and a revision to the Monticello FSAR. Section 4.5 of the submittal identified concems raised in AEC staff safety evaluation reports for GE DWR construction and operating permit l
applications for other facilities under review during the time frame of Monticello's provisional operating license process. Section 4.5.6 of amendment 21, titled "Tomado and Missile Protection - GE-BWR - Spent Fuel Storage Pool
- provided resolution for the Monticello plant to p
the AEC staff concem identified in the section title. The resolution provided reference to GE i
. Topical Report APED-5696 (November 1968), "Tomado Protection for the Spent Fuel Storage Pool," which examined (a) whether sufficient water could be removed form the pool to prevent cooling of the fuel and (b) whether missiles could potentially enter the pool and damage the stored fuel. The conclusions from APED-5696 as restated in section 4.5.6 of amendment 21 i
found that:
t The fuel poolin a General Electric BWR reactor building is designed with substantial j
capability for withstanding the effects of a tomado, as this document shows. The design of the fuelpool makes the removal of more than five feet of water due to tomado action highlyimprobable. With 25 feet of water covedng the fuel tacks, the removal oflive feet of wateris of no concem. Protection against a wide spectrum of tomado-generated missiles is provided by the water which covers the fuel tacks. It is l
shown that protection is provided against all tomado-generated missiles having a probability of hitting the pool greater than one per 1.4 billion reactorlit6 times. Typical potentialmissiles in this categoryinclude a spectrum ranging up to a 3-inch-diameter
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steel cylinder 7 feet long or a 14-inch-diameter wooden pole 12 tieet long.
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Attachment A
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i Prior to submittal of amendment 21 to the license application, reference to APED-5696 was incorporated into section 3.4.1 of the FSAR by amendment 14, dated May 15,1969, to the i
Monticello facility license application (which has subsequently been incorporated into i
Monticello USAR section 2.3.4.1). The information incorporated into the licensing basis references GE Topical Report APED-5696, "Tomado Protection for the Spent Fuel Storage Pool," as demonstrating adequate protection of the fuel pool area from tomado phenomena j
including possible effects of missiles.
Deviation from Licensina Basis Evaluations performed as part of the Monticello design basis reconstitution project have j
determined that licensing basis information, incorporated by reference, conceming the protection of stored spent fuel from the possible effects of missiles resulting from tomado i
- phenomenon, is inconsistent with design basis information for the Monticel!o Reactor Building superstructure. Specifically, information contained in GE Topical Report APED-5696, which
- has been incorporated into the Monticello licensing basis via reference, states the reactor i
building superstructure is designed to withstand tomado winds. This statement is inconsistent i
with the design basis and specific licensing basis information contained in the license application, FSAR, and USAR; in that the Reactor Building superstructure was not designed to withstand tomado winds. A review of design specifications and design information for the i
Reactor Building superstructure has been performed to resolve this inconsistency in the l
Monticello licensing basis. This review has confirmed that the Reactor Building superstructure was designed and constructed using design basis loads consistent with a Class ll structure and that the design did not include tomado loads.
i Further analysis of the Reactor Building superstructure's response to tomado phenomena has determined that the structural members will remain intact up to wind velocities of approximately 189 mph. This capability is above the current design basis for wind loading (which is i
applicable to the superstructure), but does not satisfy the tomado loading resulting from a tomado with 300 mph tangential wind speeds (which is beyond the design basis of the i
superstructure). Evaluation of the current superstructure's response to tomado phenomena has determined that the exterior aluminum siding will be blown off, the inner liner on the l
windward side of the structure will bear against the wind girts and remain intact, and the effect 1
of lateral wind loading on the windward side results in excessive deformation which is j
expected to result in structural failure for tomado winds in both the north-south and east-west wind directions. This evaluation conservatively assumed that the tomado winds impact directly perpendicular to the windward face of the structure. For any winds not impacting directly 4
perpendicular to the structure, the siding affixed to the structural members would be damaged; however, the structural members would remain intact.
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Topical report APED-5696 recognizes that damage to the Reactor Building superstructure does occur. This damage, as indicated by the report, would consist of removal of portions of the roof decking and metal siding such that tomado winds could act upon the spent fuel pool water volume as well as deliver potential missiles to the pool. The topical report concludes that I
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Attachment A February 12,1996 I
Page5 the roof docking and siding, as well as other potential low weight and large cross-section missiles, are not of concem for imparting damage to the stored spent fuel. However, the topical report does not recognize the potential for damage to the structural members of the Reactor Building superstructure as the analysis considered the structural members as being designed to withstand tomado winds.
The current licensing basis concoming the protection of the public health and safety is to prevent damage to the stored spent fuel due to the adverse consequences of tomado phenomena. Our review of this issue has concluded that contrary to the conclusion of GE topical report APED-5696, failure of the structural members of the Reactor Building l
superstructure could occur under extreme tomado conditions, such that damage to stored spent fuel could occur as a result of impact by these structural members.
Safety Sionificance of Licensina Basis Deviation Although a structural analysis of the Reactor Building superstructure indicates that the structuralintegrity of the steel superstructure is questionable during the extreme conditions of a tomado, an evaluation of radiological consequences demonstrates that the intent of the j
licensing basis has been maintained and satisfied in that 10 CFR Part 100 limits are not exceeded. Further, we have determined that the postulated event is not credible during a portion of the year and that the event is still of low probability during the portion of the year when a tomado is credible.
The maximum activity for tomado phenomena varies with geographic location. The Monticello plant is located at 45' 20' north latitude and 93* 50' west longitude in Wright County, Minnesota. The latitude of the Monticello plant places it at the northem edge of the region of maximum tomado frequency in the United States. However, only a few tomadoes have occurred in this vicinity. An average of 18 tomadoes per year since 1953 have occurred in e
Minnesota. Of this annual average number of tomadoes,14 have occurred during the months of May, June and July, with the peak month being June. No tomadoes have been observed in l
Minnesota during the months of November through February. A potential for the occurrence of a tomado does exists for the remaining months of the year. [R.A. Keen, Minnesota Weather. p.
l 57). Eight tomadoes were reported in Wright County during the period from 1916 through l
1967.. Monticello USAR section 2.3.4.1 states that the probability of a tomado striking a given point in the area of a 1* square, lying between 45' and 46' north latitude, and 93* and 94*
west longitude, can be calculated to be 5X10" per year, or one tomado every 2000 years.
1 More recent information concoming the effects of tomado phenomena can be found in i
ANSI /ANS-2.3-1983, " Standard for Estimating Tomado and Extreme Wind Characteristics at l-Nuclear Power Sites." According to Figure 3.2-1 of ANSI /ANS-2.3-1983, the Monticello plant site is in an area where the probability of experiencing tomado wind speeds of 320 mph or 4
greater is 10 per year. Similarly, Figures 3.2-2 and 3.2-3 of ANSI /ANS-2.3-1983 show that i
the probabilities of experiencing maximum tomado wind speeds of 260 mph and 200 mph or 4
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greater are 10 and 10, respectively.
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Attachment A 4
February 12,1996
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Page 6 i
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An independent calculation of the tomado strike probability has been performed using the j
more recent methods and data provided in NUREG/CR-4661,"Tomado Climatology of the Contiguous United States." The tomado strike probability was calculated as 1.59X10 per d
j year using the methods of NUREG/CR-4661. Per the methodology contained in NUREG/CR-i 4661, the probability of a tomado of wind speeds greater than 207 mph striking the Monticello site is 6.55x10 per gear, and the probability of a tomado strike with wind speeds greater than i
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i 158 mph is 1.58X10 per year.
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Based on the structure's straight-line wind capacity of approximately 189 mph, the probability of the 158 mph tomado strike would seem to be the appropriate measure of the chance of i
' exceeding the structure's capacity; however, other considerations make the probability of a l
1 207 mph or greater tomado the more correct measure. First, the wind speed given for the j
categories considered represents the vector sum of the radial, vertical, horizontal, and i
translational components of the tomado winds; the straight line winds which the structure i
l would experience during a 207 mph tomado would be less than 207 mph. Second, tomadoes i
vary in severity as they progress, and they are categorized according to the most severe level they attain; a 207 mph or greater wind speed tomado on average is only at these wind speeds l
for 24% of its path, and is at lower wind speeds for the remainder. Therefore, even if a i
tomado is categorized with wind speeds of 207 mph or greater, there is a 76% chance that it would be of a lower wind speed category at the time of impact at the plant site. Finally, the j
i determination of the ability of the structure to withstand wind speeds of up to approximately 189 mph is a conservative calculation. For these reasons,6.55x10, as the probability of a j
4 tomado of wind speeds greater than 207 mph striking the Monticello site per year, is the appropriate comparison to assess the potential for exceeding the structure's capacity.
Despite the low probability of the postulated event,if a tomado resulting in failure of the l
Reactor Building superstructure is postulated to occur such that significant damage to spent j
fuel results, it has been determined that the limits for radiation dose to the public established in 10 CFR Part 100 are not exceeded and that the spent fuel storage pool can maintain water volume. An evaluation has shown that the spent fuel storage pool is able to withstand the i-worst reasonable tomado-generated missile impact into/onto the pool structure based upon the l
acceptance criteria as defined in USAR section 12.2.1.4 for Class I structures, NUREG-0800, section 3.5.3 and NRC Regulatory Guide 1.142. No adverse impact on the pool structural integrity will occur as a result of a bounding impact of the Reactor Building crane or a bounding Impact of a section of the reactor building superstructure. The evaluation determined that no excessive spent fuel storage pool leakage is expected due to the postulated bounding missiles. Penetration of the steel liner may occur; however, water inventory can be maintained
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via the reinforced concrete structure. Should penetration of the poolliner occur, leakage would readily be detected by a Main Control Room system trouble alarm. Isolation of the liner leak detection system would be performed to isolate the leakage. instrumentation monitoring fuel pool water inventory, as well as isolation valves for the liner leak detection system, are i
located in the Class I portion of the Reactor Building and would thus be protected from the i
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Attachment A
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February 12,1996 l
Page 7 P
i adverse impacts of the tomado phenomena. Multiple diverse means of restoring pool water j
' volume are available if necessary.
T With assumed damage to stored fuel in the spent fuel pool, a calculation was performed which j
addresses the radiologicalimplications of ruptured stored spent fuel assemblies. The i
evaluation concludes "... that it would be impossible to exceed 10 CFR Part 100 limits due to i
gap radiation release from the fuelin the fuel pool following damage to the structure due to i
high winds." This analysis assumed:
1)
High winds and unstable atmospheric conditions will exist when the fission products are released from the fuel pool.
i 2)
The isotopic content of the released gap activity is proportional to the core isotopic concentrations 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown.
i 3)
The receptor (exposed person) is exposed to the entire cloud of radioactive j
material; no reduction is taken due to exposure time.
i It can readily be determined that the potential radiation dose to an individual at the site i
l exclusion area boundary as a result of damage to all fuel assemblies stored in a full storage pool is well below the limits contained in 10 CFR Part 100.
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Corrective Actions i
i The Reactor Building superstructure consists of I-beam framing covered with an outer layer of aluminum siding and an inner layer of steel siding. Adjacent to the Reactor Building superstructure north and south wall vertical 1-beam columns are located the Reactor Building i
overhead crane vertical support l-beam columns. To resolve the identified deviation from the current licensing basis, Monticello intends to perform modifications to the Reactor Building superstructure. These modifications will ensure that a failure of the structural members of the i
j Reactor Building superstructure will not occur. Thus the potential for a failure of the superstructure resulting in missiles capable of damaging stored fuel will be eliminated. The i
i modifications will consist of the following.
1)
The north and south walls of the Reactor Building superstructure are to be strengthened by forming a composite beam from the existing vertical 1-beam columns of the superstructure wall and the existing vertical I-beam columns i
which provide structural support to the Reactor Building overhead crane.
2)
The east and west walls of the Reactor Building superstructure are to be modified to provide pressure relief during postulated wind conditions in excess of approximately 160 mph. Pressure relief is to be provided by establishing seams in the inner layer of steel siding along the wind girt such that the inner 4
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Attachment A February 12,1996 Page 8 layer will tear when exposed to the excessive forces of high winds, thus relieving the windward pressure force such that structural integrity is maintained.
These modifications are to be implemented following the Monticello 1996 refueling outage,
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which is currently scheduled to be completed in May of 1996. We have evaluated the risk of introducing foreign material into reactor systems by performing this modification concurrent with refueling outage activities against the risk of the postulated tomado phenomena. We have determined that a greater risk is posed to the facility if the proposed modification activities were to be performed concurrent with outage activities; therefore, we have elected to delay the structural modifications until after the 1996 refueling outage. We estimate that the proposed modificaticn will be implemented within five (5) months of initiating construction activities.
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Attachment A February 12,1996 Page 9 REACTOR BUILDING SUPERSTRUCURE COLUMN WIND GIRT l
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TYPICAL REACTOR BUILDING PANEL CROSS SECTION FIGURE 1 9
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j Attachment B Clarification of Commitments Related to Protection of Soent Fuel frcm Potential Missiles I
By letter dated July 21,1978, Monticello responded to an NRC generic letter, dated May 17 -
L 1978, pertaining to movement of heavy loads in the vicinity of spent fuel. Question 9 of the information request enclosed with the generic letter stated:
Discuss the degree to which your facility complies with the eight (8) regulatory positions 3
delineated in Regulatory Guide 1.13 (Revision 1, December,1975) regarding Spent i
Fuel Storage Facility Design Basis.
Our July 21,1978, response stated:
"The Monticello Plant conforms with the requirement of this guide with one exception.
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There are no interlocks provided to prevent cranes from passing over stored fuel when l
fuel handling is not in progress.... Additional information on how the remaining seven regulatory positions of Regulatory Guide 1.13 are met can be found in Reference 1,13, l
& 16."
l References 1,13 and 16 refer respectively to the Monticello Updated Safety Analysis Report; i
Novsmber 22,1976 letter, " Design Report for Redundant Reactor Building Crane," L O Mayer j
to V Stello; and February 28,1977 letter, "NRC Request for Additional information on the Redundant Reactor Building Crane," L O Mayer to D L Ziemann.
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A design basis review of the heavy loads issue has identified that the above statement is not accurate. Regulatory Guide 1.13, Regulatory Position C.2 states:
"The facility should be designed (a) to keep tomadic winds and missiles generated by these winds from causing significant loss of watertight integrity of the fuel storage pool 1
and (b) to keep missiles generated by tomadic winds from contacting fuel within the pool."
' The Monticello licensing basis clearly states that the facility is designed to maintain the watertight integrity of the fuel storage pool; however, the licensing basis also states that the facility is not designed to keep tomado-generated missiles from contacting the fuel within the fuel storage pool. Amendment 21 to the Monticello Nuclear Generating Plant Final Safety Analysis Report (FSAR), section 4.5.6, responded to an AEC concem regarding spent fuel storage pool missile protection by stating the spent fuel storage pool is protected against a i
wide spectrum of tomado-generated missiles by the water which covers the fuel racks. The response to the AEC concem referenced GE topical report APED-5696, "Tomado Protection for the Spent Fuel Storage Pool", as providing the supporting technical analysis of this issue.
Reference to APED-5696 was incorporated into section 3.4.1 of the FSAR by amendment 14, dated May 15,1969, to the Monticello facility license application (which has subsequently been incorporated into Monticello USAR section 2.3.4.1), and states:
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Attachment B i
February 12,1996 Page 2
" Discussion of the expected effects of tomadoes on the fuel storage poolis included in a topical report, "Tomado Protection for the Spent Fuel Storage Pool" ( November l
l 1968) APED-5696."
The analysis provided in APED-5696 does not demonstrate that tomado-generated missiles will not contact the fuel. The analysis performed does demonstrate that:
1.
Only large, slender objects, with the tomado winds acting on the maximum cross section and the object impacting the fuel pool with its minimum cross section, could potentially impart fuel damage.
2.
Using conservative inputs, the probability of such missiles occurring was 4
demonstrated to be low,7 X 10, and even in this highly unlikely case, a wide spectrum of these missiles can hit the pool without resulting in fuel damage or liner penetration.
3.
A high degree of protection against tomado-generated missiles is provided for the spent fuel storage pool by the water which covers the fuel racks. Neither fuel damage or liner penetration is possible with any reasonable missiles.
The intent of Regulatory Guide 1.13, Regulatory Position C.2, the prevention of mechanical i
damage to stored fuel due to missiles generated by high winds, is demonstrated by the Monticello licensing documentation. We feel that this was the true intent of the statement made in our July 21,1978, letter regarding Monticello's compliance with Regulatory Guide 1.13, Regulatory Position C.2. As such, we wish to clarify the record regarding compliance to Regulatory Guide 1.13. This clarification is subject to resolution of the issues discussed in i
Attachment A of this submittal. The commitment stated in our July 21,1978, letter is modified to state the following (text revisions are indicated by bold text):
I The Monticello Plant conforms with the regulatory positions of Regulatory Guide j
1.13 with the following exceptions:
Position C.2 - Discussion of the expected effects of tornadoes on the fuel storage pool is included in a topical report, " Tornado Protection for the 4
Spent Fuel Storage Pool," ( November 1968) APED-5696. Missile contact with the fuel is not precluded; however, the stored spent fuel is protected against a wide spectrum of tornado-generated missiles by the water which covers the fuel racks or the probability has been demonstrated to be below the regulatory threshold for those missiles with the potential to damage the fuel.
Position C.3 - There are no interlocks provided to prevent cranes from passing l
over stored fuel when fuel handling is not in progress. Such interlocks are considered unnecessary for three reasons. First, it is a good and commonly
Attachment B February 12,1996 Page 3 4
known operating practice that heavy loads not be taken over critical areas; this practice is implemented administratively. Second, the travel path of heavy loads is such that they would not normally be carried over spent fuel; traveling over spent fuel would increase the travel path. Third, the redundant features of the modified crane as discussed in References 2 and 3 are such that the drop of any heavy load is highly improbable.
Additionalinformation on how the remaining seven regulatory positions of Regulatory Guide 1.13 are met can be found in References 1,2, & 3. This Information supersedes the Monticello response to item 9 submitted by letter dated July 21, 1978, " Control of Heavy Loads Near Spent Fuel," in response to NRC Generic Letter, dated May 17,1978, concerning movement of heavy loads in the vicinity of spent fuel.
References:
1 - Monticello Updated Safety Analysis Report 2 - November 22,1976, Design Report for Redundant Reactor Building Crane, L O Mayer to V Stello 3 - February 28,1977, NRC Request for Additional Information on the Redundant Reactor Building Crane, L O Mayer to D L Ziemann l
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