BSEP-97-0458, Application for Amends to Licenses DPR-71 & DPR-62,revising Description of Control Rod Assemblies,Per Description Contained in NUREG-1433,Rev 1
| ML20198P262 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/29/1997 |
| From: | Lyash J CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198P267 | List: |
| References | |
| RTR-NUREG-1433 BSEP-97-0458, BSEP-97-458, NUDOCS 9711070042 | |
| Download: ML20198P262 (20) | |
Text
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- CP&L Carolina Power & Light Company PO Box 10429 Southport. NC 28401-0429 OCT 2 01997 SERIAL: BSEP 97 0458 10 CFR 50.90 TSC 97TSB14 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 BRUNSWICK STEAh! ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50 324/ LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AhiENDhiENTS DESCRIPTION OF CONTROL ROD ASSEhiBLIES Gentlemen:
In accordance with the Code of Federal Regulations (CFR), Title 10, Parts 50.90 and 2.101, Carolina Power & Light (CP&L) Company is requesting a revision to the Technical Specifications for the Brunswick Steam Electric Plant (BSEP), Unit Nos. I and 2. The proposed license amendments revise the description of control rod assemblies in Technical Specification 5.3.2. CP&L is proposing to rep ace the existing Technical Specification 5.3.2 with the description contained in NUREG 1433, Revision 1, " Standard Technical Specifications General Electric Plants, BWR/4," dated April 1995. This proposed change is being requested to support the replacement of a portion of the BSEP Unit No. I control rod assemblies with a different design than currently used. The control rod assembly replacements are currently planned to occur during the BSEP, Unit No. I Refueling Outage 11 (i.e , B112RI). This refueling outage is scheduled to begin on hiay 2,1998; therefore, approval of the proposed license amendment for BSEP, Unit No.1 is needed by April 25,1998. ! 7 To allow time for procedure revision and orderly incorporation into copies of the Technical / Specifications, CP&L requests that the proposed license amendments, once approved by the jl NRC, be issued with an implementation date of within 30 days following issuance. CP&L is providing, in accordance with 10 CFR 50.91(b), hir. hiel Fry of the State of North Carolina a copy of the proposed license amendments. g)l) ' ('/ 0 P 9 0 9711070042 971029 ! blb!' il,'ll'a '. PDR ADOCK 05000324 P PDR l j
Document Control Desk BSEP 97 0458 / Pege 2
. Please refer any questions regarding this submittal to hir Keith R. Jury, hianager - Regulatory Affairs, at (910) 457 2783.
Sincerely, Jfcep4..yash Plant General hianager Brunswick Steam Electric Plant WRhi/wrm
Enclosures:
- 1. Basis f)r Change Request
- 2. Proposed Updated Final Safety Analysis Report Changes
- 3. 10 CFR 50.92 Evaluation
- 4. Environmental Considerations
- 5. Page Change Instructions
- 6. Typed Technical Specification Pages - Unit 1
- 7. Typed Technical Spxification Pages - Unit 2
- 8. hiarked up Techrec6 'pecification Pages - Unit '
- 9. hiarked up Technbl Specification Pages - Unit 2 Jeffrey J. Lyash, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are ollicers, employees, and agents of Carolina Power & Light Company.
, 0 , h k'(GV Notary (Scal) hty commission expirer S
- h!( lhh) .
Document Control Desk llSEP 97-0458 / P:ge 3 cc (with enclosures): U. S. Nuclear Regulatory Commission, Region 11 , ATIN: hir, Luis A. Reyes, Regional Administrator Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta,GA 30303 U. S. Nuclear Regulatory Commission ATIN: hir. Charles A. Patterson, NRC Senior Resident inspector 8470 River Road Southport,NC 28461 U. S. Nuclear Regulatory Commission ATTN: hir. David C. Trimble, Jr. (hiail Stop OWFN 141122) 11555 Rockville Pike Rockville, hiD 20852 2738 The lionorable Jo A. Sanford Chaimian - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 hir. hiel Fry Director - Division of Radiation Protection North Carolina Department of Environment, llealth, and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221
. i ENCLOSUREI BRUNSWICK STEAhi ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50 325 AND 50 324/ LICENSE NOS. DPR 71 AND DPR-62 REQUEST FOR LICENSE Ah1ENDh1ENTS DESCRIPTION OF CONTROL ROD ASSEh1BLIES BASIS FOR CHANGE REQUEST CURRENT REOUIREMENI Technical Specification 5.3.2 states:
The reactor core shall contain 137 control rod assemblies, each consisting of a cruciform array of stainless stee! tubes containing approximately 143 inches of twon carbide, ll.C, powder or hafnium absorber rods surrounded by a cruciform shaped stainless steel sheath. PROPOSED CilANGE The reactor core shall contain 137 cruciform shaped con;rol rod assemblies. The control material shall be boron carbide or hafnium absorber material as approved by the NRC. DISCUSSION The control rod assemblies currently installed in the reactors for the Brunswick Steam Electric Plant (BSEP), Unit Nos.1 and 2 were designed and fabricated by General Electric (GE) Company. Three types of control rod assemblies are currently used in BSEP, Unit Nos. I and 2. One type is a control rod assembly containing only boron carbide (i.e., B,C) absorber rods. This control rod assembly has been in use since the initial start up of both BSEP units. The second type is a hybrid control rod assembly which uses boron carbide absorber tubes and three solid hafnium rods in the outside edge of each wing. The third type is a hybrid control rod assembly which uses boron carbide absorber tubes, solid hafnium strips as replacements for select boron carbide absorber tubes, and a hafnium absorber plate at the top of each wing. During BSEP, Unit No.1 Refueling Outage 11 (i.e., B112Rl), Carolina Power & Light (CP&L) Company is planning to install hiodel CR 82hi-1 control rod assembly manufactured by Asea Brown Boveri(ABB) as replacements for some existing BSEP, Unit No. I control rod assemblies. ABB control rod assemblies of similar design have been in use in United States boiling water reactors (BWRs) for over 10 years. To support the use of the ABB control rod assemblies, CP&L proposes to adopt the description of control rod assembly used in NUREG-1433, Revision 1," Standard Technical Specifications General Electric Plants, BWR/4," dated April 1995. The improved "Stcr.dard Technical El-1
Specifications" provide a description of the number and shape of control rod assemblies, as well a stipulation that NRC approved absorber material be used in control rod assemblies. Using the approach established by NUREO 1433, the description of the control rod assemblies, previously described in the Custom Technical Specifications, is being relocated to the Updated Final Safety Analysis Report (i.e., the Updated FSAR). A proposed revision to the Updated FSAR '.o incorporate the relocated information on control rod assembly design, as well the new infonnation pertaining so the ABB control rod assembly design, is provided in Enclosure 2. Suffici:nt information relating to the design of control rod assemblies will exist in Technical Specification 5.3.2 to ensure any changes which may affect safety will require prior NRC review and approval. Since the information with a potential to affect safety is sufficiently addressed by the Technical Specifications, the criteria in 10 CFR 50.36(c)(4) for including the relocated information as Design Features is not met. Therefore, the relocated information is not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Future changes to the information being relocated to the Updated FSAR will be controlled in accordance with 10 CFR 50.59. ADDITIONAL INFORMATION: By letter dated February 20,1986, the NRC accepted Licensing Topical Report TR-UR-85-225 entitled " ASEA-ATOM Control Blades for U. S. BWRs." The report is an update to an earlier report, ASEA-ATOM Report TR-bR-82-98, Revision 1, which described two different ABB control rod assembly designs. One of the control rod assembly designs covered by Licensing Topical Report TR-UR-85-225 is designated as " Type 4." CP&L plans to install Model CR 82M-1 control assemblies, which differ slightly from the approved " Type 4" model. The differences between the " Type 4" assembly and the Model CR 82M-1 assembly are summarized below:
- 1. The " Type 4" control rod assembly wings are 316L stainless steel rather than 304L stainless steel. This provides enhanced resistance to cracking.
- 2. Cobalt content is reduced in the Model CR 82-1 compared to the " Type 4" model.
This provides a reduction in gamma radiation source term.
- 3. The nuclear exposure life of the " Type 4" control rod assembly (i.e.,42 percent Boron-10 depletion la a quarter segment, corresponding to approximately 2.5 snyt) was based on a 10 percent reduction in cold worth relative to the control rod assembly's initial worth. The nuclear exposure life of the Model CR 82M-1 is 5.1 siwt in the most limiting axial quarter segment based on GE's current definition of control rod assembly life 6.e., a 10 percent reduction in cold rod worth in any axial quarter segment relative to the original equipment control rod assembly design).
El-2
The use of 316L stainless steel, rather than 304L stainless steel, will n u .itect any physical arrangement or orientation of any system. The physical properties of the Model CR 82ht 1 centrol rod assen blies are equivalent to, or better than, the " Type 4" assemblies. The ABB Model CR 82hi 1 control rod assembly uses solid 316L stainless steel wings with short, horizontally drilled absorber holes. Compacted boron carbide powder is the primary absorber, with solid hafnium rods inserted in the absorber holes in the leading 6-inches of the absorber region of each wing. The holes are closed by a stainless steel bar welded along the outer edge of each wing. A narrow slit in the bar allows each absorber hole in each wing to communicate so that helium pressure equilibrates in each wing withors allowing significant displacement of the boron carbide powder. No sheath is used on the wings. Guide pads are used, instead of pins and rollers, on the leading end of the control rod assembly. The key physical dimensions of the ABB control rod assemblies are virtually identical to the GE models currently in use at BSEP. The key dimensions fall within the tolerances allowed for GE control rod assemblies. The velocity limiter design is identical to the GE design, and is manufactured according to GE's drawings in accordance with an agreement with GE. The weight of the ABB control rod assembly is appro::ii ately 218 pounds, as compared a weight of approximately 235 pounds for the original equipment GE control rod assemblies. The coupling socket of the ABB control rod assembly is manufactured to the same dimensions as the GE control rod assemblies currently in use at BSEP. The ABB control roussemblies use guide pads instead of the stellite pins and rollers used on most of the GE original equipment control rod assemblies. The Model CR 82M 1 control rod assembly is considered " matched worth" (i.e., having a worth within phis or minus 5 percent of the original equipment control rod assembly worth). The NRC 4 has agreed that, for " matched worth" replacements, no additional analyses are required concerning the rod drop accident or the rod withdrawal error. The actual (i.e., relative) worth differences for this absorber hole diametenitch/ depth design are +2.1 percent cold and
+1.3 percent hot. Shutdown margin is increased slightly (i.e., a conservative change) with this rod worth. In addition, with its reduced weight, but slightly increased drag, the scram time performance of the ABB Model CR 82M-1 remains about the same, or improves slightly, as compared to the GE original equipment model.
Based on an evaluation of the mechanical characteristics and neutronic behavior of the ABB control rod assemblies, CP&L has concluded that use of the ABB control rod assemblies will not increase the probability or consequences of the previously evaluated control rod drop accident or rod withdrawal errrr. t El-3 _ _ _ _ i
ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS DESCRIPTION OF CONTROL ROD ASSEMBLIES PROPOSED UPDATED FINAL SAFETY ANALYSIS REPORT CHANGES L T
BSEP 1 & 2
, , UPDATED FSAR a ring magnet attached at the bottom of the drive piston. The drive piston, piston tube, and indicator tube are all of nonmagnetic stainless steel, allowing the individual switches to be operated by the magnet as the piston
' passes. One switch is located at each position corresponding to an index tube groove, allowing indication at cach latchir.g point. An additional switch is located at each midpoint between latching points, allowing indication of the intermediate positions during drive motion. Thus, indication is provided for each three inches of travel. Switches are provided for the full-in position and full-out position.
One additional switch (an overtravel switch) is located at a position below the nortnal full-out ponition. Because the limit of down-travel is normally provided by the control rod itself as it reaches the backseat position, the index tube can pass this position and actuate the overtravel switch only if it is uncoupled from its control rod. A convenient means is thus provided to verify that the drive and control rod are coupled after installation of a drive or at any time during plant operation. 1 3.9.4.1.1.5 Flange and Cylinder Assembly The fixed components of the drive mechanism (inner cylinder and center tube) are welded to the drive flange. A sealing surface on the upper face of this flange is used to maxe th seal to the drive housing flange. Te flon-coated , stainless steel "O" rings are used for the seals. In addition to the reactor vessel seal, the two hydraulic control lines to the drive are sealed at this face. A drive can thus be replaced without removing the control lines, which are permanently welded into the housing flange. The drive flange contains the integral ball or two-way check (shuttle) valve. This valve directs reactor vessel pressure or driving pressure, whichever is higher, to the underside of the drive piston. Passages in the flange admit reactor vessel pressure to this valve f rom the annular space between the drive and drive housing. A screen intercepts foreign material in the water. Water used to operate the collet piston passes between the outar tube and the cylinder tube. The inside of the cylinder tube in honed to provide the seating surface required for the drive piston seals. Both the cylinder tube and the outer tube are welded to the drive flange. The tops of these tubes have a sliding fit to allow for differential expansion. 1 3.9.4.1.1.6 Coupling Spud. Plug, and Unlocking Tube greMer de .To72) The upper nd of the index tube is threaded to receive a couplir.g spud. The coupling ( gure 3.9.4-2) is designed to accommodate a small amount of angular misalignment between the drive and the control rod. Six spring fingers allow the couplin spud t_o ente _r t The control rod weight g W i m , M,he mating socket on the control to forcerod. pounds) is sufficient the spud fingers to enter ths s~ocket and 'pNup the lock plug, allowing the spud to enter the socket completely and the plug to snap back into place. With the lock plug in place, a force in excess of 50,000 pounds is- required to pull the coupling apart. 25 There are two means of uncoupling the control rod d ve mechaniam (CRDM) and the control rod. First, with the reactor vessel hea removed, sne lock plug may be raised against the spring force (approximately r pounds) using a Tod which extends up through the center of the control rod to an unlocking handle 3.9.4_; Amendment No. I
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(Or M% clearance between channels is 70 mils less that1 the die meter of the roller causing it to slide or skid.instead of roll-. For contr 01 rods that have spacer pads, the cjearance between channels is 70 milsXIess than the face to-face dimension of the spacer. As the rod is inserted about half way, there is a t?ndency for the control rod dEiDihto push inward on the channel, This is a control rod surface to channel surface contact. A " worst case" study indicated a possibility of a 50 mil interference. The possibility of a worst case developing is extremely remote. A statistical analysis utilizing a normal-distribution for each-of the 15 variables indicated that no interference occurs within 3 a limits. where o is
-the standard deviation in a point distribution of events _ and 3 a lies in the 0.995 percentile of probability of nonoccurrence. However, even if interference occurred. the result was negligible. About one pound of lateral force was required to deflect the channel inboard one mil. The friction force developed was an extremely small percentage of the total force available to the control rod drives.
The ]revious discussion presupposed the control rod had not moved when the fuel clannel experienced the largest magnitude of pressure drop. Analysis indicated that the rod was about 70 to 90 percent inserted. If the rod was beyond 70 percent inserted. then no interference was likely to develop because all the channel deformation was in the lower portion of the fuel channel, whereas the roller or spacer pad was at the top of the rod. It is concluded l that the main steam line break accident can pose no significant interference to *.he movement of control rods.
- 3. Jet Pumo Joints An analysis was performed to evaluate the potential leakage from within the floodable inner volume of the reactor vessel l during the recirculation line break and subsequent LPCI reflooding. The two possible sources of leakage are:
1
- a. Jet pump tnrcat to diffuser joint I
- b. Jet pump nozzle to riser joint l The jet pumo to shroud suoport joint is welded and therefore is not a potential source of leakage. The slip joints for all jet pumps leak no more than a total of 225 gpm. The jet pump bolted joint. by analysis, was shown to leak no more than 582 gpm for the pumps through which the vessel was being flooded.
The summary o' t sximum leakage rate is: Total leakage rate through all throat to
. diffuser joints 225 gpm Total leakage rate through all nozzle to riser joints 582 gpm Total maximum leakage rate' 80i gpm LPCI capacity is sized to acccamodate 3000 gcm leakage at these locations. It is concluded that the reactor vessel internals reta n su"1clent integrity durina the recirculation line creak accident to Ocw e"ocdtng of the 1cner solume of tne reactor vesse!
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BSEP 1 & 2 UPDATED FSAR CHAPTER 4 4 REACTOR 1 LIST OF FIGURES (Cont'd) FIGURE TITLE 4.4.2 CLAD TEMPERATURE VERSUS HEAT FLUX BOL 00 2 d 4.4.2-8 CLAD TEMPERATURE VERSUS HEAT FLUX - 5 YEARS U0 2 l 4.4.3-1 OPERATING MAP 4.4.4-1 FUEL ASSEMBLY INITIAL ENRICHMENT DISTRIBUTION 2.1 AVERAGE ENRICHMENT l 4.4.4-2 LOCAL POWER FACTORS 0 MWD /t 40% VOIDS l 4.4.4-3 LOCAL POWER FACTORS 10,000 MWD /t 40% VOIDS l 4.4.4 ' GROSS PEAKING FACTOR AS A FUNCTION OF EXPOSURE BRUNSWICK 1 lg +.4.H D[rolifeY2.3 velNod
'h41-2 ABB CRSEM-l C.onwol Rod a
4-xiii Amendment No. 11
BSEP 1 & 2 UPDATED FSAR In order to simplify the analysis of fuel system damage due to excessive rod internal pressure, the SRP states that rod internal pressures should remain below the nominal reactor coolant system (RCS) pressure during normal operation unless otherwise justified. GE has justified that when the rod is limited so that.the instantaneoas cladding creepout rate, due to internal rod pressure greater than RCS pressure, is not expected to exceed the instantaneous fuel swelling rate, i.e., the fuel to cladding gap does not open. This is acceptable with respect to items (a) through (d) described above, i 4,2.1.1.7 Hydraulic Loads The SRP calls for the fuel assembly holddown capability (wet weight and spring forces) to exceed worst-case hydraulic loads for normal operation including AOos. The fuel assembly is evaluated to ensure that vertical liftoff forces are not sufficient to unseat the lower tie plate from the fuel support piece to such a degree that the resulting loss of lateral fuel bundle positioning would interfere with control blade insertion. Two separate aspects of channel box deflection are considered: channel bulge and channel bow. 4.2.1.1.8 Control Rod Reactiv **
$t_b d Con:rol red reactivitt can ce'io~st -by Ieacning of certtin poison materials if the control rod cladd:ng is breached due to cracking ".aused by stress corrosion resulting f om solidification (sintering) of the boren carbide (B4C) particles in the rods 'follcwed by swelling (due to helium and lithium) of the sintered B C.4 Loss of boron due to neutron irradiation is another means of reactivity loss.
The current generic criterien defining end of control blade life is a loss of total-reactivity equal to 10 percent of initial control blade worth. Control reactivity in all circumstances, including boron carbide (B4 C) leaching and boron depletions, will be maintained such that the core sPall be capable of being rendered suberitical at any time or at any core conditions with the highest worth control rod fully withdrawn. 4.2.1.2 Fuel Rod Failure Subsections 4.2.1.2.1 through 4.2.1.2.3 apply to normal operation; Subsections 4.2.1.2.4, 4.2.1.2.5, and 4.2.1.2.7 apply to anticipated operational occurrences; and Subsections 4.2.1.2.6, 4.2.1.2.0, and 4.2.1.2.9 apply to postulated. accidents. Details with respect to limits and evaluations in the following areas are contained in GESTAR II, Section 2.2.2 (Ref . 4. 2.1 -1) . When the failure thresholds are applied to normal operation including , anticipated operational occurrences, they are used as limits since fuel failure under those conditions should not occur according to the traditional conservative interpretation of General Design Criterion 10. When these thresholds are used for postulated accidents, fuel failure is permitted, but the resulting radiological doses must be within the limits required by 10 CFR 100. 4.2.1 3 Amendment No. 10
BSEp 1 & 2 EO MGE_ 8 ~g </3 UTDATED FSAR-
- 3) Operate individually to that a failure in one positioning device does not affect the operation of any other poritioning device.
4)- k rgize individually when rapid control rod insertion (scram) is signaled so that failure of a power source external to the positioning device does not prevent the insertion of other control rods.
- 5) Lock to its control rod to prevent undesirable separation.
c vity y o to accomplish these functions: standard control'r'd assemblies containing 5 C 3- absorber rods and hybrid control rod assemblies wi " .so contain hafnium-- 4 absorber material.- Figure 3.9.4-1 is representad rod assemblies. __ both types of control Q Power distribution in the core is controlled during'the operation of the reactor by manipulation of selected patterns of control rods. The control rods are positioned in a manner which counterbalances steam void effects at he top of d results in significant power flattening.
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w pg The followingfqua nty control tests performed on @ control rod absorber p tube Hater' integrity _of the tubing _and anJ plug i_s_ verified by ultrasonic _
%[inspection a) 4:terial IA% TW T i ML detWG -{-for PDte.-typ (W$e_ wi% hilled Ab4drber$ is tAtr y _lpid PR tmut miemtion .
. b)- The boron-10 fraction of (the boron contenti of) each lot of
]~g boron-catbide is verified.
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; d c) Weldintegrityofthefinishedabsorbertubhis7erifiedbyhelium
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The control rod velocity limiter (Figures 3.9.4-1 and 3.9.4-3) is an integral Part of the bottom-assembly of each control rod. This engineered safety feature _ protects against a high reactivity insertion rate by limiting the control rod velocity in the event of a control rod drop accident. It is a one-way device;'the control rod scram velocity is not significantly affected, but the control rod dropout valacity is reduced to a permissible limit. The velocity limiter is in the form of two nearly mated conical elements that act as a.large clearance piston and baffle inside the control rod guide tube. It is effective for the length of the control rod stroke. The hydraulic drag forces on a_ control rod are approximately proportional to the square of the rod velocity and are negligible during normal rod withdrawal or rod insertion. However, during the scram stroke the rod reaches high velocity and the drag forces could become appreciable. To limit control rod velocity during dropout but not during scram, the velocity limiter is provided with a streamlined profile in the scram (upward) direction. Thu's, when the control rod is scrasused, the velocity limiter
. assembly offers little resistance to the flow of water over the smooth surface 4.6.1-2 Amendment No. 3
- _m 97 g_/og g _ g - 0F Inseth Bree types of control blade models are employed in Units 1 and 2 to accomplish these functions: the original equipment design, the " hybrid" design, and the Duralife-230 (D-230) design. The original equipment design contains B yC abr,orber rods, while the hybrid and D-230 designs contain Bf and hafnium absorber. Figure 3.9.4-1 is representative of the original equipment and hybrid designs. He D-230 design shown in Figure 3.9.4-X incorporates additional design improvements. _/
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The following supersedes the above 97FSAR-108 Changes: Inserth Up to four types of control rods are employed in Units 1 and 2 to accomplish these functions: the original equipment design, the Hybrid design, the Duralife-230 (D-230) design, and the CR82M-1 design. The first three designs consist of a cruciform array of stainless steel tubes containing 4C powder surrounded by a stainless steel sheath. The Hybrid and D-230 designs also incorporate solid hafnium rods or strip within the sheath, while the D-230 design has in addition ha.nium plate within the sheath. The CR82M-1 design is a cruciform array containing Bf powder or hafnium absorber rods in horizontally drilled holes in solid stainless steel wings. The length of the absorber-containing region in all control rod models is approximately 143 inches. Figure .1.9.4-1 is representative of the original equipment and Hybrid designs. The D-130 design is shown in Figure 4.6.1-1, and the CR82M-1 design is shown in Figure 4.6.1-2. e e
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, _ . . UPDATED FSAR, nW' .
d{ Y OYh fWi _ & Alf. hough highly nli ely, the existence of an occasional defect can be postulated. As n example, a single tube (one of 84 which make a single blade of a control to might fati prior to attaining the design boron burnup. A tube rupture could permit erosion of the boron carbide, thereby affecting the nuclear worth of the control rod; however, mechanical operation of the control rod woul' ot be icpaired. Engineering tests have demonstrated that t,ursting of the tuees results in only slight bulging of the U-shaped stainless steel sheaths which surround the tn>>es. The minor bulging of the sheaths would not impair free movement of the control rods into and out of the core. Even if all the B 4C vere lost from a single failed tube, the control worto of the rod would 7)e reduced by such a small amount that no operating or shutdown margin problem would be created. Replacement of control rods having exposures less than or equal to the limit assures that gross failure will not cecur. 4 t 4.6.2-9
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ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS DESCRIPTION OF CONTROL ROD ASSEMBLIES 10 CFR 50.92 EVALUATION Carolina Power & Light (CP&L) Company has concluded that the proposed change to the Brunswick Steam Electric Plant (BSEP), Unit Nos. I and 2 Technical Specifications revising Technical Specification 5.3.2 to replace the existing description of control rod assemblies with the description contained in NUREG-1433, Revision 1," Standard Technical Specifications General Electric Plants, BWR/4," dated April 1995, does not involve a Significant Hazards Consideration. This change is being proposed to support the planned replacement of a portion of the control rod assemblies with a different design than currently used. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.
- 1. The proposed license amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated. Relocation of the control rod assembly descriptive information from the Technical Specifications to the Updated Final Safety Analysis Report will ensure that adequate control of the infonnation is maintained.
Any changes to this design information must conform with the requirements of 10 CFR 50.59. Restricting use of control rod assembly absorber materials to those listed, or to materials that have been approved by the NRC, will ensure any changes which may affect safety require prior NRC review and approval. Since the information with a potential to affect safety is sufficiently addressed by the Technical Specifications, the criteria of 10 CFR 50.36(c)(4) for including the relocated information as Design Features are not met. Because the relocated information is not required to be in the Technical Specifications to provide adequate protection of the public health and safety, relocation of control rod assembly descriptive information will not increase either the probability or the consequences of an accident previously evaluated.
- 2. The proposed license amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated. Relocation, to the Updated Final Safety Analysis Report, of the information pertaining to the control rod assembly designs ensures that adequate control of the information will be maintained. Since the information with a potential to affect safety is sufficiently addressed by the Technical Specifications, the criteria of 10 CFR 50.36(c)(4) for including the relocated information as Design Features are not met. Because the relocated information is not required to be in the Technical Specifications to provide adequate protection of the public health and safety, the E3-1
proposed Technical Specification changes to relocate the control rod assembly design information to the Updated Final Safety Analysis Report does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed lic:nse a nendments do not involve a significant reduction in a margin of safety. As discussed in items 1 md 2 above, relocation of the control rod assembly descriptive information from the Technical Specifications to the Updated Final Safety Analysis Report will ensure that adequate control of the information is maintained. Any changes to this design information must conform with the requirements of 10 CFR 50.59.
Restricting use of'.ontrol rod assembly absorber materials to those listed, or to materials that have been approved by the NRC, will en,ure any changes which may affect safety , require prior NRC review and approval. The information with a potential to affect safety is sufficiently ad iressed by the Technical Specifications, therefore, the proposed Technical Specification changes to relocate control rod assembly design information to the Updated Final Safety Analysis Report do not involve a significant reduction in a margin of safety. E3-2
ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50 325 AND 50-324 LICENSE NOS DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS DESCRIPTION OF CONTROL ROD ASSEMBLIES ENVIRONMENTAL CONSIDERATIONS Carolina Power & Light (CP&L) Company has concludea tet the proposed change to the Brunswick Steam Electric Plant (BSEP), Unit Nos. I and 2 Tt.~hnical Specifications revising Technical Specification 5.3.2 to replace the existing description of control rod assemblies with the description contained in NUREG 1433, Revision 1," Standard Technical Specifications General Electric Plants, BWR/4," dated April 1995 is eligible for categorical exclusion from performing an environmental assessment. , in support of this determination, an evaluation of each of the three (3) criteria set forth in 10 CFR 51.22(c)(9) is provided below.
- 1. The proposed license amendments do not involve a significant hazards consideration, as shown in Enclosme 2.
- 2. The proposed license ame '.ments do not result in a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite. Tre proposed license amendments do not require any equipment or systems to perform a different type of function than they are presently designed to perform. The proposed license amendments do not alter the function of existing equipment and will ensure that the consequences of any previously evaluated accident do not increase. Therefore, CP&L has concluded that there will not be a significant increase in the types or amounts of any ellluent that may be released offsite and, as such, does not involve irreversible environmental consequences beyond those already associated with normal operation.
- 3. Relocation of the control rod assembly descriptive information from the Technical Specifications to the Updated Final Safety Analysis Report is an administrative change that does not result in an increase in individual vaumulative occupational radiation exposure.
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