ML20215C721
| ML20215C721 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/05/1986 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20215C724 | List: |
| References | |
| NUDOCS 8612150361 | |
| Download: ML20215C721 (11) | |
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Consumers power x in w s.,,,
Derector MM Nuclear Licensing SM M General Omces: 1946 West Pernell Road, Jackson. MI 49201 . (517) 788-1636 December 5, 1986 1
Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555
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DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -
TECHNICAL SPECIFICATION CHANGE REQUEST - HYBRID CONTROL RODS Attached are three (3) originals and thirty-seven (37) conformed copies of a request for change to the Big Rock Poict Technical Specifications. These proposed changes are being requested to reflect the use of new hybrid control rods manufactured by NUCOM, Incorporated and to include Reload I-2 fuel in the technical specifications. NUCOM has previously supplied cruciform shaped control rods to Yankee Rowe, Vermont Yankee and Palisades. The new control rods are currently planned to be installed during the 1987 refueling outage. l
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j Pursuant to 10CFR170.12(c) a check in the amount of $150.00 is included with i this application.
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i Kenneth W Berry Director, Nuclear Licetssing CC Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point Attachments
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- aO-l CONSUMERS POWER COMPANY Docket 50-155 Request for Change to the Technical Specifications License DPR-6 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in the Facility Operating License DPR-6, Docket 50-155, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Plant be changed as described in Section 1.0 below:
I. CHANGES 1.1 Section 5.1.1 Reword the last paragraph beginning with " Control Rods -
Either ...." to read:
" Control Rods:
Neutron Absorbing Material Solid Hafnium (Hf) or Boron Carbide (B 4C) Powder B 4 C Cladding Material 304 or 348 SS Dummy Rod / Tube Material 304 or 348 SS Assembly Sheathing Material 304 SS" 1.2 Section 5.1.2 Reword the first paragraph beginning with "The reactor core .." to read:
"The reactor core shall contain 32 control rod assemblies each composed of a cruciform array of empty or solid stainless steel tubes, boron carbide (B 4C) powder filled stainless tubes and/or solid hafnium (Hf) rods surrounded by a cruciform shaped stainless steel sheath. The individual poison rod composition made up of principal materials boron carbide and/or hafnium shall have an effective poison length of approximately 68 inches."
1.3 Section 5.2.1(b), Table 1
" Reload" Add: "I2" to the last column heading Il so that it will read " Reload" I1/12 .
1.4 Section 5.2.1(b), Table 2
" Reload" Add: "I2" to the last column heading Il so that it will read " Reload" II/I2 .
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TSCR - Hybrid Centrol R:da 2
Big Rock Point-Plant 2.0 DISCUSSION 2.1 Desian Change Discussion The proposed Technical Specification changes requested in parts 1.1 and 1.2 above are a' result from a design change to replacement control blades procured from NUCOM, Inc. by Consumers Power Company. The current General Electric (GE) Type IIA design in use at BRP has been modified in two ways by NUCOM to provide a hafnium / hybrid blade:
i) The top one quarter of the outer two poison rods have been replaced with solid hafnium rods. The solid hafnium rods are seventeen (17) inches long and have the same outside diameter as the stainless steel B4 C (boron carbide) filled tubes being replaced.
ii) The stainless steel cladding which houses the B C4powder in rod form has been changed from 304 stainless steel to high purity 348 stainless steel.
. The NUCOM hafnium / hybrid control blades built for the BRP reactor have been designed to closely match the reactivity worth of the GE type IIA control blades and to be mechanically compatible with all reactor components and control blade handling equipment. Even though there exists material differences between the two control blade designs, the sub-component and overall dimensions between the NUCOM design'and GE design are the same. (See Table I). Also, the NUCOM control blade is slightly lighter (approximately 2 lbs) than the GE control blade, and therefore scram times for the NUCOM blade should be approximately the same as or slightly faster than for the GE blade. Scram time requirements of Big Rock Point Technical Specifications 5.2.2.(a) and 5.1.3 will not be altered.
2.2 Material Acceptability Discussion l 2.2.1 Hafnium Metal i
- A topical report written by General Electric Nuclear Energy Co rp. (Report NEDE-22290-Accepted by the NRC, 08/83)
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discussed the chemical, physical, and mechanical properties of bare hafnium metal when used as a neutron absorbing material in a BWR environment. Three main concerns relating to the use of hafnium were addressed. These concerns were: (a) the increase in assembly weight associated with the use of hafnium, (b) the thermal expansion of hafnium, and (c) the corrosion resistance of unclad hafnium. The NUCOM designed blade incorporates the
! use of approximately 62 cm3 of hafnium into each assembly.
While B4 C has a density (compacted) of approximately 1.75 i gm/cm 3 , hafnium has an approximate density of 13.3 gm/cm3.
Subtracting out the displaced weight of the B 4 C then, the i mil 086-1094A-BT01-NLO4
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TSCR'- Hybrid Centrol R:da 3 Big Rock Point Plant use of hafnium in the NUCOM blade adds approximately 700 plus grams (1.6 lb) to'the total assembly weight._ However, this has no apparent effect on the final control blade weight since the NUCOM blade remains lighter than the GE blade. .With regard to the thermal expansion and irradiation growth consideration, the coefficient of-thermal expansion of hafnium is relatively half that of 304SS (the perforated sheath material). Therefore, the hafnium growth will not affect the integrity of the control blade assembly. With regard to the corrosion of hafnium in a BWR environment, GE presented data in their report showing that the corrosion behavior of hafnium in high temperature water and steam is superior to that of Zircaloy-2.
2.2.2 High Purity 348 Stainless Steel The high purity 348 stainless steel incorporated in the NUCOM modified design is more resistant to stress -
assisted intergrannular stress corrosion cracking (IGSCC) than type 304 stainless steel incorporated into the GE design. In a NUCOM Incorporated report titled " Technical Information Relating to Alternate Materials for BWR -
Control Rod - B 4C - Tube Cladding" (Attachment 2),
extensive research/ analyses results on IGSCC are summarized. The report concludes that the presence of silicon and phosphorus at grain boundaries increases the intergranular corrosion susceptibility of nonsensitized (i.e no carbide precipitates at grain boundaries) commercial austenitic alloys. The neutron absorber rod cladding IGSCC results from internal pressure build-up due to the release of helium formed by the B-10 (neutron, alpha) Li-7 reaction. If the effects of this problem are ignored, the poison rod cladding can fail (caused from the propagation of the IGSCC from the outside surface in) resulting in leaching of B4 C into the primary coolant and consequential loss of power regulation control. As mentioned above, it was concluded in the report that a decrease in resistance to IGSCC occurs when impurities are present at alloy grain boundaries. Specifically, the NUCOM report quantitatively summarized this concept by saying a sharp degradation of resistance to IGSCC occurred when the silicon level exceeded 0.2% and when the phosphorus level j exceeded 0.015%. The silicon and phosphorus levels of the f 304 SS used as the B 4C powder cladding in the GE design are 1.00% max and 0.045% max, respectively. The silicon and phosphorus levels of the 348 SS used as the B C 4 powder
! cladding in the NUCOM design are 0.2% max and 0.015% max, respectively. The 348 SS is clearly the superior alloy to l choose with respect to impurity content, and therefore j should prove to be the optimum material for maximizing mil 086-1094A-BT01-NLO4
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.TSCR'- Hybrid Csatrol Rada 4 Big R ck Paint Plant resistance to IGSCC. It should also be noted that the small grain size #9 of the 348 high purity stainless steel will add to its IGSCC. resistive qualities. The NUCOM report did not address any in-reactor performance of the
. 348 SS, however Exxon Nuclear has adopted the- use of 348 SS in the cladding design of the Dairyland Power Cooperative's Lacrosse BWR replacement fuel. -This' fuel design has been failure-free since its initiation into Lacrosse's fuel cycle in 1979. It is noteworthy to also mention that BRP monitors exposure history and B-10 depletion of each control blade and bases a prudent control blade shuffle and/or discharge program on this data.
2.2.3 Pins and Rollers The pin / roller material requirements for control blades are not specified in Big Rock Point's operating license.
However, since the pin / roller materials have been changed in the NUCOM designed control blades, the materials selection will be discussed and justified in this section for completeness. The GE Type IIA pin / roller combination.
of Haynes 25/ Stellite 3 was changed to PH13-8Mo/Inconel X-750 in the NUCOM hafnium / hybrid design. This design improvement eliminates cobalt bearing stellite material.
The qualification of these materials is based on the results of an extensive Electric Power Research Institute (EPRI) program at GE With the exception of irradiation-induced effects, the program tests simulated a typical five year duty cycle in a BWR. Results of the program, contained in EPRI NP-2329, indicate that the pin / roller combination PH13-8Mo/Inconel X-750 was shown to have as adequate wear resistance as the conventional cobalt bearings alloys for the proposed application.
2.3 Neutronic Analysis Discussion An analysis has been completed by Consumers Power Company titled:
" Big Rock Point Hybrid Control Rod Evaluation" (Attachment 3) which verifies that the control rod worth of the hafnium / hybrid design is virtually identical to the all B 4 C design presently in use. The worth of the hafnium / hybrid control blade design was calculated and compared to the all B4 C control blade design using computer codes MICBURN-1, CASMO-2E, and PDQ-7. Cross sections provided by MICBURN-1, were used by the integral transport code CASMO-2E in comparing all hafnium to all B4 C control rods at various operating conditions. However, CASMO-2E could not be used directly in the evaluation of the hybrid control blade design nor in calculating individual boron poison pin absorptions. Therefore, the cross sections generated by CASMO-2E were saved and input into the industry standard diffusion code PDQ-7 which modeled the actual hybrid control blade design. The two control blade designs were then compared in PDQ-7 using a 2 x 2 assembly configuration in both two and three dimensions with the results listed in Table 2.
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TSCR - Hybrid C:ntrol R:d3 5 Big Rock Point Plant From the PDQ-7 cases listed in TABLE 2, it is clear that the hafnium / hybrid design has almost the same worth as the standard design under all conditions.
2.4 Quality Assurance Discussion NUCOM was established in 1955 and from its beginning specialized in the development and production of control rods and ancillary equipment for the nuclear industry. They have extensive fabricating and welding experience including electron beam welding of all alloy steels, stainless steels, and other metals such as the refractory metals. They also have a well equipped metallurgical laboratory which includes x-ray and non-destructive test facilities. Their welders are certified to weld to the ASME code and NAV SHIPS 250-1500-1, and they have government certified examiners for both x-ray and non-destructive testing examinations.
Quality assurance requirements were carried out in part by a Source Verification Surveillance of NUCOM, Inc. in Great Barrington, Massachusetts. The purpose of the surveillance was to review documentation and verify control of materials for the Big Rock Point control blades. Verification inc7uded documentation reviews for material receipt inspection, testing results, and control of materials. No deficiencies were noted during the surveillance, and the results indicated that NUCOM was complying with their procedures.
The control blade assemblies were visually and dimensionally inspected upon their arrival at Big Rock Point with the results
'(foundtobequitesatisfactory.
2.5. Analysis of No Significant Hazards Consideration This Technical Specification change is being requested to include the use of NUCOM Inc. hafnium / hybrid designed control blades along with the presently used General Electric all B 4C design in the control of reactor power operation. As discussed in the above section, the control blade designs are the same with the exception of material composition. The exterior envelopes of the two designs are of the same dimensions and are mechanically compatible with all of the reactor systems and components. The representative weights of the two control blades are essentially equal. The justifications for use of hafnium, 348SS, and PH13-8Mo/Inconel X-750 have all been discussed. Also, the two control blade designs are of equal neutronic worth.
References:
(A) General Electric Nuclear Energy Corp.
Report: NEDE-22290-A (08/83)
(B) NUCOM, Inc.
Report: Technical Information Relating to Alternate for Materials for BWR-Control Rod-B.C-Tube Cladding (03/83) mil 086-1094A-BT01-NLO4 1
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e J TSCR - Hybrid C:ntrol R:d3 6 Big R:ck Paint Pirnt (C) Consumers Power Company Report: Big Rock Point Hybrid Control Rod Evaluation (10/86)
This proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the NUCOM blade is approximately equivalent in dimensions, weight, and nuclear worth (per Reference C above) with the accepted GE design, and the materials are compatible with the BWR environment (per Reference A, B above) so that the hafnium / hybrid design can be modeled the same as the standard all B4 C control blade design in use. Therefore, neither the postulated reactivity insertion accidents nor the neutron absorption characteristics /
power distri-bution control have changed. This change does not create the possibility of a new or different kind of accident from any previously evaluated because as stated above the hafnium / hybrid design can be modeled the same as the standard all B4 C design presently in use. This change does not involve a significant reduction in the margin of safety because of the approximate equality of weight and nuclear worth between the NUCOM design and accepted GE design (per Reference C above). Consequently, this proposed change does not involve a significant hazards consideration.
3.0 DISCUSSION 3.1 Technical Specification Changes Discussion The proposed Technical Change requested in parts 1.3 and 1.4 above to the Table 1 and Table 2 column headings respectively are editorial in nature to allow inclusion of the Big Rock Point Reload I-2 fuel.
3.2 Analysis of No Significant Hazards Consideration The mechanical, thermal hydraulic, and neutronic analysis for Big Rock Point Reload I-2 is the same as that for Reload I-1. The design report previously issued for Big Rock Point Reload I-1 (Exxon Nuclear Company (ENC) report XN-NF-85-38(P), Rev 0) entitled, " Design Report for Big Rock Point I-1" is applicable for Reload I-2. This reload does not contain any fuel assemblies significantly different from those previously found acceptable by the NRC. This change does not involve a significant increase in the probability or consequence of an accident previously evaluated because the limits are derived in a manner identical to that described in Exxon Nuclear Corporation (ENC) report XN-NF-79-21, revision 1, Big Rock Point LOCA Analysis using the ENC WREM NJP-BWR ECCS Evaluation Model - MAPLHGR Analysis. This report has previously been reviewed and accepted by the NRC and has been used as a basis for issuing previous reload amendments. This change does not create the possibility of a new or different kind of mil 086-1094A-BT01-NLO4
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t 2 TSCR - Hybrid Control Rods 7 Big Rock Point Plant accident from any previously evaluated because the XN-NF-79-21 report covers the required spectrum of break locations, sizes and configurations for the Big Rock Point Plant. This change does not involve a significant reduction in the margin of safety because, as stated in the XN-NF-79-21 report, reactor operation within the proposed limits assures conformance with 10CFR50.46 criteria for maximum cladding temperature, metal-water reaction and hydrogen release. Consequently, this proposed change does not involve a significant hazards consideration.
4.0 CONCLUSION
The Big Rock Point Plant Review Committee has reviewed this Technical Specification Change Request and has determined that this change does'not involve an unreviewed safety question and therefore involves no significant hazards consideration. This change has also been reviewed under the cognizance of the Nuclear Safety Board. A copy of this Technical Specification Change Request has been sent to the State of Michigan official designated to receive such Amendments to the Operating License.
Consumers Power Company requests approval of this application for license amendment as soon as possible. Six of the NUCOM control blades and twenty I-2 Exxon fuel assemblies are to be installed in the Big Rock Point BWR during the next refueling outage which is tentatively scheduled for the beginning of January, 1987.
CONSUMERS POWER COMPAhY By
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F W Buckman, Vice President Nuclear Operations Sworn and subscribed to before me this 5th day of 1)ecember 1986.
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Elaine E Budhrer, Notary Public Jackson County, Michigan My commission expires October 31, 1989 l
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TSCR - Hybrid Centrol Rods 8 Big Rock Point Plant s TABLE 1 CONTROL BLADE
- DESIGN DATA Component NUCOM - Hafnium / Hybrid Bil de General Electric - Type IIA Blade Neutron Absorber / A total of 26 rods / wing - 104 A total of 26 rods / wing - 104 rods /
Control Blade rods / blade. All rods vertical blade. All rods vertical and placed 3
Wings and placed linearly within each linearly within each wing. All B 4 C wing. All B 4 C powder encased powder encased within 304 stainless within 348 stainless steel clad- steel > cladding and vibratorily com-ing and vibratorily compacted to pacted to 1.76 1 0.13 m/cm3 . Each 1.75 1 0.15 gn/cm8 . Each wing wing consists of 10 interior empty consists of 10 interior empty rods followed by 16 B 4 C powder fill-a rods followed by 14 B 4 C powder ed rods. All rods enclosed in a filled rods with 2 exterior perforated 304 stainless steel sheath rods consisting of hafnium metal welded to a central tie rod.
rodlets (top 25%) and B C4 powder filled rods (botton 75%). All rods enclosed in a perforated 304 stainless steel sheath welded to a central tie rod.
1 Pins / Rollers PH 13-8 Mo/Inconel X-750 Haynes 25/ Stellite 3 Dimensions and Weights Neutron Absorber Rods:
B 4 C Cladding ID/OD 0.138 1 0.003 in./0.188 1 0.003 in. 0.138 i 0.003 in/0.188 1 0.002 in.
B 4 C Cladding Thickness 0.025 1 0.002 in. 0.0150 1 0.0025 in.
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TSCR - Hybrid Centrol R:ds Big Rock Point Plant
- 9 TABLE 1 (Cont'd) .
CONTROL BLADE
- DESIGN DATA i
Component NUCOM - Hafnium / Hybrid Blade General Electric - Type IIA Blade B 4 C Cladding Material High Purity 348 SS with ASTM 304 33 with Si-1.00% Max, and Grain Size 9 or finer and P - 0.045% Max Si - 0.2% Max, P - 0.015% Max and meets requirements of ASTM Spec. A269 Short B C4 Rod Length / 51.625 + 0.050 in, - 0.000 in./ N/A / N/A Active Absorber Length 51 in.(Reference)
Long B4C Rod Length / 68.625 + 0.050 in. -0.000 in./ 68.56 1 0.05 in./67.80 in. Max Active Absorber Length 68 in.(Reference) l B4 C Inventory / Density 1800 gm (Approx.)/1.75 1 0.15 1860 gm (Approx.)/1.76 1 0.13 gm/cm 3 gm/cm 3 Hafnium Rodlet OD/ 0.188 1 0.003 in./17.000 1 0.005 in. N/A / N/A Length i Hafnium Inventory 823 gm (Approx.) N/A Control Blade Assembly:
$ Sheath Wall Thickness 0.056 1 0.003 in. 0.056 1 0.003 in.
I Roller Width 0.482 i 0.003 in. 0.482 1 0.003 in.
Span of Wings 11.531 in. Max 11.531 in. Max
, Assembly Overall Length 84.750 in.(Reference) 84.765 in.(Reference)
!j Assembly Wei;ght Approx. 113 lbs. Approx. 115 lbs.
- Control Blade: Consists of 4 - wings assembled in a cruciform shape.
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e o TSCR - Hybrid Control Rods 10 Big Rock Point Plant TABLE 2 2-DIMENSIONAL 2X2 ASSEMBLY PDQ-7 CALCULATIONS 0% VOID STANDARD HYBRID DELTA K, 1,22078 1.18908 1.18922 0.00014 WORTH -
0.02184 0.02174 0.00010 25% VOID STANDARD HYBRID DELTA K, 1.19722 1.16221 1.16238 0.00017 WORTH -
0.02516 0.02504 0.00012 3-DIMENSIONAL 2X2 ASSEMBLY PDQ-7 CALCULATIONS OROD STANDARD HYBRID DELTA K, 1.20067 1.17372 1.17378 0.00006 WORTH -
0.01912 0.01908 0.00004 Where: K ,= infinite multiplication factor D
WORTH = control blade worth = K, - K, (K, ) (K, )
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