ML20036C056
| ML20036C056 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/28/1993 |
| From: | Kopp L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20036C055 | List: |
| References | |
| OLA, NUDOCS 9306090228 | |
| Download: ML20036C056 (12) | |
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.k UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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NORTHEAST NUCLEAR ENERGY COMPANY, et. al.
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Docket No. 50-336 OLA (Millstone Nuclear Power
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(Spent Fuel Pool Design)
Station, Unit 2)
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i SUPPORTING AFFIDAVIT OF DR. LAURENCE I. KOPP I, Laurence I. Kopp, being duly sworn, state:
- 1. I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Division of Systems Safety and Analysis in the Office of Nuclear Reactor Regulation, U.S. Nuclear
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Regulatory Commission. I received my Ph.D. in Nuclear Engineering from the University of r
Maryland in 1968.
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My duties include performing safety evaluations of technical specifications, core reloads, and spent fuel storage facilities, and reviewing criticality analyses of fresh and spent fuel storage racks.
I have performed reactor physics analyses relating to reactor safety, developed guides and standards applicable to the design, location, construction, and operation of power reactors, and have assisted in the development of standard procedures, methods, and j
models for safety analyses and evaluations of power reactor designs.
- 3. I have worked in the field of reactor physics since 1959. I was a senior engineer with Martin-Marietta Corporation from 1959 to 1963, and from 1963 to 1965 I was a senior scientist at the Westinghouse Astronuclear Laboratory.
I have been with the Nuclear Regulatory Commission (formerly the Atomic Energy Commission) since 1965.
9306090228 930528'
{DR ADOCK 05000336 PDR
4 I was the principal contributor for the Staff's June 4,1992 Safety Evaluation regarding Amendment No.158 to the Millstone Unit 2 facility operating license.
- 5. I am familiar with the contents of the following documents which form, in part, the t
bases of the statements I make herein:
(a) NRC Information Notice 87-43, Gaps In Neutron-Absorbing MaterialIn High-Density Spent Fuel Storage Racks, dated September 8,1987 (copy attached hereto as Exhibit A);
(b) Electric Power Research Institute ("EPRI") Report NP-6159, An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks, dated December 1988 (copy attached hereto as Exhibit B);
(c) Northeast Nuclear Energy Company ("NNECO") letter to NRC dated October 1, 1990, Millstone Unit No. 2 Spent Fuel Racks Boraflex Degradation (copy attached hereto as Exhibit C);
(d) NRC letter to NNECO dated November 15,1990, requesting additionalinformation from NNECO regarding the Boraflex degradation in the Millstone 2 SFP (copy attached hereto as Exhibit D);
(e) NNECO's response letter to NRC, dated January 4,1991 (copy attached hereto as Exhibit E);
l (f) NRC evaluation of the Boraflex degradation, dated February 7,1991 (copy attached j
hereto as Exhibit F);
(g) NNECO's letter to the.NRC dated November 21,1991, Millstone Nuclear Power -
Station, Unit No. 2, Boraflex Degradation in Spent Fuel Racks (copy attached hereto as Exhibit G);
(h) Combustion Engineering ("CE") letters to NRC, describing CE's evaluation of its -
criticality calculation errors, dated February 28,1992 and March 27,1992 (copies of these CE letters are attached hereto as Exhibits H and I);
(i) NRC Information Notice 92-21, Spent Fuel Fool Reactivity Calculations, dated March -
24,1992, and supplemental notice dated April 22,1992 (copies of these information notices are attached hereto as Exhibits J and K);
(j) Holtec Report HI-91737, Blackness Testing ofBoraflex in Selected Region-1 Cells -
ofthe Millstone-2 Spent Fuel Storage Racks) (January 1992) (copy attached hereto as Exhibit L);
(k) Holtec Report HI-92777, FuelRack Analyses For Millstone Unit 2 (With Gaps in The Bora/ lex) (April 1992) (copy attached hereto as Exhibit M);
(1) EPRI Report-TR-101986, Boraflex Test Results and Evaluation, dated February 1993 (copy attached hereto as Exhibit N);
(m) Licensee Event Report 92-003, dated March 13, 1992, submitted by NNECO; (n) NNECO's April 16,1992 amendment application and its supporting safety analyses; and (o) Declarations of Dr. Michio Kaku dated August 23,1992, and March 31,1993, filed in the above-captioned proceeding.
- 6. I have prepared this affidavit to address issues relating to CCMN's Contention 1, and to show that, for purposes of summary disposition, there are no material facts in dispute regarding these issues.
Issue: Boraflex Box Degradation In The Millstone 2 SFP
- 7. In Exhibit C (NNECO's letter to NRC dated October 1,1990, Millstone Unit No.
2 Spent Fuel Racks Boraflex Degradation), NNECO reported the results ofits neutron blackness testing of storage cells, which began on August 24, 1990. Blackness testing is the standard means for assessing the degree of degradation of a Boraflex panel in a SFP, and consists of using a neutron source and neutron detectors to measure the degree of neutron attenuation-through Boraflex panels. Increases in neutron count rates indicate missing or significantly degraded Boraflex material while low count rates confirm that Boraflex material is present.
3 Each storage cell or box within regions A and B of the Millstone 2 SFP contains four Boraflex panels.
- 8. Exhibit C states that from the 384 s:orage cells in Region 1 (now Regions A and B)-
of the SFP,420 Boraflex panels were tested and 45 panels were found to have one gap' and three panels had two gaps in their Boraflex material.
The largest single gap was 1.8 inches wide, and it was further reported that of the three panels having two gaps, the largest combined gap measurements equalled 1.9 inches.
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- 9. In Exhibit D (Staff's letter to NNECO dated November 15,1990), the Stafirequested additional information from NNECO regarding the Boraflex degradation in the Millstone 2 SFP.
i In its response dated January 4,1991,2 NNECO reported that the gaps found in the Boraflex The number of panels having one gap was subsequently corrected to 46. See 2
Exhibit L (Holtec Report HI-91737, Blackness Testing of Boraflex In Selected Region-1 Cells of the Millstone-2 Spent Fuel Storage Racks), at 3 n.1.
2 A copy of this letter is attached hereto as Exhibit E.
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panels were randomly distributed. See Exhibit E, at 2.
The random gap distribution is illustrated in Appendix A, Fig. 8, of Exhibit M (Holtec Report HI-92777, Fuel Rack Analyses For Millstone Unit 2 (With Gaps in The Boraflex).
- 10. On February 7,1991, the Staff sent to NNECO an evaluation of the Boraflex degradation,' concluding that the.95 keff criterion would not be violated, but requesting results from any future blackness tests should those tests reveal increased gap sizes.
I1. On October 15, 1991, NNECO commenced the second series of blackness tests at Millstone 2.
The results were reported in Exhibit G (NNECO's letter to the NRC dated November 21, 1991, Millstone Nuclear Power Station, Unit No. 2, Boraflex Degradation in Spent Fuel Racks). NNECO reported that the Boraflex testing consisted of (1) all of the panels previously identified to,have gap indications, (2) a large subset of panels where no gaps were encountered previously, and (3) testing of previously untested storage cells for expansion of the data base. The test results indicated that gap growth had occurred in the locations previously -
identified to have gaps. Additionally, new gaps were detected in panels where no gaps had been previously identified and new gaps were detected in the expanded cell population which had not been previously tested. One panel was found to have exceeded the 2.7 inch gap assumption utilized in the previous Combustion Engineering criticality analysis. See Exhibit F Safety Evaluation, at 1. That gap was initially determined to be between 3.0 and 3.5 inches. A panel from a second cell was found to be borderline in exceeding the 2.7 inch gap criterion. On A copy of this evaluation and its cover letter is attached hereto as Exhibit F.
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October 21,1991, NNECO commenced the activity of replacing the two Boraflex panels that exceeded the 2.7 inch criterion.
- 12. Attachment 2 to NNECO's April 16, 1992, amendment request, Spent Fuel Pool Criticality Safety Analyses, confirmed that approximately half of the Boraflex cells in Region 1 (now Regions A and B) of the SFP had been tested for the presence of gaps.4 The test data identified a Boraflex panel defect rate of 16 percent. With the exception of the two panels cited above that were replaced, the testing did not identify any gaps exceeding 2.7 inches, which corresponds to a 2% shrinkage rate. To account for any further gap' growth, NNECO's r
criticality analysis for Amendment No.158 assumed 5.65 inch gaps (slightly more than'4%
shrinkage) at the observed blackness test locations and 5.65 inch gaps randomly distributed in all of the other Boraflex panels. See Attachment 2 to NNECO's April 16,1992 amendment request, at 1.
NNECO's analysis also conservatively assumed a 4% shrinkage in width, although there is no observed evidence of such shrinkage. See Exhibit M (Holtec Report HI-92777, Fuel Rack Analyses For Millstone Unit 2 (With Gaps in The Boraflex), at 2.
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- 13. In evaluating NNECO's license amendment request again in light of Dr. Kaku's subsequent declarations, in addition to relying on plant-specific information summarized above in paragraphs 7-12, the following general information concerning Boraflex degradation is relevant:
(a) NRC Information Notice 87-43, Gaps in Neutron-Absorbing Materialin High-Density
- Exhibit L (Holtec Report HI-91737, Blackness Testing ofBoraflex in Selected Region-1 Cells of the Millstone-2 Spent Fuel Storage Racks), at 3 and Fig. 3, reflects 176 out of 384, or 46%, of the cells were tested for the presence of gaps.
Spent Fuel Storage Racks, dated September 8,1987 (copy attached hereto as Exhibit A);
(b) EPRI Report NP-6159, An Assessment ofBoraflex Performance in Spent-Nuclear-Fuel Storage Racks, dated December 1988 (copy attached hereto as Exhibit B); and (c) EPRI Report TR-101986, Boraflex Test Results and Evaluation, dated February 1993 (copy attached hereto as Exhibit N).
These reports reflect what the BISCO and EPRI-sponsored studies established regarding maximum Boraflex shrinkage, the levels of gamma exposure which produces maximum shrinkage, and the effect irradiation has on Boraflex ability to absorb neutrons. Data from the BISCO research shows that a maximum 3-4% shrinkage of Boraflex occurs at a gamma radiation saturation point of approximately 1 x 10' rads. See Exhibit B, Figure 5-2 and pp. 5-9 to 5-12.
The same maximum shrinkage of 3-4% was found in an EPRI-sponsored study of shrinkage as a function of the length of time Boraflex remains in a SFP. See id. at 3-16 to 3-21, and Figure 3-2. The 1993 EPRI report confirms that "the maximum total shrinkage of Boraflex is limited to 3 to 4 %." Exhibit N, at 6-1.
- 14. NNECO's assumptions of (a) 5.65 inch gaps at the observed blackness test locations, which corresponds to slightly more than 4% shrinkage of Boraflex and (b) 5.65 inch gaps randomly distributed in all of the other Boraflex panels, are properly conservative.
The assumption of a slightly more than 4% shrinkage rate is conservative as shown by the EPRI reports discussed above, and by the fact that, except for the two replaced Boraflex panels mentioned previously, such gaps are over twice as large as any single gap measured to date in the Millstone 2 SFP. The random distribution assumption is supported by NNECO's reported blackness testing results for the Millstone 2 SFP. See Appendix A, Fig. 8, of Exhibit M.
I Moreover, as the 1993 EPRI report discusses, Borafiex shrinkage and gap formation result in no loss of boron from the system, but merely a redistribution of boron to a slightly less optimal configuration. See Exhibit N, at 6-9. Criticality calculations, on the other hand, assume a complete loss of boron, adding an additional conservatism to the results.
- 15. Based upon the foregoing, I conclude that NNECO's assumptions submitted in 1
support of Amendment No.158 are conservative, and that NNECO's safety analyses for Amendment No.158 are valid. In reaching this determination I have fully considered Dr.
Kaku's declarations.
Issue: Gap Concentration Increasing km In The Millstone 2 SFP
- 16. Criticality is a measure of the capability of the neutron field to sustain a nuclear chain reaction. This measurement is expressed by indicating the effective multiplication factor for neutrons (keff), which is the ratio of the number of neutrons produced from fissions in each generation to the number of neutrons lost by absorption and leakage. The keff must equal 1.00 for there to be criticality. The keff is determined by mathematical calculation, and the precise keff level cannot be measured with exactitude, as discussed in the attached affidavit of Mr.
Bidinger. The.95 keff criterion, set forth in f 9.1.2 of the Comrnission's Standard Review Plan, provides adequate assurance that a safety margin will be maintained to preclude a criticality event in a SFP. The.95 keff criterion includes a number of built-in conservatisms such as: (1) the calculations are based on storage racks containing fuel of the maximum allowed j
enrichment; (2) calculational uncertainties and mechanical tolerances are included in their most adverse (highest reactivity) direction; and (3) no credit is taken for dissolved boron in the SFP water.
- 17. NNECO's analysis for Amendment No.158 demonstrated that even with an assumed concentration of 5.65 inch gaps in the central 50% of the rack height, a; opposed to the random distribution of gaps actually observed, the maximum calculated k,y was 0.9212, well within the 0.95 criterion. See Attachment 2 to NNECO's April 16,1992 license amendment request, at 3, and Table 4 of NNECO's April 16, 1992 license amendment request.
Dr. Kaku's declarations assert no facts casting doubt on the validity of NNECO's calculated k n of 0.9212 e
for the gap distribution assumption described above.
Moreover, the Oak Ridge National Laboratory ("ORNL") January 1993 report, Criticali:y Safety Calculations For Region B of the Millstone Unit No. 2 Spent Fuel Pool (attached as Exhibit A to Mr. Bidinger's affidavit), at 3 and Table 3, found that even an assumed concentration of 5.65 inch gaps at the axial midline of each Boraflex panel,phich results in the maximum reactivity of the racks, did not cause the k,y to exceed the 0.95 criterion.
Issue: Benchmark Data Supporting The Criticality Analysis Used For Amendment No.158
- 18. The 21 Babcock & Wilcox (B&W) critical experiments reported by Baldwin, et al.,
in Critical Erperiments Supporting Close Proximity Water Storage ofPower Reactor Fuel (BAW-1484-7, July 1979),5 appear to be the most representative of the Millstone 2 Region A and B spent fuel storage racks. These experiments used 0.25-inch thick Boral plate absorbers between fuel assemblies and the pertinent characteristics are summarized in the following table:
5 This benchmarking study is among those listed in Appendix C of Exhibit M. A copy of this Holtec Report HI-92777 was mailed to Dr. Kaku by NNECO counsel's letter i
dated January 26,1993.
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5,
B&W Millstone 2 1
U-235 Enrichment (wt %)
2.46 up to 4.5 Fuel rod O.D. (in.)
0.480
.44 Fuel rod pitch (in.)
0.646 0.58 Fuel rod array 14x14 14x14 Assembly spacing (in.)
1.8 0 - 1.93 Absorber thickness (in.)
0.25(Boral) 0.168(Boraflex+SS)
B-10 areal density (gm/sq cm)
.0003.005
.03 Fuel clad material Al
- Zr The use of aluminum clad instead of Zirconium is not expected to have a significant effect on -
the calculational verification. The two parameters that appear to be significantly different are the U-235 enrichment and the boron (B-10) density in the absorber panels.
NNECO demonstrates that with respect to k,y there was no significant effect of (1) fuel enrichment over -
the enrichment ranges involved in power reactor fuel, and (2) boron (B-10) density in the absorber panels. See Appendix C of Exhibit M, at A-2, A-4. Two independent methods of analysis (KENO-V.a Monte Carlo and CASMU-3 transport theory) were used in the benchmarking and showed very good agreement. See A-2, A-4, and Table 3 of Exhibit M. The 4
intercomparison between different analytical methods is an acceptable technique for validating calculational methods for nuclear criticality safety.
- 19. I, therefore, conclude that the critical experiments used in Amendment No.158 for benchmarking are sufficiently representative of the actual' Millstone 2 spent fuel racks.
Additionally, as discussed further in Mr. Bidinger's affidavit, the neutronic calculations
questioned by Dr. Kaku have been used over many years of reactor operation'and have produced J
reliable simulations of the neutron distribution, even in highly absorbing media, as evidenced by the fact that predictions of physics parameters such as control rod worths and critical boron concentrations agree well with reactor measurements.
These calculations add additional confidence to the ability to predict keffin SFPs.
- 20. On this basis, I conclude that NNECO's use of the KENO codes resulted in an acceptably conservative criticality analysis for Amendment No.158, and that the benchmarking studies account for the presence of thin neutron absorbers such as the Boraflex panels in the Millstone 2 SFP In reaching this determination I have fully considered Dr. Kaku's declarations.
Issue: Use of Vertical Buckling Term
- 21. The KENO V.a reactivity calculations for Amendment No.158 were performed in three dimensions (x, y, z) and, therefore, axial (vertical) leakage was inherently accounted for, and use of a vertical buckling term was not required. In addition, multi-group (27-group) neutron cross-sections for each material were explicitly input to the reactivity calculations. By contrast, the previous CE calculations utilized a few-group (4-group) cross-section set, which i
necessitated the input of a buckling term to compute the energy-dependent spectrum used to collapse multi-group cross-section data to the 4-group set. The CE error involved use of an i
improper buckling in this cross-section collapsing calculation. See Exhibits J and K.
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- 22. The foregoing is true and correct to the best of my knowledge, information, and belief.
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Subscribed and sworn to before
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me this 28th day of May 1993 a
/- A FAfu Laurence 1. Kopp I/ /
hab
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NotaryfPublic
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My commission expires 3fl f 95
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