ML17158C150
| ML17158C150 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/07/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17158C149 | List: |
| References | |
| NUDOCS 9705140441 | |
| Download: ML17158C150 (25) | |
Text
WASHINGTON, D.C. 2055&4001 RELATED TO AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. NPF-22 ENNSYLVANIA POWER
& LIGHT COMPANY LLEGHENY ELECTRIC COOPERATIVE INC.
SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 DOCKET NO. 50-388
~P,S AEQUI UNITED STATES NUCLEAR REGULATORY COMMISSION O
Cy
+~
~O
+**++
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
1.0 INTRODUCTION
By letter dated December 18, 1996 (Reference 1, PLA-4527);
as supplemented by letters dated February 26, 1997 (Reference 2, PLA-4572),
March 12, (Reference 3,
PLA 4582),
March 27, 1997 (Reference 4,
PLA-4595), April 3, 1997 (Reference 5,
PLA-4599), April 9, 1997 (Reference 6, PLA-4605), April 16, 1997 (Reference 7,
PLA-4611),
and April 18, 1997 (Reference 8, PLA-4613),
and April 24, 1997 (Reference 9, PLA-4620), Pennsylvania Power 8 Light Company (PPSL, the licensee) proposed changes to the Technical Specifications (TSs) for the Susquehanna Steam Electric Station, Unit 2 Cycle 9
(S2C9) which is the first 24-month operating cycle.
The requested changes would authorize the use of ATRIUM-10 fuel in Unit 2 beginning with cycle 9 under all operational Conditions (1-5) as defined in the TSs.
The proposed changes include the Safety Limits Minimum Critical Power Ratio (SLMCPR) based on the cycle-specific analysis of the mixed core of Siemens Power Corporation (SPC)
ATRIUM-10 and SPC 9x9-2 fuel parameters and other sections of the TSs relating to the use of ATRIUM-10 fuel.
Due to the limitations imposed in the approved Advanced Nuclear Fuel-B (ANFB) Critical Power Correlation (ANF-1125 (P)
(A) and its Supplements 1 and 2) and the findings in the inspection of the Application of ANFB to ATRIUM-10 for Susquehanna Reload at Siemens Power is Corporation (SPC) in February 1997, this review based on the updated information provided in References 2 through 9, and its findings relative to the minimum critical power ratio (MCPR) limits and the use of two new methodologies are applicable only to the ninth Susquehanna Unit 2 reload (S2C9).
During the staff's review of the TS changes discussed in this safety evaluation, the licensee made two exigent amendment requests.
The first requested TS change was made to permit the loading of the Atrium-10 fuel into the core and maintaining the reactor in Condition 5, refueling.
This amendment
(¹136) was approved on April 9, 1997.
The second requested, TS change was made to permit the reactor to be brought into Condition 4 (cold shutdown) and Condition 3 (hot shutdown) to permit certain testing to be conducted.
This amendment
(¹138) was approved on April 25, 1997.
Both of these TS changes authorized by the amendments noted above are being modified by the current TS revisions to enable the fuel to be used under all operational conditions and to include applicable references and safety limits.
7 ~
DOCK 05000388 140445. 970507 PDR p
Brookhaven National Laboratory (BNL) assisted the NRC staff in the review of EHF-97-010 (P), Revision 1 (provided to the Commission as an attachment to Reference 4) and prepared a technical evaluation report (TER) which is attached to this safety evaluation (SE) to support the review for the SLHCPR TS changes.
- 2. 0 EVALUATION 2.1 Mechanical Design The ATRIUM-10 fuel design is a 10x10 lattice design which contains 83 full length fuel rods, 8 part length fuel rods, and a central water channel to enhance neutron moderation.
The ATRIUM-10 fuel design was analyzed and assessed by Siemens according to the approved methodology, entitled "Generic Mechanical Design Criteria for BWR Fuel Designs," ANF-89-98(P)(A) Revision 1
and Revision 1 Supplement 1.
The staff has performed an on-site audit of ATRIUH-10 fuel at Siemens.
Although the staff discovered a procedural deficiency, we conclude that, with the correction of the deficiency, the ATRIUM-10 fuel mechanical design followed the approved methodology, and therefore, is acceptable for Susquehanna 2 Cycle 9.
2.2 Application of the ANFB Critical Power Correlation to ATRIUM-10 Fuel The review of the Siemens reload analysis for Cycle-9 of Susquehanna-2 and the application of the ANFB correlation to* the ATRIUM-10 fuel design was included in the NRC Vendor Inspection (No. 99900081/97-01) at the Siemens Power Corpor ation Facility in Richland, WA, during the week of February 9-14, 1997.
Several important concerns were identified during this review of the SSE-2 reload analysis and the application of.ANFB to the ATRIUM-10 fuel design.
First, it was noted that the local fuel rod power peaking for SSE-2 Cycle-9 fuel bundles exceeded the range of the ANFB correlation as stated in Reference 4 (local peaking
< 1.3).
'In addition, a flow dependent bias in the ANFB correlation was identified which resulted in the nonconservative overprediction of the measured critical power at low flows.
Both of these effects were outside the presently approved applicable SPC methodologies.
In response to findings of the NRC vendor inspection at the SPC during the week of February 9-14,
- 1997, PP&L has submitted a revised ANFB methodology and core flow dependent HCPR safety limits and a supporting topical report, EHF-97-010(P),
Revision 1, "Application of ANFB, to ATRIUM-10 for Susquehanna Reloads, " March 1997 (Reference
- 3) for S2C9 reload.
The detailed review was given in the attached TER (Attachment),
which was provided.by our consultant at BNL.
The staff adopts the findings and position included in this report.
Accordingly, the staff concludes that the revised methodology and newly revised core flow dependent HCPR safety limits, as proposed in the TS change, are acceptable for the SSES Unit 2 S2C9 reload.
The revisions to the TS address the staff's concerns about local fuel rod power peaking and fuel behavior at low reactor coolant recirculation flow (flow bias).
0 2.3 Technical Specification Changes The licensee requested a change to the S2C9 TSs in accordance with 10 CFR 50.90.
The proposed revisions of the TS and its associated Bases - are described below.
(1)
TS 1.2 Average Exposure and 1.3 Average Planar Linear Heat Generation Rate The proposed changes of definitions for TS 1.2 average bundle exposure and average planar exposure and for TS 1.3 average planar linear heat generation rate to reflect the use of part length rods in the ATRIUM-10 fuel assemblies as well as full length 'rods in SPC 9x9-2 fuel assemblies are acceptable since the new wordings clearly define the meaning for the new fuel assemblies used.
This change is not restricted to S2C9.
(2)
TS 2. 1.2 and 3.4. 1. 1.2 and Bases
- 2. 1. 1 and 2. 1.2 The safety limit MCPR in TS 2. 1 and its associated Bases 2.0 is proposed to change from 1.08 to the value shown in Figure 2.1.2-1
(> l. 11 depending on the core flow) for operation with two recirculation loops with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10 million ibm/hr.,
and from 1.09 to the value shown in Figure 3.4. 1. 1.2-1
(>1.22) for single loop operation (SLO) based on the cycle-specific analysis of a modified ANFB core flow dependent SLMCPR performed by SPC for S2C9 mixed core of ATRIUM-10/SPC 9x9-2 fuel (Reference 8).
The staff in conjunction with our consultant at BNL has reviewed the proposed TS and its associated Bases changes and has found them acceptable since they are based on the analyses performed using S2C9 cycle-specific inputs and approved-methodologies in Reference 3.
The details of our evaluation are provided in the attachment to this safety evaluation.
The staff noted that in the submittal dated April 3,. 1997, the licensee added a footnote for TS Section 3.4. 1. 1.2 designated with the "¹" symbol.
In recent discussions, the licensee discovered that'his footnote symbol had already been used in that TS section.
Accordingly, the licensee changed the applicable footnote symbol to "++" which is an administrative change and found to be acceptable by the staff.
(3)
TS 5.3.1 - Fuel Assemblies Section 5.3. 1 was revised to reflect the use of ATRIUM-10 fuel with a central water channel, part length fuel rods and different active fuel length from that of SPC 9x9-2.
The maximum allowed enrichment was increased from 4.0 to 4.5 weight percent U-235 which is consistent with 10 CFR 51.52.
The revised Section will be read as follows.
"...or slightly enriched uranium dioxide as fuel material and water rods or water channels....
Reload fuel shall have a maximum lattice average-enrichment of 4.5 weight percent U-235."
I
The ATRIUM-10 fuel design increases the maximum enrichment from 4.0 to 4.5 weight percent U-235 and allows a 24-month operating cycle.
The enrichment
'hange was approved, by Amendment No. 136, April 9,
- 1997, however, dose consequences of this change were not considered at that time because the plant was not permitted to startup or become critical.
The radiological consequences of design basis accidents will not be affected by the enrichment change after the fuel is used under all operational conditions, except as discussed below.
The licensee in its April 24, 1997 submittal indicated that the maximum discharge exposure for the ATRIUM-10 fuel is 48 HWd/kgU (HWD/MTU) as documented in the SPC report EMF-95-52(P),
"Mechanical Design Evaluation for Siemens Power Corporation ATRIUM-10 BWR Reload Fuel, dated July, 1995.
This burnup rate is greater than the 45 HWD/HTU value evaluated in the.
Commission Safety Evaluation dated September 12, 1995.
In addition to the information provided in the submittals by the licensee, the staff has reviewed a publication which was prepared for the NRC entitled, "Assessment of the Use of Extended Burnup Fuel in Light Water Reactors,"
NUREG/CR 5009, February 1988.
The NRC contractor, Pacific. Northwest Laboratory (PNL) of Batelle Memorial Institute, examined the changes that could result in the NRC design basis accident (DBA) assumptions, described in the various appropriate Standard Review Plan (SRP) sections and/or Regulatory Guides (RG), that could result from the use of extended burnup fuel (up to 60 HWD/MTU).
The staff finds that the only DBA that could be affected by the use of extended burnup fuel, even in a minor way, would be the potential thyroid doses that could result from a fuel handling accident (with fuel that had been subject to the maximum burnup).
PNL estimated that I-131 fuel gap activity in the peak fuel rod with 60 MWD/HTU burnup could be as high as 12'.
This value is approximately 20X higher than the value normally used by the staff in evaluating the fuel handling accidents (as per RG 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facilities for Boiling and Pressurized Water Reactors).
For the fuel handling accident, PNL concluded that the use of RG 1.25 procedures for the calculation of accident doses for extended burnup fuel may be utilized.
These procedures give conservative estimates for noble gas release fractions that are above calculated values for peak rod burnups of 60 MWD/HTU.
Iodine-131 inventory,
- however, may be up to 20X higher than that predicted by RG 1.25 procedures.
In its evaluation for the Susquehanna units issued in April 1981 (NUREG-0776),
the staff conservatively estimated offsite doses due to radionuclides released to the atmosphere from a fuel handling accident.
The staff concluded that the plant mitigative features would reduce the doses for this DBA to below the doses specified in the SRP Section 15.7.4.
In the Safety Evaluation dated September 12, 1995, the staff reanalyzed the fuel handling accident based on a
maximum fuel burnup of 45 MWD/HTU using the information from PNL discussed above.
Table 1 below was included in that Safety Evaluation..
The evaluation presented in the Table continues to bound the licensee's current proposal to use fuel with 4.5 weight percent U-235 with a max'imum burnup of 48 HWD/HTU.-
Table 1
'Radiological Consequences of Fuel Handling Design Basis Accident (rem)
~Th roid Exclusion Area
~P Zone Staff Evaluation April 1981 (NUREG-0776)
Bounding Estimates For Extended Burnup Fuel 5N Enrichment 12 14.4
<1.2 Dose Acceptance Criterion (NUREG-0800 Section 15.7.4) 75 75 The staff therefore concludes that the only potential increased doses resulting'rom the fuel handling accidents with extended burnup fuel with increased U-235 enrichment are the thyroid doses; these doses remain well within the dose limits given in NUREG-0800 and are therefore acceptable.
Based on the staff evaluation, we conclude that this revision is consistent with the staff position and thus acceptable for Susquehanna 2.
This approval is not restricted to S2C9.
(4)
TS 6.9.3.2 Core Operating Limits Report E
The proposed change is to add additional approved methodologies relating to the use of SPC ATRIUH-10 fuel assemblies.
The proposed approved methodologies are the following:
(a)
ANF-89-98(P)(A) Revision 1 and Revision 1 Supplement 1, "Generic Hechanical Design Criteria for BWR Fuel Design," Advanced Nuclear Fuel Corporation, Hay 1995.
(b)
XN-NF-81-58(P)(A) Supplements 1 and 2, "RODEX 2 Fuel Rod Thermal-Hechanical
Response
Evaluation Hodel," Hay 1986.
(c)
(d)
(e)
(g)
(h)
(m)
XN-NF-85-74(P)(A),
"RODEX 2A (BWR) Fuel Rod Thermal-Mechanical
Response
Evaluation Model," August 1986.
XN-NF-82-06(P)(A) and Supplements 2, 4, and 5 Revision 1,
"gualification of Exxon Nuclear Fuel for Extended Burnup," October 1986.
XN-NF-85-92(P)(A),
"EXXON Nuclear Uranium Dioxide/Gadolin'ia Irradiation Examination and Thermal Conductivity," November 1986.
ANF-90-082(P)(A) Revision 1 and Revision 1 Supplement 1,
. "Application of ANF Design Methodology for Fuel Assembly Reconstitution,"
May 1995.
ANF-91-048(P)(A),
"Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.
ANF-CC-33(P)(A) Supplement 2,
"HUXY : A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," January 1991.
ANF-CC-33(P)(A) Revision 1,
"HUXY :
A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option Users Manual," November 1975.
XN-NF-80-19(P)(A), Volumes 2, 2A, 2B, and 2C "Exxon Nuclear Methodology for Boiling Water Reactors; EXEM BWR ECCS Evaluation Model," September 1982.
XN-NF-80-19(P)(A), Volume 3 Revision 2 "Exxon Nuclear Methodology for Boiling Water Reactors Thermex:
Thermal Limits Methodology Summary Description," January 1987.
XN-NF-79-71(P)(A) Revision 2, Supplements 1,
2, and 3, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"
March 1986..
ANF-1358(P)(A) Volume 1, "The Loss of Feedwater Heating Transient in Boiling Water Reactors,"
September 1992.
(n),
ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4,
"COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," August 1990.
(o)
(p)
XN-NF-84-105(P)(A), Volume 1 and Volume 1 Supplement 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987.
XN-NF-84-105(P)(A), Volume 1 Supplement 4,
"XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, Void Fraction Model Comparison to Experimental Data," June-1988.
(q)* EMF-97-010(P),
Revision 1, "Application.of ANFB to ATRIUM-10 for Susquehanna Reloads,"
March 1997.
(r)* PLA-4595, "Response to NRC Request for Additional Information on Siemens'eport EMF-97-010, Revision 1, March 27, 1997.
"*" Only Applicable for S2C9 Operation.
The staff has concluded that the generic methodologies (a)-(p) are applicable to this plant-specific ATRIUM-10 fuel design.
Based on our review, we also conclude that methodologies (q) and (r) are acceptable for only S2C9 application since the proposed fuel design has been analyzed on a plant and cycle-specific basis using the NRC approved methodologies.
This application of the methodologies resolves a previous staff concern about the ANF-B correlation discussed in Amendment No.
136, dated April 9,
- 1997, and enables the plant to proceed to Conditions 2 (startup) and 1 (operation).
- Finally, the staff concludes that the licensee may use this fuel under all operational conditions for S2C9 since all applicable limits, (e.g., fuel thermal-mechanical limits, core thermal-hydraulic.limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis limits have been met.
3.0 PUBLIC COMMENTS No public comments were received from the notice of this amendment dated
'March 12, 1997 and pub1ished in the Federa1 Re<eister on March 18, 1997 (62 FR 12859).
- However, some comments were received from the public in response to a
notice, published in two local newspapers, the Berwick Press Enterprise,
- Berwick, PA, and the Wilkes Barre Times Leader, Wilkes Barre, PA, April 22-24, 1997 concerning the exigent amendment that would enable the plant to move to Conditions 4 and 3 using the Atrium-10 fuel.
The staff considered the following comments applicable to this amendment.
One comment from an individual was a request that the NRC completely review
.the new fuel design to ensure that it is as safe as the current fuel.
As discussed
- above, the NRC technical staff with assistance from Brookhaven National Laboratory has conducted an audit at Siemens Power Corporation and has completed a comprehensive review of this new fuel design and analyses which support the safe operation of the reactor.
Issues raised by the staff have resulted in the licensee proposing as noted in the revised TS to operate the fuel with conservative safety limits that provide an additional level of safety in the manner in which the fuel will be utilized to produce heat in the reactor core especially when the reactor coolant flow is at low levels.
The staff has no reason to believe that the new fuel will not be as safe as the fuel used in the reactor up to this point in time.
Another individual voiced opposition to the use of the new fuel unless there was assurance that there would be no increase in risk to the public given an accident.
Another comment was a'oncern that the use of the new fuel could potentially result in more radiation being released after an accident than
compared to that which could be released by an accident with the current reactor fuel loaded in the core.
The staff considered the fact that the new fuel reflected an enrichment increase from 4.0X to 4.5X uranium-235 and as discussed in the safety evaluation considered the maximum burnup rate provided by the licensee and the limiting accident that could produce the maximum dose to the public.
Given these facts, the staff still determined that the consequences would still be well below 10 CFR Part 100 release limits. 'his TS change and use of the ATRIUM-10 fuel was found not to result in a significant increase in the consequences of an accident previously analyzed for the plant and therefore is acceptable to the staff.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment.
The State official had no comments.,
- 5. 0 ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Receister on Ray 6, 1997 (62 FR 24669).
Accordingly, based upon the environmental assessment, the staff has determined that the issuance of this amendment will not have significant effect on the quality of the human environment.
- 6. 0 CONCLUSION The Commission has concluded, based on the conside}ations discussed
- above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Attachment:
Technical Evaluation Report No.
PLA-4527 by BNL dated March 27, 1997 Principal Contributor:
T. Huang Date:
May 7, 1997
7.0 REFERENCES
PLA-4527, Susquehanna Steam Electric Station Proposed Amendment No.
166 to License NPF-22: Unit 2 Technical Specification Changes for ATRIUM-10
- Fuel, December 18, 1996.
2.
3.
5.
6.
7.
9.
PLA-4572, Susquehanna Steam Electric Station Correc'tion to Proposed Amendment No.
166 to License NPF-22: Unit 2 Technical Specification Changes for ATRIUM-10 Fuel, February 26, 1997.
PLA-4582, Susquehanna Steam Electric Station Addendum to proposed Amendment No.
166 to License NPF-22:
Revised ANFB Methodology and Core Flow Dependent HCPR Safety Limits, March 12, 1997.
PLA-4595, Susquehanna Steam Electric Station
Response
to NRC Request for Additional Information on,SIEMENS'eport EHF-97-010, Rev.
1, March 27, 1997.
PLA-4599, Susquehanna Steam Electric Station Addendum ¹2 to Proposed Amendment No.
166 to.License NPF-22: Addition of Limiting Footnotes and a
Reference Reflecting PP&L's RAI Response, April 3, 1997.
PLA-4605, Susquehanna Steam Electric Station
Response
to NRC Request for Additional Information on PP&L's Proposed Amendment No.
166 to License No. NPF-22: Unit 2 Technical Specification Changes for ATRIUM-10 Fuel, April 9, 1997.
PLA-4611, Susquehanna Steam Electric Station Addendum to PP&L's Response to NRC Request for Additional Information on PP&L's Proposed Amendment No.
166 to License No. NPF-22: Unit 2 Technical Specification Changes for ATRIUM-10 Fuel, April 16, 1997.
PLA-4613, Susquehanna Steam Electric Station Addendum ¹3'to Proposed Amendment No.
166 to License NPF-22:
Revised
.Core Flow Dependent HCPR
'afety Limits, April 18, 1997.
PLA-4620, Susquehanna Steam Electric Station
Response
to NRC guestion on Proposed Amendment No.
166 to License NPF-22:
ATRIUM-10 Exposure Limit, April 24, 1997.
TECHNICAL EVALUATION REPORT Report
Title:
Susquehanna Steam Electric Station Proposed Amendment No. 166 to License NPF-22:
Unit-2 Technical Specification Changes for ATFJUM'-10Fuel Report Number:
PLA-4527 Report Date:
March 27, 1997 Docket No.:
50-388 Originating Organization:
Pennsylvania Power & Light Company
1.0 INTRODUCTION
The Pennsylvania Power & Light Company (PP&L) has submitted in Reference 1 the proposed changes to the Susquehanna Steam Electric Unit-2 (SSE-2) Technical Specifications for NRC review and approval.
These Technical Specification'hanges result primarily from the use of the new Siemens Power Corporation (SPC) ATRIUM>>'-10 fuel.
Specifically, these changes involve the application of the Siemens ANFB critical power correlation to the ATRIUM'-10fuel design in determining the Operating LimitMCPR, and are based on the Siemens Topical Report EMF-97-010 (Reference 2).
EMF-97-10 provides test data taken specifically to support the application of the ANFB correlation to the ATRIUMrM-10fuel design and the determination of the correlation additive constants.
The change in the ANFB correlation additive constants required for the ATRIUM'-10 fuel design affects both the Operating LimitMCPR (OLMCPR) and Safety LimitMCPR (SLMCPR).
The review of the Siemens reload analysis for Cycle-9 of Susquehanna-2 and the application of the ANFB correlation to the ATRIUM'-10fuel design was included in the NRC Vendor Inspection (No. 99900081/97-01) at the Siemens Power Corporation Facility in Richland, WA during the week of February 9-14, 1997.
Several important concerns were identified during this review of the SSE-2 reload analysis and the application of ANFB to the ATRIUM'-10fuel design.
First, it was noted that the local fuel rod power peaking for SSE-2 Cycle-9 fuel bundles exceeded the range of the ANFB correlation as stated in Reference 4 (local peaking < 1.3).
In addition, a flow dependent bias in the ANFB correlation was identified which resulted in the nonconservative overprediction of the measured
/
Attachment
0 t
critical power at low flows. Both of these effects are outside the presently approved SPC SLMCPR and PP8cL OLMCPR methodologies.
In order to address these concerns, the methodology used to determine the SLMCPR and the transient b,CPR has been revised for application to SSE-2 in References 3 and 4. The purpose of this review was to evaluate these methodology changes and insure that adequate margin is included in the SSE-2 Cycle-9 OLMCPR, This review does not include those aspects of the methodology which relate to the generic resolution of the identified ANFB concerns.
The methodology changes are summarized in Section 2, and the evaluation ofthe important technical issues raised during this review is presented in Section 3. The Technical Position is given in Section 4.
2.0
SUMMARY
OF THE REVISED OLMCPR METHODOLOGY The form of the ANFB correlation and the definition of the independent variables described in Reference 5 are not changed for application to the new SPC ATRIUM'-10fuel design.
The dependence of the ANFB correlation on the ATRIUM'-10bundle design is included by adjusting the values ofthe additive constants (used to determine the local peaking function) to match the critical power test data.
SPC has performed a
series of critical power tests and determined the ATRIUM'-10design-specific additive constants.
The ANFB correlation when used with these additive constants reproduces the measured critical power to within the correlation standard deviation.
The initial measurements consisted of a series of twelve critical power tests to determine the critical power characteristics of the ATRIUM'-10fuel bundle.
The tests were performed for an axially symmetric cosine power shape and included a set of limiting local power distributions and a range of pressures, flows and inlet subcoolings.
The tests were for a 10><10 rod array and included the ATRIUM'-10water channel and part-length rods.
The ANFB correlation predicted the measured critical power to within a standard deviation which was comparable to previous ANFB applications (Reference 5). The ECPRs were evaluated as a function ofpower, flow, pressure and inlet subcooling and no clear trends or bias were observed.
In order to insure that the ANFB correlation is applicable to asymmetric axial power shapes, SPC performed additional tests for the ATRIUM>>'-10 fuel bundle including both upskewed and
- downskewed power shapes.
A selected set of cosine tests were repeated for the upskewed and downskewed power shapes to allow comparison and identify any dependence on power shape.
These tests indicated a significant dependence ofthe critical power data on axial power shape which required a revision of the additive constants (determined based on the cosine tests) to insure a conservative critical power calculation.
Using these revised additive constants, the ANFB correlation results in a mean critical power underprediction (i.e., ECPR < 1).
3.0
SUMMARY
OF THE TECHNICALEVALUATION The SPC Topical Report EMF-97-010 (P) provides the basis for application of the ANFB critical power correlation to the ATRIUM'-10fuel design for Susquehanna reloads.
The report includes the results of the ATRIUM'-10critical power tests, derivation of the additive constants and the determination of the correlation bias and standard deviation relative to the measurement data.
The review of the SPC methodology focused on the accuracy of ANFB in reproducing the critical power test data and its applicability to the Susquehanna ATRIUM'-10reload fuel. The review included several discussions with SPC during the NRC Vendor Inspection (No. 99900081/97-01) at Siemens Power, and with PP&L and SPC during a meeting on March 26, 1997 at the NRC Headquarters in Bethesda, MD. As a result of these discussions and our review of the methodology several important technical issues were raised which required additional information and clarification from SPC and PP&L. This information was requested in Reference 6 and was provided in the PP&L responses included in References 7-10.
This evaluation is based on the material presented in the topical repott (Reference 4) and in References 7-10.
The evaluation of the major issues raised during this review are summarized in the following.
3.1 Application of the ANFB Correlation to Susquehanna ATRIUM"'-10 Reload Fuel 3.1.1 Ran e of Local Power Peakin The Reference 5 ANFB comlation is applicable to fuel rod arrays for which the local power peaking factor Fl ~l 1.3 ( ANFB-1125-P(A); Supplement 1, SER Condition 3.3(1)).
The Reference 1
ATRIUMr~-10critical power measurements were intended for a similar range of power peaking and were taken for local peakings up to Fl<<ai 1.3 (Table 6.1, Reference 4).
However, during the February 1997 NRC inspection of the SPC reload design activities," it was noted that the Susquehanna-2 Cycle-9 reload core includes several fuel bundles with local peaking factors greater than the ANFB maximum ofFl~ = 1.3. In response to this concern, SPC has indicated in Reference 4 that in order to account for the increased ANFS correlation uncertainty that occurs for high local peaking the additive constant uncertainty will be increased for rods having Fip~l + 1.3.
3.1.2 Flow-Bias in the ANFB Critical Power Predictions The ANFB correlation database includes measurements for the cosine, downskew and upskew axial power distributions.
During the February 1997 NRC inspection, it was noted that, while the cosine data does not include any clear trend versus power, pressure, inlet subcooling or bundle flow, the ANFB predictions of the upskew data indicate a nonconservative flow-dependent bias.
For the upskew
- tests, ANFB conservatively underpredicts the critical power at high flow rates and nonconservatively overpredicts the critical power at low flow rates.
In addition, ANFB tends to generally underpredict the downskew test data.
In response to this concern, in Reference 4 (Figure 6.2) SPC has determined the flow-dependent bias in the ANFB predictions.
This flow-dependent bias is based on calculation-to-measurement comparisons for the upskew tests and will be applied in all ANFB critical power calculations.
The calculated flow-bias does not take credit for the ANFB conservative underprediction at high flows.
3.1.3 Increased Additive ConstantUncertain at Low Flows The ANFB upskew tests were intended to determine the correlation dependence on axial power shape and did not include the full range of bundle flows. Consequently, in the very low flow range where test data was not available, an extrapolation ofthe high-flow upskew data was performed to determine the critical power.
In Reference 4, SPC provides an estimate of the increase in additive constant uncertainty introduced by this extrapolation at low flows.
In.Reference 8 (Response
- 4) and Reference 9 (Response 3), SPC has indicated that the SLMCPR analysis is insensitive to this uncertainty.
In Reference 9 (Response 3), SPC has conservatively increased this additive constant uncertainty by a factor of2 in the four lowest flow cases (where the uncertainty has an effect).
In three of these cases the SLMCPR was unaffected, and in the remaining case (at 50.0 Mlb/hr) the SLMCPR increased by 0.01.
SPC has indicated in Reference 10 that for conservatism the SSE-2 Cycle-9 Technical Specification SLMCPR (at 50.0 Mlb/hr) will include this additional 0.01 increase in the SLMCPR.
3.2 Safety LimitMCPR Calculation The safety limitMCPR provides the uncertainty allowance required to account for the uncertainties in the ANFB correlation and the POWERPLEX-II core monitoring system.
The Susquehanna-2 Cycle-9 SLMCPR was determined using the SPC approved methodology (Reference 11).
In the SPC methodology, a cycle-specific full core analysis'is performed in which both the fresh ATRIUM'-10 and previous cycle fuel bundles are modeled.
The flow is calculated based on the bundle-dependent pressure-drop and used in the ANFB critical power calculation.
The bundle-to-bundle difference in pressure-drop and flow resulting from the introduction of the ATRIUM'-10fuel bundles is accommodated by the full core model.
The initial analysis was performed for a relatively small set of reactor statepoints, however, in Reference 9 (Response
- 1) SPC has expanded the analysis to include the complete set of standard reactor statepoints.
The expanded set of calculations did not result in an increase in the SLMCPR.
The SLMCPR analysis was performed over a range of core flows from 108 Mlb/hr down to 30 Mlb/hr. The calculation of the critical power for each fuel bundle included the flow-dependent bias and was determined using the individual bundle flow.
The additive constant uncertainty used in the SLMCPR Monte Carlo calculation was increased to account for the increased uncertainty associated with (1) fuel rods having local peaking greater than 1.3 (discussed in Section 3.1.1),
and (2) bundle flows in the lower flow range (discussed in Section 3.1.3).
The maximum calculated SLMCPR was used to determine the Susquehanna-2 Cycle-9 operating limitMCPR.
3.3 Determination of b,CPR for AOOs The inclusion of the ANFB flow bias increases the AOO transient d,CPR for events which involve r
a reduction in the hot bundle flow.
Since the calculation of the AOO hCPR does not affect the transient dynamics, an ad-hoc correction is used to determine the effect of the ANFB flow bias on the transient MCPR.
Based on the calculated transient hot bundle flow reduction and the ANFB flow-dependent
- bias, a transient-dependent hCPR adjustment was determined.
The required ACPR adjustment to account for the ANFB flow bias has been determined for the limiting AOOs and included in the (Reference 3) Susquehanna-2 Cycle-9 Operating LimitMCPR.
V 3.4 POWERPLEX-II Core Monitoring System In the standard POWERPLEX-II core monitoring methodology, the critical power determination is based on the measured reactor statepoint variables together with the ANFB critical power correlation.
However, in the Susquehanna-2 Cycle-9 application, the ANFB flow-dependent bias is included in the SLMCPR rather than in the POWERPLEX-II ANFB critical power, calculation.
While the comparison ofthe core MCPR to the Operating LimitMCPR insures that the correct thermal margin is maintained, this approach results in an erroneous POWERPLEX-II CPR edit which incorrectly increases the MCPR of the ATRIUM'-10fuel bundles at reduced flows. Since the ATRIUM'-10 fuel bundles are the MCPR limiting bundles during Cycle-9 this will result in a nonconservative core MCPR edit.
In order to insure that this nonconservative MCPR edit is not misinterpreted it is strongly recommended that PP&L take corrective action, such as eliminating the edit or providing a warning to the staff using this edit.
4.0 TECHNICALPOSITION The Topical Report EMF-97-010 (P),
Revision 1, "Application of ANFB to ATRIUMY"'-10for Susquehanna, Reloads,"
and supporting documentation provided in References 7-10 have been reviewed in detail.
Based on this review, it is concluded that the proposed methodology and the treatment of additive constant uncertainties, as applied in the determination of the Susquehanna-2 Cycle-9 OLMCPR and SLMCPR, are acceptable subject to the condition stated in Section 3 of this f
evaluation and summarized in the following.
In order to insure that the nonconservative POWERPLEX-II MCPR edit is not misinterpreted it is strongly recommended that PP&L take corrective action, such as eliminating the edit or providing a warning to the staff using this edit (Section 3.4).
It is important to recognize that this review does not include those aspects of the methodology which relate to the generic resolution ofthe identified ANFB'oncerns.
The generic resolution of the ANFB I
concerns will be reviewed separately.
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REFERENCES 1.
"Susquehanna Steam Electric Station Proposed Amendment No. 166 to License NPF-22:
Unit-2 Technical Specification Changes for ATRIUM-10Fuel," PLA-4527, Letter, R.G. Byram (PP&L) to U.S. NRC, dated December 18, 1996.
2.
"Application ofANFB to ATRIUM-10for Susquehanna Reloads," EMF-97-010 (P), Revision 0, January 1997.
3.
"Susquehanna Steam Electric Station Proposed Amendment No.
166 to License NPF-22:
Revised ANFB Methodology and Core Flow Dependent MCPR Safety Limits," PLA-4582, Letter, R=.G. Byram (PP&L) to U.S. NRC, dated March 12, 1997.
4.
"Application ofANFB to ATRIUM-10for Susquehanna Reloads," EMF-97-010 (P), Revision 1, March 1997.
5:
"ANFB Critical Power Correlation," ANF-1125(P)(A), and Supplement 1 and Supplement 2, Siemens Power Corporation - Nuclear Division, April 19, 1990.
6, "Request for Additional Information (RAI) on Revision to License Amendment Request for Minimum Critical Power Safety Limits for Susquehanna Steam Electric Station, Unit 2 (TAC No. M97499)," Letter, C. Poslusny (NRC) to R.G. Byram (PP&L), dated April 9, 1997.
7.
"Susquehanna Steam Electric Station Response to NRC Request for Additional Information on Siemens'eport EMF-97-010, Rev. 1," PLA-4595, Letter, R.G. Byram (PP&L) to U.S. NRC, dated March 27, 1997...
8.
"Susquehanna Steam Electric Station Response to NRC Request for Additional Information on PP&L's Proposed Amendment No. 166 to License NPF-22:
Unit-2 Technical Specification Changes for ATRIUM-10 Fuel," PLA-4605, Letter, R.G. Byram (PP&L) to U.S. NRC, dated April 9, 1997.
0
(
~ '
9.
"Susquehanna Steam Electric Station Addendum to PP&L's Response to NRC Request for Additional Information on PP&L's Proposed Amendment No. 166 to License NPF-22'.
Unit-2 Technical Specification Changes for ATRIUM-10Fuel," PLA-4611, Letter, R.G. Byram (PP&L) to U.S. NRC, dated April 16, 1997.
10.
"Susquehanna Steam Electric Station Addendum to PP&L's Response to NRC Request for Additional Information on PP&L's Proposed Amendment No. 166 to License NPF-22:
Unit-2 Technical Specification Changes for ATRIUM-10Fuel," PLA-4613, Letter, R.G. Byram (PP&L) to U.S. NRC, dated April 18, 1997.
11.
"Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors,"
ANF-524 (P) (A), Revision 2, Siemens Power Corporation - Nuclear Division, April 19, 1989.
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