GO2-06-059, License Amendment Request to Adopt Origen-ARP for Calculating Fuel Pool Decay Heat Load

License Amendment Request to Adopt Origen-ARP for Calculating Fuel Pool Decay Heat Load

ML061110153
Person / Time
Site:	Columbia
Issue date:	04/17/2006
From:	Oxenford W Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-06-059
Download: ML061110153 (11)

Text

X ENERGY N NORTHWEST PeoFple Vision* Solutions P.O. Box 968

• Richland, WA

• 99352-0968 April 17, 2006 G02-06-059 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD

Reference:

Letter dated November 29, 2005 from DD Chamberlain (NRC) to JV Parrish (Energy Northwest), "Columbia Generating Station Baseline Inspection Report 05000397/2005009"

Dear sir or Madam:

Pursuant to 10 CFR 50.90, Energy Northwest requests approval for a change incorporated into the current amendment 58 of the Columbia Generating Station (Columbia) Final Safety Analysis Report (FSAR). Energy Northwest initially made this change pursuant to a 10 CFR 50.59 evaluation. The referenced inspection report documents a differing opinion on the 10 CFR 50.59 evaluation conclusion and determined the change represented a departure from a method of evaluation described in the FSAR. Therefore, Energy Northwest is seeking NRC review and approval to change the method for calculating fuel pool decay heat load from the original licensing basis methodology of ORIGEN and the Auxiliary Systems Branch Technical Position (ASBTP) 9-2, "Residual Decay Heat Energy for Light Water Reactors for Long-Term Cooling," to ORIGEN-ARP.

Since the change made to the FSAR pursuant to 10 CFR 50.59 was characterized as a violation, Energy Northwest entered this condition into the corrective action program. In accordance with the process described in Part 9900 of the NRC Inspection Manual, "Operability Determinations & Functionality Assessments for Resolution of Degraded and Nonconforming Conditions Adverse to Quality or Safety," dated September 26, 2005, Snergy Northwest has characterized this FSAR change as a nonconformance.

. I LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Page 2 of 2 Energy Northwest is seeking to resolve this nonconformance by obtaining NRC approval of the proposed change in methodology. In the interim, this condition has no impact on plant safety. This nonconformance will have no impact on plant operations until the next refueling outage scheduled for May 2007, barring an unplanned condition that would require a core offload. Energy Northwest requests approval of this amendment by January 2007 to support the timely completion of the fuel pool heat load management plan for the Spring 2007 refueling outage.

Attachment 1 provides an evaluation with a detailed description of the background, proposed change, and technical analysis. The No Significant Hazards Consideration and Environmental Consideration are also contained in Attachment 1. A copy of the revised FSAR page (amendment 58) that replaced the reference to ASBTP 9-2 with a reference to ORIGEN-ARP is provided in Attachment 2 for information.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Washington State Official.

No commitments are made in the submittal. Should you have any questions or require additicnal information regarding this matter, please contact Mr. Michael K Brandon, at (509) :377-4758.

I declare under penalty of perjury that the foregoing is true and correct. Executed orn the date of this letter.

Respezctfully, WS Oxen0 r Vice Presidetnt, Technical Services Mail Drop PE04 Attachment 1: Evaluation of the Proposed Change Attachment 2: Copy of the Affected FSAR Page cc: BS Malleft - NRC RIV BJ Benney- NRC NRR NFRC Senior Resident Inspector/988C RN Sherman - BPA/1 399 WA Horin - Winston & Strawn JO Luce - EFSEC RFZ Cowley - WDOH

LICEN SE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 1 of 7 Evaluation of the Proposed Change

1.0 DESCRIPTION

Energy Northwest is requesting NRC approval to replace the original license basis methodologies of ORIGEN and Auxiliary Systems Branch Technical Position (ASBTP) 9-2 with ORIGEN-ARP. This request will resolve the nonconformance associated with this methodology that currently exists in FSAR section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System." ORIGEN-ARP is the current and only actively supported stand-alone version of the ORIGEN code. The ORIGEN-ARP methodology is based on the original version of ORIGEN and ORIGEN-ARP has been approved by the NRC for use in the evaluation of dry cask storage systems. The use of ORIGEN-ARP provides a more rigorous solution to the problem of determining spent fuel decay heat.

2.0 PROPOSED CHANGE

Prior to FSAR amendment 58 (current amendment number), the fuel pool heat load values presented in FSAR Table 9.1-4 were calculated using ORIGEN and based on estimated refueling data. As stated in (pre-amendment 58) FSAR Section 9.1.3.2.1, the fission product decay heat loads shown in FSAR Table 9.1-4 are "in agreement with ASBTP 9-2." The FSAR recognizes that refueling specific analyses may be required to validate the maximum fuel pool temperatures are within the licensing basis acceptance criteria. Since the original ORIGEN code used to calculate fuel pool heat loads is no longer supported and has been replaced by an improved ORIGEN-ARP version, Energy Northwest is requesting to replace the methodologies of ASBTP 9-2 and ORIGEN fcr calculating spent fuel pool decay heat loads with the improved ORIGEN-ARP code.

The use of ORIGEN-ARP will facilitate future compliance with the current FSAR spent fuel pool temperature limits by improving the accuracy and ease of performing future outage specific calculations.

3.0 BACKGROUND

Since both ASBTP 9-2 and ORIGEN were discussed in the Columbia FSAR with respect to calculating decay heat, a brief discussion of each methodology is provided below.

The methodology in ASBTP 9-2 is based on experimental data published from 1958 to 1973 relating to energy release from the decay of fission products. The methodology draws heavily on the proposed ANS standard, "Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," approved by Subcommittee ANS-5, October 1971 that presents a standard fission product decay heat curve versus cooling time. The actual equation resulting from the curve fit in ASBTP 9-2 is more complex than that in the proposed ANS standard. However, in general, the results of the curve-fit equations agree with each other reasonably well (within +/-5%). In addition, the

LICEN SE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 2 of 7 terms for decay heat generation due to the heavy elements and the uncertainty factors in ASBTP 9-2 were taken directly from the above referenced ANS standard. The Foreword of the proposed standard, prepared by Subcommittee ANS-5, provides insight into the development of the methodology and states:

"It is the opinion of the generating group that the state of the knowledge of the relatively complex phenomena that contribute to the total energy release rate does not justify a detailed description of the process. From a very detailed description one might be able to determine the inventory of each radioisotope as a function of time and, knowing the decay constants and energy manifestation, calzulate the total energy deposition. Although such calculations have been made, it was the opinion of the generating group that there is not adequate knowledge of the many physical constants involved for use of this approach in a standard." (published in 1971)

Furthermore, text in ASBTP 9-2 acknowledges the lack of consistent experimental data and the differing results of various calculations but reasons that "the effect of all uncertainties can be treated ... by a suitably conservative multiplying factor." This factor is either 10% or 20% depending on the shutdown time of interest.

By contrast, the ORIGEN computer code models fissile material behavior during periods of irradiation and decay by computing time-dependent concentrations and source terms of a large number of isotopes that are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, and physical or chemical removal rates. The primary advantage of ORIGEN over earlier burnup codes was its capability to treat the full isotopic transition matrix rather than a limited number of transmutation chains. The original ORIGEN program utilized a standard cross-section library designed for the analysis of standard light water (LWR) reactor fuel. The required data input for ORIGEN consists of data relating to the specific problem to be analyzed including input for fissile isotope concentrations (i.e., bundle enrichments and uranium weight:), bundle power during irradiation, irradiation length, and length of decay period.

Hence, ORIGEN provides a fairly rigorous treatment of the problem. This assessment is supported by the NRC in Information Notice 96-39 which states:

"ORIGEN does not use empirical methods to calculate decay heat but tracks the buildup and decay of the individual fission products within the reactor core during operation and shutdown. ORIGEN also includes the effect of element transmutation from neutron capture, both in fissile isotopes and fission products.

Because ORIGEN is a rigorous calculation of all decay heat inputs, it was used in the calculations for decay heat ..."

Essentially, all features of the original ORIGEN were retained, expanded or supplemented within new computations in the development of ORIGEN-ARP. The original ORIGEN code released in 1973 was intended principally for the generic

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page a of 7 investigation of fuel cycle operations over a wide range of fuel design parameters.

Soon Users began to apply ORIGEN to problems that required greater precision and specificity than had been intended with the original data libraries (only one cross-section library for all LWR fuel with no burnup dependence). In 1980, ORIGEN2 was released providing a significant update to ORIGEN and the associated database. Like the original ORIGEN code, ORIGEN2 was designed to operate as a stand-alone calculational tool with fixed cross-section data libraries provided for several reactor models. ORIGEN2 is no longer a supported code. Rather ORIGEN-ARP has been developed as a stand-alone tool or as an interface with the SCALE code system to generate problem-specific neutronic data though interaction with other modules of SCALE. For Columbia, this neutronic data is in the form of a cross-section library specific to the bundle being evaluated.

The NRC has previously approved the use of ORIGEN2 for spent fuel pool applications.

In a Safety Evaluation (SE) dated September 21, 2001, the NRC issued an amendment to Duane Arnold for a revised thermal-hydraulic analysis of the spent fuel pool. In discussing the methodology for determining bundle decay heat, the Technical Evaluation Report referenced by the SE states:

"This program can perform decay heat calculations using either Branch Technical Position ASB 9-2 or the ORIGIN2 [sic] computer code.... The ORIGIN2 [sic]

option was used. All fuel assemblies were assumed to have been irradiated to the appropriate maximum burnup level. Based on this review, BNL [Brookhaven National Laboratory] concurs that the methodology and assumptions the licensee used to calculate the decay heat loads meet the intent of the applicable NRC guidelines."

Similarly, in an SE dated August 30, 2002, the NRC issued an amendment to V. C.

Summer for a spent fuel pool re-racking modification. The methodology used and accepted by the NRC for calculating spent fuel pool heat load in the V.C. Summer amendment was the ORIGEN2 code.

Based on the above examples, the ORIGEN2 methodology has been previously approved by the NRC for the calculation of fuel pool decay heat loads.

4.0 TECHNICAL ANALYSIS

The ORIGEN methodology, including the updated ORIGEN-ARP, provides a rigorous treatment of the physical phenomenon involved by computing time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay and physical or chemical removal rates.

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page 4 of 7 The ORIGEN series of codes is widely used throughout the world for predicting the characteristics of spent reactor fuel. The design and licensing activities associated with storage of spent fuel is dependent on the analytic characteristics (e.g., decay heat, radiation sources). Spent fuel characterization for out-of-reactor applications is routinely obtained with point (i.e., no spatial dependence) depletion/decay methods that are able to capture an almost unlimited number of nuclide inventories and associated characteristic quantities in a flexible and useful format. Perhaps the most widely used point-depletion codes are the ORIGEN series of codes developed at Oak Ridge National Laboratory (ORNL). These codes were developed by ORNL for the NRC and are currently maintained/enhanced under co-sponsorship of NRC and Department of Energy (DOE). These codes are used to predict the characteristics of spent reactor fuel and high level waste for fuel cycle evaluation and out-of-reactor analysis applications.

Like the original ORIGEN computer code, ORIGEN2 and ORIGEN-ARP were designed to operate as a stand-alone calculational tool with fixed cross section data libraries provided for several reactor models.

ORIGEN-ARP uses two modules for calculating decay heat: (1) ARP (Automatic Rapid Process) for providing a problem-specific cross-section library file for input to ORIGEN-S, and (2) ORIGEN-S for performing the depletion and decay calculations described above.

The primary difference between ORIGEN2 and ORIGEN-ARP is the different methods used to develop and supply the burnup-dependent cross-section libraries for use by the ORIGEN code module.

The ORIGEN2 code uses cross-section and fission-product yield libraries that have been developed for a limited number of specific reactor models. For BWRs, only a single library is available for use, corresponding to a 2.75 wt% enriched bundle irradiated for 4 cycles to a burnup of 27.5 MWd/kgU.

ORIGEWN-S, on the other hand, is designed to interact with the other modules of the SCALE code system to produce neutronics data based on the actual assembly design and pertinent reactor history information. The ARP module provides ORIGEN-S with a cross-section library that is based on the actual bundle design and operating parameters such as assembly type (e.g., GE 8x8, Siemens 8x8), enrichment, burnup, and operating history.

At Columbia, the SCALE code system was used to generate cross-section libraries for the GE: 8x8, Siemens (SPC) 8x8, SPC 9x9, SVEA-96, and ATRIUM-1 0 BWR fuel assembly designs that have been irradiated in the Columbia reactor core. These cross-section libraries are employed in ORIGEN-ARP calculations to predict decay heat for the various BWR fuel assemblies as a function of decay time following irradiation. In a similar fashion, SCALE was used to generate cross-section libraries for the GE 7x7 lfuel assembly design that was irradiated in the Cooper reactor core. Measurements of

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment 1 Page r; of 7 decay heat were performed on several Cooper bundles as published in EPRI report NP-4619 and NUREG/CR-5625. ORIGEN-ARP was employed to calculate the decay heat values of these same assemblies and the results were compared to the measured values in order to establish the methodology bias and uncertainties to be applied to the calculated results for predicting decay heat.

Benchmarking of the ORIGEN2 code to ORIGEN-ARP for Columbia is documented in Westinghouse Report A-GEN-BWR-90. Table 6.8 of A-GEN-BWR-90 documents the!

results of the comparison of the three codes (ORIGEN2, SAS2H/ORIGEN-S, and ORIGEN-ARP) to the measured data. The bias for decay heat from each methodology (when compared to the measured data) is within 1.2% of each other, and the standard deviation is within 0.55%. The total error at the 95% confidence level is 15.83% for ORIGEN2 and 15.76% for ORIGEN-ARP. Based on the results of the benchmarking analysis, the rigorous procedure used to determine the methodology bias and uncertainties as compared to measured data, and the fact that ORIGEN-ARP was developed to provide greater flexibility and accuracy of results, the results of ORIGEN2 and ORIGEN-ARP are essentially the same.

One difference between the assumptions and inputs evaluated in the two referenced SEs for ORIGEN2 and the intended application at Columbia is the treatment of power measurement uncertainty. In the referenced ORIGEN2 applications, a 2% factor for power measurement uncertainty was included. This would normally be applied by multiplying the specific power by 1.02. The ORIGEN-ARP bias and uncertainty terms were derived by comparing code predictions to actual measured data. Thus, the uncertainty in power measurement is inherently included in the error terms.

To ensure that the error term adequately encompasses the required power uncertainty factor, comparisons were made between the results assuming only the 2% power uncertainty factor and the results assuming no power uncertainty but applying the error assoc ated with the best estimate analysis. The decay heat with purely the bias term applied (represents best estimate) was greater than the decay heat calculated with the 1.02 multiplier on specific power for all cases. Since the best estimate approach represents the minimum error that would be applied to the decay heat results, the requirement to include a 2% factor for power measurement uncertainty is indeed satisfied by using the methodology bias procedure defined in A-GEN-BWR-90 for any confidence level.

The SCALE code system, including ORIGEN-ARP, has been classified as Category B software and installed and tested under the Energy Northwest Software QA program.

Based on this discussion, ORIGEN-ARP provides an appropriate methodology for calculating fuel pool heat loads.

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Page (i of 7 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Energy Northwest has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment" as discussed below.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The adoption of ORIGEN-ARP does not affect the probability or consequences of an accident previously evaluated. The calculation of the fuel pool heat load is used to evaluate and demonstrate the ability of the fuel pool cooling system to maintain the fuel pool temperatures within the acceptance limits specified in the Columbia FSAR. The proposed change to the methodology used to calculate the fuel pool heat load has no bearing on the probability or consequences of any previously evaluated accident. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The change involves the use of a different methodology for calculating fuel pool decay heat load. This change does not involve any new equipment, it does not change any previously approved acceptance limits, and it does not affect or alter the operation of any equipment. Therefore this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The margin of safety provided by the fuel pool cooling system is primarily defined by the difference between the maximum allowed fuel pool temperature and the boiling point of water. The margin of safety is supplemented by the ability to make up water to the spent fuel pool if boiling were to occur. The proposed change in methodology for calculating the fuel pool heat load does not alter the current temperature limits or acceptance criteria specified in the FSAR and has

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Attachment I Page 7 of 7 no effect on the ability to provide make-up water if boiling were to occur. This change will allow Energy Northwest to more accurately calculate the fuel pool heat load to provide added confidence in the ability of the fuel pool cooling system to accommodate the heat load added to the spent fuel pool during refueling activities. Therefore, this change does not involve a significant reduction in a margin of safety.

Based on the above, Energy Northwest concludes that the proposed amendment does not warrant a significant hazard consideration under the standards set forth in 10 CFR 50.92(c) and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The proposed change does not impact the compliance with any regulatory requirements. The spent fuel pool cooling system design and acceptance limits specified in the FSAR are not changed. The change allows Energy Northwest to adopt a state of the art methodology for calculating the decay heat load in the fuel pool. The proposed methodology is a change from the methodology recommended in Revision 1 of NUREG 0800, "Standard Review Plan," Section 9.1.3, dated July 1, 1981; however, the basis for the acceptability of the proposed methodology is provided in this request. This change will improve the ease and accuracy of heat load calculations and will facilitate future compliance with the existing acceptance criteria with a greater degree of certainty.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation after the proposed change, (2) operation after the change 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 the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve; (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; or, (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

LICENSE AMENDMENT REQUEST TO ADOPT ORIGEN-ARP FOR CALCULATING FUEL POOL DECAY HEAT LOAD Copy of the Affected FSAR Page

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 9.1.3.2.1 System Operation 9.1.3.2.1.1 FPC System Cooling Function The system normally cools the fuiel pool by transferring the spent fuel decay heat from the tuber side of the two fuel pool cooling heat exchangers to the RCC. Standby SW can be manually aligned as the alternate heat sink for the heat exchangers (see Section 9.1.3.2.3 for additional details). The fuel pool is maintained at or below 1250 F during normal plant operations. The fuel :ool temperature may rise above this value during refueling activities or during an anticipated operational transient of the loss of one train of the FPC system. The RHR system can also be manually aligned into several configurations to provide su lemental cooling of the fuel pool. One of these RHR configurations is the RHR/F ase mode o 5e-Section 9.1.3.2.2 for additional details).

The maximum heat load is present in the spent fue l refueling activities wn recently irradiated fuel bundles are discharged fr m the react or e to t ue e magnitude of this heat load is contingent upon t e cycle-specific refueling activities, i.e., the number of the bundles discharged, the burnup f the discharged bundles, and the decay time of each bundle when it is placed in the eiated with a planned disclarge can be calculated with ORIGEN-ARP computer cod ssuming a 2% thermal power uncertainty or other accept1o s and uncertainties.

During refueling activities, the fuel pool temperature is managed by controlling the number and schedule of fuel assemblies discharged, controlling the number of heat removal systems in service, and controlling the temperatures of the systems (RCC or SW) used to remove the heat from the FPC heat exchangers.

The fuel pool cooling system was originally designed to maintain the pool at a temperature of less than or equal to 125 0 F during refueling activities with both trains of fuel pool cooling in operation and RCC cooling water at 950 F. The decay heat load assumed for the original design basis (normal offload) was based on the original licensed power of 3323 MW-thermal, a one-year fuel cycle, and a quarter core offload with a 20-day decay period.

Since the original design, the licensed power was increased to 3486 MW-thermal and the operating cycle was revised from a one-year cycle to a two-year cycle. These changes resulted in an increased heat load in the fuel pool, particularly during refueling outages. For the current design (3486 MW-thermal and a two-year cycle), a 150'F fuel pool temperature limit (see Table 9.1-6) applies to normal refueling activities for the scenario of both trains of fuel pool sooling in operation.

The FPC system is also designed to provide sufficient cooling for an anticipated operational transient of the loss of one train of the FPC system. For this transient, the maximum bulk fuel pool water temperature is limited to 1550 F. This limit applies to both normal operation and normal refueling activities (i.e., excluding a full core offload).

__j LDCN4)5-005 9.1-23