ML20235F915
| ML20235F915 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 02/15/1989 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-66714, NUDOCS 8902230055 | |
| Download: ML20235F915 (9) | |
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David W. Cockfield Vice President, Nuclear February 15, 1989 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555
Dear Sirs:
License Change Application (LCA) 161 Ugfraded Fuel Design (TAC No. 66714)
By our letter of December 15, 1988, we committed to provide you with a Trojan-specific analysis of a locked rotor event which assumes fuel failure for fuel pins that exceed the departure from nucleate boiling (DNB) safety limit criterion.
The enclosed report satisfies that commitment and should be reviewed along with other material previously submitted under LCA 161 (i.e., refer to our letters dated November 20, 1987 and May 27, 1988).
The intent of this analysis is to supplement the locked rotor analysis provided in Chapter 15 of the Final Safety Analysis Report (FSAR) and not to replace it.
The FSAR analysis maximizes peak pressure and the cladding hot-spot temperature, while this analysis maximizes the number of fuel pins in DNB.
Although the Westinghouse position is that fuel pins in DNB do not fall, they have conservatively assumed for this analysis that all pins in DNB fail and release their fission products from the rod gaps to the coolant. The resulting doses at the site boundary, at the low-population zone boundary, and the control room were found to be within regulatory limits.
Approval of LCA 161 is needed prior to Cycle 12 operation scheduled for June 1989.
Should you have any questions, please contact S. A. Bauer at y,
(503) 464-8311.
I Sincerely, W
o TR No gg Attachment q
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Mr. John B. Martin Mr. William T. Dixon
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gy bg Regional Administrator, Region State of Oregon
]op U.S. Nuclear Regulatory Commission Department of Energy MG N
g Mr. R. C. Barr ao NRC Resident Inspector j
011 Trojan Nuclear Plant 12: SW Sarnon SPeet, Pourd. Oregon 97204
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Trojen Nuc152r Picnt Docum' int Con'rol D2sk i
Docket 50-344 February 15, 1989 License NPF-1 Attachment Page 1 of 7
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TROJAN LOCKED ROTOR " RODS-IN-DEPARTURE FROM NUCLEATE BOILING (DNB)" SAFETY EVALUATION 1.0 Introduction As a result of the Nuclear Regulatory Commission (NRC) review of l
the Trojan Nuclear Plant Fuel Upgrade Safety Evaluation (Reference 1) performed by Westinghouse and submitted to the NRC by Portland General Electric (PCE), the NRC noted that the percentage of rods experiencing departure from nucleate boiling (DNB) from a postulated reactor coolant pump (RCP) shaft seizure (Locked Rotor j
" Rods-in-DNB") was not addressed.
In response to this comment, Westinghouse stated that a specific analysis was not performed because the event was not part of Trojan's original licensing basis.
The NRC was awaro of this, but is requiring that PGE address the event.
Therefore, PGE has requested that Westinghouse perform a Trojan-specific analysis of the locked rotor " Rods-in-DNB" event.
In response to PCE's request, a Trojan-specific locked rotor
" Rods-in-DNB" analysis has been performed which includes the determination of the percentage of rods experiencing DNB and a calculation of the dose release.
The dose release is calculated since the NRC Standard Review Plan (Reference 2) calls for the assumption of fuel failure following the prediction of departure from nucleate boiling ratio (DNBR) below the minimum DNBR safety limit during a locked-rotor event.
The radiation doses resulting from this event must satisfy tho Title 10, Code of Federal Regula-tions, Part 100 (10 CFR 100) limit.
A brief description of the locked-rotor event is presented in the following section.
[For a more detailed description of the event, see Section 15.3.3 of the Trojan Final Safety Analysis Report (FSAR).)
2.0 Locked-Rotor Event Description The locked-rotor event is initiated by an instantaneous seizure of a reactor coolant pump rotor resulting in a rapid reduction of the flow in the affected reactor coolant loop.
The rapid reduction of flow initiates a reactor trip on a low-flow signal.
Following i
reactor trip, the reduction in flow causes a rapid increase in the I
reactor coolant system (RCS) and fuel rod temperatures.
Therefore, if the reactor is at full power prior to trip, DNB may not be pre-cluded for this American Nuclear Society (ANS) Condition IV event.
Any fuel damage calculated to occur must be of sufficiently limited extent that the core remains in place and intact with no loss of core cooling capability.
In addition, all rods that are predicted
Trojen Nuc1csr Pltnt Docum nt Control D:sk Docket 50-344 February 15, 1989 License NPF-1 Attachment Page 2 of 7 to experience DNB (the DNBR falls below the safety limit value) are assumed to fail and any activity release must be such that the calculated doses at the site boundary are within the 10 CFR 100 guidelines.
The following sections present a brief discussion of:
(1) the analysis of the calculation of the percentage of rods experiencing DNB, and (2) the analysis of the dose release based on the percentage of rods experiencing DNB.
3.0 Percentage of Rods Experiencing DNB Analysis Method of Analysis The Nuclear Steam Supply System (NSSS) and core transients, result-ing from a locked-rotor event, are calculated using three digital computer codes. First, the LOFTRAN code (Reference 3) is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated loop flows, the nucicar power transient, and the primary system pressure and temperature tran-sients. The FACTRAN code (Reference 4) is then used to calculate the heat flux transient based on the nuclear power and flow from LOFTRAN.
Finally, the THINC code (see Section 4.4 of the Trojan FSAR) is used to calculate the DNBR during the transient based on the heat flux from FACTRAN and flow from LOFTRAN.
This event is analyzed with the Improved Thermal Design Procedure (ITDP) as described in Reference 5.
The ITDP methodology combines the uncertainties on power, flow, temperature, etc., into the limit DNBR, thus it is primarily used for Condition II (DNB) ovents.
However, it is used here because the DNBR is calculated for this Condition IV event to determine the percentage of rods experiencing DNB for this event. The use of the ITDP methodology is acceptable for this event since all of the important parameters such as flow, temperature, pressure, power level, etc., fall within the ranges defined by the WRB-1 correlation and within the ranges used in ITDP sensitivity studies.
A description of the important assumptions follows.
1.
Initial Conditions:
Initial reactor power, pressure and RCS temperature are assumed to be at their nominal values.
Uncertainties in initial condi-tions are included in the limit DNBR as described in Reference 5.
The NSSS power is assumed to be 3,423 MWt, the RCS average temperature 584.7'F, and the RCS flow 371,700 gal-lons per minute (gpm).
Terjtn Nuclcir Plent Document C:ntrol D;;k Docket 50-344 February 15, 1989 License NPF-1 Attachment Page 3 of 7 2.
Reactivity Coefficients:
A positive moderator temperature coefficient of +5 percent milli-rho per degree Fahrenheit (pcm/*F) and a minimum Doppler-only power coefficient of reactivity are conservatively assumed.
3.
Loss of Offsite Power:
Following reactor trip, it is assumed that power is immediately lost to the intact reactor coolant pumps (a zero-second time delay).
During normal operation, power for the reactor coolant pumps is supplied through the individual buses connected to the generator and the offsite power system.
The NRC Standard Review Plan currently requires that this event be analyzed to address both with and without offsite power available condi-tions. However, only the case without offsite power available is analyzed because it is more limiting. The loss of power to the intact RCPs, in addition to the loop with the locked RCP rotor, results in a degraded core flow condition minimizing the heat removal capability of the RCS.
4.
F-delta-H:
The F-delta-H assumed in the analysis is consistent with the F-delta-H (the ITDP value of 1.56) justified in the Trojan Nuclear Plant Fuel Upgrado Safety Evaluation (Reference 1),
Results and Conclusions of Locked-Rotor Analysis Based upon the analysis of the locked-rotor event discussed above, 25 percent of the fuel rods experienced DNB.
This number is based on the percentage of rods for which the DNBR is less than the limit value. This percentage of fuel rods is used in the determination of the radiological consequences (dose release), as discussed in the following section.
It should be noted that the current Westinghouse position is that rcds-in-DNB during a locked-rotor event do not fail, as documented in Reference 6.
However, this position has not yet received NRC approval.
4.0 Radiological Evaluation Following a RCP shaft seizure with subsequent turbine and reactor trip, the steam system pressure rises, and the secondary side relief valves automatically open to release steam to the atmosphere.
Offsite power is assumed to be lost. Hence, steam l
dump to the condenser is not available. The radiological conse-quences of this event are presented belcw.
W Trojan Nuclser Pltnt Document' Control Dsck l
Docket 50-344 February 15,:1989 l
Licenst NPF-l'
' Attachment Page 4 of'7' l
Fission Product Release Assumptions The following conservative assumptions are used to evaluate the iguantity of radioactivity released to the environment following a RCP shaft seizure:
1.
As a result of the shaft seizure, 25 percent of the fuel rods in the core experience DNB.
These rods are assumed to fail,.
releasing fission products from the rod gaps to the reactor coolant.
2.
The fission product activity released to the reactor coolant is equal to 25 percent of the gap activity presented in FSAR Table 15.0-5.
3.
The equilibrium concentrations of iodine and noble gas in the primary coolant are based on one percent defective fuel as presented.in FSAR Table 11.1-2.
4.
The equilibrium concentrations of iodine in the secondary coolant are based on one percent defective fuel and a 1 gpm primary-to-secondary leak rate which is assumed to have existed during normal operation for sufficient time to establish equilibrium nuclide inventories in the secondary coolant.
5.
The iodine partition coefficient in the steam generators is 10-2 Noble gas is not assumed to partition, i.e., partition coefficient is 1.0.
6.
The total primary -Lo-secondary leak rate during the event is assumed to equal 1 gpm, which is the maximum value allowed by the Trojan Technical Specifications.
7.
The total mass of steam assumed to be released from the secondary side relief valves to the environment is as follows:
0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 600,000 pounds 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1,200,000 pounds D6se Calculation Model Activity released through the relief valves is assumed to escape at ground level.
No credit is taken for cloud depletion of radio-activity by ground deposition or by radioactive decay during transport to the nearest site boundary (662 meters) or to the low-population zone (LPZ) outer boundary (4,000 meters). The meteorological conditions and models used to determine the atmos-pheric dispersion factors used in the analysis of this event are presented in FSAR Section 2.3.
The offsite dispersion factors are presented in FSAR Table 15.0-9.
Trojan Nuclear Plant Document Control Desk Docket 50-344 February 15, 1989 License NPF-1 Attachment l
Page 5 of 7 l
l The thyroid and total-body (beta plus gamma) doses are calculated assuming immersion in a semi-infinite cloud of fission products.
The dose equations and parameters used are discussed in FSAR Section 15.0.11.
These parameters and equations are consistent with the recommendations of Regulatory Guide 1.4 (Reference 7).
Offsite Doses Offsite doses are calculated at the site boundary for the 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> period and at the LPZ for the 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> period. The doses, in radiation equivalent man (rem) are as follows:
0 Hour Site Boundary:
Thyroid 0.5 Total-body 0.02 0 Hour Low-Population Zone:
Thyroid 0.2 Total-body 0.005 These doses are within a small fraction of the 10 CFR 100 guideline (30 rem thyroid and 2.5 rem whole-body, as defined in NUREG-0800, Section 15.3.3, Reference 8).
Doses to Control Room Personnel Doses to control room personnel have been calculated for the 8-hour duration of the shaft seizure event. The gamma-body doses have been adjusted by the finite-cloud geometry factor suggested by Murphy and Campo (Reference 33 of FSAR Section 15.6).
The atmos-pheric dispersion factor at the control room air intake, corres-ponding to an activity release from the main steam safety valves, 3
assumed to equal 5.23 x 10-2 second per cubic meter (sec/m )
(Reference 9).
The control room heating, ventilating and air conditioning (HVAC) parameters are presented in FSAR Section 15.6.
The resulting control room dosos, in rem, are ns follows:
1-Train HVAC:
Thyroid 6.0 Gamma-body 0.07 Bota-skin 3.2 l
I p
Trojnn Nuciscr Pirnt Document Contral De?,k i
Docket 50-344 February 15, 1989 I
License NPF-1 Attachment Page 6 of 7 2-Train HVAC:
Thyroid 6.3 Gamma-body 0.09 Beta-skin
'4.0 These doses are within the acceptance criteria of General Design Critorion 19 and NUREG-0800, Section 6.4 (30-rem thyroid and beta-skin, 5-rem gamma) (Reference 10).
Filter Fission Product Loadings No recirculating or single-pass flitration systems are used to clean up or control the fission products released from the relief' valves as a result of an RCP shaft seizure.
Fission product iodine loadings for the control room emergency charcoal filters have been evaluated for the more severe conditions postulated for the large-break Loss-of-coolant Accident (LOCA) (FSAR Section 15.6.5.6.1) which have been found to be negligible when compared to the design capacity of the system.
(The control room emergency filter system is described in FSAR Section 6.4.)
5.0 ove-all conclusions The percentage of fuel rods expected to experience DNB as a result of the locked-rotor event is calculated to be 25 percent. The current Westinghouse position regarding these rods in DNB is that they do not fail. However, a radiological evaluation of this event is performed under the assumption that the rods in DNB do, in fact, fail and release their fission products from the rod gaps to the reactor coolant. The resulting doses at the site boundary, at the low-population zone (LPZ) boundary, and the control room were found to be within regulatory limits.
3263P.0289
,s Trejin Nucl@tr Plant Doccm:nt Control Drak Docket 50-344 February 15, 1989 License NPF-1 Attachment Page 7 of 7 REFERENCES 1.
" Plant Safety Evaluation for the Trojan Nuclear Plant Fuel Upgrade",
October 1987.
2.
" Standard Review Plan for the Review of SARs for Nuclear Plants",
NUREG-0800, Sections 15.3.3-15.3.4, USNRC, July 1981.
3.
Burnett, T. W. T., et al, "LOFTRAN Code Description", WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984 4.
Hargrove, H.
C., "FACTRAN-A Fortran-IV Code for Thermal Transients in a UO2 Fuel Rod", WCAP-7908 (Non-Proprietary), June 1972.
5.
Chelemer H.,
Boman, L.
H.,
Sharp, D.
R., " Improved Thermal Design Procedure", WCAP-8567 (Non-Proprietary), July 1975.
6.
WCAP-11157, Revision 1 " Integrity of Fuel Rods During a Locked Rotor / Shaft Accident". (Proprietary), June 1986.
7.
" Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors", Regulatory Guide 1.4 8.
" Reactor Coolant Pump Seizure and Reactor Coolant Pump Shaft Break",
NUREG-0800, Section 15.3.3 9.
Provided by G. R. Alberthal of Portland General Electric, Telecon dated February 1,1989 10.
" Control Room Habitability System", NUREG-0800, Section 6.4 3263P.0289 l
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