ML20012C736
| ML20012C736 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/10/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20012C735 | List: |
| References | |
| NUDOCS 9003230142 | |
| Download: ML20012C736 (6) | |
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UNITED STATES
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g NUCLE AR REGULATORY COMMISSION -
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W ASHINGTON, D, C. 70666
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 32 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 23 TO FACILITY OPERATING LICENSE NPF-52 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DUKE POWER COMPANY, ET AL.
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INTRODUCTION 12,)1987, (Ref.1) and supplemented June 23 and AugustDuke By letter dated June 12, 1987, (Refs. 2 and 3,
changes to the Technical Specifications to request operation up to full power with the Upper Head Injection (UHI) System removed. The VH1 system was designed for use in Catawba to enhance core cooling during the blowdown phase of the loss-of-coolant accidents (LOCA). The removal of the UHI system is to increase plant operational flexibility. The proposed changes include:
(1) 4 maintenance and leakage verification, and (2)g UHI system surveillance, deletion of Technical Specifications requirin modification of the Technical Specifications to reflect removal of the UHI system.
Similar changes have previously been approved by NRC for incorporation into the McGuire Technical Specifications (Ref. 4). The staff has reviewed the proposed Technical Specification changes and the supporting analyticc1 results, and has prepared the following evaluation.
II.
EVALUATION 1.
Peak Containment Pressure and Temperature The peak containment pressure originally calculated in the Final Safety Analysis Report (FSAR) was detemined for the design basis LOCA conservatively assuming there was no UH1 system in operation. Therefore, the current calculated peak pressure in the FSAR is still valid after the UHI system is removed. However, the main steam line break (MSLB) analysis in the FSAR for Ca determining peak containment temperature has been revised due to the concern regarding superheated steam releases described in IE Infonnation Notice 84-90.
l In the revised analysis, the limiting case MSLB was detemined to be a 0.86 n
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-o Ft break at 102% reactor power. The UHI system would not be actuated for this 38$
break because the actuation pressure of 1200 psia in the primary system could 28 not be reached. The 1200 psia actuation pressure can be reached only for No breaks at much lower power (less than 30%).
It should be noted, however, that for breaks at low power, the integrated blowdown mass and energy, maximum steam enthalpy, and therefore, the containment temperature are less than the
-o limiting case discussed above. The licensee, therefore, concluded that the 8@
deletion of the UHI system has no impact on the limiting case peak containment Sj temperature. Based on the above, the staff concurs with the licensee that om deletion of the UHI system has no effect on the peak containment temperature REo.
calculation as long as the limiting case is a MSLB at 102% power.
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c 2-2.
Minimum Containment Pressure Analysis for Perfomance Capability Studies
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Appendix K to 10 CFR 50 requires that the containment pressure used for i
evaluation of cooling effectiveness during reactor core reflood shall not exceed a pressure calculated conservatively for this purpose. The calculation must include the effect of operation of all installed containment pressure-reducing systems and processes. The corresponding reflood rate in the core will then be reduced because lessened containment pressure reduces the resistance to steam flow in the reactor coolant loops and increases the boiloff rate from the core.
The licensee has performed the containment back pressure calculation for ECCS perfomance studies using the computer code LOTIC-2. This is consistent with the FSAR. The input of mass and energy release rates corresponding to a double-ended cold leg geil10 tine break (C0 = 0.6) were calculated using the method described in the redsed FSAR Section 15.6.5, Loss-of-Coolant Accidents.
However, the composter code BASH, instead of the previously used code WREFLOOD, was employed to compute the reflood transient.
The staff has revised the licensee's input data including initial containment conditions, containment net free volume and passive heat sinks. Some of these data were revised from the FSAR to reflect the as-built plant configuration. The staff finds these data acceptable.
Based on the above, the staff finds the minimum containment pressure calculated for ECCS perfomance to be acceptable.
l 3.
Large Break Loss-of-Coolant-Accident (LOCA) Analysis The licensee provided the results of a large break LOCA analysis supporting the request for removal of the UH1 system.
In the licensee's submittal only double-ended cold leg guillotine (DECLG) breaks were analyzed since they were identified previously as limiting cases that result in the highest peak cladding temperature. The DECLG break analysis was performed with a total peaking factor of 2.32, 102% of the core power of 3411 Mwt and an assumed i
loss of offsite power at the beginning of the accident. A sensitivity study of DECLG break sizes on the effect of the peak cladding temperature was perfomed by use of Moody discharge coefficients of 1.0, 0.8 and 0.6 The results showed that the DECLG break with a discharge coefficient of 0.8 is the worst large break LOCA case, resulting in a peak cladding temperature of 1704'F. The analysis was perfomed with a modified version of the 1981 Westinghouse ECCS evaluation model (Ref 5). This evaluation model used the revised PAD fuel themal safety model (Ref. 6) for the calculation of the initial fuel conditions; the SATAN-VI code (Ref. 7) for the transient thermal hydraulic calculation during blowdown period; the WREFLOOD (Ref. 8) and BASH (Refs. 9,10 and 11) codes for the calculation of the refill and reflood transient periods; the LOCBART (Refs. 9 and 12) for the calculation of the peak cladding temperature; and the LOTIC code (Ref.13) for the calculation of the ice condenser containment pressure transient.
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. The staff has reviewed the analysis. The staff found that the approved analytical methods and computer codes (including the BASH modifications in Reference 11, which has been approved for the plant specific application) were used and the results showed that the peak cladding temperature, metal-weter reaction and clad oxidation are within the acceptance criteria imposed in 10 CFR 50.46 for LOCA analysis.
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Small Break LOCA Evaluation The small break LOCA analysis was perfomed with the approved comDuter codes.
1.e..-(1) the NOTRUMP (Refs.14 and 15) code for the calculation of the transient depressurization of the reactor coolant system, core power, water-steam mixture height and steam flow past the uncovered portion of the core, and (2) the LOCTA-IV (Ref.12) code for the peak cladding temperature analysis. The analysis was done assuming 10?% of the core power of 3411 Mwt, and a total peaking factor of 2.32. Various break sizes were perfomed and l
the results showed that the worst break size is a 4-inch diameter break which results in the highest peak cladding temperature of 1304*F. well below the acceptance criteria of 2200'F. The staff concludes that the small break LOCA analysis is acceptable since the approved method was used to show the analytical results to be within the acceptance criteria of 10 CFR 50.46.
I 5.
Transient Evaluation The licensee used the approved LOFTRAN code (Ref. 16) to reanalyze two plant transients which were (1) the inadvertent opening of a steam generator relief or safety valve and (2) the steam line break. These two transients were reanalyzed beceuse they were the only transients which were predicted to depressurize the primary system Sufficiently to actuate the UHI system. An analysis to determine whether departure from nucleate boiling (DNB) occurred was perfomed for both transients and the results confirmed that no DNB occurred and thus assured no fuel damage resulting from the transients.
The staff concludes that the licensee's transient analysis is adequate and acceptable since the approved method was used and the results were demonstrated to be acceptable since specified acceptable fuel design limits were not exceeded.
6.
Occupational Radiation Exposure The licensee has provided an estimate of the occupational radiation exposure for the UHI deletion modification. The estimate is based on anticipated stay times f or each major subtask and estimated dose rates.
This estimated total dose is 80 person-rem per unit, which is a small fraction of the 1986 annual average PWR dose of 392 person-rem per unit. The licensee is required to maintain occupational radiation exposures within regulatory limits.
Further-more, the licensee has comitted to and is implementing a program for maintaining occupational radiation exposures as low as is reasonably achievable. Therefore, the projected occuoational radiation exposures associated with the UHI deletion modification are acceptable.
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III. Technical Specification Changes The proposed chan es to the Technical Specifications (TS) associated with the removal of the UH system are discussed below.
(1) TS 3/4.3.3.8 - Fire Detection Instrumentation The requirements of TS of the fire detection instruments in the UHI building are proposed to be deleted from the previous Technical Specifications because fire protection for the UHI system is no longer needed.
(2) TS 3/4.4.6 - Reactor Coolant System Leakage The requirements of leak verification for the UH1 system related piping and valves are deleted since equipment is no longer part of the reactor coolant system.
(3)TS3/4.5.1.1-ECCS,ColdLegInjection i
The proposed change involves (1) the decrease of the operable range of water volume from 7853 and 8171 gallons to '/704 and 8004 gallons, and increase of the operable range of the nitrogen cover pressure from 385 and 481 psia to 585 and 678 psia for the ECCS cold leg accumulators. The changes are consistent with the LOCA analysis supporting the request for removal of the UH1 system.
(4)TS3/4.5.1.2-ECCS,UpperHeadInjection The Specifications associated with the maintenance of the UHI system are deleted.
(5)TS3/4.6.1and3/4.6.3-ContainmentPenetrationandValves The proposed changes would revise: (1) Table 3.6-1. secondary containment bypass leakage paths to reflect the sealing of the UHI related containment penetration, and (2) Table 3.6-2, containment isolation valves to reflect the removal of containment isolation valves associated with UHI containment penetration. Upon the removal of the UHI system, the UHI penetration will be cut and capped, and the associated piping and valves will be removed. The staff finds capping and sealing the unused containment penetration to be appropriate, and therefore, the associated TS changes are acceptable.
(6) TS 3/4.8.4 - Electrical Equioment Protection Devices Tables 3.8-1A and 3.8-1B will be revised to reflect the deletion of the UH1 system and related containment penetration conductor overcurrent protective devices.
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, The staff has reviewed the proposed Technical Specification changes and found them to be acceptable because they are consistent with the assumptions used in revised safety analyses which demonstrate that the appropriate accectance criteria are satisfied without taking credit for the UH1 system.
IV. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.32, the Comission has detemined that issuance of the amendments will have no significant impact on the environment (52 FR 39?05).
V.
CONCLUSION Notice of opportunity for a prior hearing was published in the Federal Register on August 27, 1987 (52 FR 3?365). No requests for a hearing were received. The Comission's Environmental Assessment and Finding of No i
Significant Impact was published on 0:tober 21,1987 (52 FR 39295). We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be en-dangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
VI. REFERENCES 1.
Letter, with five attachments, from H. Tucker (Duke Power Company) to NRC, dated June 12, 1987.
2.
Letter, with an attachment, from H. Tucker (Duke Power Company) to NRC, dated June 23, 1987.
3.
Letter, with an attachment, from H. Tucker (Duke Power Company) to NRC dated August 12, 1987 4.
LetterfromD. Hood (NRC)toH. Tucker (DukePowerCompany), dated May 13, 1986.
5.
" Westinghouse ECCS Evaluation Model,1981 Version," WCAP-9220-P-A, Rev.1, February 1982.
6.
" Westinghouse Revised PAD Code Thermal Safety Model,* WCAP-8720 Addendum 2, October 1982.
7.
- SATAN-VI Program:
Comprehensive Space Time Dependent Analysis of Loss of Coolant," WCAP-8170, June 1974 8.
" Calculational Model for Core Reflooding after a loss-of-Coolant Accident (WREFLOOD)," WCAP-8170, June 1974
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.. t 9.
Letter from C. E. Rossi (NRC) to E. P. Rahe, Jr. (Westinghouse).
" Acceptance for Referencing of Licensing Topical Report WCAP-10266, the 1981 Version of the Westinghouse ECCS Evaluation Model using the BASH Code," November 13, 1986.
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- 10. " BASH An Integrated Core and RCS Reflood Code for Analysis of PWR Loss-of-Coolant Accident," WCAP-10266, 1984
- 11. " BASH Methodology Enhancements " WCAP-10266, Addendum 2 March 1987.
- 13. "Long-Tenn Ice Condenser Containment LOTIC Code Supplement 1," WCAP-8355 (Supplement 1), June 1974
- 14. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code."
WCAP-1008), August 1985.
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- 15. "NOTRUMP, A Nodal Transient small Break and General Network Code,"
WCAP-10080, August 1985.
- 16. "LOFTRAN Code Description WCAP-7907, April 1984.
Principal Contributors:
Kahten N. Jabbour, PD!l-3 Sumer Sun, SRXB Chang-Yang Li, SPLB John D. Buchanan, PRPB Dated: November 10, 1987 4