ML19225D136
| ML19225D136 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/03/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19225D120 | List: |
| References | |
| TAC-08377, TAC-46962, TAC-8377, NUDOCS 7908070394 | |
| Download: ML19225D136 (7) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 20 TO FACILITY OPERATING LICENSE NO. DPR-72 FLORIDA POWER CORPORATION, ET AL CRYSTAL RIVER UNIT NO. 3 NiJCLEAR CENERATING PLANT DOCKET NO. 50-302 Introduction By application dated September 1,1977, as replaced by application dated May 25, 1979, the Florida Power Corporation (FPC or the licensee) proposed changes to Specification 3.6.2.2, Spray Additive Systen, of the Technical Specifications of Crystal River Unit 3 Nuclear Generating Plant (CR-3).
The licensee proposed a reduction in the acceptable concentration limits of sodium hydroxide in Tank BST-2 in the Chemical Additive System (CAS) of the CR-3 Reactor Building Spray System (RBSS). The licensee proposes to change the acceptable range of sodium hydroxide concentrations in Tank BST-2 from between 21.2 to 22.3 weight perdent to between 10.5 to 12.0 weight percent.
The range of acceptable values of the volume of sodiun hydroxide in Tank BST-2 is not being changed.
Condition 2.C.l4) of the CR-3 operating license required the licensee to isolate tne sodium thiosulf ate tank and its contents from the RBSS CAS until periaanent mooifications to the CAS were submitted to NRC for review and approval. This is discussed in Supplement No. 3 dated December 1976 to the Safety Evaluation (SE) dated July 1974. The isolation of Tank BST-1 from the CAS and the deletion of specifications on this tank were reviewed and approved in the SE dated Janu-a ry 4,1979.
The proposed reduction in sodium hydroxide concentration limits for Tank BST-2 implement the licensee's perinanent modifications to the CAS.
Evaluation We have reviewed and evaluated the cata provided by the licensee on the CR-3 RBSS in his letters dated September 1,1977 and May 25, 1979 By letter dated May 25, ID79, the licensee showed that the pH of the CR-3 RBSS injection spray and recir-culation (sump) spray water was between 7.9 and 11 and the potential consequences of the postulated loss of coolant accident (LOCA) were calculated to be less then the exposure guidelines of 10 CFR Part 100. The concentrations of sodium hydroxide in the CAS used in the licensee's calculations are the values pro-posed in the licensee's letter dated May 25, 1979.
427
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On the basis
- the data he has provided in his letter dated May 25, 1979, the licensee states that the CR-3 RBSS is adequate to assure acceptable spray water chemistry and potential consequences of the postulated LOCA which are less than the exposure guidelines of 10 CFR Part 100.
To show there is adequate assurance that sufficient sodium hydroxide will be added by gravity to the RBSS during a LOCA, the licensee has provided data from tests of the CAS. The licensee ran a series of measured drawdowns of the Borated Water Stor-age Tank, Tank BST-1 and Tank BST-2. These tests were made during the startup and test program of CR-3 in 1975 when Tank BST-1 was not isolated from the RBSS and containeu sodium thiosulfate.
Af ter these water tests were performed, four spring-loaded stop check valves were replaced with conventional spring check valves and the CAS was retested. The tests were run pumping water from the tanks into the fuel transfer canal. These tests were compared to calculations made by the GAI computer program " Thermal Hydraulic Analysis." Comparison of the results are given in refer-ences (1) and (2).
We conclude, based on these comparisons, that the computer pro-gram can accurately predict the performance of the CAS in the RBSS and can be used to determine the adequacy of the RBSS to naintain acceptable spray water pH.
Because the minimum pH of the RBSS spray water is less than 8.5, the evaluation of the CR-3 RBSS in Supplement No. 3 dated December 1976 of the SE dated July 1974 is no longer valid as it concerns (1) the RBSS spray water pH during the LOCA and (2) the potential consequences of the postulated LOCA. Our evaluation of the CR-3 RSSS during the LOCA concerning the above two items, based on the lat-est data on the CR-3 RBSS supplied by the licensee, is below.
In addition, the evaluation in Supplement No. 3 above did not include the potential consequences due to leakage from safeguards equipment outside containment.
This equipment cut-side containment, Decay Heat Removal and RBSS spray pumps and piping, circulates potentially highly radioactive water from the containment sump back to containment.
Leakage from these systems outside containment will contribute to the potential consequences of the postulated LOCA. The potential consequences given in this evaluation will include a contribution from this pathway.
The requirements on post-accident spray water chemistry are discussed in Standard Review Plan (SRP) 6.5.2.
The pH of the RBSS spray water during injection and recir-culation should be between 8.5 and 11.0.
As given in the Tables in the licensee's May 25,1979 letter, the pH of the RBSS spray during the LOCA is between 7.9 and 11.0.
This includes the case of failure of one of the two valves at Tank BST-2 and all of the sodium hydroxide from the tank entering one of two operating RBSS trains which results in the maximum pH in the spray.
For the most restrictive single active failure in the RBSS, in tems of the potential consequences of the postulated LOCA, the loss of a spray pump, the minimum pH of the RBSS spray is 8.1.
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The minimum pH of the RBSS spray water is below 8.5 because the maximura pH was not allowed to be greater than 11.0.
The requirement that the maximum spray water pH must be less than 11.0 is more important than the requirement that the minimum pH should be 8.5 or greater. Keeping the spray water pH no greater than 11 will prevent caustic stress corrosion cracking and degradation of the RBSS piping during a LOCA.
Keeping the spray water pH no less than 8.5 will in-hibit chloride stress corrosion cracking and degradation of the RBSS piping and will provide maximum spray effectiveness at preventing radiciocines released to the containment atmosphere during a LOCA from being released outside to the environment. Allowing the spray water pH to be less than 8.5 will reduce the ef fectiveness of the spray water to remove radioiodine from the containment atmosphere and to retain this radiciodine in the water. However, with the pH of the RBSS spray water above 7, the water will still inhibit chloride stress corrosion cracking in the RBSS piping and will still remove radioiodine from the containment atmosphere and retain it in the water. The effectivness of the spray to remove radioiodines from the contsinment atmosphere and retain it in the water and, thus, to reduce the potential consequences of the postulated LOCA increases rapidly with pH between 7 and 8.5.
We have calculated the potential consequences of the postulated LOCA at CR-3 with the proposed changes to the RBSS CAS. This is for a minimum RBSS spray water oH of 8.
The potential consequences and the assump-tions made to calculate them are given in Table 1 and Table 2, respectively.
The potential consequences of the postulated LOCA are well within the exposure guidelines of 10 CFR Part 100. The potential consequences include a contribu-tion due to leakage f rom safeguard equipment located outside containment. This contribution to potential consequences had not been included in previous evalua-tions of the postulated LOCA; however, the licensee does have specifications lim-iting the maximum acceptable leakage from this equipment outside contairment.
Corrpliance with Specification 4.5.2.e.5 (Decay Heat Removal System) and 4.6.2.1.b (RRSql provide assurance that the leakage rates assumed for the two systems during the postulated LOCA will not be exceeded. Because the potential consequences of the postulated LOCA are within the exposure guidelines of 10 CFR Part 100, the potential consequences are acceptable and the proposed Specification 3.6.2.2 is acceptable as written.
Moderator Dilution Based on an operating experience it had been determined that inadvertent injection of the contents of BST-2 into the reactor coolant system must be considered as a potential moderator dilution accident.
Currently procedural restrictions are imposed to preclude this accident while the issue is being re-viewed.
However, during review of the spray additive system discussed above, it was determined that the potential for boron dilution by this means existed at times during Mode 4 operation with the decay heat system lined up for recirculatinn.
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By letter dated July 3,1979, the licensee proposed a Technical Specification 1
change which would require a shutdnwn nargin of >2.2% ak/k instead of 10 ak/k during flode 4.
This requirement was also proposed for mode 5 for consistency of operation.
In support of this change the licensee presented an analysis which demonstrates that with this shutdown margin the operator has at least 15 minutes to take action before the reactor core could become critical.
This neets the current criteria in Standard Review Plan 15.4.6.
We have reviewed the licensee's analysis and proposed change, tc the Technical.
Spectrications and find them acceptable.
Environmental Considerations We have determined that the am ent does not authorize a change in e"?uent types, an increase in total aru s of effluents or an increase in power level and therefore will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration cnd environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We conclude on the basis of the above considerations that the proposed changes to Specification 3.6.2.2 of the CR-3 Technical Specificationsis acceptable as wri tten.
We also have concluded, based on the considerations above, that:
(l's because the amendment does not involve a significant increase in the probabi Sity or cor.-
sequences of accidents previously considered and does not involve a significant decrease in a safety margin, the 6mendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, ar.d (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health _nd safety of the public.
References:
- 1. Ely, R. F. Jr., " Borated Water Storace Tank Drawdown Transient Analysis, Revision 1," for Crystal River Unit 3 Nuclear Generating Plant, Florida Power Corporation, 30 January 1976.
- 2. Ely, R. F. Jr., " Hydraulic Analysis of Piping Networks Using PIPF Corputer Program," Topical Report GAI-TR-105, December,1976.
Dated: July 3,1979 427 107
5 TABLE 1 POTENTIAL OFFSITE DOSES OF THE POSTULATED LOSS-OF-COOLANT ACCIDENT Two Hour Course of Accidents Exclusion Boundary Low Population Zone (1340 Meters)
(8047 Meters)
Accident T hyroid Whole Body T hyroid Whole Body (Rem)
(Rem)
(Rem)
(Rem)
Loss of Coolant' Leakage thru containment 150.
3.0 27.
0.7 t
Leakage outside contai nment 4.8
.012 4.1
.004 I
i Post-LOCA I
Hydrogen Purge
<1
<1 Dose 427 l08
TABLE 2 ASSUMPTIONS FOR THE POSTULATED LOSS-OF-COOLANT ACCIDENT Hydrogen Purge Dose Analysis Using Regulatory Guide 1.7 assumptions, the licensee has calculated a hydrogen
~~
purge dose of approximately 0.1 Rem at the Low Population Zone. Our independent calculations are in substantial agreement with this incremental dose.
l.oss-of-Coolant Accident Regulatory Guide 1.4, Revision 2,1974 Power (MWt) 2544 3
6 Containment Vclume (f t )
2 x 10 6
Volume In Sump (lbs.)
3.8 x 10 Distribution of Radioiodines (%)
elemental 91 organic 4
particulate 5
Through Containment Leakage Design Containment Leak Rate (%/ day) 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.125 Spray fall height (f t) 96 Spray flow rate (gpm) 1500 Spray reduction limits elemental iodine 50 Partition Coefficient elemental iodine 1600'(pH=8)
Spray Removal rates (hrs )
elemental iodine 7.05 organic iodine 0.0 particulate iodine 0.45 Unsprayed region ( % )
25.
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(1 6
I II TABLE 2 (Cont'd)
T 1
1 Leakage Outside Containment (gph),
Reactor Building Spray System 12 Decay Heat Removat System 12 Charcoal Filter Efficiency (%)
elemental iodine 90 organi-iodine 70 part:culate iodine 90 Percent of lodine Released (%)
10 Start of Recirculation Af ter LOCA (Hr) 0.67 3
Atmospheric Dispersion Factors (sec/m )
-4 0
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (1340 meter *,)
2.2 x 10
-5 0
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (8047 meters) 1.0 x 10
-6 8
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.8 x 10
-6 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 2.8 x 10 24
-7 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 7.5 x 10 96 e
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El l10
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