ML19329A233
| ML19329A233 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/16/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329A232 | List: |
| References | |
| NUDOCS 7912300286 | |
| Download: ML19329A233 (10) | |
Text
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4 SAFETY EVALUATION REPORT OCONEE UNIT 1 DOCKET No. 50-269 O
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' SAFETY EVALUATION REPORT OCONEE UNIT 1 DOCKET No. 50-269 TABLE OF CONTENTS Page
1.0 INTRODUCTION
1 2.0 ECCS RE-EVALUATION.............................
1 2.1 Single Failure Criterion.................
4 2.2 Containment Pressure.....................
5 2.3 Long-Term Boron Concentration............
5 2.4 Submerged Valves.........................
6 2.5 Partial Loop Analyses....................
6 3.0 TECllNICAL SPECIFICATIONS.......................
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4.0 CONCLUSION
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5.0 REFERENCES
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1 I
1.0 INTRODUCTION
d the Atomic Energy Commis'sio issued an Or er for Modification of License (Reference 2) implementing the require-On December 27, 1974, Core ments of 10 CFR 50.46, " Acceptance Criteria and Emergency One of Cooling Systems for Light Water Nuclear Power Reactors."
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the requirements of the Order was that the licensce s a d
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a re-evaluation of ECCS cooling performance ca i
i The Order also required that the evaluation shall be accompanied by such proposed changes in Technical Specificat of 10 CFR 50.46.
h luation or license amendment as may be necessary to implement t e eva 27, 1974, Duke Power As required by our Order of December Company (the licensee) has submitted an ECCS re-evaluation a results.
related Technical Specifications.
Specifications were submitted in References 1 Also discussed in Section 2.0 of this Safety Evaluation Report.Section 2.0 a j
fic il areas of single failures, long-term boron concent l k and the containment pressure calculation.
l results of the staff review of the proposed Oco and references, respectively.
2.0 ECCS RE-EVALUATION l
The background of the staff review of the B&W ECCS evaluation ff SER for this and its application to Oconee is described in the staissued in connection w f acility dated December 27, 1974, The bases for acceptance of the Order for Modification of License.
in the principal portions of the evaluation model are se Supplement hich are to the Status Report of November 1974 (Reference 6) w 27, 1974 SER. The December 27, 1974 i
referenced in the DecemberSER also describes the various cha Together, the December 27, 1974 SER and the of the B&W model. Status Report and its Supplement describe an ac f the model.
aluation model and the basis for the staff's accep l
The aluation report properly conforms to the accepted mode.
licensee's July 9,1975 submittal (Reference 1) contains docu by reference to B&W Topical Reports of the revised ECCS m 27,'1974 SER)
(with the modifications described in our December (Reference 7 and a generic break spectrum appropriate to Oconee 1 included in and d, respectively). In addition, Duke Power Company break for this July 9th submittal a separate analysis of the worst I
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Oconee Unit 1, using the following plant-specific parameters:
Power" level = 1.02 x 2568 Mwt.
The generic analyses in BAW-10103 a.
used 1.02 x 2772 Mwt.
b.
Initial average fuel temperature assumed reflects the reload core (T = 3030*F for 18 kw/ft with 580*F sink temperature).
The generic analyses used'T = 3050*F.
Different pin dimensions were employed to reflect fuel changes, c.
d.
Core flood tank line resistance was changed to reflect the as-built value for Oconee Unit 1 (6.5 versus 7.75 in generic analyses).
System enthalpies and steam generator heat loads were changed e.
to. reflect the lower power level of 2568 Mwt.'
j f.
Initial pin pressures and oxide layer thicknesses were changed to reflect the different fuel in Oconee 1.
The generic analysis in BAW-10103 identified the worst break size as the 8.55 ft2 double-ended cold leg break at the pump discharge
=
e e w a e summarizes the results of the with a C j
D LOCA limit analyses which determine the allow'able linear heat rate l
limits as a function of elevation in the core for Oconec Unit 1:
Elevation LOCA Peak Cladding Max. Local Time of (ft)
Limit Temperature (*F)
Oxidation Rupture (kw/ft)
Ruptured Unruptured
(%)
(sec)
Node Node Oconee 1 2
16.0 1882 1930 3.40 10.90 4.
17.5 1975 1978 3.17 12.39 6
18.0 2066 2146 5.46 15.55 8,
17.0 1743 2110
. 5.19 15.01 10*
-16.0 1642 1931 2.93 39.20 i
- See discussion in text.
1 The maximum core-wide metal-water reaction for Oconee 1 was calculated to be 0.557 percent, a value which is below the allowable limit of 1 percent.
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As shown in the tabulation, the calculated values for the peak clad temperature and local metal-water reaction were below the allowable limits specified in 10 CFR 50.46 of 2200*F and 17 percent, respectively. BAW-10103 has also shown that the core geometry remains amenable to cooling and that long-term core cooling can be established.
The staff noted during its review of BAW-10103 that the LOCA limit calculation at the 10-foot elevation in the core showed reflood rates below 1 inch /second at 251 seconds into the accident (Section 7.2.5).
Appendix K to 10 CFR 50.46 requires that when reflood rates are less than 1 inch /second, heat transfer calculations shall be based on the assumption that cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling or rupture as such blockage might affect both local steam flow and heat transfer. As indicated by the staff in References 5 and 6, a steam cooling model for reflood rates less than 1 inch /second was not submitted by B&W for staff review. The steam cooling model submitted by B&W in BAW-10103 is therefore con-sidered to be a proposed model change requiring further staff review and ACRS consideration. Accordingly, B&W was informed that until the proposed steam cooling model is reviewed, the heat transfer calculation at the 10-foot elevation during the period of steam cooling specified in BAW-10103 must be further justified.
In lieu of using their proposed steam cooling model, B&W has submitted the results of calculations at the 10-foot elevation using adiabatic heatup during the steam cooling period, where this period is defined by B&W as the time when the reflood rate first goes below 1 inch /
second to the., time that REFLOOD predicts the 10-foot elevation is covered by solid water.
The new calculated peak cladding temperature, local metal-water reaction and core-wide metal-water reaction at the 10-foot elevation are 1946*F, 3.02%, and.647%, respectively. These values remain below the allowable limits of 10 CFR 50.46 and are acceptable to the staff.
Until a steam cooling model has been accepted by the staff, these values will serve as the LOCA results for Oconee 1 at the 10-foot elevation.
As indicated in a previous paragraph, Duke Power Company elected to provide a plant-specific calculation for Oconee Unit 1 utilizing selected as-built data. We have reviewed the input changes used (relative to BAW-10103) and believe them appropriate for Oconec Unit 1.
Our rev.ew of other plant-specific assumptions discussed in the following paragraphs regarding the Oconee 1 adalyses addressed the areas of single failure criterion, long-term boron concentration, potential submerged equipment, partial loop operation, ECCS valve interlocks, and the containment pressure calculation.
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2.1 Single Failure Criterion Appendix K' to 10 CFR 50 of the Commission's regulations requires that the combination of ECCS subsystems to be assumed ope' ative r
shall be those available af ter the most damaging single failure Babcock and Wilcox has assumed of ECCS equipment has occurred.
all containment cooling systems operating to minimize containment pressure.and has separately assumed the loss of one diesel to We concluded in Reference 5 that the minimize ECCS cooling.
application of the single failure criterion was to be confirmed during subsequent plant reviews.
A review of Oconee' 1' piping and inctrumentation diagrams indicated that the spurious actuation of certain motor-operated valves could A spurious affect the appropriate single. failure assumptions.
actuation of core flooding tank (CFT) vent valves CF-5 or CF-6 The rate at which this 4
would result in a decrease in CFT pressure.
decrease occurs is controlled by a preset needle throttling valve downstream of the electrically-operated valve (CF-5 or CF-6).
The predetermined position of the needle valve is provided by manually turning the local handwheel such an amount as to limit the rate at A recent which a depressurization of the CFT could take place.
test at Oconee indicated that the tested valve setting allowed 17 minutes for the CFT pressure to decay from 625. psi to the low pressure alarm, 580 psi,when the electrically-operated valves were opened._ Since it is clear that CFT pressure is important to mitigating the consequences of a LOCA, a Station Technical Specification must be adopted, either for the position of the menual throttling valves or for the moto -operated valves. Since presetting the throttling valve by turning the' handwheel an amount equal to the aforementioned depressurization rate does not appear to be sufficiently accurate to serve as a safeguard against an uncontrolled CFT blowdown, we will require that the normally closed motor-operated
-valves CF-5 and CF-6 have their power disconnected and their associated breakers locked open.
- A. review was also conducted of the electrical schematics for ECCS It was determined that a single failure of motor-operated valves.
valve interlocks 'could not affect the appropriate single failure assumptions.
To further. minimize the potential for a water. hammer due to the discharge of ECC water into a dry line, the. staff requires that valves LP-21, LP-22,- HP-24 and HP-25 be lef t-in the open position This maintains the ECCS lines' filled with during normal operation.
-a continual s%**eJ%&L,
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s supply of water from the BWST due to.the available static head Such a configuration will also eliminate built into the design..
A the need for one automatic safety action in the event of a LOC ;
that is,- the automatic opening of these valves to provideIn addition, Duke water to the ECCS and Building Spray pumps.
Power Company will be required to adopt a Technical Specification whereby a monthly venting of ECCS pump casings and high points in ECCS lines will.be performed to ensure that no air pockets have Such venting must also be performed prior to any ECCS formed.
flow tests.
2.2 Containment Pressure The ECCS' containment pressure calcu,lations for Oconee Class plants were performed generically by B&W for reactors of this type asT g
described in Reference 8.
5 and model and published the results of this review in Refe
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We ret,cired that justification of the plant-6.
for ECCS evaluations.
dependent input parameters used in the containment analyses be A containment pressure submitted for our review of each plant.
calculation specific to Oconee 1 was submitted in Reference 1.
Justification for the containment input data was submitted forThis justif 10, 1975 (Reference 11).
Oconee 1 on Octoberallows comparison of the actual containment pa Duke Power Company has with those assumed in References 1 and 8. evaluated th d
and operation of the containment heat-removal systems with regar This evaluation was to the conservatism for the ECCS analysis.The containment heat removal based on as-built design information.
systems wcre assumed to operate at their maximum capacities, a minimum operation values for the spray water and service water The containment pressure analysis was temperatures were assumed.
demonstrated to be conservative for Oconee Unit 1.
We.have concluded - tliat the plant-dependent information used f ECCS containment pressure analysis for Oconee 1 is reasonably conservative and, therefore, the calculated cont
' regulations.
Long-Term Boron Cc,ncentrction 2.3 The NRC staff ;has reviewed. the proposed procedures-and the system Ldesigned for preventing. excessive boric acid buildups in the reacto Duke
= vessel during the;1ong-term. cooling period a h
1would allow adequate boron dilution during t e These procedures will comply with: the single failure criterion'.'
h ept described -
= will employ a hot' leg drain network similar to t e conc
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4
4 in BAW-10 s.
To crpl:;y c cingl2 fcilure ', af moda Duka P vsr to th2 existing DilR dcsign during Comp:ny will mak3 modifictticn2The proposal consists of the additien the next refueling outage.
One of two drain lines from the decay heat drop line to the sump.
line(installed upstream of the DllR isolation valves)[will include two qualified motor-operated valves.,The other line installed downstream of the DilR isolation valves) will include one qualified The Licensee will be required to install a motor-operated valve.
flow rate measuring device to confirm that a minimum of 40 gpm are With available following a LOCA, and to facilitate system tests.
the addition of this fbw device, this proposal is acceptable to A final review of. the installed system will be conducted the staff.
before startup.
-2.4 Submerged Valves The applicant has conducted a review of equipment arrangement to determine if any valve motors inside the containment will become submerged following a LOCA.
Based on this review, no valves were identified which would be flooded and which would affect short-term or long-term ECCS functions or containment isolation.
t 2.5 Partial Loop Analyses To. allow an operating configuration with less than four reactor coolant p tmps on the line (partial loop), the staff requires an analysis of the predicted consequences of a LOCA occurring during the proposed partial loop operating mode (s).
Duke Power Company submitted an analysis for partial loop operation with one idle reactor coolant pump (three pumps operating) in Reference 9.
Using a reduced power icvel of 77% of rated power, B&W performed this analysis 2
assuminb the worst-case break (8.55 ft DE, CD = 1) and maximum LIIGR (18.0 kw/f t) from the 4-pump analysis discussed in Section 2.0. The worst break selected was located in the active leg of the partially idle loop.
Placing the break at the discharge of the pump in an active cold leg of the partially idle loop (instead of at the discharge of,the pump in an active cold leg of the fully active loop) yicids the most degraded positive flow through the core during the first half of the blowdown and results in higher cladding temperatures. The maximum cladding temperature for the one-idle-pump mode of operation was 1766*F. A staff review'of all input assumptions and conclusions resulted in a set of inquiries which were answered in References 4 and 10.
The results of a new analysis was submitted to reflect a more appropriate value of initial pin pressure. The original partial loop analysis in Reference 9 used an initial pin pressure of 1600 psi. As was demonstrated in the time-in-life sensitivity study in Reference 8, the worst pin pressure for this analysis should have been 760 psi.
The maximum cladding temperature for the re-analysis is 1784*F, a value which is within the criterion of l
10 CFR 50.46. Therefore, this analysis may be used to support Duke Power Company's proposed operation with one idle reactor coolant pump.
Since an' analysis of ECCS cooling performance with one idle reactor coolant pump in each loop has not been submitted, power operation in this configuration must be limited by Technical Specifications to
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-24 hours.
-le loop operation (i.e., operat, ion.with two idle pumps in one L.
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notifying the Commission. Each proposal for n scheduled single loop test will be considered on a case-by-case basis.
3.0 TECHNICAL SPECIFICATIONS T
We have reviewed the Technical Specifications proposed by Duke Power Company.to assure that operation of Oconee 1 is within the limits imposed by the Final Acceptance Criteria. These criteria permit an increase in the allowable heat generation rate from 15 to 16 kw/f t at the 10-foot elevation compared to the Interim Acceptance Criteria currently in effect.- For Unit 1, the LOCA-
. related heat generation limits occur.in the reload fuel (Batch 4).
The limits for this fuel are not changed from those now in effect for the bottom half of the core.
Although the proposed rod insertion limits for Unit 1 are not changed from those which are currently in effect, the burnup at which Group 7 withdrawal begins van modified from 250 1 10 EFPD to 245 + 10 EFFD.
This has made necessary a reduction in the allowabic positive axial imbalance from 14% to 10% at full power. Ue find the revised Technical Specification (See Reference 4) to be acceptable.
4.0 CONCUUSIONS The staff has completed its review of the Oconee 1 ECCS performance re-analyses and has concluded:
1 The proposed Technical Specificationsare based on a LOCA E
a.
analysis performed in accordance with Appendix K to 10 CFR 50.
l b.
The ECCS minimum containment pressure calculations were performed in accordance with Appendix K to 10 CFR 50.
l The singic failure criterion will be satisfied provided that c.
the modifications specified in subsection 2.1 of this Safety Evaluation Report are implemented.
d.
The proposed procedures for long-term cooling af ter a LOCA are acceptab' to the staff.
The impicmentation of these procedures d.ing the next refueling outage is required to l
provide assurance that.the ECCS can be operated in a manner which would prevent excessive boric acid concentration from occurring.
The proposed mode of reactor operation with one idle reactor e.
coolant pump is supported by a LOCA analysis performed in accordance with Appendix I to 10 CFR 50.
Operation with one idle pump in each loop is restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Requests for single loop operation will be reviewcd on a case-by-case basis.
l a
5.0 REFERENCES
Letter from William O. Parker, Jr. to Mr. Angelo Giambusso 1.
dated July 9, 1975.
" Order for Modification of License, " Letter from Robert A. Purple 2.
'to Mr. Austin C. Thies dated December 27, 1974.
Letter from William O. Parker, Jr. to Mr. Benard C. Rusche 3.
dated November 10, 1975.
- Letter f rom William O. Parker, Jr. to Mr. Benard C. Rusche 4.
dated October 31, 1975.
5.
" Status Report by.the Directorate of Licensing in the Matter of Babcock and Wilcox ECCS Evaluation Model Conformance to 10 CFR 50, Appendix k," dated October 1974.
6.
" Supplement 1 to the Status Report by the Directorate of Licensing in the Matter of Babcock and Wilcox ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K," dated November 13, 1974.
B. M. Dunn, et al., "B&W's ECCS Evaluation Model," BAW-10104, 7.
Babcock and Wilcox, :by 1975.
8.
R. C. Jones, et al., "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS," BAW-10103, Babcock and Wilcox, June 1975.
9.
Letter from William O. Parker, Jr. to Mr. Angelo Giambusso dated August 1, 1975.
10.
Letter from Kenneth E. Suhrke to Mr. A. Schwencer dated December 15, 1975.
11.
Letter from William O. Parker, Jr. to Mr. Benard C..Rusche dated October 10, 1975.
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