ML20024H539
| ML20024H539 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/14/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20024H537 | List: |
| References | |
| NUDOCS 9106040411 | |
| Download: ML20024H539 (11) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO STEAM GENERATOR TUBE RUPTURE LICENSE NOS. NPF-35 AND NPF-52 DUKE POWER COMPANY CATAWBA NUCLEAR ST/ TION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414
1.0 BACKGROUND
Following the Ginna steam generator tube rupture (SGTR) event on January 25, 1982, the SGTR subgroup of the Westinghouse Owners Group (WOG) submitted WCAP-10698, "SGTR Analysis fiethodology to Determine the Margin to Steam l
Generator Overfill," dated December 1984 (Reference 1) for NRC staff review, which also references WCAP-10698, Supplement 1, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident."
In our evaluation of these WCAP's (Reference 2), the staff concluded that the WOG provided an acceptable and conservative methodolony for the generic SGTR analysis, but that five specific and crucial parameters and assumptions used in the analysis may vary significantly from plant to plant, altering the steam generator overfill and radiological dose results.
Duke Power Company chose to analyze the SGTR accident independently, with the key difference being the use of the RETRAN computer code for Catawba 1 and 2 as opposed to the LOFTRAN computer code as used by Westinghouse. The staff concluded that each member of the SGTR Subgroup and that all Westinghouse NT0L's (near-term operating licenses) were required to submit plant-specific information as follows before use of the methodolgy from WCAP-10698 could be applied on a plant-specific basis:
(1)
"Each utility in the SGTR subgroup must confirm that they have in place simulators and training programs which provides the required assurance that the necessary actions and times can be taken consistent with those l
assumed for the WCAP-10698 desian basis analysis. Demonstration runs l
should be performed to show that the accident can be mitigated within l
a period o' time compatible with overall prevention, using design basis assumptions regarding available equipment, and to demonstrate that the operator action times assumed in the analysis are realistic.
(2)
"A site specific SGTR radiation offsite consequence analysis which assumes the most severe failure identified in WCAP-10698, Supplement 1.
l The analysis should be perfomed using the methodology in SRP Section l
15.6.3, as supplemented by the guidance in Reference 3.
(3)
"An evaluation of the structural adequacy of the main steam lines and associated supports under water-filled conditions as a result of SGTR overfill.
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"A list of systems, components and instrumentation which are credited for accident mitigation in the plant specific SGTR E0P(s).
Specify whether each system and component specified is safety grade.
For primary and secondary PORVs and control valves specify the valve motive power and state whether the motive power and valve controls are safety grade. For non-safety grade systems and components state whether safety grade backups are available which can be expected to function or provide the desired information within a time period compatible with prevention of SGTR overfill or justify that non-safety grade components can be utilized for the design basis event.
Provide a list of all radiation monitors that could be utilized for identification of the accident and the ruptured steam generator and specify the quality and reliability of this instrumen-tation if possible.
If the E0Ps specify steam generator sampling as a means of ruptured SG identification, provide the expected time period for obtaining the sample results and discuss the effect on the duration of the accident.
(5) " A survey of plant primary and ' balance-of-plant' systems design to determine the compatibility with the bounding plant anaysis in WCAP-10698. Major design differences should be noted. The worst single failure should be identified if different from the WCAP-10698 analysis and the effect of the difference on the margin of overfill should be provided."
Duke Power Company has provided, via letters dated December 7,1987, and August 8 and 24, 1988, the plant-specific information required by the staff for the Catawba 1 and 2 plant, in accordance with license conditions 16 (Unit 1) and 10 (Unit 2) of license numbers NPF-35 and NPF-52.
License Condition 2.C(16) states:
' rior to startup, following the second refueling outage, Duke Power Company shall submit for NRC review and approval an analysis which demonstrates that the steam generator single-tube rupture analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses.
Consistent with the analytical assumptions, Duke Power Company shall propose any necessary changes to Appendix A to this license."
CatawbaUnit2LicenseCondition2.C(10)requiresactionsidenticaltothose required by License Condition 2.C(16) for Unit 1.
However, Unit 2 is required to complete these actions prior to its startup following the first refueling outage while Unit 1 is required to complate them prior to its startup following the second refueling' outage.
The following is our review of the SGTP accident analysis for the Catawba 1 and 2 Final Safety Analysis Re cr+ (FSAR), consistent with those five requirements previously set forth by the staff for application of the WOG generic methodology on a plNt-specific basis.
2.0 INTRODUCTION
Two different scenarios have been examined by the licensee for the SGTR accident analysis.
These are (1) the SGTR scenario conducive to steam generator (SG) overfill, and (2) the SGTR scenario resulting in a maximized of fsite dose based on a conservative analysis.
The significance of the consequence for the SG overfill scenario is the pos?ibility for an offsite radioactive release through the SG safety valves and the possibility of creating a main steam line break.
For the maximum offsite dose scenario, the concern is in exceeding regulatory of fsite dose limits as specified in 10 CFR 100.11.
Both scenarios have been evaluated under design-basis conditions using the methodology for SGIR in WCAp-10698 and WCAP-10698 Supplement 1 (Reference 1 and 3, respectively).
2.0 EVALUATION Or SG DVERTILL SCENARIO Catawba 1 and 2 are four-loop Westinghouse designed plants with SGs Model D3 type and SGs Model D5 type, respectively.
Fcr the design-basis SGTR most conducive toward overfill, the worst-case single failure is a failed-closed power-operated relief valve (PORV) on an intact SG. The failed-closed SG PORV is the most limiting because it extends the primary cooldown time, increating the opportunity for overfill of the fauited SG.
In the overfill scenario, the conservative assumption is that a loss of offsite power (loop) occurs with the reactor trip, The SGTR is nodeled as the double-ended, guillotine break of a single tube, consistent with SRP 15.6.3.
Primary flow of the break is maximized with the break location assumed above the tube sheet, at the SG tube bundle exit.
This break location results in higher primary to secondary leakage.
This is a conservative break location with respect to the margin to SG overfill.
Westinghouse performed a sensitivity analysis (Table 4.1-1, Reference 1) to identify the most limiting reference plant with respect to the time to overfill due to the relatively large tube diameter and the relatively small secondary steam volume associated with the SG.
The three loop plant type with a 12 foot core, high pressure safety injection system, and Model 51 stean generators was identified to be the most limiting plant type.
Duke Power Company did not perform a detailed SGTR overfill analysis. They compared the results and plant parameters of the reference plant with the Catawba 1 and 2 plant.
Catawba steam generators have smaller tube diameters and larger steam volumes than the reference plant. The reference plant inside tube diameter is 0.755, "while the Catawba inside tube diameter is 0.664."
The steam volume for the reference plant is 3658 cubic ft.
compared to 3848 cubic f t, for Catawba Unit 1 and 3883 cubic f t, for Catawba Unit 2.
Since the reference plant has adequate margin to overfill, Catawba 1 and 2 will have more margin to overfill.
In addition, Catawba Unit ? has lowered the full power normal operatina water level in the SGs, resulting in an even larger nargin to overfill.
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The generic analysis in Reference 1 is applicable to Catawba. The licensee has identified a few minor desegn changes in their plants, however, they would not affect the results of the analysis and are bound by the generic a naly si s.
The staff has reviewed the licensee's submittal and examined the generic analysis applicability and concludes that there would be an adequate margin to overfill for design-basis SGTR event for Catawba 1 and 2.
4.0 OPERATOR ACTION TIMES The operator action times relevant to the licensee's SGTR analysis are discussed in this section.
The staff's evaluation (Reference 2) of the Westinghouse Owners' Group WCAP-10698 stipulates plant-specific criteria for assessing operator action times in the event of an SGTR. Those criteria were employed to evaluate the information provided by Duke Power Company regarding operator action times for an SGTR at C6tawba Units 1 and 2.
The evaluation is based on the licensee's submittals dated December 7, 1987 (Reference 12), August 8, 1988 (Reference 13), August 24, 1988 (Reference 7), June 29, 1990 (Reference 14),
and August 20, 1990 (Reference 15).
The staff's evaluation by criterion follows:
Criterion 1.
Provide simulator and emergency operating procedure training related to a potential SGTR.
The licensee documented by letter dated June 29, 1990 that onsite simulator and E0P training relevant to an SGTR are provided.
The staff finds that the licensee has satisfied Criterion 1.
Criterion 2.
Ensure that the plant-specific operator response times are consistent with those assumed in WCAP-10698.
Duke Power Company did not record specific operator action times during demonstration runs regarding an SGTR.
Instead, the licensee documented by checklist whether the results of demonstration runs were bot nd by the operator action times assuned in WCAP-10698.
By letters dated December 7, 1987, August 8, 1988 and August 20, 1990, the licensee provide / the results of the plant-specific demonstration runs for an SGTR. The results are summarized below.
OPERATOR ACTION WCCAP 10698 TIMES (minutes)
DEMONSTRATED o
Identify and isolate 10*
met ruptured steam generator o
Initiate cooldown after 5
met isolation
o initiate depressurization 2
met after cooldown completed o
Initiate safety injection 1
(SI) termination
- Ten minutes after reactor trip or the time prior to reaching 330 narrow range level in the ruptured steam generator (whichever is greater).
- The licensee's August 8, 1988, letter clarified that this step was not included in the demonstration run because (1) the Catawba emergency procedures use a simultaneous cooldown and depressurization rather than a sequential method assumed in Table 2.3-1 of WCAp-10698 in the Westinghouse Owners' Group generic onalysis and (2) as long as pressure equilibrium is maintained across the break, resulting from depressurizing the RCS to the ruptured SG pressure, no overfill consequences are associated with 51 termination.
Sased on the information and explanations discussed above, the staff finds that the licensee has satisfied this criterion.
Criterion 3.
Utilizing typical control room staff as participants in demonstration runs, show that the operator action times assumed in the SGTR analysis are realistic.
The licensee's August 8,1988 submittel irdicated that three demonstration runs were completed involving a total of 14 operators.
Tnose operators represented 20* of operators at the plant in June 1988.
During the conference call of July 9, 1990, the staff raised a concern that the number of operators participating in the simulator runs were too few to demonstrate that the results are realistic for Catawba operators.
By letter dated August 20, 1990, the licensee responded to the staff's concern and committed to complete additional demonstration runs representing at least 80i of its licensed operators by August 30, 1991. This y
is a confirmatory issue.
Baseo on the results of completed demonstration runs and the licensee's commitment, the staff concludes that this criterion has been satisfactorily addressed.
Criterian 4 Coinplete demonstration runs to show that the postulated SGTR occident can be mitigated within a perico of time compatible with overfill prevention, using design basis assumptions regarding available equipment.
As notec under Criterion 2, the licensee's demonstrated times were bounded by times assumed in WCAP-10698, and the staff finds that the licensee has satisfied Criterion 4
Criterion 5 If the emergency operating procedures (E0Ps) specify SG t
sampling as e means of identifying the SG with the ruptured tube, provide the expected time period for obtaining the sample results and discuss the effect on the duration of the accident.
The licensee's December 7, 1487 letter indicates that the E0Ps for Catawba Units 1 and 2 do not specif3 SG sampling as a means of identifying a ruptured SG.
Instead, steam line monitors and SG level are the means of identifying a ruptured SG.
Based on this information, the staff finds Criterion 5 is satisfied.
Summary The staf+ has reviewed Duke Power Company's responses regarding operator action times for an SGTR, concluding that the licensee's demonstrated times for Catawba operators, and commitment to conduct additional demonstration runs, are satisfactory.
5.0 EVALU
TION-0F aCENARIO FOR MAXIMUM.0FFSITE RADIOLOGICAL CONSEOUENCES Duke Power Comaany performed a conservativa analysis of the postulated steam generator tube rupture assuming the loss of offsite power at reactor trip.
The licensee (DPC) has used the Standard Technical Specifications (STSs) for concentrations of radioactivity in primary and secondary coolant at Catawba Units 1 anc 2.
These STSs are:
nominal 1.0 and 0.1 microcuries dose equivalent I-131 (DE I-131) per gram in primary and secondary coolant water, respectively, with a shutdown limit in the event of an " iodine spike" (Reference 4) of 60 microcuries per gram in primery coolant.
DPC has calculated the consequences of a postulated steam generator tube rupture (SGTR) accicent which could release radiciodine with coolant water to the atmosphere.
The licensee's calculated doses using the costulated accident scenario and conservative assumotions as described in SRP 15.6.3, Revision 1, meet the staff's guideline values for acceptance of the STS's at a particular site (Reference 9 and Table 1.)
The staff has independently estimated the limiting thyroid doses at the exclusion area boundary that would result from the atmospheric releases.
Under the assumptions in the SRP, the dose due to the release of primary coolant directly to the atmosphere results in about 80% of the total dose; the remainder results from the release of secondary coolant.
For it itsmeteorologicaldispersionvalueof3.8x10gcalcu}ations,thestaffused sec/m (Reference 10), the ICRP No. 30 thyroid dose factor for I-131 (Reference 11), the licensee's estimated release of 195,000 pounds of primary coolant to the atmosphere over a perica of 45 minutes during a design basis SGTR accident (Reference 9), and the licensee's estimate of an average flashing fraction of 0.16 (of water to steam in the atmosphere).
In accordance with the SRP, two cases were assumed:
a coincident iodine spike and a pre-existing iodine spike.
The staff's results for these two cases are also shown in Table 1.
In either case, the staff finds that the resulting doses are within NPC quideline values and that, therefore, the use of the Standard Technical Specifications for radioactivity in the trirarv anc seccrcary coolant at Catawba Units I and 7 are acceptable.
7 6.0 RETRAN0?/ MOD 00A COMPUTER CODE The licensee used the RETRAN0?/ MOD 004 to analyze the SGTR event as opoosed to the LOFTRAN computer code as used by Westinghouse for the reference plant analysis (Reference 1).
This version of RETRAN has been approved by the NRC for generic licensing applications (Reference 8).
The licensee used the Catawba Unit 2 RETRAN model based on its similarity to the post reactor trip level behavior of the referenced plant.
This behavior results in later identification of the ruptured SG and therefore a longer duration for primary-to-secondary leakage.
The base plant model was reinitialized to appropriate conditions to maximize heat input ari time required for cooldown.
The licensee performed a conservative SGTR analysis using an approved version of RETRAN02/ MOD 004 and we find it acceptable as applied in this licensina application.
7.0 EVALUATION OF WCAP-1069R (WOG GENERIC SGTR ANALYSISI. APPLICABILITY TO CATAWBA 1 AND P The five issues discussed in Section 1.0 of this report have been addressed by the licensee.
The staff has evaluated the licenste's responses to these issues as discussed below:
(1) Duke Power Company confirmed that Catawba 1 and 2 have placed the simulator and training programs necessary to assure that operator actions and operator action times can be conducted under design-basis conditions consistently with those actions and tines assumed in the Catawba 1 and ?
SGTR accident analysis.
Demonstration runs have been performed by the licensee to show that the ar..ent can be mitigated within a period of time comoatible with SG overfill prevention using design-basis assumotions.
The licensee postulated an SGTR scenario to train and retrain the group of operators.
Tne scenario included a reactor trip, manual safety iniection due to the SGTR, a total loss of offsite power and the failed PORV on "A" SG.
The licensee outlined a set of criteria for the critical action times and the operators met those criteria during their + raining.
The staff concludes that the operator action times from the demonstration runs already performed are close enough to the times asnam d in the analysis and that enough conservatism has been incorporated into the overall SGTR analysis to assure the successful mitigation of an 53TR accident. We find their training programs acceptable.
(2) Duke Power Company performed a site-soecific SGTR radiation off-site consequences analysis assuming the worst-case single failure applicable to the Catawba 1 and 2 plant, using the methodology in SRP, Section 15.6.3.
This analysis is discussed in Section 5.0 of this SER.
(3) The licensee performed a structural stress analysis for main steam lines with all supports active under water-filled conditions. The licensee used the AMSE Code, Section 3 for Class 2 piping analysis approach and found all pipe stresses to be within the ASME Code allowables. The staff finds the evaluation of structural adequacy of the main steam lines under water-filled conditions to be acceptable.
(4) Duke Power Company has provided a list of systems, components, and instrumentation which are credited for accident mitigation in the plant-specific SGTR Emergency procedures. Most of the principal equipment is safety grade except the MFW and AFW flow control valves and the isolation valves for the steam supply to the AFW pump turbine. The backup equipment in each case is safety grade. The isolation function of the MFW control valves is safety grade. The radiation monitors are high quality instrumentatior. i t are not safety grade. Main steam line radiation monitors could be used to identify the ruptured steam generator on the appropriate Catawba Units. The normal motive power for the primary and secondary PORVs is the instrument air system which is not safety grade.
Safety grade backup eir motive power is available for two primary PORVs and all secondary PORVs.
The primary and secondary PORV block valves are electric motor operated and are supplied with emergency power from the diesel generators. The primary and ser.ondary PORVs and block valves all have safety grade control power.
The staff has reviewed the information provided and has concluded that the licensee does in f act credit mostly safety-grade equipment for identification and mitigation of the SGTR event.
Verification of the ability to mitigate the event in accordance with the assumptions in the accident analysis is incorporated into the procedures for the SGTR simulator runs.
Because the results of the simulator runs will be consistent with the accident analysis, with mostly safety-grade equipment available to the operator from the time of accident initiation, the staff is reasonably assured of SGTR identification within the times assumed in the analysis (consistent with item (1)) without the availability of condenser air ejector radiation monitors. The staff finds the list of equipment used for SGTR mitigation to be acceptable.
(5) The compatibility of the Catawba 1 and 2 SGTR analysis with the bounding plant analysis (Reference 1) has been determined by the licensee and incorporated into the plant-specific analysis. The generic analysis in Reference 1 is applicable to Catawba 1 & 2.
The licensee has identified a few design variations with respect to the reference plant which do not affect the results of SGTR analysis.
Since the Catawba 1 and 2 plant has adequate margin to overfill, we conclude that the design variations would have no impact on the SGTR event.
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-9 8.0 CONTL US ION _S Based on the information provided by the licensee, we find that Duke Power Company has met license conditions 16 (Unit 1) and 10 (Unit ?) of license numbers NPF-35 and NPF-5? respectively.
The methodology and analysis for the Catawba 1 and 2 SGTR accident analysis are acceptable subject to completion of the commitment included in the licensee's August 20, 1990 letter regarding additional demonstration runs for operator response times.
The licensee anticipated completion of this commitment by August 30, 1991.
9.0 REFERENCES
1.
WCAP-10698, "SGTP Analysis Methodology to Determine the Margin to Steam Generator Overfill," dated December 1984 2.
Letter from Charles E. Rossi (NRC) to Allan E. Ladieu (Westinghouse Owners Group, SGTR Subgroup), " Acceptance for Referencing of License Tooical Report WCAP-10698 'SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,' December 1984," dated March 30, 1987 3.
WCAP-10698, Supplement 1, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," dated May 1985.
4 NUREG-0800, Standard Review Plan, Section 15.6.3., " Radiological Consequences of Steam Generator Tube Failure (PWR)," Rev. 2, dated July 1981.
5.
"WOG Emeroency Response Guidelines (ERGS)," Rev lA, dated Jul_y 1987 6.
Letter from H. Berkow (NRC) to A. Ladieu (Westinghouse Owners Group, SGTR Subgroup), "Accectance for Referencing of Licensing Topical Report WCAP-10698 Supplement 1, Evaluation of Offsite Radiation Doses for a SGTR Accident," dated December 17, 1985.
7 Letter from H. B, Tucker (Duke Power Company) to USNRC, " Catawba l
Nuclear Station, Units 1 and ?, Docket No.
50-413/a14. License No.
NPF-35/5?, Steam Generator Tube Rupture Analysis," dated August 24, 1988 8.
Letter from A. C. Thadani (NRC) to R. Furia (GPU Nuclear Corporation),
" Acceptance for Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions ?
and 3 Regarding RETRANOP/M0DC03 anc M00004," dated October 19, 1988.
9.
Letter from H. B. Tucker. Duke Power Company, to USNRC, " Catawba Nuclear Station, Units 1 and 2, Steam Generator Tube Rupture Analysis," dated December 8, 1989.
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Safetv Evaluation Report related to the operation of Catawba Nuclear Station, Units 1 and 2, NUREG-0954, dated February 1983, 11.
ICRP Publication 30, limits for intakes of Radioactivity by Workers, International Comission on Radiological Protection, Pergamon Press, Oxford, U.
K., dated July 1978.
12.
Letter from H. B. Tucker (DPC) to USNRC, " Steam Generator Tube Rupture Analysis," dated December 7, 1987, 13.
Letter from H. B. Tucker (DPC) to USNRC, " Steam Generator Tube Rupture (SGTR) Analysis," dated August 8, 1988.
14 Letter from H. B. Tucker (DPC) to USNRC, " Steam Generator Tube Rupture,"
dated June 29, 1990, 15.
Letter from 4. B. Tucker (DPC) to USNRC, Steam Generator Tube Rupture Analysis," dated August 20, 1990.
Frincipal Centributors:
K. Desai J. Martin G. West Dated: May 14, 1991
Table 1 l
STEAM GENERATOR TllBE RUPTURE ACCIDEf4T THYR 010 DOSES (REM)
Licensee NRC Staff Acceptance At EAB At LPZ At EAB Criterion With pre-existing 91 3
127 300 lodine spike With coincident 26 1
23 30 lodine spike Notes:
1.
Two hour dose at EAB (Exclusion Area Boundary).
2.
Course of accident dose at LPZ (Low Population Zone) boundary.
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May 14, 1991 D_is t ribution L
Docket File NRC PDR local POR PDII-3 R/F S. Varga G. Lainas B. Clayton R. Martin OGC E. Jordan ACRS (10)
L. Reyes, Ril 1
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