ML20136E890
| ML20136E890 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/10/1997 |
| From: | Polich T NRC (Affiliation Not Assigned) |
| To: | Terry C TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| Shared Package | |
| ML20136E894 | List: |
| References | |
| GL-88-20, TAC-M74397, TAC-M88982, NUDOCS 9703130361 | |
| Download: ML20136E890 (79) | |
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.r Mr. C. Lance Terry.
i Based on the " Step 1" review, staff conclude that the CPSES IPE has met the intent of Generic Letter 88-20. A more detailed review, a " Step 2" review will not be conducted by the NRC staff.
Sincerely, ORIGINAL SICNED BY:
Timothy J. Polich, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation t
I Docket Nos 445 and 50-446
Enclosure:
Staff Evaluation Report l
w/ attachment L
l cc w/ encl:
See next page l-t DISTRIBUTION:
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AGody SVarga ERodrick, RES MWHodges, RES Document Name:
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l DATE-3 //b /97 3 /'7 /97 l
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UNITED STATES i
r NUCLEAR REGULATORY COMMISSION i
WASHINGTON D.C. 30006 4001 l
March 10,1997 l
Mr. C. Lance Terry i
TU Electric i
Group Vice President, Nuclear Attn: Regulatory Affairs Department P. O. Box 1002 l
Glen Rose, TX 76043
SUBJECT:
REVIEW 0F COMANCHE PEAK STEAM ELECTRIC STATION INDIVIDUAL PLANT EXAMINATION SUBNITTAL - INTERNAL EVENTS (TAC NOS. M74397 AND M88982) 1
Dear Mr. Terry:
i Enclosed is our Staff Evaluation Report (SER) of the Comanche Peak Steam flectric Station (CPSES) Individual Plant Examination (IPE) submittal for i
internal events and internal flood.
A " Step 1" review was performed which examined the IPE results for their
" reasonableness" considering the design and operation of CPSES. The staff employed Brookhaven National Laboratory to review the front-end and the back-and analyses of the IPE submittal, and Sandia National Laboratory to review i
the human reliability analysis (HRA). The contractor's Technical Evaluation i
Report (TER) is attached as Appendix A to the SER. This contractor TER was reviewed by the IPE " Senior Review Board" (SRB) as part of the Office of Nuclear Regulatory Research (RES), quality assurance process. The SRB is d
comprised of RES staff and consultants at Sandia and Brookhaven National l
Laboratories with Probabilistic Risk Assessment (PRA) expertise. The Nuclear i
Regulatory Commission (NRC) project manager and senior resident inspector for l
CPSES also attended the SRB meeting.
Your IFE submittal estimated the total core damage frequency (C0F) for the CPSES as 5.7E-5/ reactor-year for internally initiated events, including internal flooding. The CPSES CDF compares reasonably with that of other Westinghouse plants. Transients contribute 54% (includes loss of offsite Power / station blackout 28% and anticipated transients without scram, 9%),
j internal floods 23%, loss of coolant accidents (LOCA) 17%, steam generator tube rupture (SGTR) 6%, and interfacing systems LOCA <1%.
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pump seal LOCAs appear in sequences that contribute 29% to the CDF, a majority of which stem from station blackout (SBO).
Your submittal indicated that the results of the CPSES IPE were evaluated using the reporting criteria from Generic Letter (GL) 88-20, followed by a qualitative analysis. Since the CPSES IPE core damage frequency profile by i
initiating event is relatively flat you concluded that there is no plant specific vulnerability at CPSES. However, several plant enhancements were identified and implemented.
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NRC FILE CENTER COPY i
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Mr. C. Lance Terry l
TU Electric Company Comanche Peak, Units 1 and 2 CC' j
Senior Resident Inspector Honorable Dale McPherson U.S. Nuclear Regulatory Commission County Judge P. O. Box 1029 P. O. Box 851 Granbury, TX 76048.
Glen Rose, TX 76043 l
Regional Administrator, Region IV Office of the Governor U.S. Nuclear Regulatory Commission ATTN: John Howard, Director 611 Ryan Plaza Drive, Suite 400 Environmental and Natural 4
Arlington, TX 76011 Resources Policy P. 0. Box 12428 4
i Mrs. Juanita Ellis, President Austin, TX 78711 Citizens Association for Sound Energy i
1426 South Polk Arthur C. Tate, Director Dallas, TX 75224 Division of Compliance & Inspection i
Bureau of Radiation Control Mr. Roger D. Walker Texas Department of Health i
TU Electric 1100 West 49th Street Regulatory Affairs Manager Austin, TX 78756-3189 P. O. Box 1002 Glen Rose, TX 76043 Texas Utilities Electric Company c/o Bethesda Licensing 3 Metro Center, Suite 610 Bethesda, MD 20814 George L. Edgar, Esq.
Morgan, Lewis & Bockius 1800 M Street, N.W.
Washington, DC 20036-5869
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Mr. C. Lance Terry !
Based on the " Step 1" review, staff conclude that the CPSES IPE has met the intent of GL 88-20. A more detailed review, a " Step 2" review will not be conducted by the NRC staff.
Sincerely, Timothy J. Polich, Project Ma. nager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation i
l.
Docket Nos. 50-445 and 50-446
Enclosure:
Staff Evaluation Report w/ attachment cc w/ enc 1: See next page e
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COMANCHE PEAK STEAM ELECTRIC STATION INDIVIDUAL PLANT EXAMINATION STAFF EVALUATION REPORT J
1 I.
INTRODUCTION i
On August 28, and October 30, 1992, the Texas Utilities Electric Company (TU) provided the Comanche Peak Steam Electric Station (CPSES) Individual Plant i
Examination (IPE) submittal, front-end and back-end analyses respectively, in response to Generic Letter (GL) 88-20 and associated supplements. On January 23, and September 18, 1996, the staff sent questions to the licensee l
requesting additional information (RAI). The licensee responded in letters dated June 14, and October 24, 1996.
A " Step 1" review of the CPSES IPE submittal was performed and involved the efforts of Brookhaven National Laboratory in the front-end and the back-end i
analyses, and Sandia National Laboratory in the human reliability analysis i
(HRA). The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities. Therefore, the review considered (1) the completeness of the information and (2) the reasonableness of the results given the CPSES design, operation, and history. A more detailed review, a I
" Step 2" review, was not performed for this IPE submittal. Details of the l
contractor's findings are in the attached technical evaluation report (Appendix A) of this staff evaluation report (SER).
In accordance with GL 88-20, CPSES proposed to resolve Unresolved Safety Issue l
(USI) A-t5, " Shutdown Decay Heat Removal Requirements." No other specific USIs.or generic safety issues were proposed for resolution as part of the CPSES IPE.
II. EVALUATION CPSES is a Westinghouse 4 loop pressurized water reactor (PWR) with a large dry containment. The licensee's IPE submittal estimated the total CDF for the i-CPSES as 5.7E-5/ reactor-year for internally initiated events, including i
internal flooding. The CPSES CDF compares reasonably with that of other i
Westinghouse plants. Transients contribute 54% (includes loss of offsite power / station blackout 28% and anticipated transients without scram, 9%),
j internal floods 23%, loss of coolant accidents (LOCA) 17%, steam generator tube rupture (SGTR) 6%, and interfacing systems LOCA <1%.
Reactor coolant pump seal LOCAs appear in sequences that contribute 29% to the CDF, a majority of which stem from station blackout (SBO). The licensee's Level 1 analysis examined the significant initiating events and dominant accident sequences, and it appears complete.
'f ENCLOSURE 1
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1 The IPE has considered the impact of common cause failure (CCF) through the use of plant specific component CCF factors based on data from the PLG Inc.
mneric CCF database. The licensee's approach allows those CCFs that remain
< n the database after a data screening process, to be applied to CPSES.
In addressing the staff's concerns regarding the low Multiple Greek Letter (MGL)
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parameters used in the CCF analysis the licensee performed a sensitivity analysis wherein the 95th percentile CCF values were used. The licensee j
indicated that the total CDF increased by less than 2% as a result of the use of these CCF values.
i Based on the licensee's IPE process use'd to search for decay heat removal
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(DHR) vulnerabilities, and review of the CPSES plant-specific features, the staff finds the licensee's DHR evaluation consistent with the intent of the j
USI A-45, Decay Heat Removal Reliability, resolution.
The licensee perfomed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events. The licensee indicated that the methods used were the " HRA l
calculator methodology", which is based on a number of other Electric Power j
Research Institute (EPRI) methods, and the " direct estimation method."
In the l
HRA the licensee searched for pre-initiator events (latent human errors) including miscalibration and restoration faults. A significant number of pre-initiator events were considered. The following are some vf the operator j
actions the licensee identified as important in the estimate of the CDF:
Failure to align the suction of the centrifugal charging and safety l
injection pumps to the discharge of the residual heat removal pumps for l
recirculation.
Failure to recover an emergency diesel generator that has failed to l
start.
i Diesel Generator (CP-1-MEDGEE-01 or CP-1-MEDGEE-02) inadvertently disabled due to latent human error.
Failure to isolate break flow on steam generator tube rupture after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Turbine driven auxiliary feedwater pump unavailable due to latent human error.
However, there appear to be some limitations in tha HRA. Regarding post-initiator event analysis, the licensee used screening human error probabilities (HEPs) in the range of 0.05 to 1.0, to determine the most important human events. While these values may be acceptable as screening values, a potential problem arises from the fact that only lo human actions were assigned more realistic values, leaving the remaining HEPs at their screening values. The resulting magnitudes of the 10 HEPS were not unreasonable; however, there exists a possible distortion of the results from leaving the screening values in the dominant sequences.
In response to RAIs
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I the licensee argued that the guidance that they used from the Systematic Human Reliability Procedure (SHARP) method for " fine" screening analysis provided i
them with the capability to prevent distortions in the relative ordering of l
sequences from the use of overly conservative HEPs. However, the staff is not able to confirm whether or not the approach provided reasonable assurance that l
the ordering of sequences is appropriate.
i As part of the HRA the CPSES IPE credited local repair of various equipment components and systems, including diesel generators and pumps. The licensee j'
stated that the repair activities credited in the CPSES IPE relies on data for l
recovery of equipment from a generic Electric Power Research Institute i
database. The licensee indicated that the data from the EPRI document are
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based on " normal" operation and that during an emergency condition the induced stress is expected to be a motivating factor and hence the. data is l
" conservative." It is not clear that this assumption is valid since the success of equipment repair depends on many important plant-specific factors j
such as the type of failure, time needed for diagnosis, time needed for repair j
(which may range from a very few hours to several days), crew competing tasks
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under different accident conditions, and crew availability (especially when i
multiple-repairs are credited). These factors do not appear to have been i
taken into consideration in the CPSES IPE in modeling and estimating equipment repair probabilities under accident conditions.
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The licensee, however, as part of its response to a staff
- request for i
additional information, performed a sensitivity analysis that involved j
setting the repair / recovery values to guaranteed failure. The result was a CDF increase of 70% due to internal events (60% from failure to recover diesel generator and 10% from failure to recover pumps) and 55% for flooding. The i
licensee indicated that the dominant sequences (loss of offsite power and flooding) remained dominant and relative contributions from all other initiators remained in about the same range as in the original quantification.
i While the staff believes that credit taken for repair of equipment does have an effect on the IPE's results (as shown above) it does not expect the impact to be as great as identified by setting the failure to recover probability to one. The staff believes, however, that the licensee should have provided a 1more sound technical basis for modeling equipment repair in the CPSES PRA.
However, regardless of these limitations, it appears that the licensee in their systematic examination gained an understanding of the quantitative
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impact of human performance on core damage and radioactive material release 4
frequencies such that a potential vulnerability was not overlooked.
i The licensee evaluated and quantified the results of the severe accident j
progression through the use of a containment event tree and considered i
uncertainties in containment response through the use of sensitivity ar41yses, 4
The licensee's back-end analysis considered important severe accident j
phenomena, and it appears to be complete. Among the CPSES conditional i
containment failure probabilities, the licensee estimated that early l
containment failure is 1%, late containment failure is 51% with overpressurization (due to accumulation of non-condensible gases) or basemat i
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l melt-through being the primary contributors, and bypass is 8% (SGTR, 8% and i
interfacing systems LOCA, <1%). According to the licensee, the containment remains intact about 40% of the time. Early radiological releases are l
dominated by transient sequences and late releases are dominated by LOCA i
sequences. -
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The licensee's response to containment performance improvement program j
recommiendations is consistent with the intent of GL 88-20 and the associated Supplement 3.
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Some insights and unique plant safety features identified at CPSES by the j
licensee are:
1.
Ability to perfom bleed and feed cooling with one power operated relief valve (PORV).
PORVs are provided with large nitrogen accumulators l
2.
4 high pressure (2 charging and 2 safety injection) emergency core cooling system pumps to provide reactor coolant system (RCS) injection i
and makeup flow following a safety injection signal.
3.
Cross connection capability exists between the CPSES units 1 and 2 for the service water (SW) systems and the component cooling water ( CW) i systems.
4 4.
Ability to use SW system as a source of water supply to the auxiliary j
feedwater pumps.
l 5.
Four-hour battery capacity without load shedding.
i 6.
Credit is given in the IPE for local manual operation of the turbine i
driven auxiliary feedwater pump (TDAFW).
7.
Switchover from injection to recirculation during LOCAs is automatic; i
however, during very small, small and medium LOCAs, establishment of high pressure recirculation from the sump requires manual actions of the operators to align the discharge of the residual heat removal pumps to j
the suction of the safety injection and/or the charging pumps.
8.
A number of systems have a dependence on heating ventilation and air-conditioning and Instrument Air.
g.
The CPSES has a large reactor cavity floor area and a cavity design that allows water to flow into the reactor cavity. This facilitates debris coolability in the reactor cavity.
The licensee indicated that the results of the CPSES IPE were evaluated using the reporting criteria from Generic Letter 88-20, followed by a qualitative l
analysis. They stated that since the core damage frequency profile by initiating event is relatively flat they concluded that there is no plant a
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I specific vulnerability at CPSES. The licensee identified several plant enhancements listed below that were implemented and reflected (except for j
item 7) in the IPE:
l 1.
Upgraded'the procedure for manual control of the TDAFW upon loss of air l
and dc power to a higher level procedure.
l 2.
Procedure revised to include verification of the availability of the CCW j
'before switching from injection to recirculation.
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3.
Procedure revised to instruct the operator to locally manually throttle j
the valve for normal charging flow in order to divert flow to the 4
reactor cooling pump seals.
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4.
Procedure revised to instruct the operators to start the standby safety chilled water train upon auto-start of the auxiliary feedwater pumps to j.
prevent overheating of the motors.
l 5.
Revised procedure for the operator to restart the mair, feedwater on i
failure of the auxiliary feedwater, preferentially using the flowpath j
that contains valves controlled from the control room instead of the
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flowpath containing valves that are locally manually controlled, i
6.
Eliminated the blank flange from the CCW cross-connect to allow the CCW l
system to be cross-connected between units using valves.
l 7.
Replaced the seals for the reactor coolant pumps with high temperature i
seals. The IPE seal LOCA model is based on the characteristics of the j
original seals.
t III. CONCLUSION Based on the above findings, the staff notes that:
(I) the licensee's IPE is l
complete with regard to the information requested by GL 88-20 (and associated guidance NUREG-1335), and (2) the IPE results are reasonable given the CPSES 1
design, operation, and history. As a rest.lt, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, sr.d therefore, that the CPSES IPE has met the intent of GL 88-20. The staff, however, identified limitations in the licensee's HRA during severe accidents that will limit the use of the IPE for purposes other than GL 88-20. The licensee has not indicated their intent to use the IPE as a "living PRA." Regardless, the staff encourages the licensee to address the limitations identified for the CPSES IPE identified above in order to make it a valuable tool for other applications.
It should be noted that the staff's review primarily focused on the licensee's ability to examine the CPSES for severe accident vulnerabilities. Although certain aspects of the IPE were explored in more detail than others, the review is not intended to validate the accuracy r* the licensee's detailed
i e
a findings (or quantification estimates) that steemed from the examination.
Therefore, this SER does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of GL 88-20.
Attachment:
Appendix A Principal Contributor:
E. Rodrick, RES Date: March 10, 1997 i
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APPENDIX A j
COMANCHE PEAK STEAM ELECTRIC STATION INDIVIDUAL PLANT EXAMINATION l
l TECHNICAL EVALUATION REPORT 4
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4 ATTAQMENT 4