ML20140C337

Forwards Staff Evaluation for Licensee IPE Submittal for Internal Events & Internal Flooding & Contractor TERs

ML20140C337
Person / Time
Site:	Clinton
Issue date:	03/27/1997
From:	Pickett D NRC (Affiliation Not Assigned)
To:	Telthorst P ILLINOIS POWER CO.
Shared Package
ML20140B649	List: ML20140B642 ML20140B652 ML20140B658 ML20140C337
References
TAC-M74396, NUDOCS 9704030150
Download: ML20140C337 (12)

Text

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p%% &W y t UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30066 4 001 o% ...../ .

March 27, 1997 Mr. Paul J. Telthorst Director - Licensing Clinton Power Station i P. O. Box 678 i Mail Code V920 i Clinton, IL 61727  !

SUBJECT:

STAFF EVALUATION OF CLINTON POWER STATION INDIVIDUAL PLANT j EXAMINATION - INTERNAL EVENTS (TAC NO. M74396)

Dear Mr. Telthorst:

j Enclosed is the staff's evaluation for the Clinton Power Station individual

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plant examinati6n (IPE) submittal for iriternal events and. internal flooding.

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The contractor's technical evaluation reports (TER_s) are included with the evaluation. i The staif performed a " Step 1" review during which it examined the IPE results 4 for their " reasonableness," taking into consideration the design and operation of Clinton Power Station. Science & Engineering Associates, Inc., Concord Associates, and Scientech, Inc reviewed the front-end analysis, human reliability analysis, and back-end analysis, respectively, of the IPE submittal. Thair TERs are enclosed as Appendices A, B, and C. respectively, to the staff's evaluation. These TERs were reviewed by the IPE Senior Review Board (SRB) as part of the Office of Nuclear Regulatory Research (RES) quality assurance process. The SRB consists of RES staff and consultants at Sandia ,

and Brochten National Laboratories with probabilistic risk assessment (PRA) l expertise.

In the IPE, you estimated a total core damage frequen~cy (CDF) of about 3E-5/ reactor-year, including a contribution from internal flooding of about 2E-6/ reactor-year. Transients contribute 52 percent, station blackout 37 i percent, internal flooding 6 percent, loss-of-coolant accidents 4 percent, and anticipated transient without scram 1 percent.

To identify vulnerabilities, you addressed the following questions: (1) Are there new or unusual means by which core damage or containment failure could occur compared to those identified in other PRAs: (2) Do the results suggest /

that the plant CDF would not be able to meet the NRC's subsidiary safety goal for core damage (IE-04/ reactor-year); and (3) Are there any systems,

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components, or operator actions that control the core damage result (i.e., J greater than 90 ercent)? Your examination did not lead to the identification ks--D/

of any vulnerabi ities. Several plant improvements, however, were identified l and implemented.

NHC m g6H 2 CUPY 970403o150 970327 PDR ADOCK 0500o461 P PDR

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P. Telthorst-

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l . i-t On the basis of the " Step 1" review, we conclude'that you have met the intent of Generic Letter 88-20.--The staff does not recommend that a more detailed.

" Step 2" review be conducted. We note, however, that identified weaknesses in

the IPE may limit its use for'any other regulatory purposes.

If you have any questions regarding the enclosed evaluation, please feel free 1

to contact me at (301) 415-1364.

i  : Sincerely,1 Original signed by:

Douglas V. Pickett, Senior Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-461

Enclosures:

As stated cc: See next page l

DISTRIBUTION: W/ encl i Docket File OGC PUBLIC RHernan PD3-3 R/F JRoe ,

ACRS EAdensam (EGA1)

JCaldwell, RIII GMarcus MHodges, RES i

DOCUMENT NAME: G:\CLINTON\CLI74396.LTR To receive a copy of this document, indicate in the box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy 0FFICE LA:PD3-3, E PM:PD3-3 .c c:  ;

NAME CBoylell) DPickettF' l DATE 03/a?/97 03tL3/97

0FFICIAL RECORD COPY

._ m P. Telthorst ,

On the basis of the " Step 1" review, we conclude that you have met the intent of Generic Letter 88-20. The staff does not recommend that a more detailed.

" Step 2" review be conducted. We note, however, that identified weaknesses in the IPE may limit its use for any other regulatory purposes.

If you have any questions regarding the enclosed evaluation, please feel free to contact me at (301) 415-1364. .

Sincerely.

Original signed by:

Douglas V. Pickett. Senior Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-461

Enclosures:

As stated cc: See next page DISTRIBUTION- w/ encl Docket File OGC PUBLIC RHernan PD3-3 R/F JRoe ACRS EAdensam (EGA1)

JCaldwell. RIII GMarcus MHodges. RES DOCUMENT NAME: G:\CLINTON\CLI74396.LTR Ta receive a copy of this document, Indicate in the box: "C" = Copy wthout enclosures "E" = Copy wth enclosures "N" = No copy 0FFICE LA:PD3_3, lE PM:PD3-3 .c c:

NAME CBoyle/ k) DPickett F DATE 03/2?/97 03tL3/97 0FtlCIAL RECORD COPY

. . . _ _ _ . - . _ _ _ _ - _ . _ _ _ _ . ~ _ . _ _ _ . _ _ _ _ . _ . . . _ . _.- _. . _ . _ _ . _ _ _ . . _ _

P. Telthorst '

On the basis of the " Step 1" review, we conclude that you have met the intent  !

( of Generic Letter 88-20. The staff does not recommend that a more detailed, s l

" Step 2" review be conducted. We note, however, that identified weaknesses in the IPE may limit its use for any other regulatory purposes.

l If you have any questions regarding the enclosed evaluation, please feel free to contact me at (301) 415-1364.

Sincerely.

l w =h V f i l Douglas V. Pickett. Senior Project Manager Project Directorate III-3 t

Division of Reactor Projects III/IV l Office of Nuclear Reactor Regulation -

Docket No. 50-461 i

Enclosures:

As stated cc: See next page t

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l Mr. Paul J. Telthorst Clinton Power Station Illinois Power Company Unit No. 1 cc:

Mr. Wilfred Connell Vice President Illinois Department Clinton Power Station of Nuclear Safety Post Office Box 678 Office of Nuclear Facility Safety Clinton, Illinois 61727 1035 Outer Park Drive Mr. Daniel P. Thompson Manager Nuclear Station Engineering Department Clinton Power Station Post Office Bdx 678

! Clinton, Illinois 61727 Resident Inspector )

l U.S. Nuclear Regulatory Commission i RR#3, Box 229 A <

l Clinton Illinois 61727 l l Mr. R. T. Hill l Licensing Services Manager j General Electric Company 175 Curtner Avenue, M/C 481 San Jose, California 95125 Regional Administrator Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road l Lisle Illinois 60532-4351 l Chairman of DeWitt County c/o County Clerk's Office DeWitt County Courthouse Clinton, Illinois 61727 Mr. J. W. Blattner Project Manager Sargent & Lundy Engineers 55 East Monroe Street Chicago, Illinois 60603

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CLINTON NUCLEAR POWER PLANT INDIVIDUAL PLANT EXAMINATION STAFF EVALUATION REPORT l

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l ENCLOSURE

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1. INTRODUCTION On September 23, 1932, Illinois Power (licensee) submitted tne Q11nton Nuclear

, Power Plant individual plant examination (IPE) in response to Generic Letter

. (GL) 88-20 and associated supplements. On July 21, 1995, the staff sent a request for additional information to the licensee. The licensee responded in a letter dated November 22, 1995.

The staff performed a " Step 1" review of the Clinton IPE submittal. As part of this review, Science & Engineering Associates, Inc., Scientech, Inc., and Concord Associates reviewed the front-end a'nalysis, back-end' analysis, and the human reliability analysis (HRA), respectively. 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 Clinton design, operation, and history. A more detailed review, a " Step 2" review, was not performed as part of this IPE submittal. Details of the con' tractors' findings are given in the technical evaluation reports (Appendices A, B, and C) attached to this staff

. evaluation report (SER).

i In accordance with GL 88-20,'the licensee proposed to resolve Unresolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements." No other USIs or generic safety issues were proposed for resolution as part of the Clinton IPE.

II. EVALUATION Clinton Nuclear Power Station is a boiling water reactor (BWR) 6 with a Mark III containment. The licensee estimated a core damage frequency (CDF) of about 3E-5/ reactor-year from internally initiated events, including a contribution from internal flooding of about 2E-6/ reactor year. The Clinton CDF compares reasonably with that of other BWR 6 plants. Transients i ' contribute 52 percent, station blackout -37 percent, internal flooding 6 -

percent, loss-of-coolant accidents 4 percent, and anticipated transient without scram 1 percent.

" On the basis of the licensee's Fussell-Vesely importance analysis, the most-important contributors to the estimated CDF sequences are associated with loss

of offsite power, failure of the high pressure core spray (HPCS) system, failure of the reactor core isolation coolant (RCIC)* system, transients l without isolation, transients with isolation, failure of the fire protection system, and failure of t.he automatic depessurization system (ADS).

The licensee performed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery-of-failure events. The licensee identified the following operator actions as important in the estimate of the CDF: failure to recover offsite power in 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, failure to recover HPCS, failure to recover RCIC, failure to manually initiate ADS, failure to recover Division 2 within the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and failure to manually initiate service water.

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The licensee's Level 1 analysis appears to have examined the significant initiating events and dominant accident sequences. However, it does not appear that the licensee incorporated plant-specific experience in the calculation of equipment failure rates and initiating event frequencies.

Responding to staff's'RAI, the licensee stated that very limited plant-specific data exists because Clinton'is a relatively new plant. The licensee cited examples showing that plant-specific failure data are comparable to or better than the generic data used. However, plant-specific data could have been used to update the generic data by means of a Bayesian process.

Insufficient incorporation of plant experience may not have a significant  !

', impact on the total CDF but it may have a significant impact on' the relative contributions of the various sequences and failure events and, therefore, it may limit the ability-to identify plant-specific insights and improvements.  ;

Also, the Clinton IPE credited local repair of various equipment components and systems (including diesel generators, pumps, valves, and instrumenfation) and has taken credit for up-to-two component / system repair activities. The licensee stated that the repair activities credited in the Clinton IPE were based on actual emergency exercises; however, the quantification of these repair activities is based on a generic Electric Power Research Institute database. ,

The licensee, as part of its response to staff's RAI, performed a sensitivity analysis that involved removal of all credited equipment repair recoveries, except thos.e involving offsite power, direct current power, and operator actions performed from the control room. The result was~a'n insignificant increase of the CDF from about 3E-05/ reactor-year to about 4E-05/ reactor-year, a factor of 1.5. The relative contributions of individua,1 accident sequences also were not significantly altered. In no instances were increases in individual accident sequence frequencies greater than a factor of 2.4.

Although two new sequences.were introduced, their frequencies were less than IE-08/ reactor-year. .

l The success of equipm'ent repair depends on many important plant-specific factors such as the type of failure, time needed for diagnosis, time needed i for repair'(which may range'from a very few hours to several days), crew competing tasks under different accident conditions, and crew availability (especially when multiple-repairs are credited). These factors do not appear to have been taken into consideration in the Clinton IPE in modeling and estimating equipment repair probabilities under accident conditions.

Therefore, although the staff concludes that the credit.taken for repair of equipment did not have a significant effect on the IPE's results, the staff believes that insufficient examination of important aspects of human performance under severe accident conditions may have limited the ability to identify factors critical to the accomplishment of the actions modeled.

The Clinton IPE results are in line with those for similar BWR 6 plants in

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both the most dominant initiators and the most dominant accident sequences.

Therefore, the staff believes it is unlikely that the limitations identified have impacted the licensee's overall conclusions from the IPE and its capability to identify vulnerabilities. However, they may have limited its ability to gain insights and identify improvements.

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0n the basis of the licensee's IPE process used.to search for decay' heat removal (DHR) vulnerabilities, and review of Clinton plant-specific features,

' the staff finds the licensee's DHR evaluation consistent with the intent of the resolution of USI A-45 " Shutdown Decay Heat Removal Requirements."

The licensee evaluated and quantified the results of the severe accident progression. The back-end analysis results are as follows: early containment

, failure (including isolation failure) will occur about 3 percent of the time, late containment failures will occur about 2 percent of the time, and bypass will occur less than 1 percent of the time. The containment. remains intact 95 percent of the time.

The staff has also identified a weakness in the Clinton IPE back-end analysis.

The licensee, responding to staff's request for additional information, did not provide a justification for the very small probability of containment failure from hydrogen ignition after power recovery under station blackout conditions, or for no containment failure from excessive fuel-coolant interaction. Although, as stated in the IPE, Clinton has "the largest free air volume and suppression pool Volume to rated thermal power of any domestic Mark III," it is not clear whether the result that the containment remains intact 95 percent of the time is skewed because of the licensee's optimistic '

approach to the examination of hydrogen combustion and ex-vessel steam explosion phenomena. It is noted, however', that although certain phenomena may have been~ treated optimistically, the licensee has considered all relevant phenomena. ' The licensee developed containment event trees to characterize containment response to severe accidents and examined several severe accident l phenomena in detail for their applicability to Clinton. Also, the licensee '

used the Modular Analysis Accident Program to ' analyze representative sequences and addressed phenomenological uncertainties of accident progression through sensitivity studies. These findings coupled with the fact that the Clinton containment is substantially stronger than any other Mark III containment -

leads the s.taff to conclude that the licensee did not miss a vulnerability -

with respect Clinton's containment.  ;

The licensee's r.esponse to' containment performance improvement program  ;

recommendations is consistent with the intent of GL 88-20 and.its '

Supplement 3. ,

Some insights and uni'que plant safety features identified by the licensee are the following:

1. four-hour battery lifetime

2. motor-driven feedwater pump

3. three safety-related divisions of core cooling, each of which has its I own emergency diesel generator and cooling water

4. ability of the emergency core cooling system pumps to operate when,the suppression pool is saturated  !

5. ability to cr,oss-connect the fire protection system for core injection l

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6. a cont'ainment with the largest free air volume and suppression pool volume with respect to rated thermal power of any domestic Mark III.

To identify vulnerabilities the licensee addressed the following questions:. )

(1) Are-there new or unusual means by which core damage or containment failure could occur. compared to those identified in other PRAs? (2)-Do the results suggest that the plant CDF would not be able to meet the'NRC's subsidiary safety goal for core damage (IE-04/ reactor-year)? (3) Are there any systems, components, or operator actions that control the core damage result (i.e., greater.than 90 percent). The licensee's examination did not lead to the' identification of any vulnerabilities. Several plant improvements, however, were identified. The following improvements have been implemented: I

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l. operator training to emphasize the importance of maintaining offsite l power-  !

2. operator training to emphasize the importance of manual ADS initiation

3. modification of the HPCS surveil, lance procedure to test the suppression pool suction path

4. installation of a bypass line to allow easier use of the fi.re protection system for vessel makeup

5. operator training to emphasize the importance of maintaining offsite power.to prevent offsite releases

6. operator training to emphasize the importance of alternate current power recovery .

7. operator training to emphasize the importance of manually isolating the containment bypass path into the fuel pool cooling / cleanup line during station blackout

8. operator training to emphasize the significance of scram system hardware failures as related to release frequencies III. CONCLUSION Based on the above findings, the staff notes thati (1) the licensee's IPE is complete with regard to the information requested by Generic Letter 88-20 (and associated guidance in NUREG-1335), and (2) the IPE results are reasonable given the Clinton Nuclear Power Station. design, operation, and history. As a '

result, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident .

. vulnerabilities, and therefore, that the Clinton Nuclear Power Plant IPE has met the intent of Generic Letter 88-20. However, the staff noted limitations in the. licensee's IPE associated with (1) the extensive use of generic data, (2) the' credit taken for equipment' repair, (3) the credit taken for containment performance under hydrogen combustion upon power recovery

-(negligible failure probability), and (4) the credit taken for containment j 4

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' 8 performance under ex-vessel steam-explosion conditions (no containment failure). These limitations may limit the IPE's usefulness for other regulatory , applications.

It should be noted that the staff's review primarily focused on the licensee's ability to examine Clinton Nuclear Power Station 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'of the licensee's detailed findings (or quantification e timates) that stemmed,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.

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l l i APPENDIX A FRONT-END TECHNICAL EVALUATION REPORT S

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