ML18058B428

Vols 1,2 & 3 of Palisades Nuclear Plant Ipe.

ML18058B428
Person / Time
Site:	Palisades
Issue date:	11/30/1992
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18058B427	List: ML18058B426 ML18058B428
References
NUDOCS 9302120096
Download: ML18058B428 (14)

Text

PALISADES INDIVIDUAL PLANT EXAMINATION TABLE OP CONTENTS VOLUME 1

1.0 DESCRIPTION

OF APPROACH 2.0 FRONT-ENO ANALYSIS 2.1 SYSTEMS ANALYSIS 2.1 FIGURES 2.2 SEQUENCE DEFINITION 2~3 SEQUENCE QUANTIFICATION 2 *4 RESULTS *

2.5 REFERENCES

APPENDIX A - FLOODING APPENDIX B - OHR VOLUME 2

3. 0 BACK-.END ANALYSIS 3.1 PHYSICAL DESCRIPTION

3. 2 SEVERE ACCIDENT. ISSUES

3.2 REFERENCES

3.2 FIGURES 3.3 PLANT DAMAGE STATES

3.3 REFERENCES

3.3 FIGURES 3.4 CONTAINMENT RESPONSE

3.4 REFERENCES

3.4 FIGURES 3.5 CONTAINMENT EVENT TREE

3.5 REFERENCES

• -\ ---*1

• \~.~*/(/*

/ .

3.5 FIGURES VOLUME 3 3.6 CET QUANTIFICATION/ANALYSIS

3.6 REFERENCES

APPENDIX 3.6 PDS BE PROBABILITY SHEETS 3.7 RELEASE CATEGORIES

3.7 REFERENCES

4.0 UTILITY PARTICIPATION 5.0 UNIQUE SAEFTY FEATURES & INSIGHTS 6.0

SUMMARY

AND CONCLUSIONS

~- .,,.,

.: ~~:~~~-;:~;t~.*~.

TABLE OF CONTENTS SECTION 1.0 1.0 PESCRIPTION OF APPROACH ************

• 1. 0-1 1.1 GENERAL METHODOLOGY . . . . . . . . . . . . . . . .

• 1. 0-1 1.2 INFORMATION ASSEMBLY * . . . . . . * * * . . . . . .

• 1.0-2 1.3

SUMMARY

OF MAJOR FINDINGS . . . . . . . . . . . * *

• 1.0-3 1.3.1 FRONT-END ANALYSIS * . . . * * * . . * * . . . 1.0-3 1.3.2 BACK-END ANALYSIS . . . . . . . . . . . . .

• 1.0-4

1.4 CONCLUSION

S . . . . . . . . . . ........ *

• l. 0-6 1.5 TABLES AND FIGURES . . . . . ** . . * * * . . . . * *

• 1.0-8

EXECUTIVE

1.0 DESCRIPTION

OF APPROACH The methodology for completing the Palisades IPE is described in this section. It ensures that the IPE objectives and provisions of NRC Generic Letter 88-20 are met. The objectives for the Palisades IPE are to satisfy the recommendations of the Generic Letter and provide valuable input to facilitate closure of severe accident issues at Pal~sades. Generic Letter 88-20 recommendations are:

l} develop an overali appreciation of severe accident behavior;

2) understand the most likely severe accident sequences that could occur at Palisades;

3) gain a more quantitative understanding of the overall probabilities of core damage and fission product releases; and

4) if necessary, reduce the overall probability of core damage and radioactive material releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.

In addition to these recommendations, the intent of Generic Letter 88-20 is to cause significant utility participation in performance of the IPE. Utility participation increases the utility's knowledge of severe accidents and aids in the resolution of the severe accident issues at the plant.

1.1 GENERAL METHODOLOGY The Palisades IPE was directed by Consumers Power personnel from both the Palisades Plant and the General Office. The. IPE was performed by CPCo personnel with consulting services provided by TENERA, L.P.; Gabor, Kenton and Associates; and ABB Impell.

Performing the IPE in this manner enabled CPCo personnel to gain the knowledge derived from performing the IPE and to apply the results of the IPE to procedures, training, and modifications.

Consulting services were utilized to provide general guidance.

Further details of the IPE organization a~e provided in Section 4.0.

The process was divided into front-end and a back-end analyses.

The front-end is a level l PRA which includes the development of of logic models (event trees and fault trees), human reliability analysis, data base development, acc~dent sequence quantification and an internal flooding analysis. A level 2 PRA was performed for the back-end analysis. The level 2 includes analysis of containment structure and systems reliability, physical processes 1.0-1

of accident progression, and source term characterization. The detailed methodology for the front-end analysis is discussed in Section 2.0 and the back-end analysis methodology is discussed in Section 3.0. These sections provide the details for the:

1) methods used to prepare the fault and event trees;

2) success criteria for each fault or event tree branch;

3) method used to account for dependencies; and

4) quantification process applied.

Also a detailed analysis of the decay heat removal systems at Palisades was performed to address Unresolved Safety Issue (USI)

A-45 "Shutdown Decay Heat Removal Requirements" and is discussed in Section 2.4.5 and Appendix B.

The results of the Palisades IPE for the front-end and back-end analyses are discussed in Sections 2. 4 and 3. 6. Those results were used to determine if any vulnerabilities exist. The following questions were used as the screening criteria for vulnerabil~ties:

1. Do the Palisades IPE results meet the NRC's safety goal for core damage?

2. Are the results for core damage sequences or containment performance consistent with other PRAs?

3. Does the probability of sequences characterized as having large releases exceed 10% of the core damage frequency?

Overall, the answers to these questions are;

1. Yes, the Palisades IPE results do meet the NRC's safety goal for core damage.

2. Yes, the results for core damage sequences and contain-ment performance are consistent with other PRAs.

3. No the probability of sequences characterized as having large releases does not exceed 10% of the core damage frequency.

1.2 INFORMATION ASSEMBLY The details of the specific information and data used for the front-end and back-end analyses are discussed in Sections 2.0 and 3.0. Each of these sections discusses the generic or plant specific data utilized, the design data utilized from existing sources (FSAR) or new analysis, and any IPEs or PRAs that were used for 1.0-2

comparison purposes. The last part of those sections includes references to specific data or reports such as plant procedures and drawings or NRC guidelines and NUREGs.

The IPE team performed walkdowns to obtain the latest and most accurate information available in regard to the as built configuration of the plant. All modifications installed before July 1, 1992 are included in the IPE. During the walkdowns, particular attention was given to the development of the human reliability analysis. Results of the walkdowns, reviews of plant procedures and discussions with operations and training personnel provide the basis for the IPE modeling.

An overview description of the major plant equipment is provided in Table 1-1 and Figure 1-1 (a simplified plant schematic drawing).

Palisades has a two-loop Combustion Engineering nuclear steam supply system which produces 2560 MWth and consists of the reactor,.

a pressurizer with two PORVs (each capable of initiating and sustaining feed and bleed cooling) and three safety relief valves, two steam generators, and four primary coolant pumps. The primary system is contained in a large dry pre-stressed concrete containment building designed by the Bechtel Power Corporation.

The containment structure is lined with a 1/4" carbon steel liner.

The concrete is reinforced and po~t-tensioned~

The secondary system has a turbine/generator manufactured by the Westinghouse Electric Corporation which has an electrical output of 845 MWe (max).

• The plant is located on Lake Michigan six miles south.of South Haven, Michigan. Construction started on August 25, 1966, and*

commercial operation began on December 31, 1971.

1.3

SUMMARY

OF MAJOR FINDINGS 1.3.1 FRONT-END ANALYSIS The level 1 analysis resulted in a total Core Damage Frequency (CDF) of 5.07E-5/yr excluding flood initiators. Internal flood

• initiators are estimated to contribute 3.0E-7/yr to the COF. The initiating event contributions to the core damage frequency are shown in Figure 1-2. The dominant core damage sequences are:

1) The most probable contributor to the total core damage frequency (at l.2E-OS) is a transient event with failure of secondary cooling and once through cooling. Seventy-eight percent of the sequence frequency is due to station blackout.

2) The second most probable contributor to the total core damage frequency (at l.*lE-05) is also a transient event with a loss of all secondary cooling and once through 1.0-3

cooling. In this case, once through coolipg is successful in the injection phase but fails in recirculation. The important initiators are loss of service water (34%) and loss of offsite power (26%).

3) The third most probable contributor to the total core damage frequency (at 7.SE-06) is a small break LOCA with failure of high pressure injection in the injection pha-e.

4) The fourth most probable contributor to the total core damage frequency (at 7.lE-06) is a small break LOCA with failure of high pressure injection in the recirculation phase.

5) The fifth most probable contributor to the total core damage frequency (at 6.2E-06) is transient failure of secondary cooling and once through cooling due to the unavailability of the PORVs. The dominant initiator is turbine trip with the main condenser available (45%).

6) The sixth most probable contributor to the total core damage frequency (at 1. SE-06) is an anticipated transient without scram (ATWS) sequence. The sequence is due to an electrical scram failure with successful relief valve opening but with an adverse moderator temperature coefficient.

7) The seventh most probable contributor to the total core damage frequency (at 1.lE-06) is an ATWS sequence due to a mechanical scram failure with successful relief valve opening anQ. an adverse moderator temperature coefficient.

In general, the results of the core damage assessment compare favorably with results of previous risk assessments. No vulnerabilities were identified in the front-end analysis.

1. 3. 2 BACK-END ANALYSIS In preparation for determining the back-end analysis input, . a bridge event tree was developed and quantified to extend the core damage sequences to plant damage states. The plant damage states combine the core damage results with the .status of containment systems (sprays and coolers). This- methodology was used to separate containment system performance quantification from the phenomenological issues such as direct containment heating or core-concrete interaction assessed in the containment event tree.

Results of the event tree quantification are shown in Figure 1-2.

The two most probable plant damage states shown in Figure 1-3 for Palisades are small break LOCAs (BEGP) and transients (TEJP) that have both containment sprays and containment air coolers initially available. BEGP and TEJP combined account for approximately 40% of 1.0-4

the plant damage states. The next most probable plant damage state (TEJW) results from transients with no containment systems available and the safety injection refueling water not injected into containment (TEJW) accounts for 18% of the plant damage states. The next most probable plant damage state * (TEJV) is transients with no containment systems available and the safety injection refueling water injected into containment. TEJV accounts for approximately 13% of the plant damage states. In approximately 75% of the plant damage states, the containment systems are initially available. Based on these results, we believe that the reliability of the containment systems is acceptable. There were no unusual or unanticipated issues identified in this analysis.

The back-end phenomenological issues were evaluated using the Containment Event Tree (CET). The CET was created to allow each sequence result to represent a unique source term characterization.

The fission product release information presented in Section 3.7, Table 3. 7. 21-1 will be used as the input to the Level 3 PRA in which the risk to the health and~safety of the public due to severe accidents will be determined.

The largest source term categories resulting from the CET analysis are the early bypass or containment isolation failure scenarios which overall are small contributors to the core damage frequency.

Larger source terms were identified in cases where the core is recovered in the vessel but the containment fails near the time of vessel failure; and, for upward debris dispersal cases with early containment failure. The results of the ..CET quantification indicate that for standard containment failure modes, the source term composition is consistent with that found in PRA studies of other PWRs.

The CET quantification also includes a non-standard containment isolation failure mode that is unique to Palisades. It is the potential for the.core debris, after it has exited the vessel at low pressure, .to relocate to the engineered safeguards rooms. This failure mode contributes approximately 30% to the probability of early containment failure results in an early containment failure contribution which is higher than would be expected of a large dry containment. The impact of this failure mechanism is related to two unique features of the Palisades plant. The two features are the location of the sump, which makes this unique containment failure possible, and the existence of a passive cavity flooding system. The first is the fact that the sump for engineered safeguards pumps is located directly below the reactor cavity (see Figure 1-4). The second feature is the passive cavity flooding system which will cause the reactor cavity to be flooded whenever the containment sprays are activated for more than a few minutes.

The cavity floodin~ system provides a means to prevent core debris from exiting the vessel if core damage occ~rs. We believe that the location of the sump and the its impact on containment performance constitutes a significant insight for the Palisades plant.

1.0-s

1.4 CONCLUSION

S With respect to the general plant response to initiating events, the plant system performance is acceptable and in line with what would be expected when it is compared to previous risk assessments.

The structural issue in regard to the location of the containment sump to the engineered safeguards system pumps does warrant

• -*--additional consideration. Several possible corrective actions are being considered. These can be characterized as being in one of two groups:

1) Actions to reduce . the probability of core damage .

sequences which contribute most _strongly to the potential for core relocation, and

2) Actions to delay or prevent relocation of the core debris from the reactor cavity to the sump.

• Sensitivity studies to establish possible benefits from each of the options considered are ongoing. However, to a certain extent, the relative impact of, or potential for, a core relocation is driven by the uncertainties associated with several of the key accident progression phenomena. The appropriateness of actions taken or the degree of benefit to be gained from those actions initiated to reduce* the plant risk in this area will be increased as improved knowledge of severe accident phenomena is gained. The risk of a severe accident that will affect the safety of the public is no greater than that for other nuclear power plant~ and is within the guidelines provided by the NRC. Therefore, at this time we plan no plant modifications.

1.0-6

S~ry of Design Features

1. Hlgh*Preasure Injection **

b.

Two centrifugal low heed Injection p.111)9

• Three 2500*psig positive displacement charging ~*

. - ,, c. Charging JlUl1)S can take suction from Sefety Injection Refueling Weter tank or boric acid storage tents.

d. P~ r~lre coq>0nent cooling ... tar efter reclrculetlon ectuetlon.

2. Low*Pressure Injection ** Two low pressure Injection~ deliver flow when RCS Is below 200 pslg.

b. Reclrculetlon llOde autonietlcelly stops low pressure Injection JlUl1)S but ellows 11111,.,.l restert with suction from the conteirrnent s""'.

c. P~ r~ire coq>0nent cooling weter efter recirculation actuation.

3. Auxiliary F~ter

• • Two 100 percent 110tor*drlven ~ end one 100 percent turbfne*drlven ~*

b. P~ take suction from condensate storage tank <CST) with a minillUll 100 000 gallon Inventory.

4. E11111r9ency Power System ** Two 2400V AC class 1E buses, each feeding 480V AC class 1E buses end inotor control centers.

b. One diesel generator for each 2400V AC bus.

c. Two trains of de power are supplied from the Inverters and 2 unit batteries.

5. C~t Cooling Water **

b.

Three 1110tor*drlven purflS end two heat Cools RHR he*t exchangers; RCP inotors exchanvers.

and thenaal barriers, prf1111ry cool Ing for ESF and charging pu!Jll.

c. One of 3 ~ can provide sufficient flow but bo~h heat exchanQers are rea.ilred.

6. Service Water **

b.

Three 111Dtor*drlven ~ discharge to a ring heeder.

Cools ccqionent cool Ing heat exchanvers, cont.alnnent

. fan coolers, diesel generator coolers, ESF rOOlll cooling, backup cooling to ESF ~. alternate suction source to auxiliary feedwater ~*

c. One cum can s~ly sufficient flow.

7. Contairment Structure **b. Large, dry, prestressed concrete.

1.6 million cubic foot volUllt.

c. 55 PSI; design pressure.

8. Contalnnent Sprey **

b.

Three motor*drlven pull)S, two spray headers.

Heat exchangers downstre111 of ~ provide recirculation heat removel.

c. All three pu11>9 discharge to one header.

d. Water supplied by refueling water storage ~anlc or contalnnent SUIP*

• • Recirculation mode takes suction on contalrrnent s""' end discharges to the RCS or HPI suction.

f. P~ end heat exchangers require corrponent cooling water efter recirculation actuation.

9. Contalrwent Fan Coolers ** Four fan cooler units (three In service)

• 11lnl11U1 of 1 needed for post*accldent heat rel!IOV8l.

b. Fan uni ts shift to low speed on SI st;nel.

c. Coolers require service water.

Table 1.1 Sl.llllllry of design features: Palisades.

1.0-1

...... oll'lul(80f')

T_...

r - , _ - - - - - T e aCSCald 1A91 Hft-*

(0..T~a-.)

LPllUla*A

~

0 I

r-~-----TeaCSO.W Hft*A IApA 00

<JIG*C Palisades Nuclear Plant Schematic FIGURE 1-1

• 11NITIATING EVENTSj TRANS W MC (15.5%)

SBLOCA (29.4%)

~

0 I

'° LOIA (4.3%)

OTHER LOCAs (1.5%)

OTHER TRANS (6.3%)

LOOP (25.5%)

Figure 1-2 IPALISADES !PE

jPLANTDAMAGESTATES I TEJP (18~6%)

.. WCA ~

• • .... TRANSIBNT ...**

• • .. NQ SI INJ ..*

WCA NO ********.....

' NO PORV

/

..........- TRANS

"***.... .......- NOSI SI RBCIRC OTHERS (3.2%) *..*...*

0

....I 0

BEGS (1.4%)

DEJS (2.2%) . TEJW (17 .8%)

OP DEJP (2.6%) OPPSITB POWBR TEJV (13.4%)

Figure 1-3 I PALISADES /PE I

,~,

\

REFUELING POOL ACCESS

.____.TU l"----

B~E_:----...._, ________

ESF SUMP AUXILLARY BUILDINO ESF RECIRCULATION SUCTION PIPE PALISADES REACTOR CAVITY-ENGINEERED SAFEGUARDS SUMP ARRANGEMENT FIGURE 1-5 1.0-11