ML20045H716

Rev 1 to TS Change Request 198 to License DPR-16,changing Drywell Temp to Be Consistent W/Saturation Temp for Proposed Revised Drywell Pressure of 44 Psig

ML20045H716
Person / Time
Site:	Oyster Creek
Issue date:	07/12/1993
From:	J. J. Barton GENERAL PUBLIC UTILITIES CORP.
To:
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ML20045H715	List: ML20045H714 ML20045H716 ML20045H717
References
NUDOCS 9307210173
Download: ML20045H716 (29)

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4 GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION facility Operating License No. DPR-16 Technical Specification Change Request No. 198, Revision 1 Docket No. S0-219 Applicant submits, by this Technical Specification Change Request No.198, Revision 1, to the Oyster Creek Nuclear Generating Station Technical Specifications, a proposed change to page 5.2-1.

By T . J . Ba'~o V e Pr 'd nt and Director 0 r k Sworn and Subscribed to before me this/ 2 day of c 1993.

WA 1- W gNotary Public of NJ JUDITH M. CROWE Notary Public cf NewJersey My Commission Emires - //M/ 9 P l

l 9307210173 930712 ADOCK 05000219 7gjj  ;

PDR p PDR tg j i

'<' ', C321-93-2150 Page R Pursuant to 10 CFR 50.91 (b) (1), a copy of this revision to the Reference-1 change request has been sent to the State of New Jersey Department of Environmental Protection.

Very truly yours j J. J Barto Vice reside t and Director e

JJB/RTZ/ pip cc: Administrator, Region 1 NRC Senior Resident Inspector, Oyster Creek Oyster Creek NRC Project Manager l

t .

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of ) Docket No. 50-219 GPU Nuclear Corporation )

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No..

198, Revision 1, for Oyster Creek Nuclear Generating Station Technical Specifications, filed with the U.S. Nuclear Regulatory Commission on July 12',

1993, has this day of July 12, 1993, been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the United States mail, addressed as follows:

The Honorable Louis A. Amato Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 /

r e im y

J. J. ll aijfon ice P % ident and Director 03 Creek

i' +

0YSTER CREEK NUCLEAR GENERATING STATION

. . FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 198, REVISION 1 Applicant hereby requests the Commission to change Appendix A to the above captioned license as below, and pursuant to 10 CFR 50.92, an analysis concerning the determination of no significant hazards considerations is also presented:

I.0 SECTION TO BE CHANGED Section 5.2.

2.0 EXTENT OF CHANGE ,

Revise Technical Specification 5.2.A and add basis statement.

3.0 CHANGES RE0 VESTED The requested change is shown on attached Technical Specifications page 5.2-1. Related changes to Technical Specifications Bases are required on pages 4.5-12 and 4.5-16. In addition, editorial changes unrelated.to the basis of this request are needed on pages 3.4-8 and 3.5-8. l 4.0 PURPOSE The Oyster Creek drywell internal design pressure is presently 62 psig at a temperature of 175 F. This design pressure value is based on loss-of-coolant accident (LOCA) simulation tests which were conducted to confirm the adequacy of the pressure suppression containment design of the Bodega Bay plant (Ref. 1). However, a comparison of Oyster Creek and Bodega Bay containment design features shows that the Oyster Creek i drywell pressure should be less than that for Bodega Bay (Ref. 6). '

Corrosion in the drywell shell has prompted GPUN to establish an Oyster .

Creek specific design pressure. This new value would be used for any future drywell repair decisions. To develop such a design pressure, state- of-the-art analytical tools were used in conjunction with experimental data. This evaluation includes a recalculation of the reactor vessel blowdown into the drywell as well as the corresponding j containment response.

Reactor vessel blowdown was calculated using both TRACG and RELAP5.

TRACG is a GE computer code that has been qualified for use in evaluating  !

boiling water reactor (BWR) LOCA response (Ref. 8-10). RELAP5 (Ref. 2) is a computer code which has been developed to simulate light water reactor transients as well as large and small break loss-of-coolant accidents. Therefore, both of these computer codes are appropriate for

'his analysis.

The containment response to the reactor vessel blowdown was calculated using M3CPT and CONTEMPT. M3CPT is a GE computer code used to evaluate the short term containment response to a design basis LOCA. It is the I

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.same code that was used to evaluate the Oyster Creek LOCA containment 1 pressure response for the Mark I Long Term Program analysis (Ref. 5).

. CONTEMPT (Ref. 3) is a nuclear reactor containment analysis code which is used to evaluate pressure temperature response to mass and energy inputs (blowdown of reactor vessel). These containment codes in conjunction with the vessel blowdown results provide a complete method for establishing an Oyster Creek specific design pressure.

The results of this evaluation show that the peak drywell pressure following a design basis loss of coolant accident (DBLOCA) is 38.1 psig with a corresponding saturation temperature of 285 F. To establish a '

design pressure value, an additional 15% margin is added to give a value of 44.0 psig. The drywell temperature which corresponds to this pressure e

is 292 F.

In addition to the drywell pressure and temperature change, unrelated revisions are needed to two Bases pages. The last paragraph in Section 3.4 Bases on Page 3.4-8 is revised to clarify that the containment spray system may be inoperable when primary containment integrity is not '

required. Also, on page 3.5-8 of Section 3.5 Bases, an editorial change is necessary to properly state the 2 psig external design pressure of the drywell .

5.0 ORIGINAL DESIGN PRESSURE ,

The Oyster Creek drywell design pressure was originally established at 62 psig. This pressure was first established as a design value for the Bodega Bay plant and later specified for Oyster Creek. The Bodega Bay <

design pressure value is based on LOCA simulation tests. These tests were conducted to confirm the adequacy of the pressure suppression containment design of the Bodega Bay BWR, The tests showed that the maximum drywell pressure for those tests which were representative of the Bodega Bay design was 52 psig. An additional 10 psi was added for margin when establishing the Bodega Bay design pressure of 62 psig.

This value was assigned to Oyster Creek even though there are major differences in the design of the two plants. The differences between the i plants are such that the peak drywell pressure for Oyster Creek is less i than that for Bodega Bay. The Oyster Creek Updated Final Safety Analysis Report (FSAR) correlates the Bodega Bay test values for peak drywell pressure as a function of the ratio for drywell to wetwell-vent area to break area (Ref. 6, Fig. 6.2-6). This particular plant parameter plays a key role in determining the peak drywell pressure. The larger this ratio is, the greater the impact of the suppression system on reducing peak drywell pressure. This correlation produced an estimate of the peak Oyster Creek drywell pressure to be 37 psig.

Additionally, the Oyster Creek FSAR (Section 6.2.1.3) presents calculated results of the Oyster Creek response to a DBLOCA. The FSAR states that '

this model tends to overpredict maximum containment pressures when compared with Bodega Bay and Humbolt Bay pressure suppression tests (Ref. 6, Figs. 6.2-8 and 6.2-9). The result of this analytical model when applied to Oyster Creek was a peak drywell pressure of 33 psig.

Both of these peak drywell pressures presented in the FSAR are less than the 52 psig value established for Bodega Bay. It was thus previously recognized that the 62 psig design value was significantly larger than that which would be adequate for Oyster Creek.

, . 6.0 RE-EVALUATION OF THE DRYWELL DESIGN PRESSUR_E

'To establish an appropriate design pressure and corresponding l

. , temperature for the Oyster Creek drywell, it is necessary to simulate containment response to the DBLOCA. For peak drywell pressure, this is the double-ended guillotine break of a recirculation loop pipe. The simulation must, therefore, include a reactor vessel model of this I accident.

In addition, it is necessary to simulate the Mark I pressure suppression '

containment. From this simulation, the drywell pressure response to the double-ended guillotine break can be determined.  ;

6.1 Methods In order to simulate the peak drywell pressure for Oyster Creek, four computer codes were used (refer to Fig. 1). The first two of these include the GE BWR version of TRAC (TRACG, Refs. 8, 9,10) and RELAPS MOD 3 (Ref. 2). These codes were independently used to calculate the Oyster Creek reactor vessel blowdown for the DBLOCA. .

The results from these analyses were then used as input to a second set of codes. These codes were used to evaluate the drywell pressure and temperature response to the blowdown. The first of. l these, M3CPT, (GE containment code) was used to predict the containment response to the TRACG blowdown. The second code used in the containment analysis is CONTEMPT /EI28C. CONTEMPT was used to evaluate the containment response to both the RELAP5 and TRACG '

blowdowns.

6.2 TRACG Best Estimate Vessel Blowdown Model TRACG was used to establish a best estimate blowdown for Oyster Creek. A multi-node reactor vessel model was developed for Oyster Creek as part of the revised 10 CFR 50 Appendix K analysis program.

This same model was used for this evaluation (Ref. 4). The nodalization was performed with vessel geometry as well as governing phenomena in mind.

The initial conditions assumed (Table 1) are the same as those used for the OC Mark I long Term Program containment analysis (Ref. 5).

They are also consistent with how the plant is currently operated. i The resulting mass and energy release are provided in Table 2 and Figure 2. This represents the TRACG best estimate blowdown. ,

In order to insure confidence in the TRACG break flow model, a comparative analysis was performed. This analysis (Ref. 4) compared the TRACG best estimate results with a number of actual blowdown  :

tests. The tests included:

• Simpie vessel blowdown tests (PSTF)

• Scaled integral BWR tests (TLTA, FIST, FIX II)

Full size reactor vessel tests (Marviken)

Each blowdown was divided for analytical purposes, into two regimes.

The first regime represents the period during which the vessel conditions at the break are subcooled liquid. The second represents

i l' ',

a two-phase condition at the break. An average break mass flow rate '

was obtained in each flow regime for the test data and the TRACG prediction. From these values, a break flow multiplier was developed for each regime. The multipliers were defined to be the ratio of the measured to predicted average break flow rate. The maximum multiplier (multiplier-1.25) was then applied to the besi.-

estimate TRACG braak flow from the Oyster Creek LOCA analysis-(Table 2). .This produced what will be referred to as the TRACG best estimate blowdown for Oyster Creek. It should be noted that this multiplier will increase the total mass and energy into the containment by 25%. This is not physically possible since the source of this mass is the reactor vessel which transports a clearly defined quantity of mass and energy through the break. The multiplier addresses uncertainty of TRACG prediction in the rate of transport only.

6.3 RELAPS Best Estimate Vessel Blowdown Model To independently confirm the TRACG blowdown results, a RELAPS MOD 3 blowdown model was developed. This model was nodalized with the same considerations as those used for the TRACG model. The blowdown results are provided in Figure 3 and Table 3. This will be referred to as the RELAPS best estimate blowdown. A graphical comparison with TRACG is shown in Figure 4. The results presented in Figure 4 show that TRACG predicts a somewhat higher peak flow rate out of.the vessel. However, the RELAP5 code predicts a larger initial rate of

• change of blowdown flow. Therefore, the impact on the containment response is expected to be different.

The RELAPS blowdown model was compared with actual test data from  ;

Marviken tests (Ref. 7). This comparison was used to establish a multiplier for the RELAPS blowdown. As was the case for the TRACG blowdown, the multiplier addresses uncertainty in the rate of break '

mass flow rate only. As a result, the integrated mass and energy into the containment model is in excess of what would actually occur. For the RELAPS blowdown, this multiplier is conservatively set at 1.30.

6.4 Containment Model (M3CPT)

The containment response to the TRACG blowdown was evaluated with the GE code M3CPT. This code is used to evaluate short term DBLOCA response of the containment. M3CPT was used in the evaluation of the Oyster Creek LOCA containment pressure response for the Mark I long Term Program analysis (Ref. 5).

Three separate cases were run using this code. The initial conditions for these cases are provided in Table 4. Case 1 is the same set of conditions used in the Mark I Long Term Program analysis. The vent system downcomers are assumed to be submerged 4.06 feet (Ref. 5) below the suppression pool surface. This value corresponds to the highest water level allowed for continuous operation of the plant. This assumption increases the drywell pressure required to clear the vents and thus increases the peak ,

pressure. It is also consistent with the volume assumed for the '

torus vapor space. This volume is minimized by setting the l l

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. suppression pool at the high level. The non-condensibles which are swept into this air space will produce a higher torus back pressure because of the smaller available volume. Finally, a zero pressure differential'between the drywell and the wetwell is-assumed. This <

assumption is consistent with allowed plant operation. In. addition, j this will contribute to the maximum water leg length inside of the -

vent's downcomers. With the drywell at an initially higher pressure than the wetwell, water in the downcomer will be partially forced-out of the pipe. This reduces the water leg inside of the downcomer. With a zero pressure differential, water will not be forced out of the pipe.

Case 2 is a variation of Case 1. The only difference is that the initialcontainmentpressureisincreapedto16.1psig,andthe wetwell air space is reduced by 400 ft . The increased pressure corresponds to the highest operating pressure expected (high drywell pressure alarm setpoint) under normal conditions for Oyster Creek.

(This maximizes the mass of non-condensible gases in the containment). The reduction of the air space volume is'an added conservatism.

Both Cases 1 and 2 are run using the best estimate blowdown calculated by the TRACG computer model. Case 3 is identical to Case 2 except that 1.25 times (refer to Section 6.2 discussion of i multiplier) the TRACG best estimate blowdown is used. i The results of these cases (Ref. 4) are provided in Figures 5 through 7. The peak calculated pressures are provided in Table 5.

6.5 CONTEMPT CONTAINMENT MODEL The CONTEMPT computer code was used to evaluate the M3CPT results. .

This is accomplished by running the three previously described cases l with the TRACG blowdown. The results are then compared with those i calculated by M3CPT. In addition, the code was used to compare the impact of the different blowdown models on the containment response. ,

This is accomplished by running the three cases previously described R with each blowdown. The CONTEMPT results were then compared for each blowdown.

The results obtained using the TRACG blowdown (Figs. 8 to 10) show good agreement with that calculated by M3CPT (Figs. 5 to 7). A graphical comparison shows that both models exhibit similar pressure profiles. This indicates that the containment's pressure suppression phenomenon is modeled properly. The peak pressures are compared with M3CPT in Table 6. The comparison shows that CONTEMPT calculates a slightly lower pressure than M3CPT. It is concluded-that the models are in good agreement.

The same three cases were run using the RELAPS blowdown results, however, a 1.3 multiplier was used (refer to Section 6.3). As can be seen from figures 11 to 13 (Table 7), the peak pressure occurs.

somewhat earlier than for the TRACG blowdown. This is a result of the containment's dynamic response to the different blowdowns depicted in Figure 4.

l

l The peak drywell pressures calculated using these different methods are in good agreement and confirm what was described in the FSAR

. . (discussed previously). It is concluded that following a DBLOCA, the peak drywell pressure will not exceed 38.1 psig with a corresponding saturation temperature of 285'F. Therefore, after applying a 15%

margin, the design pressure for the OCNGS drywell can be adequately established at 44.0 psig with a corresponding saturation temperature l of 292 F. l l

7.0 DETERMINATION We have determined that the proposed Technical Specification change -

involves no significant hazards considerations as discussed below.

1. The change will not involve a significant increase in the probability or consequence of any accident previously evaluated.

The change in drywell design pressure and. corresponding temperature j has no effect on the probability of~ loss of coolant accidents which the containment is designed to help mitigate. The consequence of the design basis LOCA is not changed since adequate structural _;

integrity is maintained. The drywell design pressure and ,

corresponding temperature ~ values of 44 psig and 292 F are-greater.

than the calculated peak pressure and temperature of 38.1 psig and 285*F.

2. The proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

The primary containment functions to minimize the release of radioactive materials during a loss of coolant accident. The change >

in drywell design pressure and corresponding temperature will -l continue to ensure this function is maintained. Since the containment mitigates not initiates LOCAs, new or different accidents are not created.

3. A significant reduction in margin of safety is not involved. r The margin of safety for drywell structural integrity is based upon compliance with ASME code limits at a given design pressure and -

corresponding temperature. The drywell design pressure change to 44 psig at 292 F maintains the margin of safety since the vessel will still be required to comply with ASME code limits. The change in design pressure reflects a reduction in uncertainties and conservatisms which resulted in the design pressure of 62 psig at 175 F. Therefore, it is concluded that the drywell design pressure and temperature corresponding change will not reduce the margin of safety.

8.0 CONCLUSION

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~

The peak drywell pressure is calculated conservatively to be 38.1 psig. i This value should be used with an additional 15% margin added to give a drywell design pressure of 44 psig with a corresponding saturation '

temperature of 292 F. The change in design pressure and corresponding l temperature will not impact plant safety following any design basis l accident. '

9.0 REFERENCES

l

. . 1. " Preliminary Hazards Summary Report, Bodega Bay Atomic Park Unit No. 1", Docket No. 50-205, December 28, 1962.

2. NUREG/CR-4312 EGG-2396 - Appendix A RELAPS Input Data Requirements i Prepared for Release of RELAP5 M003 EG&G Idaho Inc., Idaho Falls, ID, January 1990.

3. CONTEMPT EI/28C - A Computer Program for Predicting Containment Pressure - Temperature Transients.

4. GENE 770-07-1090 February 1991, '0yster Creek LOCA Drywell Pressure Response'.

5. NED0-24572, '0yster Creek Plant Unique ~ Load Definition, July 1982'.

6. GCNGS FSAR

7. Intermountain Technologies Division SAIC, ' Transmittal of RELAP5-MARVIKEN Comparison Summary', DCS-636-90 from Don Slaughterbeck to N. Trikouros.

8. NUREG/CR-4127-1, EPRI NP-3987-1, GEAP-30875-1, "BWR Full Integral Simulation Test (FIST) Program: TRAC-BWR Model Development, Volume 1

- Numerical Methods", July 1985. ,

9. NUREG/CR-4127-2, EPRI NP-3987-2, GEAP-30875-2, "BWR Full Integral

• i Simulation Test (FIST) Program: TRAC-BWR Model Development, Volume 2

- Models", August 1985.

10. NUREG/CR-4127-3, EPRI NP-3987-3, GEAP-30875-3, "BWR Full Integral Simulation Test (FIST) Program: TRAC-BWR Model Development, Volume 3 ,

- Developmental Assessment", September 1985.

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,,. ,, TRBLE 1 1

IRACG/RELAP5' INITIAL CONDITIONS TP.ACG RELAPS Reactor Power (% Rated)- 102 100 Dome Pressure (PSIA) 1035 1035-Reactor Core Flow (MLB/HR) 61.0 55.56

• Steam Flow (MLB/HR) 7.395 7.506 Core Inlet Temperature (*F) 525.0 512.0 Feedwater Temperature (*F) 317.0 316.5 Break Area (FT )

Vessel Side Area 3.109 3.11 (limiting flow area just upstream of break)

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Pipe Side Area 3.149 at break 3.11 at break-1.547 at flow venturi- 1.55 at flow venturi

• Does not include the core bypass flow.

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, . . TABLE 2

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TRAOG BEST-ESTIMATE BREAK MASS FLOW RATE (WITMOUT MULTIPLIER) AND ASSOCIATED ~

r ENTHALPY FOR INPUT TO M3CPT05 AND CONTEMPT CONTAINMENT RESPONSE ANALYSIS-.

TIME (SEC) MASS FLOW RATE (LBM/SEC) ENTMALPY (BTU /LBM) i 0.0 0. 519.3 0.49365 42045 518.1 t

0.65958 40896 518.2 0.85506 41883 518.1 -

1.073 40381 518.2 1.671 38102 ~518.1' 2.1645 34566 518.1 2.9523 29654 518.9 ,

3.9009 24818 520.8 5.1141 23000 523.9 5.9771 22997 525.8 .

6.9483 22553 526.8 ,

7.9597 22133 527.7 9.0245 21335 529.9 9.9242 20052 534.7- [

12.137 15376 572.1 14.937 9476 745.6 20.936 6945 714.4 25.192 4165 832.4 30.0 854 871.2 NOTE: M3CPT05 ACCEPTS 20 BREAK MASS FLCW RATE AND ENTHALPY POINTS.

TABLE ?t

.PELAPS BREAK FLOW RATE IWITHOUT MULTIPLIER) AND ASSOCIATED ENTHALPY FOR INPUT TO CONTEMPT CONTA7NMENT RESPONSE ANALYSIS 1 i TIME (SEC) MASS FLOW RATE (LBM/SEC) ENTHALPY (BTU /LBM) 0 0 0 0.1. 23970 527.7 I 0.2 31710 528.9 .

0.3 30110 525.9 0.4 31640 526.7 0.6 32860 527.6 1.1 32930 528.2 1.3 33300 529.6 1.5 32580 527.4 .

2.1 31670 530.4 2.6 30060 534.5 3.2 -28380 534.6  :

3.6 27700 538.8 4 27100 540.5 4.7 25900 544.3 5 25070 547.5-6 22360 544.6 7 19670 550.3 8 17550 553.5 9 16520 553.7 10 15890 548.1 11 15380 545.4 12 14750 547.8 l 13 14000 558.3 14 11930 596.2 15 11570 583.9 16 11200 586.2 17 10450 581.5 19 8612 608.1 19 7498 614.1 20 6656 625.8 21 5924 638.2 22 5376 639.4 23 4946 631.4 24 4469 632.9 25 4006 640.6 26 3471 672.2 27 3078 687. '

28 2768 692.4 29 2681 652.9 30 2111 768.5 l

i

, . , TABLE 4

• -* KEY CONTAINMENT PARAMETERS  ;

CASE 1 CASES 2 & 3 ,

1. WETWELL AIRSPACE AND SUPPRESSION POOL r i

Wetwell Airspace Free Volume (FT ) 121,400 121,000 Initial Wetwell Airspace Pressure (PSIA) 14.7 16.1 Initial Wetwell  ;

Airspace Temperature (*F) 77.5 77.5 Initial Wetwell Airspace Relative Humidity (%) 100 100 t Suppression Pool Volume at HWL (FT ) 92,000 92,000 Initial Suppression Pool Temperature (*F) 77.5 77.5

2. DRYWELL AND VENT SYSTEM ,

3 180,000 180,000 Drywell Free Volume (FT ) ,

Initial Drywell Pressure (PSIA) 14.7 16.1 Initial Drywell Temperature (*F) 135.0 135.0 Initial Drywell Relative l Humidity (%) 20.0 20.0 Number of Downcomers 120 120 i

Inside Diameter of Each Downcomer (FT) 1.958 1.958 Downcomer Submergence 4.06 4.06 Total Downcomer Loss Coefficient (Including entrance, exit, turning and friction losses) 5.06 5.06 i

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TABLE'S ,

. 4

SUMMARY

OF PEAK DRYWELL PRESSURES r

TRACG BREAK FLOW CASES /M3CPT CONTAINMENT MODEL

. PEAK DRYWELL FLOW RATE PRESSURE TIME CASE ?to. MULTIPLIER (PSIG) (SEC) ,

1 1.0 30.5 -9.5 2 1.0 32.9 9.7 3 1.25 -

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FLOW RATE CASE MULTTPLIER M3CPT/TRACG CONTEMPT /TRACG  !

1 10 30.5 paig / 9.5 sec 28.8 psig / 8.9 sec 2 1.0 32.9 psig / 9.7 sec 31.4 psig / 9.2 sec  ;

3 1.25 38.1 psig / 3.0 sec 36.8 poig / 3.1 sec 9

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i FLOW RATE FLOW RATE CASE MULTIPLIER TRACG MULTIPLIER PELAP5  !

1 1.0 20.8 psig/8.9 see 1.0 29.6 psig/5.2'sec.

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