ML20148D134

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Forwards Addl Info Re Exemptions from Requirements Previously Unacceptable Covering Personnel Airlock,Bellows Leakages & Water Testing for Feedwater Check Valves.W/ Oversize Drawings
ML20148D134
Person / Time
Site: Cooper 
Issue date: 10/30/1978
From: Pilant J
NEBRASKA PUBLIC POWER DISTRICT
To: Ippolito T
Office of Nuclear Reactor Regulation
References
TAC-01909, TAC-11040, TAC-1909, NUDOCS 7811020253
Download: ML20148D134 (11)
v • d • e


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THIS DOCUMENT CONTAINS P00R QUAUTY PAGES

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GENER AL OFFICC wf

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October 30, 1978 Director, Nuclear Reactor Regulation Attention : Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Appendix J Exemption Request / Additional Information Cooper Nuclear Station NRC Docket No. 50-298, DPR-46

Dear Mr. Ippolito:

A letter dated September 16, 1977 from V. Stello to the Nebraska Public Power District transmitted Amendment 38 to the facility Operating License for Cooper Nuclear Station.

This amendment consisted of changes to the Technical Specifications relating to exemptions from the requirements of 10 CFR Part 50, Appendix J.

Originally NPPD had requested five exemptions of which three were found acceptable to the Commission.

This letter provides additional information relating to the two exemptions which were not acceptable.

Exemption 1:

"The personnel airlock door would be tested at intervals no longer than one year at 58 psig (Pa) and at 3 psig after each opening during the one year interval between the 58 psig cests."

In Mr. Stello's letter of September 16, 1977, the following additional information was requested:

1.

Acceptance criteria for the reduced pressure tests which correlate the personne! airlock ieakage rate at 3 psig to the leakage rate which would be expected at 58 psig.

2.

Acceptance criteria which relates bellows leakage rate at 5 psig test pressure to the leakage rate which would be experienced at 58 psig pressure.

The acceptance criteria for both the personnel airlock and bellows leakages are taken from ASME Section XI, Winter 1976 Addendum, Article IWV-3000 " Test Procedures", paragraph IWV-3420 which states:

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Mr. Thomas Ippolito October 30, 1978 Page 2 "When leakage tests are made in such cases using pressures lower than function maximum pressure differential, the observed leakage shall be adjusted to - function maximum pressure differential value by calculation appropriate to the test media and the ratio between test and function pressure differential assuming leakage to be directly proportional to the pressure differential to the one-half power."

In our letter of September 10, 1975 to Mr. K. R. Goller, NPPD requested an exemption to allow conducting the airlock integrated leak tests at one year intervals rather than at the 6 months intervals required by Appendix J.

The Staff's safety evaluation transmitted September 16,1977 9tated that insufficient justification was provided by NPPD in support of a yearly test interval.

Although our September 10 letter stated that the airlock would be tested at 58 psig yearly, and at 3 psig after each opening, we neglected to specify that the containment airlocks would be leak tested at least every 6 months at 3 psig. The justification for not performing the 6 month test at the full pressure of 58 psig remains as stated in our September 10, 1975 letter.

No changes are contemplated from the existing testing requirements.

Exemption 2:

"The feedwater check valves would be tested with water rather than air or nitrogen."

In our letter of September 10, 1975, we also requested an exemption from Appendix J requirements so that Cooper Nuclear Station could continue to leak test the feedwater check valves using water rather than air or nitrogen.

In the Commission's letter of February 17, 1977, NPPD was requested to demonstrate that the feedwater check valves would remain filled with water during and after a postulated loss of coolant accident and that the fission products intrained in the liquid leakage would not result in additional radiological dose such that the total accident dose would exceed 10 CFR Part 100 guidelines.

Enclosed please find the results of an analysis performed to demonstrate the above.

Should you have any questions or require additional information, please contact me.

In addition to one signed original, 39 copies are also submitted for your use.

Sincerely yours, W4 M._hilant Director of Licensing and Quality Assurance IMP /jw:str24/8 Enclosure

Mr. Thomas Ippolito October 30, 1978 Page 3 STATE OF NEBRASKA )

) ss PLATTE COUNTY

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Jay M. Pflant, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this information on behalf of Nebraska Public Power District; and that the statements in said application are true to the best of his knowledge and belief.

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May M. Pilant Subscribed in my presence and sworn to before me this 1 day of October, 1978.

/ NOTARY PUBLIC/

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My Commission expires e StuttAL 50fMT. StMe et lettssas I

MARLYN R. H0HNDoM d

a # ley Cases.14. Ost 14,1000

APPENDIX J FEEDWATER CHECK VALVE ANALYSIS Purpose 1.

To demonstrate that the feedwater system checkvalves and the HPCI and RCIC-to-FW system check valves remain covered with water during and after a LOCA.

2.

To evaluate.the radiological doses resulting from feedeater check valve leakage after a LOCA and determine if 10 C1'R 100 guidelines are exceeded.

Conclusions 1.

The feedwater system check valves will remain covered with water.

2.

The feedwater system check valve leakage following a LOCA will not result t

in 10 CFR 100 guidelines being exceeded.

3.

The 30 day thyroid dose at the low population zone (LPZ) is the limiting factor associated with valve leakage following a LOCA.

4.

The 30 day thyroid dose resulting from a liquid release at the 100 m elevation, even assuming an unlimited leak rate at time equal to zero, will not exceed the allowable dose margin.

(See Figure 2)

Assumptions and Data Input 1.

Power level 2381 x 1.05 = 2500 MWt (FSAR) 2.

Activity concentrations in suppression pool:

50% of Halogens 0% of Noble Gases 3.

Suppression pool liquid volume:

87,650 ft3 (FSAR) 4.

10% of Liquid flashes to steam:

(NRC Standard Review Plan 15.6.5) l 5.

No mixing credit 6.

Filter efficiency = 99% (FSAR) l 7.

Stack height = 100 m I

8.

Activity concentration in drywell (US NRC Regulatory Guide 1.3) 25% of Halogens 100% of Noble Gases 9.

Drywell volume 145,320 ft3 (FSAR) l l

.. _ -. _ = - -.

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. Assumptions and Data Input (continued) 3 10.

X/Q Values (sec/m )

Stack Release Ground Release 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> @ EA 2.76x'10[6

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2.38 x 10 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> @ LPZ 5.68 x 10

1. 6 x 10,4

-6 1.02 2 10 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> @ LPZ 4.47 x 10

-5 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> @ LPZ 1.40 x 10,6 1.70 x 10 7

-6 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> @ LPZ 7.08 x 10,7 6.70 x 10 -6 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> @ LPZ 2.69 x 10 3.38 x 10 (EA = Exclusion area or site boundary, LPZ = Low Populatica Zone) 11.

LOCA Radiological ~ Doses from containment leakage, Rem (FSAR Table XIV-9-1.)

EA LPZ Whole Body 0.4 1.06 Thyroid 31.3 89.2 12.

Mixing in the feedwater line does not occur until the original volume of water above the check valve has leaked out.

13.

Dose margin allowable for valve leakage:

300 - 89.2 = 210.8 Rem; 30 day thyroid dose at LPZ Discussion 1.

Feedwater Check Valves Remain Covered A calculation was performed to give a basis for using engineering judge-ment in determining whether air or water remains on top of the.feedwater system check valves following a LOCA.

The calculation assumed all sensible heat in the pipe is transferred to the fluid and added the change of internal energv of the fluid going from liquid to vapor as a result of the j

depressurization.

The following iniormation was developed (See attached piping drawings 2509-1 and 2509-2, l

243 Ft 12" Nom OD Pipe wall t = 1.125

' 56 Ft 18" Nom OD Pipe wall t = 1.562 Carbon Steel Pipe Vol. of ga'ter contained between vessel nozzle and first check valves

= 214 Ft 420 F = Initial FW Temp = Initial Wall Temp The conclusion that water remains in the pipes on top of the feedwater system check valves is based on four considerations.

First, calculation.

shows that 7333 lbm of the original 11298 lbm is vaporized leaving a total iG t

s of 3965 lbm of liquid above the first feedwater check valves.

Second, the feedwater check valves are at the lowest point in the portion of the system being considered.

Third, no consideration was given to additional water in the system due to ECCS (HPCI and RCIC) initiation or the delay in loss of feedwater flow associated with the reactor levei 2 isolation signal which essentially trips the feedwater pump turbines some time after the LOCA initiation.

Finally, the superior sealing qualities of the tilting disc type feedwater check valves were not considered as the full acci-dent pressure causes the valve discs to slam shut and ride up the inclined seating surface to improve the seal (see attached drawing 920-3).

No credit is taken for the ability to isolate a leak in the feedwater system from the control room at the feedwater pump discharge valve (including the discharge valve bypass valve). The potential for failure of non-Class I feedwater piping at Cooper Nuclear Station is considered to be negligable due to 100% radiographic inspection of this piping after installation and the continued inspection of all welds back to the feedwater pump discharge valves as pert the approved ISI program.

The same conclusion applies to HPCI and RCIC-to-feedwater system check valves since they are outboard and at a lower elevation than the feedwater check valves addressed above.

2.

Off Site Dose Determination The allowable leak rate for water is determined by assuming the allowable dose equals 300 Rem minus the dose from normal containment leakage as defined in FSAR Table XIV-9-1.

Consistent with the FSAR, all combinations of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 day allowables were evaluated for whole body and thyroid at both site boundary and low population zone (LPZ).

The results of the study showed the 30 day thyroid dose to be controlling at the LPZ for all leaks starting at time = 0 and later.

As illustrated in assumption 13 the resultant dose marF n allowable from valve leakage is 210.8 i

above, Rem.

Leak rates vs leak time are plotted in Figure 1 for liquid ground release (no SGTS Filtration).

The curve shews a leak rate which will result in 210.8 Rem thyroid for a 30 day LPZ release.

The time on the graphs is the time at which the valve starts to leak fission products developed during the LOCA.

Following depressurization there remains 3357 lbm of liquid above the second feedwater check valve in each feedwater line.

Pasttestingojthe feedwater check valves at CNS give an average leakage rate of 8.3 ft /hr.

At containment accident pressure it would take 421 minutes for this original nonradioactivevolumeofliquidtogeakfromthecheckvalves. Figure 1 shows that at a leak rate of 8.3 ft /hr., fission product leakage would be required to start at 228 minutes to exceed the dose margin allowable.

Leakage rates through the smaller HPCI and RCIC-to-feedwater system check valves would be considerably lower than through the feedwater check valves and would therefore result in an offsite dose much less than that considered above.

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