ML20011D401

Proposed Tech Specs Assuring Allowed Leakage from RHR Sys Does Not Adversely Affect Control Room Habitability

ML20011D401
Person / Time
Site:	Fort Calhoun
Issue date:	12/20/1989
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20011D399	List: ML20011D398 ML20011D400 ML20011D401
References
NUDOCS 8912270123
Download: ML20011D401 (8)

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.. %Fa5Gb" 13.0 SURVEILLANCE REOUIREMENTS 1

3.16 Recirculation Heat Removal System Intearity TestiDS Anolicab111tv B Applies to determination of the integrity of the shutdown cooling system and associated components.

Ob.iective f

To verify that the leakage from the recirculation heat removal system components is within acceptable limits.

Sgri;ifications

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(1) 4. 1ht artion ?h hutdown cooling system that is outside the containment shul , tested at 250'asig or a refueling N{J

• interval,

b. 51,ing f rom valves HCV-383-3 a .d W 383-4 to the discharge 5 ilation valves of the safety 1& Nn pumps and containment

..? pray pumpr shi.11- be hvdrostaticar ested at no less than 100 psig A.t the-test.ing frequency spe W esd in (1)a. above.

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c. Osual inspection of the system's cceponents shall be performed et the frequency specified in (1)a. ibove to uncover any -

significant leakage. The leakage shall be measured by colled nor ad weighing or i>y any other equivalent method.

(2) a. The maximum allowable leakage from the recirculation heat removal system's components (which include valve stems, nges, and pump seals) to atmosphere shall not exceed 1243 flg/hr cm under the normal hydrostatic head from the SIRW tank.

b. Repairs shall be made as required to maintain leakage within the acceptable limits.

Basis The limiting leakage rates from the shutdown cooling system are judgment values based primarily on assuring that the components could operate without mechanical failure for a. period on the order of 200 days after a design basis accident. The test pressure (250 psig) achieved either by ncrmal system operation or by hydrostatic testing gives an adequate marginoved)thehighestpressurewithinthesystemafteradesignbasis accident. Similarly, the hydrostatic test pressure for the return lines-from the containment to the shutdown cooling system (100 psig) gives an adequate margin over the highest pressure within the lines after a design basis accident.

3-84 8912270123 891220 PDR ' ADOCK 05000295 P. PDC

.. DECff d p 3.0 $1!Ryi}LLANCE REQUIREMENTS .

.3.16 Recirculation llent Removai System Inteority Testino (Continued)

A shutdown coolin system leakage of 12433 cm /hr will limit off-site exposures due.to eakage to insignificant levels relative to those calculated for direct leakage from the containment in the design basis .

accident. The safety injection system pump rooms are equipped with individual charcoal filters which are placed into operation by means of-

~ switches in the control room. The radiation detectors in the auxiliary building exhaust duct are used to detect high. radiation level. Leakage

.to the regenerant spent safety injection tanks. (gystem pump room l Additional sumps makeup water willtobe the returned to the containment sum) inventory can te readily accommodated via the charging pum)s from.

eit1er the SIRW tank or the concentrated boric acid storage tan (s.

In case of failure to meet the acceptance criteria for leakage from the shutdown cooling system or the associated components, it may be possible to effect repairs within a short time. If so, it is considered unnecessary and unjustified to shutdown the reactor. The times allowed for repairs are consistent with the times developed for other engineered safeguards components.

References (1) USAR,' Section 9.3 i

(2) USAR, Section 6.2 l

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DESCRIPTION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATIONS Description of amendment reauest:

The current Technical Specification allows a maximum leak rate of one gallon per minute (1 gpm) for the Residual Heat. Removal-(RHR) system. Yet, if a design basis LOCA would occur, the Thyroid dose to control room operators would exceed GDC-19 limits as further defined by SRP 6.4, even if the proposed ventilation filters design basis of 99% efficiency was met. Therefore, to achieve maximum protection to the operating staff and validate the design ba::is for the control room air filter system modification, a maximum RHR leak rate of 1243 cc/hr must be incorporated into Section 3.16 of the Technical Specifications. The Technical Specification change will close out item 5 of Safety Analysis for Operability (SAO) # 89-02.

This change is proposed to assure allowed leakage from the Residual Heat Removal System does not adversely effect control room habitability.

Surveillance Requirement 3.16(2)a. now allows a leakage rate of one gallon per minute. This value is to be changed to 1243 cc/hr, based on an analysis by Combustion Engineering l ; this is the maximum expected from the safety injection system at Fort Calhoun.

I The current Technical Specification allows a maximum leak rate of one gallon per minute (1 gpm) for the Residual Heat Removal system.

Justification:

1 An analysis completed by Combustion Engineering shows the maximum leakage expected from the safety injection system at Fort Calhoun Station is 1243 ,

cc/hr.  ;

1 Documentation from Surveillance Tests (ST-RHR-1 & 2) has shown minimal leakage I for the past ten years.

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l Changing Section 3.16 of the Technical Specifications would decrease the I radiological significance to the genegal public and control room operators. An analysis completed.by Stone & Webster shows the integrated doses to the general public and control room staff is within regulatory requirements.

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1 Leakage From Engineered Safety Features (ESF) Components' Outside Containment (0-PEC-135 Rev. 0); p. 81-88 2- Control Room Habitability Evaluation for NUREG-0737, Item III.D.3.4, November 1989.

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NO SIGNIFICANT HAZARDS CONSIDERATIONS The proposed change to the Technical Specificaticns does not involve a significant hazards considerations because the operation of Fort Calhoun in accordance with this amendment would not:

1. Involve a significant increase-in the orobability or consequences of an ,

accident previously evaluated. The proposed amendment does not involve a modification to nuclear systems. It does involve a change in .

i Surveillance Testing requirements in that the allowable leakage from the Residual Heat Removal system is decreased.. This results in a lower potential dose to control room operators in the event of a design basis accident. Therefore, this change does not increase the probability or ,

consequence of a previously evaluated accident.

i 2. Create the possibility of a new or different kind of accident from any ,

l accident previously evaluated. It has been determined that a new or

. different kind of accident is not created, since no new or different moda L of operation is proposed for the plant. The continued use of the '

L existing Technical Specifications' administrative controls prevent the possibility of a new or different kind of accident.

3. Involve a significant- reduction in a margin of. safety. The proposed change to-the Technical Specifications reduces the maximum allowable-leakage from the Residual Heat Removal system to 1243 cc/hr. Therefore, the. radiological consequence to the plant staff and the general public is L reduced. The potential doses are well within the requirements of 10 CFR 100.11, and does not increase the Thyroid dose to the general public, thus, increases the margin of radiological safety to both plant staff and the general public.

Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration as defined by 10CFR50.92 and the proposed changes will not result in a condition which significantly alters the impact of the station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion sst forth in 10CFR51.22(e)(9) and pursuant to 10CFR51.22(b) no environmental impact or environmental assessment need to be prepared.

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LEAKAGE FROM RHR SYSTEM W#iMnB4%isST-RHR11% # { WMW@w'?WST-RHR;2 < 5%!?p#P

%DATE05 N WLlikAGE P CTT ' "TFR -! MDATEP: 3ItEAKAGE ' ;5TJ @FRi.

i 31-OcT-77 2 DEoPS/SEc 406 1415 1-OcT-77 NoNE 404 1839 21-Nov-78 52 ML/ MIN 398 1343 18-SEP-81 NoNE 530 775 8-FEs-83 NoNE- 775 719 4-MAR-83 NoNE 775' 738 2-MAY-84 NoNE 778 273 4-MAR-84 0.33 ML/ MIN 778 285 27-FEs-80 81 DROPS / MIN 531 1261 20-JAN-80 NoNE 531' 1280 31-DEc-85 0.2 DROPS / MIN 1049 787 1-OcT-85 NoNE 1049 794 13-APR-87 NoNE 1605 18178 11-MAR-87 NoNE 1605 38 l'

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BEFORETHEUNITEDSTATES NUCLEAR REGULATORY COMMISSION

.i In the Matter'of ) i

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Omaha Public Power District ) Docket No. 50-285-(Fort Calhoun Station-- -)-

Unit-No.1) )

i APPLICATION FOR AMEhW.ENT e OF

' OPERATING LICENSE

- Pursuant:to Section 50.90 of the regulations of the U. S. Nuclear Regulatory Connission ("the Commission"), Omaha Public Power District, holder

?of Facility Operating License No.-DPR-40, herewith requests that Section 3.16.  !

7 0f the; Technical Specifications set forth in- Appendix A to that License be'  ;

amended to assure that allowed leakage rate does not adversely affect control-  !

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room habitability.  ;

j The proposed changes in Technical Specifications are discussed in Attachment Atto this A:elication. A discussion, Justification and no -l Significant Hazards Consideration Analysis, which demonstrates that the-

proposed changes do not involve significant hazards considerations, is appended i

'1 in Attachment _B. The. proposed changes in specifications would not authorize ~  ;

any change in the types or any increase in the amounts of effluents or a change i Lin thelauthorized power level of the facility. - i WHEREFORE, Applicant respectfully requests that Sections 3.16 of Appendix

' A to Facility Operating License No. DPR-40 be amended in the form attached

.. - hereto as Attachment A.

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l A copy of this Application, including its attachments, has been submitted to the Director - Nebraska State Division of Radiological Health, as required' n"" t' by.10 CFR 50.91.

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OMAHA PUBLIC POWER DISTRICT 1

By / .

l- DMsion' Manager Nuclear Operations- l

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$dbscribed and sworn to before me this Jora day of December.,1989.

l Oh Notary rablic  !

GENER NDTM State or gana,n' J T. GLEAson i - - -

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