ML20245K648

Forwards Revised Tech Spec Bases Section 3/4.7.9,adding New Paragraph to Define Sealed Sources

ML20245K648
Person / Time
Site:	Beaver Valley
Issue date:	05/03/1989
From:	Tam P Office of Nuclear Reactor Regulation
To:	Sieber J DUQUESNE LIGHT CO.
References
TAC-72913, TAC-72914, NUDOCS 8905050112
Download: ML20245K648 (5)

Text

- May 3, 1989 DI STRIBUTION Docket No. 50-334 t Docket" File B Grimes 50-412 NRC and Local PDRs T. Meek (4)

SVarga WJones BBoger DHagan SNorris EButcher Mr. J. D. Sieber, Vice President PTam ACRS (10)

Nuclear Group OGC GPA/PA ARM / LFM3 Duquesne Light Company EJordan Post Office Box 4 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY UNITS 1 AND 2 - CLARIFICATION OF TECHNICAL SPECIFICATION BASIS ON SEALED SOURCES (TAC NOS. 72913 AND 72914)

By letter dated April 5,1989 you requested that Technical Specifications Bases sections 3/4.7.9 for both units be revised by adding a r.ew paragraph defining " sealed sources."

We have reviewed your proposed paragraph and :onclude that it is identical to one in section B 3/4.7.10 of the Standard Technical Specifications for Westinghouse Pressurized Water Reactors (NUREG-0452, Revision 4), and is thus acceptable. Enclosed please find the revised pages to replace existing pages of the same number in the Technical Specifications of both units. This completes'our efforts.

Sincerely, signed by Peter S. Tam, Senior Project Manager Project Directorate I-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page

[ TAC 72913 AND 914]

LA:PDI-4 PM:PDI-4 PD:PDI-4 JStolz M SNorris h b PTam:cb

1. / 3 /89 / 3 /89 3 /89 8905050112 890503 ti PDR ADOCK 05000334 P PDC

Mr. J. Sieber . Beaver Valley Power Station  !

-Duquesne Light Company Units 1 2 CC:

Jay E. Silberg, Esquire Bureau of Radiation' Protection Shaw, Pittman, Potts and Trowbridge Pennsylvania Department of 2300 N Street, N.W. Environmental Resources Washington, DC 20037 ATTN: R. Janati Post Office Box 2063 Kenny Grada, Manager Harrisburg, Pennsylvania 17120 Nuclear Safety Duquesne Light Company Mayor of the Borrough of P. O. Box 4 Shippingport Shippingport, Pennsylvania 15077 Post Office Box 3 Shippingport, Pennsylvania 15077 John A. Lee, Esquire Ashley C. Schannauer Duquesne Light Company Assistant City Solicitor One Oxford Centre City of Pittsburgh 301 Grant Street 313 City-County Building Pittsburgh, Pennsylvania 15279 Pittsburgh, Pennsylvania 15219 W.F. Carmichael, Commissioner Regional Administrator, Region I Department of Labor. U.S. Nuclear Regulatory Commission 1800 Washington Street East 475 Allendale Road Charleston, West Virginia 25305 King of Prussia, Pennsylvania 19406 John D. Borrows Resident Inspector-Director, Utilities Department U.S. Nuclear Regulatory Commission Public Utilities Commission Post Office Box 181 180 East Broad Street Shippingport, Pennsylvania 15077 Columbus, Ohio. 43266-0573 Director, Pennsylvania Emergency Management Agency Post Office Box 3321 Harrisburg, Pennsylvania 17105-3321

- - _ - _ - - - _-- . . _ - _ _ _ __. - _ a

PLANT SYSTEMS

, BASES 3/4.7.7 CONTROL ROOM EMERGENCY HABITABILITY SYSTEM The OPERABILITY of the control room emergency habitability system ensures that the control room will remain habitable for operations personnel during and following all credible accident conditions. The ambient air temperature is controlled to prevent exceeding the allowable equipment qualification temperature for the equipment and instrumentation in the control room. The OPERABILITY of this system

.in conjunction with control room design provisions is. based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

3/4.7.8 SUPPLEMENTAL LEAK COLLECTION AND RELEASE SYSTEM (SLCRS)

The OPERABILITY of the SLCRS provides for the filtering of postulated radioactive effluents resulting from a Fuel Handling Accident (FHA) and form leakage of LOSS OF COOLANT ACCIDENT (LOCA) activity from systems outside of the Reactor Containment building, such as Engineered Safeguards Features (ESP) equipment, prior to their release to the environment. This system also collects potential leakage of LOCA activity from the Reactor Containment building i penetrations into the contiguous areas ventilated by the SLCRS except for the Main Steam Valve Room and Emergency Air Lock. The operation of this system was assumed in calculating the postulated offsite doses in the analysis for a FHA. System operation was also assumed in that portion of the Design Basis Accident (DBA) LOCA analysis which addressed ESF leakage following the LOCA, however, no credit for SLCRS operation was taken in the DBA LOCA analysis for collection and filtration of Reactor Containment building leakage even though an unquantifiable amount of contiguous area penetration leakage would in fact be collected and filtered. Based on the results of the analyses, the SLCRS must be OPERABLE to ensure that ESF leakage following the postulated DBA LOCA and leakage resulting from a FHA will not exceed 10 CFR 100 limits.

3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source contamination ensure that the total body oe individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. -The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. Leakage of-sources excluded from the requirements of this specification represent less than one maximum permissible body burden for total body irradiation if the source material is inhaled or-ingested.

l t Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)

Deleted BEAVER VALLEY - UNIT 1 B 3/4 7-5 Amendment No, 1$Q, 132

{crrect4cr 10tter dated 9'/9/n7 Letter dated 5/3/89 ,

I~

3/4.7 PLANT SYSTEMS ILASIs

_3/4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. Leakage of sources excluded from the requirements of this specification represent less than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

I Sealed sources are classified into three groups according to their use, l with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are )

required to be tested more often than those which are not. Sealed sources ~

which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded i mechanism.

I 3/4.7.10 and 3/4.7.11 RESIDUAL HEAT REMOVAL SYSTEM (RHR)

Deleted 3/4.7.12 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is main-tained during and following a seismic or other similar event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on i nonsafety-related systems and then only if their failure or failure of the '

system on which they are installed, wculd have nc adverse effect on any safety-related system.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed ma:/ be used as a new reference point to determine the next inspection.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified OPERABLE by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically, susceptible snubbers are those which are of a specific make or model and have the same BEAVER VALLEY - UNIT 2 B 3/4 7-5 Letter dated 5/ 3 /89

PLANT SYSTEMS 1

BASES j SNUBBERS (Continued) design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation and vibration.

When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested during plant shutdowns at refueling or 18 month intervals not to exceed two (2) years.

Observed failures of these sample snubbers shall require functional testing of additional unit's.

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc...). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber life review are not intended to affect plant operation.

3/4.7.13 STANDBY SERVICE WATER SYSTEM (SWE)

The OPERABILITY of the SWE ensures that sufficient cooling capacity is available to bring the reactor to a cold shutdown condition in the event that a barge explosion at the station's intake structure or any other extremely I

remote event would render all of the normal Service Water System supply pumps inoperable.

l BEAVER VALLEY - UNIT 2 B 3/4 7-6 Letter dated 5/ 3 /89 f

I C_____________. .__ i