ML20095G990
| ML20095G990 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/17/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8408280281 | |
| Download: ML20095G990 (5) | |
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DUKE POWER GOhmtNY P.O. ISOX 3318D GIKAllLOTFIC, N.C. 28242 HALII. TUCKER TELEPIIONE mm rarement (704) 373-4531 August 17, 1984 m m...o m.
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station, Unit 1 Docket No. 50-413
Dear Mr. Denton:
My letter of July 31, 1984 submitted a proposed amendment to the Technical Specifications for Facility Operating License No. NPF-24 for Catawba Unit 1.
The purpose of this letter is to supplement the previous discussion of the emergency or exigent circumstances involved in the requested amendment and to amend the previously requested change to the Bases for Specification 3/4 7.1.2.
As previously identified, portions of the current Technical Specification surveillance requirements for the auxiliary feedwater pumps can not be met.
These current requirements are nominal design conditions as reflected in FSAR Tables 10.4.9-1 and -2, and do not reflect the test configuration for the pumps. As stated in the attached revised bases, the proposed surveillance requirements reflect the minimum auxiliary feedwater flow required for 0
cooldown to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power as previously stated in FSAR Section 10.4.9.1.
On or about July 16, 1984, Catawba station person 1el concluded that auxiliary feedwater pump test criteria in Technict1 Specification 4.7.1.2.1.a could not be met. Duke's Design Engi eering Department then reviewed the Auxiliary Feedwater System minimum iIow require.ments and determined corresponding flows through the test Loop to the upper surge tank (FSAR Figure 10.4.9-1).
After the required station and Nuclear Safety Review Board reviews, the amendment was filed on July 31, 1984. Thus in the Licensees' view, the amendment request was developed, reviewed and filed in a timely manner.
Catawba Unit 1 is currently in Mode 5 and is scheduled to enter Mode 4 on August 26, 1984 and Mode 3 on September 5, 1984.
If the requested amendment does not become effective until 30 days after publication in the Federal Register (assumed publication date of August 20, 1984) then the unit would incur at least a 14 day schedule delay and entry into Mode 3 could not proceed as scheduled. There is a substantial probability that a day-for-day delay would result in the startup schedule of Catawba Unit 1.
Such delay will have a substantial financial and economic impact 8408200281 840817
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. Mr. Harold R Denton, Director-August 17, 1984 Page 2 For each day's delay'in the commercial operation'of Catawba Unit 1, costs of approximately $500,000 a day will accrue against the Licensees' customers. Duke itself owns only a small portion (25 percent) of the l-Catawba Plant and the remainder is owned by small cooperative or municipal j
electric systems.
As provided in the foregoing d'iscussion, Licensees have demonstrated, in accordance with 10 CFR 50.91, that a timely application was filed and that emergency or exigent circumstances exist.
It is therefore requested that the previously requested amendment to Catawba Unit 1 Technical Specification 4.7.1.2.1 and the attached amended change to the Bases for this specification be granted on or before September 5,1984.
Very truly yours, 44 b Hal B. Tucker ROS: sib Attachment cc: Mr. James P. O'Reilly, Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 Mr. Jesse L. Riley l
Carolina Environmental Study Group 854 Henley Place l
Charlotte, North Carolina 28207 l
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Mr. Harold;R._ Denton, Director August 17', 1984--
Page 3 l
cc:
Palmetto Alliance 21356 Devine Street.
~ Columbia, South Carolina 29205 Heyward G. Shealy, Chief Bu_reau'of Radiological Health South Carolina Department of Health & Environmental Control 2600 Bull Street Columbus, South Carolina 29201
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- Mr ! Harold.R.- Denton, Director August'17, 1984 Page 4.
HAL.B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Re'gulatory Commission this revision-
- to the Catawba Nuclear Station Technical Specifications, Appendix A to License No. NPF-24; and that all statements and matters set forth therein ~are true and correct to the best of his knowledge, dk Hal B. Tucker,- Vice President Subscribed and sworn to before me this 17 day of August',1984.
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PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that_the Reactor Coolant System can be coolad down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power.
etRad 420 T84-5::5 :!::tri: ::t:rdr';;n iliary feedwater capable of deliver-ing a total feedwater flow of 500*gpm at a pressuresof 1210 psig to the entrance of the steam generators. Th; ;t;. dri.e..
.,ili;;y feed ;ter p.;;;
4 fi; ;;s;;;M if d;11;;rirg-: t:t:? ':;t::ter #!:r :#,1000 7 t ! ;' ereu-cd-1210 p;ig 0; th;.atr;;;; ;f th; ;;;;.; g;nerater;. This capacity is suffi-cient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.
AWWk 3/4.7.1.3 SPECIFIC ACTIVITY
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The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of f
10 ClR Part 100 dose guideline values in the event of a steam line rupture.
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This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
3/4.7.1.4 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to:
(1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.
l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION I
l The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum I
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allowable fracture toughness stress limits. The limitations of 70*F and i
200 psig are based on a steam generator RT f 10*F and are sufficient to prevent brittle fracture.
NOT CATAWBA - UNIT 1 8 3/4 7-2
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