ML20236D779

Discusses Util 861107 Proposed Tech Spec Supporting Steam Generator Tube Rupture Analysis.Inconsistencies Re Time Interval Allowed for Inoperability & Operability W/Isolated Atmospheric Relief Valves Due to Leakage Discussed

ML20236D779
Person / Time
Site:	Wolf Creek
Issue date:	03/16/1989
From:	Calvo J Office of Nuclear Reactor Regulation
To:	Withers B WOLF CREEK NUCLEAR OPERATING CORP.
References
TAC-63950, NUDOCS 8903230263
Download: ML20236D779 (7)

Text

_ _ _ _ _ _ -_

C)

1. 'o g UNITED STATES

!" g NUCLEAR REGULATORY COMMISSION

+

E WASHINGTON. D. C. 20555 8, March 16, 1989 Docket No. 50-482 Mr. Bart D. Withers -

Prssident and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, Kansas 66839

SUBJECT:

PROPOSED TECHNICAL SPECIFICATION SUPPORTING STEAM GENERATOR TUBE RUPTURE ANALYSIS (TAC N0. 63950)

License Condition 2.C(11) of Facility Operating License No. NPF-42 for the Wolf Creek Generating Station, Unit I required the submittal of a steam generator tube rupture (SGTR) analysis and supporting Technical Specification (TS) changes. The required analysis and supporting TS changes were submitted by the Wolf Creek Nuclear Operating Corporation (WCNOC) by letter dated November 7, 1986.

Since the SGTR analysis takes credit for atmospheric relief valves (ARVs) to mitigate the design-basis event, new TS for the ARVs have been submitted.

Based on the analysis results, one ARV has the capacity to cool down the plant during a design-basis SGTR. As proposed, WCNOC would require three of the four ARVs to be operable at all times. If one of the three is inoperable, a 30-day interim period is proposed to return that valve to " operable" status. Should the manual block valve be closed upstream of an ARY only because of ARY leakage, WCNOC proposes that the ARV wculd still be considered " operable."

The staff has considered the TS as proposed, but finds certain inconsistencies in need of resolution. These are in the areas of: (1) the time interval allowed for inoperability, ano (2) operability with an isolated ARY due to excessive leakage.

TIME INTERVAL ALLOWED FOR IN0PERABILITY WCNOC has proposed a 30-day allowable outage time (A0T) when one of the three ARVs is inoperable. For recent 4-loep Westinghouse plants, the TS for ARVs have required all four ARVs to be operable, with a 7-day A0T when one valve is inoperable and a 72-hour A0T when two valves are inoperable. The WCNOC proposal differs from recent Westinghouse 4-loop plants in both'the number of ARVs required to be operable and the A0T with one inoperable. The staff would accept a proposed TS that only requires three ARVs to be operable since this would allow one ARV to be unavailable due to a steam generator tube rupture associated with the loop for an operable ARV, one ARV to be unavailable due to a random undetected single failure, leaving one operable ARV which is sufficient for mitigation of the tube rupture accident. Hcwever, the staff

~

8903230263 890316 PDR ADOCK 05000482 P i FDC k a

se Mr. Bart D. Withers concludes that a 30-day A0T is excessive if one of the three ARVs, required to be operable, is inoperable. Therefore, a more appropriate time interval should  !

be considered. For instance, the staff would consider a 7-day time period to i be reasonable, i OPERABILITY WITH AN. ISOLATED-ARV DUE TO EXCESSIVE LEAKAGE WCNOC has proposed that if the local manual block valve is closed due to excessive ARY leakage, the ARV is considered to be operable on the basis that i the block valve could be reopened to permit the ARV to perform its function to manually control steam generator pressure for the SG tube rupture accident.

The basis for this proposal is that a similar action requirement exists for the PORVs that could be isolated by the closure of their block valves due to excessive PORV seat leakage and that there are no restrictions _ on continued power operation under these conditions.

The staff considers that the unrestricted time limit applicable to the PORV shall not apply to the inoperable ARY if its block. valve is closed. This is because the reliability of the PORY block valve, which is remotely operated, is not equalled by the ARV block valve, which must be manually and locally operated.

Therefore, we find that, for an ARV that is inoperable because of excessive t seat leakage only and its associated block valve closed, a 30-day time period, similar to that currently proposed for an inoperable ARV, is acceptable.

The enclosure provides the copy of the staff accepted South Texas Technical Specifications on ARV operability and their Bases which may be used as guidance i for an alternate Technical Specification. ,

The alternative Technical Specifications discussed above represents a staff-pcsition relating to the SGTR accident mitigatiun capability. We believe that such Technical Specifications are necessary to provide sufficient reliability .

to the atmospheric relief valves'. In order to expedite resolution of the license condition, we request that a response including revised Technical Specifications be provided within 30 days of the date of this letter.

Due to both the significance of the changes being sought relative to your initial Technical Specification proposal and the extended elapsed time from your original November 7, 1986 submittal to the present time, the NRC staff is closing out its Technical Assignment Control (TAC) tracking of this activity.

When you respond to the staff position described herein, we request that you specifically state that you are superseding your submittal of November 7, 1986. This will require you to submit a new application for license amendment-including a revised safety evaluation and significant hazards consideration, Upon receipt of your submittal we will renotice your application in the Federal Register and assigt. it a new TAC tracking number.

s,.

]!

a w

. .;, O Mr. Bart D. Withers ,

^  !

The reporting and/or recordkeeping requirements contained in this letter )

affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, 0JUGl@l SlGsf,D_E1108.EA@"

l ' Jose A. Calvo, Director

Project Directorate - IV )

Division of Reactor Projects - III,

.IV, V and Special Projects

, , Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page I

4 l

i i

1 4

I DISTRIBUTION l Docket File D. Pickett l NRC PDR 0GC l Local PDR E. Jordan l PD4 Reading B. Grimes G. Holahan ACRS (10)

L. Rubenstein PD4 Plant File J. Calvo P. Noonan a

PD4/ PD4/P % R PD4/D D PNoo DPickEtt:bj JCalvo 034(o/89 03/ %0/89 03//la/89 1

)~

j'.

~, ,

i l

I l

Mr. Bart D. Withers . The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required i under P.L.96-511. l l

Sincerely,  ;

' G, l

Jose A..Calvo, Director Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulatior

Enclosure:

. As stated cc w/ enclosure: I See next page l

1 1

O

l Mr. Bart D. Withers .

Wolf Creek Generating Station Wolf Creek Nuclear Operating Corporation Unit No. I cc:

Jay Silberg, Esq. . Mr. Gerald Allen i Shaw, Pittman, Potts & Trowbridge Public Health Physicist 1800 M Street, NW Bureau of Air Quality & Radiation i Washington, D.C. 20036 Control Division of Environment Chris R. Rogers, P.E. Kansas Department of Health

, Manager, Electric Department and Environment Forbes Field Building 321 '

Public Service Commission P. O. Box 360 Topeka, Kansas 66620 .

Jefferson City, Missouri 65102 l Mr. Gary Boyer, Plant Manager l Regional Administrator, Region III Wolf Creek Nuclear Operating Corp.,

U.S. Nuclear Regulatory Commission P. 0. Box 411 799 Roosevelt Road Burlington, KansL3 66839 l Glen Ellyn, Illincis 60137 )

Regional Administrator, Region IV ,

Senior Resident Inspector / Wolf Creek U.S. Nuclear Regulatory Commission  !

c/o V. S. Nuclear Regulatory Commission Office of Executive Director i

P. O. Box 311 for Operations Burlington, Kansas 66839 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 l Mr. Robert Elliot, Chief Engineer l Utilities Division Mr. Otto Maynard, Manager Licensing i Kansas Corporation Commission Wolf Creek Nuclear Operating Corp.

4th Floor - State Office Building P. O. Box 411 Topeka, Kansas 66612-1571 Burlington, Kansas 66839 i

t 1

i 1

s i

• A h

PLANT SYSTEMS

~~ ~

ATMOSPHERIC STEAM __ RELIEF VALVES LIMITING CONDITION FOR OPERATION j l

3.7.1.6 At least four atmospheric steam relief valves and associated manual controls shall be OPERABLE. -

APPLICABILITY: MODES 1, 2, 3, and 4.* .

ACTION:

a. With one less than the required atmospheric steam relief valves OPERABLE, restore the required atmospheric steam relief valves to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required RCS/ l RHR loops in operation for decay heat removal.

b. With two less than the required atmospheric relief valves OPERABLE, re-store at least three atmospheric relief valves to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and place the required RCS/RHR  ;

loops in operation for decay heat removal. l 1

SURVEILLANCE REOU_IREMENTS 4.7.1.6 Each atmospheric relief valve and associated manual controls shall be demonstrated OPERABLE prior to startup following any refueling shutdown or COLD SHUTDOWN of 30 days or longer, verify that all valves will open and close l

fully by operations of manual controls.

l l

l

• When steam generators are being used for decay heat removal.

SOUTH TEXAS - UNIT 1 3/4 7-10

~

PLANT SYSTEMS BASES 3/4.7.1.6 ATMOSPHERIC STEAM RELIEF VALVES The atmospheric steam relief valves are required for decay heat removal and safe cooldown in accordance with Branch Technical Position RSB 5-1. In the safety. analyses, operation of the atmospheric steam relief valves is assumed .

in accident analyses for mitigation of small break LOCA, feedwater line break, J loss of normal feedwater and loss-of-offsite power. )

l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNDT of 10*F and are sufficient to prevent brittle fracture.

j 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suf-ficient cooling capacity is available for continued operation of. safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the 1C- assumptions used in the safety analyses.

3/4.7.4 ESSENTIAL COOLING WATER SYSTEM l The OPERABILITY of the Essential Cooling Water System ensures-that suffi-

! cient cooling capacity is available for continued operation of safety-related

. equipment during normal and accident conditions. The redundant cooling capac-

!. ity of this system, assuming a single failure, is consistent with the assump-tions used in the safety analyses.

f

! 3/4.7.5 T UL_,IMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available either: (1) provide normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits.

The limitations on minimum water level and maximum t'emperature are based on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommend-ations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants,"

March 1974.

3/4.7.6 (Not used) i SOUTH TEXAS - UNIT 1 B 3/4 7-3 l