ML18227C739

Regarding Overpressurization Events During Cold Shutdown, Letter Requesting to Evaluate System Designs to Determine Susceptibility & to Notify NRC within 20 Days to Provide Requested Info within 60 Days or Schedule Time to Provide

ML18227C739
Person / Time
Site:	Turkey Point
Issue date:	08/11/1976
From:	Lear G Office of Nuclear Reactor Regulation
To:	Robert E. Uhrig Florida Power & Light Co
References
Download: ML18227C739 (8)

Text

Docket Nos. 50-250 and 5f-261 Florida Power 8 Light Company ATTN:

Dr. Robert E.. Uhrig Vfcj President

%. 0.

Box 013100

'famf, Florida 33101 Gentlemen:

RE:

TURKEY POINT ttUCLEAR PLANT, UNITS NO.

3 At(0'4 A number of reported instances of reactor vessel overpressurization in Pressurized Hater Reactor (PHR) facilities have occurred in which the Technical Specifications implementing 10 CFR Part 50 Appendix G

limitations have been exceeded.

The ma3orfty of cases have occur red during cold shutdown fn which the primary system has been in water solid conditions.

These overpressur ization events have been initiated by a variety of causes, including the following:

(1)

Isolation of RHP. system/letdo'~e system while charging to a water solid primary system, (2)

Thermal expansion fol'lowing the starting of a primary coolant pump due to stored thermal energy in steam generators, (3 )

Inadvertent actdation of safety injection accumulators, and (4)

Initiation of operation of a reactor coolant pump or a high pressure safety injection pump.

In essentially all of the events reported, a single personnel

error, equipment malfunction or procedural deficiency has been sufficient to cause the event.

He believe that appropriate steps should be taken promptly by all PM lfcensees to minimize the likelihood of additional occurrences of re c r

vessel overpressurfzation.

To 'that end we recently canpleted a serie of meetings with several P1$ R licensees and HSSS suppliers in which we d cussed the reported overpressurizatfon events and assessed the mea es that are currently being employed to either avoid or reduce the probability of similar occur rences, or to control the pressure transient to less than Appendix 6 limits.

Examples of those measures identified by the various liceiisees ar e as folio~is:

OPTIC II SURNAME~

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Complete avoidance of water solid conditions by either maintaining a pressurizer steam bubble or by providing a 'low pressure nitrogen blanket in the pressurizer when a steam bubble cannot be maintained, (2)

Disabling High Pressure Injection and Safety Infection pumps by disconnecting electrical power supplies when at low primary system temperatures, (3 )

Installation of dual setpoint pressurizer po~~r relief valve( s) to provide protection against exceeding Appendix G limits while at lo<1'r imary system temperatures, a

(4)

IIinlmlsatlon of tlma at mater solid condltlons and upgrading plant:/,

procedures to include appropriate warnings and cautions when such',;;

operations are necessary, and i'5)

Installation of reliqf valves in charging pump discharge lines with a setpoint to provide protection against, exceeding Appendix 6 limit< >

~

It was noted in our discussions with the PMR licensees that, for the majority of those plants involved, not a'll potential overpressurization

'vents would be prevented by the measures they had identified and that some of the remaining measures may have undesirable effects on reactor safety.

Based on the information gathered to date, we have concluded that all PAR licensees should evaluate their system designs to determine the susceptibil ity to overpressurization events.

Speci fically, you shoul d provide the following; (1)

An analysis of the Reactor Coolant System (RCS) response to pressure transients that can occur during startup and shutdown.

Any design modifications determined to be necessary to preclude exceeding Appendix 6 limits are to be incorporated in this analysis.

The analysis should include a plot of pressure as a function of time until termination of the event.

The analysis should assume the most limiting initial conditions (e.g.,

one RHR train operating or available for letdown, other components in nodal operation when the system is water solid such as pressurizer heaters and charging

pumps, and one or more reactor coolant pumps in operation) with the worst single failure or operator error as the initiating event.

Justification should be provided for the choice of limiting con-ditions and worst single failur e or operator error assumed in the

analysis, OFFICK 3m'URNAMK3R'AT K3F Form hEC-318 (ReT. 9-33) hZCM 0240 k U'r oovKRNI4KNT.FRINTIND DFFIcKI 1074 025 Iom

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A description of those design modifications determined to be necessary, including equipm>>ent performance specifications and system operational sequences.

The design basis used in the choice of equipment should be included, and (3)

A schedule for the prompt implementation of the proposed design modi fications.

The basic criteria to be applied in determining the adequacy oj Over-pressurization protection are that no single equipment failure or

'single operator error will result in Appendix G limitations being exceeded.

For those situations in which the necessary design changes identified cannot be implemented within the next few months, you should identify short-term measures to reduce the likelihood that overpressurization events will occur in the interim period until the permanent design changes can be made..

Short term measur es should be identified separately for immediate implementation within the terms and conditions of your license.

Short tern>>

measures might include some combination of, but would not be limited to, the following suggestions:

(1)

Procedural changes to minimize the time in which the primary system is in a water solid condition, (2 )

Upgrading existing plant procedures and administr ative controls to assure that appropriate warnings and cautions are included to alert the operator whenever the potential for primary system overpressurization

exists, (3)

Provide alarms and/or indications to alert the operator whenever primary system pressure increases toward Appendix G limits, (4 )

Introducing tem>porary plant modifications for pressure relief, and (5)

Assignment of additional personnel to monitor plant operations when water solid.

ORRICe~

SURNAME&

DATE~

POIID hEC 318 (Rev. 9.33) hECM 0240 4 U, BI OOVORNMCNT PRINT>>NO ORR>>CIII 1174 524 IOO

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IN Florida Power and Light Company ktodifications to preclude or minimize the probability of reactor vessel overpressurization events are plant dependent and the examples given may or may not be adaptable to your specific system design.

Consideration must also be given to the potential effects of both the short term and long term measures you consider to assure that other aspects of nuclear safety are not compromised.

To verify compliance with Appendix G pressure-temperature limits durinq startup and shutdown, you should assure that the appropriate instrumen-tation is installed to provide a continuous permanent record over the full range of both pressure and temperature.

This instrumentation should be in service during long periods of cold shutdown as well as during star tup and shutdown operations.

Reliance upon the plant computer to reconstruct a pressure transient is not considered sufficient because of the likelihood of computer downtime especially during plant shutdown conditions.

We request that you notify us within 20 days after receipt of this letter that you will provide all the information requested within 60 days or explain

<vhy you cannot meet this schedule and provide the schedule that you will meet.

This request for generic information was approved by GAO under a blanket clearance number B-180225 (R0072); this clearance expires July 31, 1977.

Sincerely, George Lear, Chief Operating Reactors Branch 03 Division of Operating Reactors cc:

Hr. Jack R.

Newman, Esquire Lowenstein, Newman, Reis 8 Peel rad 1025 Connecticut

Avenue, N. W.

Sui te 1214 Washington, D. C.

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