ML20140G286

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Forwards Proposed Tech Spec Changes to License DPR-59, Revising Suppression Pool Water Temp Limits,Per NRC ,For Review.Changes Will Be Effective Unless Notified within 20 Days of Ltr Date.Safety Evaluation Also Encl
ML20140G286
Person / Time
Site: FitzPatrick, 05000000
Issue date: 07/15/1975
From: Goller K
Office of Nuclear Reactor Regulation
To: Berry G
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
Shared Package
ML20140F372 List: ... further results
References
FOIA-85-665 NUDOCS 8604020089
Download: ML20140G286 (8)
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I y '{ UNITED ST ATES NUCLEAR REGULATORY COMMISSION - 1 C-h w AsHINGTON D. C. 20555 ' * ),, .-

July 15, 1975 ( y.

Docket No. 50-333 ,

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Power Authority of the State 3[

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of New York 'Ig 6-

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A'ITN: Mr. George T. Berry t ,

General Manager and y* d Chief Engineer J

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10 Columbus Circle ,

New York, New York 10019 g(_

Gentlemen: f '

Our letter to you of February 13, 1975, discussed the Steam Vent Clearing Phenomenon and the Steam Quenching Vibration Phenomenon at various BWR plants with Mark I Containments. We also requested that you initiate action in accordance with a prescribed schedule of major events set forth in this letter. The first action to be accomplished was your ,

submittal of proposed Technical Specifications to revise the suppression pool water temperature limits. Your letter of March 31, 1975, proposed no such changes to the Technical Specifications.

Because of the potential adverse effects on public health and safety of continued plant operation in accordance with existing Technical Specifications related to this matter, we believe that appropriate changes to these Technical Specifications are needed to assure that the integrity of the pressure suppression pool of your facility continues to be maintained. Accordingly, unless you inform us in writing within 20 days of the date"of this letter that you do not agree with this course of action, including your reasons. we_ plan _to_init.iate. steps

'to issue the encios'ed chance to the Technical Specifications of the FitzPatrick Plant. A copy of our related Safety Evaluatfoii i~s' enclosed.

Sincerely, l

Karl R. Goller, Assistant Director for Operating Reactors Division of Reactor Licensing

Enclosures:

1. Proposed Chac.ges to Technical Specifications
2. Safety Evaluation ec: See next page ,

8604020089 860114 PDR FDIA FIRESTD85-665 PDR b' 2 b-

l i h r Authority of the State July IS, 1975 ,

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  • ot New York cc w/ enclosures: Oswego City Library Scott L. Lilly, t .m . d ..e. .1 Power Authority of the 120 East Second Street Oswego, New York 13126 State of Imv York 10 Columbus Circle Mr. Robert P. Jones, Supervisor Neu York, New York 10019 Town of Scriba

/rvin E. Upton, Esquire R. D. #4 LeBoeuf, Lt.T.b, Leiby and MacRae Oswego, New York 13126 1757 N Street, t.v.

Ucchington, D. C. 20036 Mr. Alvin L. Karkau Chairman, County Legislature Lauman llar' tin, Esquire County Office Building 46 East Bridge Street Senior Vice President Oswego, New York 13126 and General Counsci Niagara Mohawk Corporation -

300 Eric boulevard West Dr. William E. Seymour Syracuse, New York 13202 Staff Coordinator New Yod hate Mode heny Mr. Z. Chilazi Council Power Authority of the' New York State Department of State of New York Commerce 10 Colmabus Circle 112 State Street New York, New York 10019 Albany, New York 12207 J. Bruce thcDonald, Deputy Mr. Paul Arbesman

  1. ' ^8 N et- o1 a r of ed Pa Commerce and Counsci to the New York, New York 10007 Atomic Enerr,y Council 99 Faching.ron Avenue Anthony Z. Roisman, Esquire Albany, New York 12210 Berlin, Roisman & Kessler 1712 N Street, NU Ecology Action Washington, D.C. 20036 c/o Richard Colds.nith

! Syracuse University Colicge of Law E. I. White llall Campus Syracuse, New York 13210 1.!s. Suzanne Ueber R.D. #3, West Lake Roao Oswego, New York 13126

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PROPOSED CHANGES 'IO THE TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333

'1he enclosed pages 165, 166, 187, 188, and 188a are proposed as replacement pages to the Appendix A Technical Specifications. '1he changed areas on the revised pages are shown by a marginal line.

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~3.7 LIMITING CONDITIONS FOR OPERA- 4.7 SURVEILLANCE REQUIREMENTS TION 4.7 CONTAINMENT SYSTEMS 3.7 CONTAINMENT SYSTEMS Applicability:

. Applicability:

Applies to the primary and secondary Applies to the operating status of the containment integrity.

primary and secondary containment systems. Obiective: )

Obiective: To verify the integrity of the primary..

and secondary containment systems.

To assure the integrity of the primary and secondary containment systems. Specification:- ,

Specification: A. Primary Containment A. Primary Containment 1. The pressure suppression chamber water level and temperature

1. The volume and temperature of shall be checked once per day.

the water in the pressure The accessible interior surfaces suppression chamber shall at all of the drywell and above the tiraes, except as specified in water line of the pressure l Specification 3.5.F.2, be suppression chamber shall be maintained within the following inspected at each refueling limits: outage for evidence of deterioration. Whenever there is ,

I a. Muimum water volume 110,100 ft 3 indication of relief valve operation or corresponding to a vent submergence testing which adds heat to the suppression -

level of 56 in. pool, the pool temperature shall be continually monitored and also observed

b. Minimum water volume 105,600 ft 3 and logged every 5 minutes until the corresponding to a vent- submergence heat addition is terminated. Wh'enever level of 4 ft 2 in, there is indication of relief valve operation with the temperature of the 4
c. Maximum water temperature suppression pool reaching 160F or more and the primary coolant system pressure (1) During normal power operation greater than 200 psig, an external visual maximum water temperature shall examination of the suppression chamber shall ,

be 95F. be conducted before resuming power operat lon.

165

3 7 (cent'd) JAFNPP 4 7 (cont'd) .

(2) During testing which adds heat to the suppression pool, the water temperature shall not exceed 10F above the normal power operation limit specified in (1) above. In connection with such testing, the pool temperature must be reduced to below the normal power operation J limit specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(3) The reactor shall be scrammed from any operating condition if the pool temperature reaches 110F, Power 2. The primary containment operation shall not be resumed urtil integrity shall be demonstrated the pool temperature is reduced below as follows:

the normal power operation limit

' specified in (1) above. a. Type A Test (primary Containment Integrated (4) During reactor isolation conditions, Leakage Rate Test) the reactor pressure vessel shall be ,

- depressurized to less than 200 psig (1.) Containment inspection at normal cooldown rates if the pool shall be performed as a temperature reaches 120F. prerequisite to the performance of Type A

2. Primary containment integrity shall be maintained tests. During the at all times when the reactor is critical or when period between the the reactor water temperature is above 2120F, and i

initiation of the fuel is in the reactor vessel, except while containment inspection performing lowpower physics tests at atmospheric and the performance of pressure at power IcVels not to exceed 5 MWt. the Type A test, no repairs or adjustments shall be made.

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JAFNPP 3.7 BASES excursion did occur, the reactor building and Standby Gas Treatment A. Primary Containment System, which shall be operational during this time, offers a The integrity of the primary sufficient barrier to keep offsite containment and operation of the doses well within 10CFR100.

Emerqency Core Cooling Systems in combination limit the offaite doses The pressure suppression pool water )

to values less than those suggested provides the heat sink for the in 10CFR100 in the event of a break ,

Reactor Coolant System energy in the Reactor Coolant System release following a postulated

piping. Thus, containment integrity rupture of the system. The pressure is specified whenever the potential suppression chamber water volume for violation of the R4 actor Coolant must absorb the associated decay and System integrity exists. Concern structural sensible heat released about such a violation exists during reactor coolant system whenever the reactor is critical and blowdown from 1,020 psig.

above atmospheric pressure. An exception is made to this Since all of the gases in the I requirement during initial core drywell are purged into t ne pressure loading and while the low power test suppression chamber air ; pace during progran is being conducted during a loss of coolant accident, the initial startup and ready access to pressure resulting from isothermal the reactor vessel is required. compression plus the vapor pressure There will be no pressure on the of the liquid must not e"cced system at this time, which will 56 psig, the suppression chauber ,

greatly reduce the chances of a pipe denign pressure. The design volume break. The re.ictor may be taken of the suppression chamber (water j critical during this period; and air) was obtained by considering

! however, restrictive operating that the total volume of reactor procedures, RSCS and kWM, will be in coolant to be condensed is effect again to minimize the discharged to the suppression probability of an accident chanber and that the drywell volume occurring. Procedures and the rod is purged to the suppression chamber worth minimiser would limit control (Section 5.2) .

worth to less than 1.5 percent k.

In the unlikely event that an i

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Thiing the mini _ mum or maximum wat.er temg.er.it ure and a maxiuum initi.ai voler..es given'in the specification, temler.s t ur e af 4SOF, a t emperat u re contain nent pressure during the of 145ar in achieved, which is well design basis accident is below the 1700F temperature which is approximately 45 psig which is below used f or complete condensation.

the d.' sign of 56 psig. L x imum water volume of 110,100 ft3 results For an initial miximum suppression in a downcomer submergence of 56 in chamber water temperature of 950F and the minimum volume of and assu:ning the normal complement 105,600 ft3 results in a submergence of containment cooling pumps (two approximately 6 in. less. The LPCI pumps and two Rl!R service water .

majority of the Bodega tests (9) pumps) containment pressure is not were run with a submerged length of required to maintain adequate net 4 ft and with complete condensation. positive suction head (NPS!!) for the

. Thus, with respect to downcomer core spray LPCI and !!PCI pumps. )

submergence, this specification is adequate. Limiting suppression pool temperature to 105 0F during RCIC, l The maximum temperature at the end  !!PCI , or relief valve operation, of blowdown tested durir.g the when decay heat and stored energy llumbold t Bay (10) and Bodega Bay are removed from the primary system tests was 1700F, and this is by discharging reactor steam ,

, ' conservatively taken to be the limit directly to the suppression chamber for complete condensation of the assures adequate margin for a reactor coolant, although potential blowdown any time during condensation would occur for RCIC,  !!PCI, or relief valve temperatures above 1700F. opera tion.

Should it be necessary to drain the Experimental data indicates that excessive suppression chamber, this should steam condensing loads can be avoided if only be done when there is no the peak temperature of the suppression requirement for Emergency Core pool is maintained below 1600F during any ,

Cooling Systems operability as period of relief valve operation with sonic explained in basis 3.5.F.

conditions at the discharge exit.

Using a 40*F rise (Section 5.2 FSAR) Specifications have been placed on the in the suppression chamber water envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

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3.7 BASES (Cont'd) JAFNPP J In addition to the licits on tecperature of -,

the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inad-vertently opens or sticks open. These procedures include: (1) use of all available means to close the valve, (2) initiate .

suppression pool water cooling heat ~ l exchangers, (3) initiate reactor shutdown,

! and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure

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mixing and uniformity of energy insertion to the pool. ,

Because of the large volume and thermal capacity of the suppression pool, the

. volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature I trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur

{ provides assurance that no significant damage was encountered. Particular attention j should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

If a loss-of coolant accident were e to occur when the reactor water '

l temperature is below 330*F, the containment pressure will not exceed , s the 56 psig design pressure, even if -

no condensation were to occur. The maximum allowable pool temperature, 188a '

whenever the reactor is above 212*F, shall be governed by this

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