ML17341A517

Requests Addl Info Re Pressurized Thermal Shock to Reactor Pressure Vessels,Per Review of PWR Owners Group 810515 & Licensees 810522 Responses to NRC

ML17341A517
Person / Time
Site:	Turkey Point
Issue date:	08/21/1981
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
References
NUDOCS 8109140193
Download: ML17341A517 (15)

Text

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'AUGUST 2 g f981 Docket Ho. 50-251 Dr. Robert E. Uhrig, Vice President Advanced Systems and Technology Florida Power and Light Company P. 0. Box 529100 Miami, Florida 33152 AUG P. 8)98)t "0"0 Qw, XO ac no Aw Gl Qlm Qw OO dm OlQlQ QQw

SUBJECT:

PRESSURIZED THERhtAL SHOCK TO REACTOR PRESSURE VESSELS Me have reviewed the PhfR Owners'roups responses of May 15, 1981 and the licensees'esponses of Hay 22, 1981 to our letter dated April 20, 1981 concerning the subject issue.

The EPRI work which bears on the issue was included in the licensees'esponses.

On the basis of our independent review, of the plants where neutron irradiation has significantly reduced the fracture toughness of the reactor pressure vessels (RPVs), all plants could survive a severe overcooling event for at least another year of full power operation.

However, we believe that additional action should be taken now to resolve the long-term problems.

This belief is based upon our analyses which indicate that reductions in fracture toughness for some RPVs are approaching levels of concern.

It is also based in part on the fact that any proposed corrective action must allow adequate lead time for planning, review, approval, procurement and installation.

These conclusions were recently discussed with the PNR Owners Groups on July 28-30, 1981.

At those meetings, the Owners Groups reviewed the programs underplay at the three PflR vendors which are designed to scope the magnitude and applicability of the generic problem and to be completed by late 1981.

The three programs appeared to contain the necessary elements for resolution of the problem on a generic basis and the NRC plans to make full use of the reports due by the end of the year.

Hhile the vendors and Owners Groups are to be commended and encouraged in addressing the generic issue, there is also a need for plant-specific information for your plant.

Based on current vessel reference temperature and/or system characteristics, we have identified Ft. Calhoun, Robinson 2, San Onofre 1, htaine Yankee, Oconee 1, Turkey Point 4, Calvert Cliffs 1 and Three Nile Island 1 as plants from which we require additional information at this time.

The staff has used the time-dependent pressure and temperature data from the Harch 20, 1978 Rancho Seco transient as a starting point for our evaluation of this issue because:

(1) it is the most severe overcooling event experienced to date in an operating plant; (2) it is a real, as

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Dr. Robert E. Uhrig opposed to a postulated, event; and (3) it was severe enough that it could challenge the RPV when combined with physically reasonable.values of ir-radiated fracture toughness and initial crack size.

In future reviews the staff plans to use the steam line break accident or other appropriate transient/accident in order to estimate minimum operational times available before plant modifications are required.

Using calculated RPY steel mechanical properties, credible initial flaw

sizes, reasonable thermal-hydraulic parameters, and a simplified pr essure-temperature transient similar to that observed dur ing the Rancho Seco event, the staff has concluded that all operating plants could safely survive such an event at the present time and for at least an additional year of full power operation.

However, because of the required lead times for future actions, the margins in time for long term operation are not

large, and there is considerable uncertainty in the probability that similar or more severe transients may occur. It is clear that positive action must be initiated soon for those plants with significantly high transition temperatures.

As indicated above, several such plants have been selected by the staff, based on estimates of the current reference temperature for the nil ductility transition (RT

) of the RPVs.

HDT The need to initiate further action at this time is emphasized by the recognition that implementation of any proposed fixes or remedial actions must allow for adequate lead time.

Because long-term solutions may require a year or more, you should explore short-term approaches as well.

Although clear, concise instructions should be provided to operators to reduce the likelihood of repressurization during overcooling transients, the NRC staff believes that reliance on operator actions to prevent, repressurization during an overcooling transient will be very difficult to )ustify as an acceptable long-term solution to the problem.

In accordance with 10 CFR 50.54(f) of the Commission's regulations, you are requested to submit written statements, signed under oath or affirmation, to enable the Commission tn determine whether or not your license should be modi-fied, suspended or revoked.

Specifically, you are requested to submit the following information to the NRC within 60 days from the date of this letter:

(1) Provide the RT values of the critical welds and plates (or for-HOT gings) in your vessel for:

(a) initial (as-built) conditions and location (e.g., 1/4 T) and (b) current conditions (include fluence level) at the RPV inside carbon steel surface.

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Dr. Robert E. Uhrig 3w (2) At what rate is RT increasing for these welds and plate materia17 HDT (3) What value of RT for the critical welds and plate material do HDT you consider appropriate as a limit for continued operation7 (4) What is the basis for your proposed limitV (6)

Provide a listing of operator actions which are required for your plant to prevent pressurized thermal shock and to ensure vessel integrity.

Include a description of the circumstances in which these operator actions are required to be taken.

Included in this summary should be the specific pressure, temperature and level values for:

a) high pressure in)ection (HPI) termination criteria presently used at your facility, b) HPI throttling criteria and instruction presently used at your facility and c) criteria for throttling feedwater presently used at your facility.

For each required operator action, give the information available to the operator and the time available for his decision and the required action.

State how each required operator action is incorporated in plant operating procedures and in training and requalification training programs.

You are also requested to submit a plan for Turkey Point, Unit No. 4 to the NRC within 160 days of the date of this letter that will define actions and schedules for resolution of this issue and analyses supporting continued operation.

We request that you include consideration and evalua-tion of the following possible actions:

(1) reduction of further neutron radiation damage at the beltline by replacement of outer fuel assemblies with dum~ assemblies or other fuel management changes; (2) reduction of the thermal shock severity by increasing the ECC water temperature; (3) recovery of RPV toughness by in-place annealing (include the basis for demonstrating that your plant meets the requirements in 10 CFR 60 Appendix G IV C);

(4) design of a control system to mitigate the initial thermal shock and control repressurization.

For these, as well as for arly other alternative appr oaches, provide implementation schedules that would assure continuance of adequate safety margins.

In the interest of efficient evaluation of your submittal, we request that you include with the above plan, a response to the enclosed request t n.

Dr. Robert E. Uhrig Due to the nature of this review, and the past review effort that has been

expended, we consider the above schedules to be reasonable; however, inform us within 30 days if you anticipate conflicts with previous commitments with either submittal and a basis for any delay.

We also expect participation by the appropriate PHR Owners Group and NSSS vendors in developing solutions to the problem.

Sincerely, Griginai signed biI'arrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosure:

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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY VSGPO: ISSI~S60

DISTRI BUTIOH!

Docket File WRC PDR L PDR TEPA HSIC ORB85 Rdg ORDP4 Rdg ORB81 Rdg ORB83 Rdg Gray Files (8) 889 Owners Group CE Owners Group Westinghouse Owners Group HDenton DEisenhut RVollner SHanauer RWattson THurley FSchr oeder OELD (S)

AEOD IE (7)

ACRS (10)

GRAB SEPB NConner CHarwood RDiggs HHughes PWoolley I I KKniel NAnderson-RJohnson

'Clifford, ',,

BSheron DCrutchfield EThron

}fHazelton RKlecker GLainas HRandall JHartore THovak JStolz SVarga RAClark CTramnell DNeighbors SNowicki RSnalder RJacobs HSilver HGrotenhuis PHagner BRequa DJaffe GVissing RIngrav>

PKreutzer EHylton CParrish HSmith

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Robert E.

Uhr ig Florida Power and Light Company CC:

hr. Robert Lowenstein, Esquire Lowenstein,

newman, Reis and Axelrad 1025 Connecticut

Avenue, N.W.

Suite 1214 l<ashington, D.,C.

20036.

Environmental and Urban Affairs Library Florida International University Miami, Florida 33199 t1r.

Norman A. Coll, Esquire

Steel, Hector and Davis 1400 Southeast First National Bank Building Miami, Florida 33131 Hr. Henry Yaeger, Plant Hanager Turkey Point Plant Florida Power and'ight Company W. 0.

Box 013100 Miami, Florida 33101 Hr. Jack Shreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Resid nt Inspector

,Turkey Point nuclear Generating Station U.

S. Nuclear Regulatory Commission Post Office Box 1207 Homestead; Florida 33030

Encl osur e REQUEST fOR ADDITIONAL INFORMATION l.

Geometry Geometrical description including design and as-built (when available) dimensions of the core, assemblies, shroud/baffle, thermal shield, downcomer, vessel, cavity, and surrounding shield and/or support structure.

2.

Material Descri tion Region-wise material composition and material isotopic number densities (atoms/barn-cm) for the core, near-core regions and RPV, suitable for neutron transport calculations.

3.

Neutron Source Present and expected EOL:

a)

Assembly-wise and core power history (EFPY).

b)

Rod-wise and core power history (EFPY) for peripheral assemblies.

c)

Core average axial power history. distribution.

4.

Vessel Fluence a)

Description of available calculations of the vessel fluence including fluence values, locations, and corresponding power histories (EFPY),

including 1/4T, 1/2T and 3/4T through the RPV.

b)

Description of available capsule-inferred vessel fluences including fluence values, locations, and corresponding power histories (EFPY).

5 Surveillance Ca sules a)

Capsule materials, radial and axial dimensions and locations.

b)

Capsule fluence measurements, together with the accumulated power history (EFPY) and a description of the lead factors used to extra-polate the measurements to the peak wall fluence location.

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~ 6.

Y.esse.l; Me.l ds, A'xial'nd azimuthalocations; of vessel weld-seams with respect to the: core..

Overlay of "current fluence map with weld locations.

Identify, the critical.'elds,. vertical and circumferential, and give the, we~ld,'wire heat, numbers Give weld chemistry for the critical welds'.,

For each welidl wire heat number, report the estimated mean copper..'content the. range and the standard deviation, based on all the: reported'measurements for. that wel'd wire heat.

The welds may be surveil'lance wel'dm'eats. for your vessel or others, nozzle dropouts that contai'n a wel'dweTd metal qualification data, or archive material.

Iin~ the, absence.

of. any information, assume that copper'ontent is at i't's; upper limit (0.35 percent. when using R.G. 1.99Rev.

1) and that, the. nickel'ontent is, hi'gh'.

7'.

Systems. Analysi,s a>));-Provide:

a Tist of transients or accidents by class

{for, example:

excessive feedwater,, operating: transients which result from mul tiple fai>liures including. control system failures and/or aerator error, steam 1'ine; break and small'reak LOCA) which could lead to i'nside vessel fluid temperatures of 300 F or lower'.

Provide any Failure Nodes and Effects A'nalyses, {FNEAs) of control"systems currently available or reference any such" analyses already submitted.

Provide the analysis of the most limiti'ng transient or accident with regard to ~essel thermal shock con-si'derations.

Estimate: the'. frequency of occurrence of this event and provide'he basis for thi's estimate.

Discuss the assumptions made regarding reactor operator actions.

I(

b):

Identify the computer programs used to calculate the limiting transient or accident.

I'ndicate the degree to which the computer programs used have been verified and any other additional verification requi red to demonstrate that the computer'rogram models adequately treat the identi-fied important physical mo'del's (i.e.,

ECC mixing, heat transfer, and repressurizati'on).

6.

Yessel Welds Axial and azimuthal locations of vessel weld-seams wi th respect to the core.

Overlay of current fluence map with weld locations.

Identify the critical welds, vertical and circumferential, and give the weld wire heat

numbers, Give weld chemistry for the critical welds.

For each weld wi re heat nuniber, report the estimated mean copper content, the range and the standard deviation, based on all he reported measurements for that weld wire heat.

The welds may be surveillance weldments for your vessel or others, nozzle dropouts that contain a weld, weld metal qualification data, or archive material.

In the absence of any information, assume that copper content is at i:s upper limit (0.35 percent when using R.G. 1.99, Rev.

1) and that the nickel content is high.

7.

Systems Analysis If a)

Provide a list of ransients or accidents by class (for example:

excessive feedwater, operating ransients which result

,'rom multiple

.ailures including c'ontrol system failures and/or aerator error, steam line break and small break LOCA) which could lead to inside vessel fluid temperatures of 300 F or lower.

Provide any Failure Modes and Effects Analyses (FMEAs) of control systems currently available or reference any such ana3yses already submi.ted.

Provide the analysis of the most limiting transient or accident with regard to vessel thermal shock con-siderations.

Estimate the frequency of occurrence of this event and provi de the basis for this estimate.

Discuss the assumptions made regardi'ng reac.or operator actions.

b)

Identify the computer programs used to calcula

.e the limiting transient or accident.

Indicate the degree to which the computer programs used have been verified and any other additional verification requi red to

. demonstrate that the computer program models adequately treat the identi-fied important physical models (i.e.,

ECC mixing, heat transfer, and repressurization).

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