ML20127G609

Summary of 740220 & 21 Meeting W/Util,Bechtel & GE in Bethesda,Md Re Prompt Relief Trip & Stated Valves

ML20127G609
Person / Time
Site:	Monticello
Issue date:	03/14/1974
From:	James Shea US ATOMIC ENERGY COMMISSION (AEC)
To:	US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20127G595	List: ML20127G591 ML20127G609 ML20127G621
References
NUDOCS 9211170391
Download: ML20127G609 (7)

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.-g DOCKET NO: 50-263 DATE: MAR 141974 LICENSEE: Northern States Power Company FACILITY: Monticello Nuclear Generating Station MIhVrES OF MEETING - PROMPT RELIEF TRIP (PRT) AND REPLACEME!rt OF SPRING-LOADED SAFETY VALVES WITH SAFETY / RELIEF VALVES - MONTICELLO BWR Representatives of Northern States Power Company (NSP), Bechtel, Cencral i Electric Company (GE), and AEC Regulatory met in Bethesda, Maryland, on February 20 and 21,1974, to review the proposal submitted by NSP's j letter dated January 23,1974, to:

1. replace spring-loaded safety valves by pilot-operated safety / relief valves.

2. install four new 10" pressure relief lines between the four new safety / relief valves and the torus suppression water.

3. install a conceptually new prompt relief valve trip (PRT)  !

system. i i

( 4. adopt more realistic control rod scram times and change the i Technical Specifications to reflect the change.

5. consider reduction of analytical uncertainty factors; 1.e. ,

substitute Operational Conservatism Factors (OCF) for Design Conservatism Factors (DCF) . ,

l A list of attendees in enclosed.

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It had been established by the participants, prior to the meeting, that l

priority attention should be given during the meeting to decisions affecting the proposed Monticallo plant modifications that are to be l ,

made during the next plant outage currently scheduled to begin on i March 15, 1974. Accordingly, since items 4 and 5 above rein.te to cal-culational input assumptions and technical specification changes, matters I that need not be decided within the next two weeks, discussion related l

to these items was incidental. Our evaluation of the PRT system) initiated 9211170391 740314

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j following the January 23, 1974 NSP submittal that proposed installation

of a PRT system on the Monticello nuclear power plant, and of the l analytical methods used in the same submittal but described in greater detail in a GE topical report (NED0-10002 " Analytical Methods of Plant

j. Transient Evaluations for the General Electric Boiling Water Reactor",

dated February 1973), is tentatively scheduled to be completed in .

) approximately six months. u 1 a The PRT system is designed to electrically open relief valves before

- pressure activation at a nominal 1080 psig pressure, following closure of the turbine stop valves or f ast closure of the turbine control valves, to prevent pressure transients in excess of ASME Code allowances  !

(1375 psig) and/or nuclear excursions that could result in excessive fuel clad temperature (MC11FRs less than the design limit of 1). i Anticipatory signals to open relief valves earlier during an _ over-pressure transient cannot be approved at this time, pending completion of our evaluation of PRT system and calculational methods, to justify reduction of reactor coolant system overpressure relief capacity. NSP i had proposed (USP submittal dated January 23, 1974) that the four safety i

valves at Monticello be removed and that full power operation be allowed with the four existing safety / relief valves and two new safety / relief  !

valves for a total of six valves where eight (four safety / relief and four safety valves) are presently required by Technical Specifications. .

It was emphasized by CE representatives, with supportive references to e the HSP submittal dated January 23,1974, " Permanent Plant Changes to ,

l Accor=nodate Equilibrium Core Scram Reactivity Insertion Characteristics',

' that the PRT modification was necessary to satisfy the fuel clad thermal design requirement whereby the minhnn critical heat flux ratio must

, be greater than 1.0 with t!.ne input assumptions specified in the report.

It was stated that additional pressure activated relief valves would

not satisfy this thermal design criterion assuming worst case control rod scram reactivity insertion (curve D) with design conservatism factors (DCF) .

- ~

Af ter a discussion of calculational input parameters other than the scram reactivity, such as Doppler and void coefficients and review of selected analytical results, the meeting participants agreed to the following as a basis for a timely AEC-Licensing evaluation that could

' . allow the proposed codifications to proceed on schedule:

1. Existing safety valve flow capacities will be maintained as a minimum by either:

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a. retaining two of the existing four safety- valves in addition to the six safety / relief valves, omct > _

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b. installing four safety / relief valves during the outage, if they can be obtained, in addition to the four existing safety / relief valves presently installed - but removing i all four presently installed safety valves, or

c. installing at least three new safety / relief valves during the outage, one less than plan (b) above, if this is all '

that can be obtained at this time. (Three safety / relief valves each rated at 800,000 lb/hr flow capacity are '

equivalent to the four safety valves which have individual flow capacities of 600,000 lb/hr.)

2. An analysis for fuel cycle 3 with DCFs as assumed in the HSP proposal using scram reactivity curve " Curve B" will be pro-  ;

vided for reactor operation which is expected to allow operation at rated conditions without dependence on PRT to prevent excessive fuel thermal cycle during a " turbine trip without steam bypass to the condenser" transient. It was noted that j the "B" curve would only be applicable at BOC 3 and for a j limited period of operation, to be determined by GE, after which further changes to calculational input assumptions j (use OCFs instead of DCFs if AEC review of this change is {

4 completed and accepted) or new power restrictions may be '

required unicas the Directorate of Licensing has approved the January 23, 1974 USP proposal to rely on the PRT and six safety / relief valves to prevent excessive reactor coolant '

pressure and/or fuel thermal duty.

The preference of those present at the meeting, pending completion of the Directorate of Licensing evaluation of the USP January 23, 1974 l

subnittal, was to install four new safety / relief valves in addition j to the four valves originally installed and remove the four safety l valves. However, at the time of the meeting, there was only assurance that two of the valves would be delivered in time for installation and

_4 limited optimism that one or possibly two more could be delivered and installed prior to plant startup in May 1974.

I It was noted that, providing reactor safety is not compromised, authorization l

to activate the PRT system upon completion of the installation need not l await final evaluation of system performance by the Directorate of

! Licensing, currently scheduled for September 1974, but should be granted as soon as possible. Approval to connect and activate the PRT system prior to need would permit confirmation of design adequacy and component reliability to prevent overpressurization or excessive fuel thermal duty

during the most severe but infrequent operational transients that can occur. This matter. it was agreed, can be resolved before the PRT -

I l ins ta11ation is eampleted.

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99 6 @pb All of the stresses in the existing steam lines and the 10" relief line j pp, piping between the new safety / relief valves and the torus suppression g

• pool are claimed to be within acceptable code values and steam line F A movement resulting from reactive forces during safety / relief valve

' gj cpening is negligible due to the large mass of the steam line and the short distance in the relief line piping to the first pipe band.

D Computer results, discussed by Bechtcl, to support this claim will be g provided by NSP in a suppleam to the January 23, 1974 submittal.

According to present analytical results, three new snubbers are required on the steam line but further stress calculations with various numbers l t and arrangement of snubbers could alter this number.

(

( The adequacy of the torus to withstand relief valve operation over the 40-year licensed lifetime of the plant was discussed. CE representatives C committed that design adequacy will be demonstrated selectively by

' installing instrunintation to measure torvu responses during relief

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valve openings at selected typical plants. The concern relatas to ,

y the 40-year license and the possibility that the lifetime is shorter than 40 years, based on revised calculational input data. 7t h not a V problem inaadiately related to reactor safety.

4

• Die consequences of MCHFRs SL following " turbine trip without bypass" 4 were discussed since this is the basic justification for PRT: 1.e.,

g to prevent MCHFR $ 1. Since the period following turbine trip when

' lj MCHFR $ 1 is relatively short and the core remains flooded, it is possible that clad temperatures do not reach damage chresholds even though HCILFR $ 1. When queried to establish the importance of main- l tcining MCHFRs greater than 1 during this transient, GE responded that they were not prepared at this time to go into details such as trancient clad temperatures while below MCHFR of 1.

,z l ( Steady state calculational methods have beco verified by measurements j l g at many BWRs, but transient performance has nut been demonstrated adequately to date Jua to the infrequency of planned transient tests I and unplanned circumstances during such tests that have marrad the  !

interpretation of results. Because of the severity of the abnormal t transients, tests are necessarily not of a repetitive type. Fast da::a acquisition systems with memory units to capture data during unplanned transients have been considered and are being considered by CE, but there are no firm plans to move in this direction at this time. Con-i firmation of design adequacy by plant measurements could reduce some l- of the uncertainties in the GE calculational methods that are currently under review within the Directorate of Licensing.

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y;V 4 5-The meeting was concluded with the following summary of the understandings:

1. Existing pressure relief steam flow capacity will be maintained until the Directorate of Licensing evaluation of the NSP proposal to reduce total relief capacity is completed.

2. Initial operation during cycle 3 will be without dependence on the PRI system.

3. Initial operation during cycle 3 may be with or without the PRT system in service, but Directorate of Licensin:; spproval is required to activate the proposed system and it must be established that activnLion will not reduce the rollability of other systems.

4. USP will provide additional information, prior to completion of plant modifications in !!ay 1974, to satisfy the primary conditions (items 1 and 2 above) necessary for Directorate of Licensin:.; approval to resume power production up to rated power level.

James J. Shea Operating Reactors Branich (12

, Directorate of Licensing Enclosure :

List of Attendens m,c, , L:0RB' #2 L:0&p,f2

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Form AEC-51s tRev.9 53) AECM 0240 u.s. cons egutet paistana ortset 1971 443 506 1

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MW Distribution Docket File AEC PDR Local PDR M. Voth, NSP L Reading

~12 %adkrg RP/TR Assistant Directors T. J. Carter, L:0R RP/TR Branch Chiefs J. M. Hendrie, L:TR J. J. Shea, L:0RB #2 R. Bevan, L:EP-4 J. Gallo, OGC R0 (3)

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