ML19327A660

Discusses Differing Professional View Re Plant Containment Isolation Valves

ML19327A660
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	07/20/1989
From:	Licciardo R Office of Nuclear Reactor Regulation
To:	Miraglia F Office of Nuclear Reactor Regulation
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References
TAC-73427, NUDOCS 8909140111
Download: ML19327A660 (2)

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July 20, 1989 HEMORANDUM FOR: Frank J. Miraglia, Associate Director for. Inspection and Enforcement

.FROM:

Robert B. A. Licciardo, Reactor Engineer Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERING PROFESSIONAL VIEW (DPV) CONCERNING CONTAINMENT ISOLATION VALVES AT ZION E

On May 11, 1989, The writer submitted a memo on the subject.

. Differing Professional View Concerning a)

Issuance Of SER To Zion 1/2 Allowing Full Power Operation With Open 42" Containment Isolation Valves b)

Methodology Used for Calculating Related Offsite Doses

.By memo of May 11, 1989, from F. J. Miraglia to R. Licciardo, the writer was asked to clarify certain aspects of the regulatory positions used 'ti the analyses including the time to failure used in LOCA analyses and mechanisms for the-transport of fission products from the primary (system) to the.

containment.

The writer was also asked to provide a view as to the safety significance of

'.the Amendment proposed by management and the safety significance of my concern regarding LOCA analyses.

In response to the above request, I am pleased to submit the enclosed document which analyzes for your specific concerns and presents the related conclusions in Section 4.

Regarding the safety significance of the existing Zion Amendment proposed by -

management.. Use of that Amendment and required Regulatory Guide 1.4 criteria would result in a contribution to thyroic dose over seven (7) secs. of 158,000 rem; using DNBR failure criteria with 10% fission product gap release would reduce this to 64,000 rem. Use of DNBR failure and equilibrium gap activity only would contribute 27,000 rem.

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It would take a fuel failure of only 0.2% of the existing rods releasing 10% gap activity only to increase offsite doses to 10 CFR 100 limits.

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'It must be recognized that allowing the containment purge valves to remain open for seven (7) secs. following a LOCA, multiplies by 194,000 the amount

.of fission product that would otherwise be release by leakage over the same period of seven (7) secs. from an isolated containment.

It becomes a direct contradiction of the regulatory-need for effective containment and limited lea kage.

-In sunnary:

Proceeding with the existing Amendment proposed by management would be in direct violation of regulatory requirements.

The writer's SCR of May 11 issued with his DPV of that date remains the writer's safety conclusions and recommendations in this matter i.e.:

"The 42" valves at Zion should remain closed in Modes 1, 2, 3 and 4 because the consequences of the offsite dose to thyroid (from iodine) during a LOCA is unacceptably high; whole body dose has not been evaluated.

The least value for offsite dose to the thyroid which may be proposed within the existing l

licensing basis is 64,000 rem.

The conventional. treatment of BTP CSB 6-4 which assumes that fuel failure does l-not occur over the first 5-15 seconds after a LOCA and thereby that only RCS operating inventory of fission products is released to the containment, and i

then to the environment, cannot in general be sustained against thermal hydraulic analyses for containment response, and licensing basis requirements I

(including criteria) for the calculation for, and the occurrence of, fuel I

failure and the quantification and treatment of the resulting source terms."

Robert B. A. Licciardo l

Registered Professional Engineer California Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. M 015380

Enclosure:

As stated l

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C. Rossi 1-F. Congel H. Smith l

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,AN EVALUATION OF THE CRITERIA FOR.

AND THE CALCULATION OF OFFSITE DOSES DERIVING FROM

.OPEN CONTAINMENT PURGE VALVES DURING l

A LOCA AT ZION UNITS 1 & 2

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i DATED JULY 20, 1989 PREPARED BY ROBERT B. A. LICCIARD0 i

REGISTERED PROFESSIONAL ENGINEER CALIFORNIA i

NUCLEAR ENGINEERING LICENSE NO. NU 003056 MECHANICAL ENGINEERING LICENSE NO. M015380 l

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INTRODUCTION On May'11,'1989, the writer. submitted a memo on the subject:

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. DIFFERING PROFESSIONAL VIEW CONCERNING a)

Issu'ance Of SER to Zion 1/2 Allowing Full! Power Operation With-Open 42" Containment Isolation Valves.

b)

Methodology Used For Calculating Related 0ffsite Doses.

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By memo of May-11, 1989, from F. J. Miraglia to R. Licciardo, the writer was

. asked'to clarify certain aspects of the regulatory positions used in his.

analysis including:

a) Time,to failure used'in LOCA analysis and b) mechanisms for the transport of fission products from the primary (system) to the contain -

ment..The writer was also asked to provide his view as to the safety significance of the Amendment proposed by management, and.the safety significance of his concerns regarding LOCA analysis, i

This material was prepared in response to that request and is in adjunct to his D.P.V which is attached to this document as Attachment 1.

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1-FISSION PRODUCT RELEASE FROM' FUEL AND CONTAINMENT USED IN l

em ACCIDENT ANALYSES................................................-

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't 1.1 Radiological Source Terms Within,the Core...................

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't 112J LOCA: ' Reg.. Guide 1.4 Criteria: ' Application to Zion.......-

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LOCA.

BTP CSB 6-4 B5.

Criteria.............................

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to' l ' 3.- 1 Characteristics of Fuel: Failure Giving' Fission o

ProductJRelease During Postulated' Accidents.........

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1.3.2 Characteristics of Fission Product Released 1

From Failed Fuel During Postulated Accidents........ 12 '

1.3.2.1 General....................................

_1-12 1.3.2.2 -Regulatory Guide 1.25......................

1-13 1.3.2.3 Regulatory Guide 1.77.......................

1-15 1

1.3.2.4 S umma ry....................................

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1.4 BTP CBS 6-4 B5 Criteria:

Application to Zion..............

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2 0FFSITE DOSE CCNSEQUENCES:

SUMMARY

2-1 vo 2.1 Basis for Calculations.....................................

2-1 2.2 Offsite Doses..............................................

2-2 2.2.1 RG 1.4 Source Terms Released Immediately on LOCA....

2-2 2.2.2 10% Gap Activity Released on DNBR...................

2-2 2.2.3 Equilibrium Gap Activity Released on DNBR...........

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2.2.4; RCS @ 60 pC/gm Activityi Released to Containment l

Immediately on a LOCA.....................'..........

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2.2.5'.RCS @ 60 pC/gm Activity: ' Released Progressively to O/

e Containment on RCS Discharge from a LOCA............

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. APPENDIX X EVALUATIONS, FUEL FAILURE; AND FISSION PRODUCT RELEASE.-

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3.1l Preliminary...................................

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13.2 Review.....................................................

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- 3.3.~ Summary....................................................

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' CONCLUSION'S....................................................a

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1 -FISSION PRODUCT RELEASED FROM FUEL AND CONTAINMENT USED IN ACCIDENT ANALYSES 1.1 Radiolooical Source Terms Within The' Core Exhibit l'shows core and gap activities for Zion for iodine, f

Calculated levels of iodine in the fuel clad gap are given to show a total-I-131'EQU of 24.09 x 105 curies Total iodine in the core as I-131 EQU is~15.79 x 107 curies.

1.2 LOCA:

Reg. Guide 1.4 Criteria:

Application to Zion x

Branch Technical Position CSB 6-4 (Ref. 25) states that:

"The sizing of the purge lines in most plants have been based on the need to control the containment atmosphere during refueling operations. This need has resulted in very large lines penetrating the containment (about 42 inches in diameter).

Since these lines are' normally the only ones pro-vided that will permit some degree of control aver the containment atmos-phere to facilitate'persennel access, some plants have used them for con-tainment purging during normal plant operation.

Under such conditions, calculated accident doses could be significant.

Therefore, the use of these large containment puroe and vent lines should be restricted to cold i

shutdown conditions and refueling operations and they must be sealed closed in all other operational modes.

i The design and use of the purge and vent lines should be based on the l

premise of achieving acceptable calculated offsite radiological consequences and assuring emergency core cooling (ECCS) effectiveness l

is not degraded by a reduction in the containment backpressure.

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o Purce system' designs that are acceptable for use on a nonroutine basis-during normal plant operation can be achieved by providing additional p

puroe lines.

The size of these lines should be limited such that in the event of a loss-of-coolant accident, assumi'ng the purge valves are open I

and subsequentiv close, the radiological consequences calculated in accor-dance with Regulatory Guidos 1.3 and 1.4 would not exceed the 10 CFR Part 100 auideline values.

Also the maximum time for valve closure should not exceed five seconds to assure that the purge valves would be closed before the onset of fuel failures following a LOCA.

Similar concerns apply to vent system designs."

This is interpreted by the writer as specifying that the large 42" purge and vent lines (PVLs) should be closed except in Modes 5 and 6.

And if purging is necessary in Modes 1, 2, 3 and 4, then smaller lines (8" and 10") should be considered and the source term to be used for evaluating offsite dose is that of Reg. Guide 1.4 which uses TID 14844 source terms as the fission product available for release to containment.

RG 1.4.C Regulatory Position (Ref. 30) requires the following under related subsecticn !!0.:

"la. Twenty-five percent of the eouilibrium radioactive iodine inventory developed from maximum full power operation of the core should be assumed to be immediately available for leakage from the primary reactor containment.

Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 percent in the form of particulate iodine, and 4 percent of this 25 percent in the form of organic iodides."

i.e., 25% of the radioactive iodine inventory from exhibit 1 is specified to be immediately available inside primary containment for leakage to the atmosphere.

For Zion this would represent approximately 25 percent of 15.79 x 107 curies of I-131 EQU in the core i.e., 3.9 x 107 curies immediately available inside containment for leakage to atmosphere.

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"Ic. The ef fects of radiological decay during holdup in the containment'or other buildings should be taken into account."

g With half lives for iodine (I) varying from 3.16 x 103 secs for I-134 to 6.95 x 105 secs for I-131, released immediately on a LOCA, and a time to valve closure of seven (7) seconds, there is no tims for significant r

radioactive decay of'any iodine isotope before it is discharged to atmosphere.

It is to be noted that the actual first stage of fission product release 4

during a LOCA occurs with the infringement of DNBR for the fuel rod, leading to overheating of the clad and fuel failure according to SRP 4.2 (Ref. 26) by perforation (or loss of hermeticity).

For Zion, this is specified to occur 0.1 sec's into the event in the Appendix K evaluation of the LOCA event; the off-site calculations for this submittal have been made for a DNBR infringement of 1/2 sec. and are therefore less conservative.

"Id. The reduction in the amount of radioactive material available for leakage to the environment by containment sprays, recirculating filter systems, or other engineered safety features may be taken into 1

account, but the amount of reduction in concentration of radioactive materials should be evaluated on an individual case basis."

During the first 7 seconds, there are no engineered safety features (ESF) fission product clean up systems available for reducing fission product 3

content prior to discharge to the environment.

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containment sprays are initiated after 45 secs.

Any filtration systems on the 42" inlet and outlet penetrations are not designed to ESF requirements.

Recirculating filter systems provided by }/ for fission product control of j

containment atmosphere during normal operations are not ESF equipment.

Containment volume of 2 million cubic feet originally containing 144,000 j

I lbs of air reduces fission product discharged from the RCS by prior dilu-tion through mixing.

Exhibits 3 and 4, and 3A and 4A show the circumstances for containment and the discharging reactor coolant system.

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The containment has an initial mass of air of 144,000 lbs (at atmospheric g

pressure).

On a LOCA, the initial rate of discharge from the RCS into l

containment'is 75,000 lbs/sec and over a period of seven (7) seconds prior to' containment valve closure, a total of 270,000 lbs is so discharged.

H This' increases total mass in containment to 420,000 lbs, increasing total pressure in containment to 23.7 psig; at the same time a total mass of 15,000 lbs [ valves fully open] to 2,860 lbs (valves partly open) of mixed containment inventory is discharged to the atmosphere.

L If it is assumed that all fission product released from the core is immediately available to containment as in RG 1.4, then total mixing of this product should be assumed to occur on initiation of the LOCA.

(The data presented show the results for a release second after the LOCA, but the differences are not significant for the intent of this submittal.) As a result, containment inventory discharged contains a uniform concer,tration of a decreasing curie content over the first 7 seconds, and the net result is a release to outside containment of 4.38%

of the source term fission product inventory Q, released from the core on occurrence of the LOCA.

(A reduced amount of 1.57% is released for partly closed valves). Exhibit 2A shows that for the RG 1.4 source term, this gives a total release from containment over the first 7 seconds of 1.7 x 106 curies direct to atmosphere.

Related offsite dose is 490,000 rem for 2 x fully open valves.

Partially open valves reduce this to the value shown in Exhibit 2 of 612,000 curies and 156,000 rem.

It should be reccgnized that the thermel-hydraulic, including energy conditions, are such that fluid is discharging from both the RCS and the containment at very high energy levels, with associated pressure levels giving sonic discharge velocities into containment of the order of 1000 fps.

Under these conditions it takes only hundredths of a seconds for RCS fluid to reach the containment isolation valves from the RCS system.

This is no comparison with the very low transport rates from the top of a fuel pool to containment isolation valves for a fuel handling accident inside 1-4

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containment as discussed in Section 1.3.3.5 of this sube t :ta. values of up to 15 secs, have been considered appropriate for these :.9eumstances, o

If is assumed that the core fission product source term is 'ipstead uniformly mixed with the RCS Fluid prior to its dicharge to containment, (less con-servative than R.G. 1.4) curie content discharged to atmosphere is reduced F

from 4.38% Q-to 1.9% Q where Q is the total term source released from the core by the LOCA and related source terms and related offsite doses are reduced by the same amount.

j These are nct unrealistic assumptions, for conservative purposes. The L

1.0CA causes sudden pressure drops in the RCS, to saturation pressures for the prevailing temperatures of the RCS inventory, causing steam release from violent boiling throughout the system.

This would cause substantial vibration of the fuel' rods and movement of the prevailing damaged UO2 pellets, facilitating the mass transfer of fission product gases to and through the gap to the locally faulted cladding, followed by blowdown through the clad defects at high rate; because of the prevailing pressure drops, between the gap and the core.

Over the.first seven seconds of the event, heat is being tranferrred from the core to containment by steam formation at the core and subsequent mass transfer to the RCS system and break, and discharge to the containment, at the very high rates discussed earlier in this subsection.

Since fission j

product gases are released from the cladding, (and probably at the hottest sections) the transport of fission products released from the gap would be within the same steam and entrained liquid transport system to the break and then containment.

i Within containment, unless special provisions have been made, there is no guarantee that a certain percentage of high concentrations of fission produ:t inventory being released by RCS discharge is not being bypassed directly to the open containment isolation valves from its main path to f

l principal containment volume.

In this sense, assuming an immediate f

release of all fission product to the containment on DNBR would help of fset the potential non-conservatism of this bypass.

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le. ine primary reactor containment should be assumed to leak at the leak

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rate incorporated or to be incorporated as a technical specification requirement at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> L

[0.1 percent per day], and at 50 percent of this leak rate for the h

remaining duration of the accident.

Peak accident pressure is the F

maximum pressure defined in the technical specifications for containment leak testing."

Except for dilution through mixing discussed under Id above, there is complete bypass of containment for 7 secs through the 2 x 42" open valves.

The magnitude of disci.arge to the environment with related offsite doses L

has be;n discussed under Id above.

In reviewing these figures, it should be recognized that for a normal leakage of 0.1%/ day from containment, 8 x 10-% of Containment Inventory (Q), would be released in the same time frame of 7 seconds. When compared with 4.38%, this represents a dose reduction factor of 541,000 and would reduce the 7 second dose from 489,000 rem to 0.9 rem.

Over a two hour time frame, and making allowance for 38 seconds without spray, followed by an iodine removal coefficient of 54/hr with a maximum reduction factor of 100, gives an approximate reduction in discharge by a factor of 32,000 leading to a calculated dose of 15 rem.

These reduction factors in offsite dose of 489,000 for the first seven seconds by effective early containment at 0.1%/ day, and of 32,000 in the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by effective containment at 0.1% per day and an iodine cleanup factor of 100, manifest the real significance of effective containment and containment spray in fission product containment.

x 1.3 LOCA:

BTP CSB 6-4, BS Criteria The Reg.1.4 source terms of 1.2 above, are based upon the Regulatory requirement of 10 CFR 100.11, (a) footnote 1 (Ref. 36) that:

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"The fission product release assumed for these calculations should be based I

upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible.

Such accidents have generally been assumed to result in substantia' meltdown of the core with' subsequent release of appreciable quantities of fission products."

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t However, Branch Technical Position CSB 6-4 (Ref. 25) provides another basis to justify containment purge design and which is less conservative than the Regulatory position.

This is given in related section B-5, as:

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"5.

The following analyses should be performed to justify the containment puroe system design:

a.

An analysis of the radiological consequences of a loss-of-coolant accident.

The analysis should be done for a spectrum of break sizes, and the instrumentation and setpoints that will actuate the purge valves closed should be identified.

The source term used in the radiological calculations should be based on a calculation under the terms of Appendix K to determine the extent of fuel failure and the concomitant release of fission products, and the fission product activity in the primary coolant.

A pre-l existing iodine spike shobld be considered in determining primary l

coolant activity. The volume of containment in which fission products are mixed should be justified, and the fission products l

1 rom the above sources should be assumed to be released through the open purge valves during the maximum interval required for valve closure.

The radiological consequences should be within i

10 CFR Part 100 guideline values."

To gain further regulatory interpretation of the meaning of fuel failure within this context, the writer's DPV (Ref. 42) refers to SRP 4.2 FUEL SYSTEM DESIGN, I (AREAS OF REVIEW), 2nd para. (Ref. 26) which states that, in respect of postulated accidents:

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"Tne obiettives of the fuel system safety review are to provide assurance P

that (a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences, (b) fuel system damage is never-so severe as to prevent control rod insertion when it is required, t

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(c) the number of fuel rod failures is not underestimated for postulated L'

accidents, and (d) coolability is always maintained.

"Not damaged," as f

used in the above statement, means that fuel rods do not fail, that fuel i

system dimensions remain within operational tolerances, and that~ functional H

capabilities are not reduced below those assumed in the safety analysis.

This objective implements General Design Criterion 10 (Ref. 38), and the

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design _ limits that accomplish this are called Specified Acceptable Fuel k'

Design Limits (SAFDLs).

" Fuel rod failure means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached.

Fuel rod failures must be accounted for in the dose analysis reouired by 10 CFR Part 100 (Ref. 2) for postulated accidents."

The underscored lines show that fue rod failure in the context of this i

paragraph must be evaluated for postulated accidents and this evaluation must be conservative.

Fuel Rod Failure means that the fuel rod leaks and that the first fission product barrier (the cladding) has therefore been breached; these fai. lures must be accounted for in the dose analysis required by 10 CFR Part 100 (Ref. 36) for postulated accidents.

Coolability is addressed as a separate criterion.

1.3.1 Characteristics of Fuel Failure Giving Fission Product Release During Postulated Accidents Regulatory clarification of fuel rod failure is given in SRP 4.2.II.A.2.

(Ref 26) This is abstracted as follows for the circumstances of postulated accidents in particular:

"2.

FUEL RCD FAILURE This subsection applies to [normai eperation--anticipated-operationsi occurrences--and] postulated ^ accidents.

[ Paragraphs-fa)-threagh-fe3-address 1-8

f siiere-meenanisms-that are-mete-iimiting-during-nermai-operatderi nc-the infermatien-te-ee-reviewed-shecid-be-contained-in-Sectien-4 f-ef-tre-Safety Ansiysis-Repert:) Paragraphs (d) through (h) address failure me: hem sms that are more limiting durino (anticipated operational occurrences anc) costulated accidents, -[and-the-informatien-to-be reviewed-wiii esesiiy-be-centained-in Ehapter-15-of-the-S af ety-Anniysis-Repert:--Paragraph-fi)-s hocid-be-eedress ed i n-Se c ti en-4: f-e f-t he-S af e ty-Ansiys i s-Repo rt-becaus e-i t-i s-not-addre s s ed eisewhere)

To meet the reouirements of [(a3-Senerai-Besign-Eriterien-10-as-it-reintes-to Specified-Acceptabie Fuel Design Limits for normal operation, including antici pated-eperatienti-eceerrences--and-fe)) 10 CFR Part 100 as it relates to fission product releases for postulated accidents, fuel rod failure criteria should be given for all known fuel rod failure mechanisms.

Fuel rod failure is defined as the' loss of fuel rod hermeticity.

[Aitheegh we-recognize-that-it-is-net pes sibie-te-aveid-sii-f eei-red-f aiieres-and-that-eleanep-systems-are-instaiied te-handie-a-smaii-number-ef-iesking-reds--it-is-the-ebjective-ef-the review-te assere-that-feei-cees net-faii-dee-to-specific-causes-dering normai-eperation ane-anticipated-eperatiensi-eecerrencest) Fuel rod failures are permitted duri,n_g postulated accidents, but they must be accounted for in the dose analysis.

Fuel rod failures can be caused by overheating, pellet / cladding interaction (PCI), hydriding, cladding collapse, bursting, mechanical fracturing, and fretting.

Fuel failure criteria should address the following to be complete.

Only those failure mechanisms that are more limiting for postulated accidents are abstracted here:

(d) 0_verheating of Cladding:

It has been traditional practice to assume that f ailures will not occur if the thermal margin criteria (DNBR for PWRs [and EPR-fer-BWRs3] are satisfied.

[ Th e-re v i ew-e f-th es e - c ri te ri a-i s - de t ai l e d-i n SRP-Sectien-4:4:--Fer-normai-eperatien-and-anticipated operationai-occer-renees--vieistien-of-the-thermai-margin-criterie-is-not permitted:] For postulated accidents, the total number of fuel rods that exceed the cri teria has been assumed to fail for radiological dose calculation purposes.

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- Although a thermal maroin criterion is sufficient to deenstrate the avoid-ance of overheating from a deficient cooling mechanism, it is not a necessary n

condition (i.e., DNB is not a failure mechanism) and other mechanistic methods may be acceptable.

There is at present little experience with other approaches, but new positions recorzendino different criteria should address cladding temperature, pressure, time duration, oxidation, and embrittlement, l

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(e) Overheatino of Fuel Pellets:

[it-has sise-been-traditionai practice-te assume-that-failere wiii cecer-if-centeriine-melting-takes piace:--This anaiysis shocid-be performed-for-the-maximam-linear-heat generatien-rate a nywne re-i n-th e-c o r e;-i nci ndi ng-sii-he t-s pe ts-a nd-het-channei-f acter s;- and l

snenie-acteent-fer-the effects-of-burnap-and-compositien-en-the-melting peint:--Fer-nermai-eperation-and-anticipated-eperationai ocentrences; i

centeriine-melting-is-net permitted ] For postulated accidents, the total number of rods that experience centerline melting should be assumed to fail for radioloolcal dose calculation purposes.

[The-centeriine-melting-cri-terien-was establishee-to-assure-that-axiai-et radiai reiecation-of-moiten f e e t - weni d-ne i t he r-aii ew-mei te n-f e el-te-c ome-i nt e-c ents e t-wi th-the-ei ad ding ner predece-ieesi-het-spets ] The assumption that centerline melting results l

in fuel failure is conservative.

(f) Excessive Fuel EnthalDy:

[for-s* severe

• reactivity-initiated-accident-(RIA) in-a-BWR-at-cere-or-iew power;-fuei-faiiere-is-assumed-to-occur-if-the-radi-niiy-averaged-feei-red-enthaipy-is greater-than-370-cai/g-st-any-axial-ieca-tion:] For full-power RIAs in a BWR and all RIAs in a PWR, the thermal mar-gin criteria (DNBR and CPR) are used as fuel failure criteria to meet the guidelines of Regulatory Guide 1.77 (Ref. 6) as it relates to fuel rod failure.

[The-170-tai /g-enthaipy-criterien-is primeriiy-intended-te address-eisdding-everheating-effects;-but-it-aise-indirect 4y-address peliet/ciadding-inte* actions-(PEi3:] Other criteria may be more appropriate for an RIA, but continued approval of [this-enthaipy-criterien-and-the ther-mal maroin criteria may be given until generic studies yield improvements.

(g) Pellet /Claddino Interaction:

There is no current criterion for fuel failure resulting from PCI, and the design basis can only be stated generally.

Two related criteria should be applied, but they are not sufficient to preclude 1-10 kr

r.I-L PCI failures.

(1) Th'e uniform strain of the cladding should not exceed 1%.

[in-this-centexti eniform-strain-(einstie-and-ineinstic3-is-defined-as f

transient-indeced-deformatien-with gage-iengths-corresponeing-to-cladding i

dimensions;-steady-state-creepdown-and-irradiation growth-are-exeinded ]

j-Althouch observing this strain limit may preclude some PCI failures, it will not preclude the corrosion-assisted failures that occur at low strains, nor will it preclude highly localized overstrain failures.

(2) Fuel melting

[

should be avoided.

The large volume increase associated with melting may L

cause a pellet with a molten center to exert a stress on the cladding.

Such a PCI is avoided by avoiding fuel melting.

Note that this same cri terion was invoked in paracraph (e) to ensure that overheating of t,h_e_

claddino'would not occur.

l 1

l f

(h) Bursting:

To meet the requirements of Appendix K of 10 CFR Part 50 (Ref.

9) as it relates to incidence of rupture during a LOCA, [a reptere-tem-peratere-cerreistien-mest-be used-in-the-t06A-E665-anaiysisr] Zircaloy j

cladding will burst (ructure) under certain combinations of temperature, heatino rate, and differential pressure. [Aitheegh-feei-seppliers-may-ese cifferent-reptere-temperatere-vs-differentiai pressere-cerves--an-accept-0 abie-cerve-shecid-be-simiiar-to-the-ene-described-in-Reft-10:]

(i) Mechanical Fracturing:

A mechanical fracture refers to a defect in a fuel rod causea by an externally applied force such as a hydraulic load or a i

I load derived from core plate motion.

Claddina integrity may be assumed if the applied stress is less than 90% of the irradiated yield stress at the l

appropriate temperature.

Other proposed limits must be justified.

Results from seismic and LOCA analysis (Appendix A to this SRP section) may show that failures by this mechanism will not occur for less severe events."

l l

Summary:

Failure Mechanisms include:

(a)

Infringment of DNBR criteria during postulated accidents which causes overheating of the cladding of the fuel rod, and is assumed to cause failure l

1-11 1

m

[.

f

(

of the clad,' and release of contained fission products from the gap as

[

a source term for the calculation of radiological doses.

+

h (b)

If postulated accident conditions cause calculated values of fuel pellet t

L temperature.to reach the melting point for the uranium dioxide at the centerline of. the pellet, it is assumed that all such rods shall fail (and h

release fission products from the pellets - as well as the gap) for the L

. calculation of radiological doses.

L 1.3.2 Characteristics of Fission Product Released From Failed Fuel'During Postulated Accidents l,'

I l

L 3. 2.1 General Fission product release as source terms for postulated accidents relevant to the above fuel f ailure criteria are specified as:

SRP 4.2,Section I, last paragraph (Ref. 26) states.that:

"All fuel damage criteria are described in SRP Section 4.2.

For those cri-teria that involve ONBR or CPR limits, specific thermal-hydraulic criteria are given in SRP Section 4.4.

The available radioactive fission product inventory in fuel rods (i.e., the gap inventory expressed as a release fraction) is provided to the Accident Evaluation Branch for use in estimat-ing the radiological consequences of plant releases."

SRP 4.2.C.3(h) (Ref. 26) states that:

" Fission Product Inventory:

To meet the guidelines of Regulatory Guides 1.3, 1.4, 1.25 and 1.77 [Refs--6--28-30] as they relate to fission product release, the available radioactive fission product inventory in fuel rods (i.e., the gap inventory) is presently specified by the assumptions in those Regulatory Guides.

These assumptions should be used until improved calcu-lational methous are approved by CPB [see-Ref--Si]."

1-12 i

b(.,

n The ::riteria from these Reg Guides are considered separately in the following

' subsections of this submittal in order to examine for general guidelines which L'

may be applied to BTP CSB 6-4 B5 Criteria.

s 1.3.2.2 Regulatory Guide (RG) 1.25:

Assumptions Used for Evaluating the Potential

[

Radiological Consequences of a Fuel Handling Accident in the fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors RG 1.25 (Ref. 31) covers the Fuel Handling Accident inside containment.

RG 1.25 page-25.1 under Section B, second para, provides for an immediate release of all activity from the fuel rod gap of the damage rods:

"The number and exposure histories of fuel assemblies assumed to be damageci determine the total amount of radioactive material available for immediate release into the water during a fuel handling accident."

The same Section B, fourth para. provides that:

c "Only that fraction of the fission products which migrates from the fuel matrix to the gap and plenum regions during normal operation would be avail-able for immediate release into the water in the event of clad damage.

(Migration of fission products is a function of several variables including operating temperature, burnup, and isotopic half life taken into considera-tion in establishing the release fractions listed in this guide.)"

RG 1.25 also assumes that 10% of the total radioactive iodine in the rod (with calculated peak activity) is contained in the gap for release.

(See page 25.2, Item C.I.d):

"All of the gap activity in the damaged rods is released and consists of 10% of the total noble gases other than Kr-85, 30% of the Kr-85, and 10%

of the total radioactive iodine in the rods at tpe time of the accident.

1-13

~-

gm 1

(.f p

l i

Released iodine rises to the surface of the related pool with a decontamination f actor of 100, provided a minimum depth of 25 f t exists, and gap pressure is no greater than 1200 psig.

Subsequent treatment of the source term is' typified E

by tne guidelines of SRP 15.7.4 Radiological Consequences of Fuel Handling Accidents (Ref. 28) which requires (under Section III.4, second and third para's_that:

"The reviewer should assess the time required tu isolate the containment.

This should include the instrument line sampling time (where appropriate),

(;

detector response time and containment purge isolation valve actuation and closure time.

The containment is considered isolated only when the purge isolation valves are fully closed.

The applicant's analysis should be reviewed regarding the travel time of any activity release starting from its release point above the refueling cavity or transfer canal and including travel time in ducts or ventilation systems up to the inner containment purge isolation valve."

"The time required for the release to reach the inner isolation valve is compared to the time required to isolate'the containment.

If the time required for the release to reach the isolation valve is longer than the time required to isolate containment, then essentially no release to the atmosphere occurs, and the reviewer's assessment should reflect this.

If the time required' for the release to reach the isolation valve is less than that required to isolate containment, and no mixing or dilution credit can be given, the reviewer should assume that the entire activity release escapes from the containment in evaluating the consequences.

l Claims for credit for dilution or mixing of a release due to natural or forced convection inside containment are reviewed and assessed.

References

[4] and [5] should be consulted and used by the reviewer for guidance in estimating dilution and mixing. Where mixing and dilution can be demon-strated within containment, the radiological consequences will be reduced by the degree of mixing and dilucion occurring prior to containment isolation."

3-14

pay E

a 9

e L

heiated references [4] and (5) are:

"4.

Evaluation of Fission Product Release and Transport for a Fuel Handling Accident by G. Burley, Radiological Safety Branch, Division of Reactor Licensing, revised October 5, 1971.

l 5.

Industrial Ventilation /A Manual of. Recommended Practice - American Conference of Governmental Industrial Hygienists."

These circumstances relate to a set of containment environmental conditions in which mixing energy is' virtually absent, being provided by low energy contain-ment purge and exhaust ventilation fans, and virtually no additional energy from the very small mass of fission product gas released from the damaged fuel 4

elements, after travelling through a minimum depth of 23 ft.

Under certain conditions, this could provide for the total activity released (after decon-taminatiun in the po::1) to be discharged directly to atmosphere outside containment.

i for Zion, the funaamental set of values for the thermal hydraulic parameters covering the above circumstances, are completely different to those governing the release and disbursement of fission products to the environment from a i.0CA.

1.3.2.3 Regulatory Guide 1.77:

Assumptions Used for Evaluating a Control Rod Ejection Accident For Pressurized Water Reactors j

i Fundamentally, this Guide provides for an evaluation of the Thermal Hydraulic l

and Power conditions within the core, during the accident, to determine a) the

)

extent of DNBR infringement and b) the amount of fuel exceeding the. initiation temperature of fuel melt (approximately 5150'F).

For Source Terms, RG 1.77, Appendix B1 (Ref. 32) proposes that:

"a.

The case resulting in the largest source term sho'uld be selected for evaluation, y

l 1-15

gn

'\\

t b,

The nuclide inventory in the fuel elements potentially breached L

should be ' calculated, and it sho,u.ld be assumed that all gaseous l

constituents in the fuel-clad caps are released.

I l

[

c, The amount of activity accumulated in the fuel-clad gap should be assumed to be 10% of the iodines and 10% of the noble gases accumulated at the end of core life, assuming continuous maximum full power operation, d.

No allowance should be given for activity decay prior to accident initiation, regardless of the reactor status for the selected case.

.e.

The nuclide inventory of the fraction of the fuel which reaches or exceeds the initiation temperature for fuel melting (typically 2842 C) at any time during the course of the accident should be calculated, and 100% of the noble gases and 25% of the iodine

~

contained in this fraction should be assumed to be available for release from the containment."

Summarily:

The source term from molten fuel is the same as for RG 1.4.

The source term release from the gap is the same as for the fuel handling accident.

The subsequent effects of the release path on the ultimate source terms from containment are evaluated for each of two release paths, as if the other did not exist.

These release paths are:

(1) By effectively immediate release of all source terms to containment to be followed by the following cleanup and decay provisions which are the same as those normally accounted for in a LOCA in RG 1.4 (Ref. 30).

RG 1.77, App. B1 (Ref. 32) provides that:

"f.

The effects of radiological decay during holdup in the containment or other buildings should be taken into account.

1-16

n p

ig l

!s i

g.

The reduction in the amount of radioactive material available for l

leakage to the environment by containment sprays, recirculating filter systems, or other engineered safety features may be taken into account, but the amount of reduction in concentration of radioactive materials should be evaluated on a case-by-case basis.

['

h.

The-primary reactor containment should be assumed to leak at the leak rate incorporated or.to be incorporated as a technical specification requirement at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident.

i-Peak accident pressure is the maximum pressure defined in the technic 61 specifications for containment leak testing."

o Additionally SRP 15.4.8,Section III.3 (Ref. 27), further specifies that:

"For releases via the containment building, 100% of the noble gases and 25% of the iodines contained in the fuel which is estimated to reach initiation of melting are assumed to be available for release from the containment."

Summarily:

For the release path to containment, these are effectively the provisions of RG 1.4 in respect of the treatment of Fission Product Source Terms after release from the core.

(2) By release of fission products to the secondary system as per RG 1.77, Appendix B, Items li, j and k (Ref. 32).

There are not considered in this submittal,.as they do not apply to a release to containment.

1.3.2.4 Summary (of General Positions on Fission Product Releases Deriving from RG 1.25 and 1.77)

(a) For f ailure of fuel cladding by either DNBR infringement or fuel handling accident:

n 1-17

i for iodine, 10% of the fuel rod inventory is released from the gap.

For the control rod ejection accident this release is assumed to be available r

immediately inside containment for leakage.

(b) For f ailure by centerline melting of the fuel pellet:

25% of the iodine inventory of any fuel rod which reaches or exceeds the initiation temperature of fuel melting is assumed to be immediately available inside containment for release.

This is the same assumption applied in RG 1.4 for fuel melt deriving from a LOCA.

I J LOCA:

BTP CSS 6-4/85 Criteria: Application to Zion Zion Fuel temperatures during normal operation at maximum power prior to a LOCA vary from 2500 F to 4100 for approximately 15% of the core (Exhibit 23). There-will be a substantial increase in temperature of the whole core over a period of up to 7 seconds following a LOCA and Exhibit 6 shows the related average cladding temperatures.

Considering the correlation of fission product release as a function of temperature shown in Exhibit 22, there is a high probability of a substantial increase in fission product activity in the gap over that of the equilibrium activity level represented on Exhibit 1, during these first seven (7) seconds of the accident, so that an increase in gap activity level from the equilibrium values shown in Exhibit 1 to the value of 10% used in the other postulated accidents is not an unreasonably conservative regulatory position to adopt for this event.

On this basis, the iodine source term deriving from fuel rod failure by overheating of the fuel cladding by DNB infringement at Zion at 0.1 second into the event would be 157.9 x 105 curies of I-131 EQU and is the value adopted by the writer in conformance to the related BTP.

In respect of fuel rod failure by centerline melting, the Zion FSAR (Ref. 33) does not provide detailed information on fuel pellet tempera-l tures except for the general statement that the safety injecti u system prevents core meltdown Ref. 33, page 14.3-46, Revision 1 second para.; provision for l

related fission product release from melted fuel rods is therefore not necessary for this evaluation to the guidance of the related BTP.

1-18

f L

l' On the basis of BTP CSB 6-4, B5 therefore, a total iodine fission product release of 157.9 x 105 curie I-131 EQU f rom the core, would be available to insiae containment at 0.1 second into the LOCA.

By reference to the conditions inside containment discussed in detail in Section 1.2, items Id and le above, it can be shown that, the release of 157.9 x 105 curies of I-131 EQU from the core as a source term will result in the discharge of 692,000 curies of I-131 EQU to atmosphere with an offsite dose of 176,000 rem with 2 x 42" fully open for 7 seconds, see Exhibit 2A, item 5.

With valves partly closed this is reduced to 249,000 curies 1-131 EQU and 63,400 rem, see Exhibit 2 item.5.

It is noted that in its recent revision to the FSAR (Ref. 34 ) page 14.3-38 Revision 1.

y has calculated an offsite dose from the LOCA on a non-Reg. Guide 1.4 basis, by also using the entire inventory of fission products contained in the pellet cladding gap, but has assumed the equilibrium values only, as listed in Exhibit 1.

This is equal to 24.09 x 105 I-131 EQU which is 1.52% of the core activity as compared with the 10% exemplified in other NRC criteria and used by the writer.

Effective doses that would be obtained using equilibrium gap activity only are also presented in Exhibits 2A and 2 under items 4 and show offsite doses to thyroid are reauced to 27,000 rem for 2 fullopen valves and 9,700 rem for 2 partially closed vahes, 1-19

gn n

ji.

i o

2. OFFS 11E DOSE CONSEQUENCES:

SUMMARY

7 2.1 Basis for Calculations J

Based on discussions in section 1, radiological releases and related offsite consequences are shown in Exhibit 2A item 6 for 2 x 42" fully.open (90 ) valves and Exhibit 2 item 6 for 2 x 42" values at a limited opening of 50.

All calculations are based on' valves closing in 7 seconds from commencement of a LOCA.

Doses are based upon valves being in the open position for a full 7 seconds as required by the SRP.

Valves will be required by technical specifica-tions to close within seven (7) seconds of commencement of the LOCA.

For the sake of example only, source terms are restricted to iodine in terms of I-131 EQU, and thyroid dose only has been calculated.

Dose is calculated at the site boundary (exclusion distance) of 415 meters.

Each dose is calculated independently of each other and are to be added to the LOCA leakage dose (over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) of 123 ree as appropriate.

An additional dose due to RCS inventory discharged into the containment would also need to be added, for all non-RG 1.4 calculations.

These are given in Exhibits 2A and 2 under items 2 at 132 rem for 2x fully open valves, and 48 rem for 2 partially opened valves.

For the diffusion coefficient, a value of 5 x 10 4 sec/cm3 applicable to leakage conditions over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period has been used.

In fact we have a high energy puff release of 7 seconds giving a patential finite cloud in travel to the enclosure boundary instead of a low leakage release diffusing into a cloud; as a result, the offsite dose under actual conditions is likely to be increased.

For the 0-2 hour leakage, the licensee has used a more conservative value than the NRC of 9.2 x 10 4 sec/cm3 and this would increase dose by a factor o' 1.84.

2-1

-m

.y-q, g{

cy h

h, 2.2' 0ffsite Doses.

l%

2.2.1 RG 1.4 Source Terms Released Immediately on LOCA Exhibit 2A, item 6, shows that for fully (90*) open 42" valves, the offsite g

[

dose for;a RG 1.4 source term is calculated at'489,000 rem.

And Exhibit 2, item 6, shows that for partially (50 ) open 42" values, these doses are

'i reduced to 156,000 rem.

L 2.2.2 10% Gap Activity Released on DNBR T

(.

Exhibit 2A-(item 5) shows offsite doses reduced to 176,000 rem for fully open valves, and Exhibit 2 (item 5) shows reduction to 63.000 rem for partially open valves.

't 2.2.3 -Equilibrium Gap Activity Released on DNBR-g Exhibit 2A (item 4) shows offsite dose is reduced to 27,000 rem for fully open L

valves and Exhibit 2 (item 4) shows reduction to 9,700 rem for valves partially open.

'2.2.4-RCS @ 60 pc/gm Activity; All Released To Containment Immediately On A LOCA.

Exhibit 2A (item 2) shows offsite dose contribution is 132 rem for fully open valves and Exhibit 2 (item 2) shows a reduction to 48 rem for partially open valves.

This activity release is equivalent to DNBR infringement of only.08% of the S

, fuel in the core.

RCS @ 60 pc/gm Activity; Released Progressively To Containment On RCS 2.2.5 Discharge From A LOCA Exhibit 2A (item 3) shows offsite dose contribution is 58 rem and Exhibit 2 (item 3) shows a reduction to 21 rem for partially open valves.

]

2-2

[W f

y 2.2 Conclusions n

I

.(1) According to Reg. Guide 1.4 criteria the offsite doses are completely L'

unacceptable.

L..

. LOCA calculations for Zion show no fuel melt; however, for DNBR infringe-(2) ment only, an evaluation of offsite dose based on release of 10% gap activity from 100% fuel still shows completely unacceptable circumstances.

L :

Although this is in conformance with SRP 6-4, BTP, CSB B5 criteria, it is not in conformance with 10 CFR 100.11 (a) footnote 1 requirements which states that:

l "The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible.

Such' accidents have generally I

been assumed to result in substantial meltdown of the core with i

subsequent release of appreciable quantities of fission products."

1 (3) Partially closing the valve to 50 from 90* is not successful in reducing the offsite dose to acceptable values.

(4) With valves partially open at 50 ; fuel failures by DNBR infringement on a LOCA would have to be limited to 0.2% of the core to limit total doses to 10 CFR 100 limits.

l l

i 1

2-3

4R 3 APPENDIX K EVALUATIONS, FUEL FAILURE, AND FISSION PRODUCT RELEASE 1

10 CFR 50.46 (Ref. 37), acceptance criteria for emergency core cooling system l

for light water nuclear power *eactors, requires that during a LOCA, cladding temperatures, cladding oxidatir1. and hydrogen generation, are limited and such L

that the core remains amenable to cooling in the short run froro the initial break through reflood, and also for long term post accident cooling.

10 CFR 50.46 does not include a requirement to evaluate the earliest time at which fission products could be released by local failure of the fuel cladding as fuel rod conditions rapidly change, challenge and exceed the limiting features of design which ensures fuel clad (and rod integrity) under Normal Operating Conditions and Transient Occurrences.

These limiting features are described as specified acceptable fuel design limits (SAFDLs) and are required under 10 CFR Part 50, Appendix A, Criterion 10.

A principal feature of the Appendix K evaluation is that it is designed to identify that rupture which causes a maximum post rupture cladding temperature within the fuel assembly being evaluated; and it is this time to rupture which is reported in the FSAR.

The Appendix K evaluation is not designed to report the earliest rupture that can occur.

3.1 Preliminary In evaluating 10 CFR 50.46 criteria through the use of the Appendix K evalua-tion model (Ref 39), licensees are required to undertake a detailed evaluation of the items to be discussed below throughout the complete LOCA event, i.e.,

from time 0 through 50-60 seconds, to determine that the clad rupture meeting the Appendix K criteria does not occur in the first 10-15 seconds of the event, ano which is the region of special interest for this review.

In the time avail-able for this research, a search of the UFSAR and the related reference mate-rial on the docket does not disclose many of essential the details of this calculation (Ref's 1-24).

We therefore use the limited information available to draw conclusions.

3-1

z 1 i e'

3.2 Review Appendix K calculations are undertaken on that fuel element assembly which ultimstely provides the maximam clad temperature after (post) clad rupture.

[

Generic work by W (Ref. 17) proposes that maximum calculated temperatures (post rupture) occurs in the low burn up (third region) fuel assembly, t

F Exhibit 6 shows the average clad temperatures deriving from Appendix K calcula-tions from the Zion FSAR, Figure 14 F. 2-19a, (Ref. 33).

This shows that on I

infringement of DNBR at 1/10' second, average clad temperature increases very rap 1aly from a normal operating value of 720 F to at least 1350 F, and then to 1750 F, over a total period of seven seconds; thereafter temperature reduces rapidly to 1000 F at about 15 secs, from which it sharply increases ultimately

' to approx 2200 F.

Exhibit 10 shows that W fuels are designed to require a yield strength of 45,000 psi a minimum for normal operations, and an ultimate tensile strength of 57,000 psi as a damage limit, as specified acceptable fuel design limits (SAFDL).

Exhibit 11 shows that as temperatures increase above 850 F, the available mechanical properties can be reduced below both these limits so that fuel clad cannot therefore be considered reliable in terms of protection against fission product release.

Exhibit 10 also shows that W fuels require a design limit of 1% on cladding strain as a design limit, and 1.7% as a damage limit.

The work of this Sec-tion 3 will show how both these limits can be exceeded inside the seven seconds on infringement of CNBR during the course of a LOCA, so that again, fuel clad cannot be considered reliable in terms of protection against fission product release.

Exhibit 15, shows how a temperature range of 1350 -1750 F traverses a range of Zircalloy metallurgical phases (transitions), a to (a + p) to p phases, during which ys = UTS and structural stability under stress is dependent upon mechan-ical/ strength properties which are a function of temperature and related time and stress at temperature. Under the circumstance of the transient expected l

3-2 l

y

?

y 5

from Appenoix K calculations with rapid changes of both temperature and stress, their is a need for empirical tests to determine swelling and burst (rupture) characteristics under these same dynamic conditions.

Exhibit 15 represents results_from such a series of tests (Ref. 13).

I Such conditions are also repr.sented in Exhibit 16 for Engineering Hoop Stress f

and temperature at rupture, for particular heating rates, and in conjunction with the information in Exhibit 20 on related rates of circumferential strain on rupture, at the given rupture temperatures.

What are the expected operating pressure differentials across the clad under thest LOCA conditions:

l Reference information shows that internal clad pressure under normally operat-f ing conditions is of the order of 1400 psig for new fuel and expected to increase to 2250 psig at the end of the 3rd cycle (for the fuel). On this j

basis, we evaluate a grp pressure of 1500 psig at approxim?tely 1/3 burnup into i

the first cycle, at which burnup maximum calculated clad tempe "atures are expected on a LOCA.

t It is propos9d that, immediately on a LOCA as clad temperature increases to j

1350", gap pressure will increase by 20%, to 1800 psig.

Exhibit 12 shows that j

at this time, core pressure has reduced to 1500 psig giving a pressure drop across the clad of 300 psi which according to Exhibit 13 will give a noop stress of approximately 2460 psi, f

At 7 seconds into the event, clad temperature has increased further to 1750'f, a-total increase of 1030'F from the normal operating condition.

From this, it can be proposed that gap pretsure for the complete rod can increase by 36% over j

its normal operating value to 2100 psig.

Exhibit 12 shows that at this time, core pressure has reduced to 950 psig so that the pressure drop across the l

clad is now 2100-950 1.e., 1150 psi which according to Exhibit 13 will give a j

hoop stress of 9400 psi.

When the above values of pressure and temperature are plotted on a particular Hoop Stress vs Burst Temp curve (Exhibit 14) from reference 1, at one see the 3-3

L clao ooes not rupture, but at seven seconds the clad is well into the rupture regirre.

.In its calculation of clad strain during Appendix K calculations, y uses results from tests by Hardy (Ref, 13).

Exhibit 15 is a set of results from one such l

test at 100'C/sec heat up rate (the heat up rate between 720'F and 1750'r in l

7 seconds = 150F'/second (or 64C'/second)).

This exhibit shows that these Appendix K values over the first 7 seconds bracket the range from zero (0) expansion at 1350'F to the burst regime at 1750'F.

In respect to these values, l

y has assumed that if clad strain reaches 10%, the clad will rupture; see Exhibit 18 from Ref. 3.

Note that the SAFDLs of 1% and 1.7% on cladding strain can both be exceeoed in the first seven seconds of DNBR infringement in the course of the LOCA.

The NRC, in its clad strain and rupture models uses the data shown in Exhibit 16 to determine when rupture is likely to occur for given rates of increase in l

temperature.

It is proposed by the NRC that the 28'C/S (=50F'/second) test L

points apply also to larger values (of rate of temperature increase).

Exhibit 16 shows that the Appendix K values again bracket the complete set of l

experimental data and significantly at the higher temperatures of the trar.sient.

Exhibit 20 shows the circumferential strain that can occur at given rupture l

temperatures, and the curve proposed by the NRC for Appendix K calculations.

i t

Prime Facie; maximum strain gives maximum blockage leading to maximum calcu-j lated temperatures for cladding after the burst.

In fact, W has established that maximum post rupture cladding temperature does not necessarily occur with j

a maximum circumferential strain at rupture, due apparently to direct radiation f

influences from fuel rods exposed by rupture at lesser values.

Providing rup-ture is expected by the data of Exhibit 16, the related strain is to be given i

by the NRC curve on Exhibit 20 (or lesser value giving maximum temperature).

[

l It should be noted that with this information there would be a very high prob-ability of rupture at 1750 F down to 15000F, with the probability decreasing, but still present at lower temperature.

l l

Note that Exhibits 16 and 20 do show that fuel temperatures and pressures could rupture the cladding over a whole range of conditions.

However, the purpose of l

3-4 1

i

C,

-n

'3<

the Appendix A evaluation is to identify that particular rupture which would nave tne most conservative effect with respect to meeting the requirements of 10 CFR 50.46 and for this end, it models, and uses factors, to conservatively calculate values for the related parameters.

Its purpose is not to determine and identify when failure by bursting (rupture) first occurs as an otherwise evaluation of when fission product is first released. An example can be seen f roni Exhibit 16.

The test points can show marked deviations from what are apparently best estimate curves for the various rates of temperature increase.

For conservatism in estimating the first occurrence of fuel rupture, one would have presumed the use of a boundary curve at the lower temperatures and pressures of each heating rate and Exhibit 20 would not have been required.

Note that Exhibit 15 does show that even though rupture may not occur with a detailed re-evaluation, cladding strain is most likely to exceed the 1% strain used by W (Ref 33 P. 3.2-39) as a SAFDL to meet the regulatory requirements of Ref. 38.

The writer would be concerned about the relevance of the hoop stress, strain /

rupture data of Exhibits 16 and 20 to the power generation and heat trans-fer conditions inside a reactor.

These tests were done on electrically resist-ance heated cladding tubes.

They do not simulate the heat transfer from central fuel rod pellets at high temperatures through a realistic gas gap of varying geometry, fuel pellet-clad contact, and pellet fracture / fragmentation to a cladding which is 12 f t long and which is likely to have a much smaller ratio of rupture length to clad length and gap volume than the test specimens.

The most revealing feature of Exhibit 16 is the data from the only test under-taken under much inore realistic conditions, on a nuclear fuel rod using Zircalloy cladding in the TREAT reactor at ORNL; this information shows ruptures

'at very much reduced stress levels than the rest of the data.

3.3 Summary 1.

Conditions within the core as currently evaluated by the Appendix K model, show that over the first seven (7) seconds following a LOCA, the following significant events occur:

3-5 e

neg i

c p

(

1.1 DNER for the whole core is infringed at 1/10 sec requiring gap activity at 10% core inventory for the whole core to be assumed as f

[

a source inside containment.

l t

1. 2 The temperature of the fuel clad, and the pressure drops across the same fuel clad, infringe specified acceptable fuel design limits (SADL) for normal operation and operational occurrences, required by l

[

'10 CFR 50 Appendix A, Criterion 10.

Fuel rod failure must therefore

{

[

be assumed for conservative calculations of offsite dose.

p I

1. 3 The temperature of the fuel clad and the related pressure drops show conditions in which lubstantial deformation of the fuel clad by strain, can exceed the oesign and damage SAFDL values for cladding strain.

Fuel rod f ailure must therefore be assumed for conservative calculations of offsite dose.

1.4 The temperature of the fuel clad and the related pressure drops show conditions which could result in fuel rupture.

This conclusion would need to be subject to detailed verification using the Appendix K model.

1.5 For Zion, fuel rods do not reach the melting point of the fuel pellets so that under minimum engineered safeguard conditions, additional fission product release from the fuel rods would not occur.

2.

The writer proposes that the purpose of Appendix K is to identify that particular rupture which would have the most conservative effect with p

respect to meeting the requirement of 10 CFR 50.46 and for this end it models, and uses factors, to calculate valces for the related purposes.

The purpose is not to determine and identify when failure by bursting (rupture) first occurs as an otherwise evaluation of when fission product is first released from the fuel summary a LOCA.

3-6

7, l

4 CONCLUSIONS 1.

Conditions within the core as currently evaluated by the Appendix K model, show that over the first seven (7) seconds following a LOCA, the following

[

significar.t events occur:

p 1.1 DNBR for the whole core is infringed at 1/10 sec requiring gap i.

activity at 10% core inventory for the whole core to be assumed as

j..

a source insido containment.

2.2 The temperature of the fuel clad, and the pressure drops across the saroe fuel clad, infringe specified acceptable fuel design limits (SADL) for normal operation and operational occurrences, required by 10 CFR 50 Appendix A, Criterion 10.

Fuel rod failure must therefore be assumed for conservative calculations of offsite dose.

1.3 The temperature of the fuel clad and the related pressure drops show conditions in which substantial deformation of the fuel clad by strain, can exceed the design and damage SAFDL values for cladding strain.

Fuel rod failure must therefore be assumed for conservative caiculations of offsite dose.

1.4 The temperature of the fuel clad and the related pressure drops show conditions which could result in fuel rupture.

This conclusion would need to be subject to detailed verification using the Appendix K model-.

1. 5 For Zion, fuel rods do not reach the melting point of the fuel pellets so that under minimum engineered safeguard conditions, additional fission product release from the fuel rods would not occur.

2.

The writer proposes that the purpose of Appendix K is to identify that particular rupture which would have the most conservative effect with respect to meeting the requirement of 10 CFR 50.46 and for this end it models, and uses factors, to calculate values for the related purposes.

4-1

4

'4 4

g The' purpose is not to determine and identify when failure by bursting I

(rupture) first occurs as an otherwise evaluation of when fission product is first released from the fuel summary a LOCA.

3.

As a result of the above 3.1 Fission product release from the fuel gap is a realistic considera-tion over the first seven seconds and prudent conservatism at this

{

time should consider release from the whole core.

3 3.2 Reg Guide 1.4 deriving from Regulatory Requirement 20 CFR 100 requires consideration of substantial molten fuel as a design for the source term.

4.

The writer proposes that Regulatory philosophy recognized the possibility l

of Beyond Design Basis Events as the realism of a substantial commercial i

industry and therefore required protection against this occurrence and f

made provision in the Regulations for this purpose.

'f Considering the energy exchanges occurrir,g in the core, and the insight of j

the Appendix K evaluations, it is not difficult to foresee significant fuel melt with potential additional substantive release of fission

[

i products from the fuel pellets over this time frame.

The question of the separate consideration of the timing of this additional contribution to I

the suurce term inside containment however must be moot.

Uncontrollable i

release through open 42 inch CIVs is out of the question so that steps taken to correct that problem by effective isolation do resolve the unanswered philosophical question as to when fission products released by fuel melt should be more realictically and conservatively established.

4.1 A review of available fuel failure c'*iteria, and the thermal-hydraulics aspects of the movement of fission gases from the clad l

to the environment over the first seven seconds of the event shows f

that:

4-2 l

74 Y.

t (a) The assumption of an immediate relcase to the containment is the'only available conservative basis for use at this tinie, and that (b)' The physics of the large energy releases from the core clad through the RCS to containment, and through the open isolation l

l; valver, shows effective mass transfer of fission product release l

l from the clad to the environment within the same (7) secs.

i b

5.

Fully open purge valves for a period of seven (7) secs discharge

[

1.7 x 108 curies of 1282 EQV to the environment giving an offsite dose l

r cf 489,000 rem to thyroid, i

t An isolated containment leaking at the safety analyses and TS limit of O 1% over 24 hrs, releases 3.14 curies of 1282 EQV over the same seven seconds with a contribution to offsite dose of 0.9 rem.

l

'i t

l The effectiveness of containment isolation and effective leak tightness in achieving a clean up factor of $41,000 over the first seven seconds of I

the LOCA is manifest.

h 6.

The offsite dose to thyroid for fully (90') open 42" valves using RG 1.4 source terms is calculated at 489,000 rem.

For partially (50') open 42" valves, these doses are reduced to 156,000 rem.

Reduction of source terms

[

from RG 3.4 to 10% gap activity released on DNBR infringement reduces i

offsite dose to 176,000 rem for fully open valves with a reduction to j

63,000 rem for partially open valves.

l Since the allowable limit for thyroid under 10 CFR 100 is 300 rem for 2 hrs at the Exclusion Boundary, these circumstances are unacceptable.

Therefore the 42" valves at Zion 1 and 2 should remain closed in f

Operational Modes 1, 2, 3 and 4.

l 7.

The stress / temperature relationships used to calculate fuel clad rupture to 10 CFR 50.46 are derived from test environments which are substantively

{

non-realistic when compared with actual fuel rod conditions in a reactor 4-3

j y

i during a LOCA.

The only in-reactor tests known to the writer at this Ln time with,the closest simulation of a real fuel condition gives ruptures

. ~

at very much reduced pressures for given rupture temperatures.

This L-comparison needs'to be revisited to more thoroughly evaluate the reasons

(,

for the differences and thereby improve our detailed knowledge of the total heat transfer environment which can lead to improvements in the q

calculational models of the fuel assemblies used in the Appendix K evalua-l tions.

This can help in a improved definition of the limiting features of l

the circumstances and lead to ways and means of improving fuel clad design

_i and performance for these circumstances.

l l

Y l

i l

f i

l.

l l

4-4 l

l l

1 p

v i

L i

L j.

I REFERENCES t

1.

Letter f rom E. P. Rahe (W) to James R. Miller (NRC):

Subject:

WCAP 9220-P A.

Rev. 1 " Westinghouse ECCS Evaluation Model 1981 Version."

r 2.

Safety Evaluation Report on the 1981 Version of the Westinghouse large Break ECCS Evaluation Model.

[

3.

Westinghouse Electric Corporation:

Locta IV Program, Loss of Coolant f

L iransient Analysis.

Bordelon et al.. WCAP8301.

June 1974.

l l

4.

Letter f rom E. P. Rahe (W) to M. Lauben:

Subject:

Millstone Large Break j

Results for Increasing Fuel Burnup.

July 20, 1981.

(

5.

Letter from T. M. Anderson to J. Stolz:

Subject:

Additions to WCAp9220.

I

. 6.

Letter from T. M. Anderson (W) to J. R. Miller, dated May 15,19B1.

Proposed 1951 version of the Westinghouse Appendix K Evaluation Model.

r 7.

Rahe, E. P., " Westinghouse ECCS Evaluation Model 1981 Version," WCAP-9220-p-A i

Rev.1 (Prcprietary), WCAP-9221-P-A Rev.1 (Non-Proprietary).

I 8.

USNRC:

Safety Evaluation Report on the 1981 Version of the Westinghouse f

Large Break ECCS Evaluation Model.

9.

Letter from J. Stoltz (NRC) to T. Anderson (W);

Subject:

Safety Evaluation

'l of WCAP-9220, " Westinghouse ECCS Evaluation Model, February 1978 Version."

February 1918.

I

- 10.

" Topical Report - Performance of Zircaloy Clad Fuel Rods During a Simulated f

I Loss-of-Coolant Accident - Single Rod Tests," WCAP-7379-L, Volume !

(Proprietary) and Volume II (Non proprietary), September 1969.

1 t

r e

g,,

LN'

+

+

M's

.e g

b t.

11., iopical Report - Performance of Zircaloy' Claa Fuel Rods During a i

L' Simulated Loss-of-Coolant Accident - Multi-Rod Tests," WCAP-7495-L, I

Prcptietary, Volume I - Test setup and Results; Volume II Analyses of.

Results, j

12

'Lorenz, Hobson & Parker.

0RNL.

Fuel Rod Failure Under loss-of-Coolant i

J,:

L Conditions In TREAT.

Nuclear Technology, Vol II, May 1971-August 1971.

g 13.

D. G. Hardy:

High Temperature Expans' ion and Rupture Behavior of Zircaloy Tubing, Atomic Energy of Canada Ltd.

Topical Meeting on Water Reactor

[

Safety, Salt Lake City, Utah, March 26-28, 1973.

L1 r

14.

Letter from G. F. Owsley (EXXON) to J. R. Miller (NRC);

Subject:

Acceptance-l',

for Referencing of Topical Report XN-76-47(P), dated Nov. 5, 1981.

15.

D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for-LOCA Analysis," USNRC Report NUREG-0630, April 1980.

I 16.

D. H. Risher, et al., " Safety Analysis for Revised Fuel Rod Internal Dressure Design Bases," Westinghouse Report WCAP-8963-P A, L

' August 31, 1978.

I' 17.

Letter from C. Eichelding (W) to J.F. Stolz (NRC) concerning " Review of WCAP-8963(P)," dated May 18, 1977.

b L

18. - Letter from T. M. Anderson (W) to J. F. Stolz (NRC); Subject WCAP-8963-P-A.

Safety Analyses for the Revised Fuel Rod Internal Pressure Design Basis (Prop).

L.

19.-

LedterfromJ.F.Stolz,USNRC, tot.M. Anderson, Westinghouse,

Subject:

?

SafetyEvQ,dtionofWCAP-8720,datedFebruary9,1979.

20.

W. D. Leech, et al., " Revised PAD Code Thermal Safety Model," Westinghouse Report WCAP8720, Addenda 2, October 1982.

2 l'

[

t

y i ',..

21.

Safety Evaluati.cn of the Westinghouse Electric Corporation Topical Report f

WCAP-9500.

" Reference Core Report 17 x 17 Optimized fuel Assembly."

j I

May 1981, Core Performance Branch, Reactor Systems Branch.

L 22.

Letter from T. N. Anderson (W) to J. R. Miller (NRC);

Subject:

Proprietary L

Responses to the NRC request for additional information on WCAP-9500 E

(Proprietary).

J,anuary 22, 1981.

23.

Letter from Robert L. Tedesco, to Westinghouse Electric Corporation, Attention:

T. M. Anderson;

Subject:

Acceptance for Referencing of

~

I Licensing Tepical Report WCAP 9500.

P 1

24. WCAP-8785:

" Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations."

25.

NUREG-0800 Standard Review Plan for the Review of Safety Analyses Report.

i l.

for Nuclear Power Plants.

July 1981, Section 6.2.4, Containment Isolation System, Branch Technical Position CSB 6-4, Containment Purging During Normal Plant Operations.

l i

26.

NUREG-0800, Standard Review Plan, July, 1981, Section 4.2, Fuel System f

1 Design.

l l

27.

NUREG-0800, Standard Review Plan, July 1981, Section 15.4.8.

Radiological Consequences of a Control Rod Ejection Accident (PWR) Appendix A.

i l

28.

NUREG-0800, Standard Review Plan, July 1981, Section 15.7.4.

Radiological-Consequences of Fuel Hand 1!ng Accidents.

i l

29.

USNRC, NUREG-75/077, The Role of Fission Gas Release in Reactor Licensing, f

November 1975.

30.

USNRC, Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized i

Water Reactors.

+

3 i'

V

'o r

31.

USNRC, Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the F

. Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."

32.

USNRC, Regulatory Guide 1.77, " Assumptions Used for Evaluating A Control RodEjectionAccidentforPressurizedWaterReactors,May1974.

[

33.

ZION 1 & 2: Updated Final Safety Analyses Report.

Commonwealth Edison.

i 34.

Deleted 35.

U.S. Government Printing office, " Reactor Design " Criterion 10 Appendix A, " General Design criteria for Nuclear Power Plants," Part 50 Title M [nergy, Code of Federal Regulations, January 1,1979.

n 36.

Ibid., " Reactor Site Criteria," Part 100.

37.

Ibid, "Acceptan:e Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," Part 50.46.

38.

Ibid., " Reactor Design", Criterion 10, Appendix A, part 50.

39.

Ibid., Appendix K Evaluation, Part 50.

40. Memo f rom R. Licciardo (NRC) to T. E. Murley (NRC): Subject; Differing Professional View concerning (A) Zion 1/2 Containment Isolation Valves and (B) Methodology Used for Calculating Related Offsite Doses.

Dated May 18,

1939, 4

1 l

7 r-i EXHIBITS OF EACKGROUND INFORMATION RELATED TO DIFFERING PROFESSIONAL VIEW CONCERNING a)

Issvente of SER te Zion 1/2 allowing full power operation with open 42" containment isolation valves, b)

Methocology used for calculating related of fsite doses.

e

1 c

hv1 j

ZION l

CORE AND GAP ACTIVITIES (10 DINE ONLY)

Assumptions:

Operation at 3391 kWt for 500 days Equilibrium Curies Percent in the of Core Curles

! 131 EQU Activity inthg) Gap 1 131 E00 Core 7) 7 (X 10 (X10 )

(X 10 x 10 in the Gap Isotope 1-131 8.35 8.35 2.3 19.2 19,2 l-13I 12.75 46 0.26 3.3

.12 l

1 133 15.C9 5.16 0.79 16.1 4.08 1-134 23.01

.39 0.16 3.8

.06 1-135 17.05 1.43 0.43 7.5

.63 E9 7009

i L

2 Rev. 1 ZION: LOCA DURING CONTAINMENT PURGE l

USINC 2x42" PENETRATIONS - VALVES OPEN 50' THn0!D DOSE AT SITE B0UNDARY RESULTING ONLY FROM DISCHARGE TO CONTAINkENT OUTSIDE DURING CLOSURE (LOCA LEAKAGE LOSE (0VER 2 HRS) = +123 REMS)

Site /Excl.

i Iten:

Curies Discharged Boundary Dose No.

Source Radiolopical Sources i 131 EQ (Thyroid (REM) 2 Licenser I 131 EQ. 60 ue/gm in 73.5 18.7 RCS 50% cleanup in cont.

All releaseo to 1

containn.ent on LOCA 2

RL I 131 EO 60 ue/gm in 188 48 i

RCS. Allreleasedto

~

cent on LOCA + 0.5 secs.

5

[ Total = 0.119 x 10 curies) 3 RL I 131 EQ; 60 uc/gm in RCS. 82 21 l

~

Released progressively to cont. With RCS discharge i

4 RL 1 131 E0; eouiv gap 38,000 9676 activity (F$AR calc.)

e

[24.09 x 10 curies of I 131 E0 into cent.

on LOCA + 0.5 secs.]

5 RL i 131 E0; SRP Gap activity 248,950 63.400 at 10: Total Activity (SRPcalc.)[157.9x10 5

curies of I 131 E0 into cont on LOCA + 0.5 secs.)

6 RL I 131 E0; Reg. Guide 1.4 611,500 155.700 at25%TogalActivity

[390x10 curies of i

1 131 EQ into cont, on LOCA) l X = 5 x 1C"# sec/m8 for 0-2 hrs, at nnnimum exclusion distance of 415 meters

[NRC]U

[ Licensee has used 9 x 10'# sec/m3 forSARs]

1 l

7 3 :

t I

{

ZION 1 & 2 l

CONTAINMENT INVENTORIES l

DURING LOCA BLOW DOWN l

I I

b

- RCS Mass Discharge Rate i

/

Into Containment r

$ Cumulative Discharge of l

2 100 RCS Into Contelnment 400 x 103

[

& Cumulative Moss of Alt

.El and MCS Discharge j

f a

i j_

/

D

,0 r

r l

f g

300 x 103 g l

l f

ac T

I f

]

[

B

~

~

s T

y 80 2

1 q

( /

j,

= x 10 L) 4k 40 _n

[

s db \\

30 i

100 x 108 20 i i

\\

'i=w

\\

a i

n.,

10

- - ~

3 N

Il P f

0 0

O 4

8 12 16 20 24 28 Time After Break - Seconds s

I l

l w

y yw- - -

ww-y-ee-

,-v--e,--w

---e.-v.w--w-=ww,.wwwww w

w-o--we,w-m-r-e-e wn + ew w w. e-w-www - w w-wv e-w----ww-w w,www-i--

w-vw,*ww,---w w--g--ww,=ww

-v.ywv-wwe

4

~..

ZION 1 ft 2 CONTAINMENT THERMAL HYDRAULIC CONDITIONS i

FISSION PRODUCT INVENTORIES 2 x 42" Lines l

Valves Open Only 80' l

~

Instead of 90' Fully Open

.l At 7 Secs i

t i

154,480 lbs Air 272.100 LbsRCS 428,820 Lbs Press a 23.70 psig Fission Product Inventory

= 0.484 x Q Released at 0.5 sens l

Discharge Rates Cumulative Totals Discharged Air + RCS Inventory Air + RCS Inventory 1023.88 Lbs/sec i

5379 Lbs

(.237% inv.)

i Fission Product Inventory Fission Product inventory 1.568% of Q

.237% Q/sec (Q = Fission Product inventory Released at t = 0.5 secs) s

'l

1 Ei o-FISSION PRODUCT DISCHARGED YO OUTSIDE CONTAINMENT EFFECT OF ASSUMPTIONS ON FISSION FRODUCT RELEASE TO CONTAINMENT 2 x 42" lines.

Valves open 50' 1

i Given 0 = total inventory of fission products in RCS at T=0.5 secs after LOCA If Q is released instantaneously to the total containment volume:

i I

i Fission proouct inventory discharged outside containment i

over 7 secs = 1.5681 Q If 0 is released over time with RCS iiiver. tory and based on a uniform j

distribution within the inventory:

Fission product inventory discharged outside containment i

over 7 secs = 0.561% Q l

i i

f g

b s

I e

i i

6 l..

.o i

r I-2A Rev. 1 ri ZION:

LOCA DUR1hG C0hTAINMENT PURGE j

USIt;G 2x4P" PENETRATIONS - VALVES FULLY Ol'EN (90')

i THYROID 00SE AT SITE BOUNDARY RESULTING ONLY fkOM DISCHARGE TO CONTAINMENT OUTSIDE DURING CLOSURE (LOCA LEAL 3GE DOSE (OVER 2 HRS) = +123 REMS)

Site /Excl.

Boundary) Dose (REl Item Curies Discharged (Thyroid

?!c.

Source Radiological Sources i 131 E0 1

Licensee I 131 E0. 60 uc/gm in RCS 204.3 52

~

501 cleanup in cont.

r All released to containment on LOCA 2

' RL 1 131 E0, 60 uc/gm in 522 132 RCS. All released to cont.

on LOCA + 0.5 secs.5

[ Total = 0.119 x 10 curies) 3 RL i 131 E0; 60 uc/gm in ACS.

227 58 i

~~

Released progressively to r

cont. with RCS discharge i

4 RL 1131E0;couivgapactigity 105,600 26,878 (FSARcalc.)[24.09x10 curies of I 131 E0 into cont.

I onLOCA+0.5 secs.)

5 RL i 131 E0; SRP Gap activity 691,520 176,010 calc. ) [157.9 x 10{ty (FSAR et 10% Total Activ curies i

of I 131 EQ into cont. on LOCA + 0.5 secs.]

l 6

RL I 131 E0; Reg. Guice 1.4 1,698,592 488.911 l

at25%TogalActivity

[390x10 curies of l

1 131 EQ into cont, on LOCA)

X = 5 x 10-# sec/m3 for 0-P hrs, at minimum exclusion distance of 415 meters

[NRC)O

[ Licensee has used 9 x 10"4 sec/m3 forSARs) w

x c

i i

[\\

Zion 1 ft 2 CONTAINMENT INVENTORIES DURING LOCA BLOW DOWN l

i Il 4- - - ncs mese oischeres not.

/

Into Containment i.

(

g Cumulettve Dischstge of

~~~

r 400x1Wd~h 100 nCS Into Containment J

.E l

[

& Cumulative Mass of Alt and RCS Discharge f

l I

i

r l

a r

r 300 x 103 g 1

f-I y'

.l xg 70 1

8 f)

.0

~ M x iM I j 1

50 ii j

u i

m u

40

~

k a

d k i

ur db \\

\\

100 x 108 b

i Il

\\

20 f

N ab w

10 T

Ill 1%am n 0

4 8

12 16 20 24 28 Time Af ter Broek - Seconds i

-.,-...-.--...-.-..._.....--.-..,.-.._._....--___.-_._.._.._.--..1

4A f 'c' Zi@N 1 ft 2 CONTAINMENT. THERMAL HYDRAULIC CONDITIONS F!SSION PRODUCT INVENTORIES 2 x 42" Lines Fully Open At 7 Secs N'

t i

154,400 Lbs Air 282.474 Lbs RCS.

416,934 Lbs I

Press a 23.79 psig i

I l

i Fission Product Inventory l

= 0.956 x Q Reloosed j

at 0.5 secs Discharge Rate f

Air + RCS Inventory Cumulative Totals Discharged 2000 lbs/sec i

Air + RCS Inventory

(.082% inv.)

i i

15026 lbs-i Fission Product inventory l

Fission Product inventory

.682% O/sec j

4.38% of G

-l (Q = Fission Product inventory Released at t = 0.5 secs) e

• • • - **",fl..

SA FIS$10N FF0 DUCT ?!SCHARGED TO OUTSIDE CONTAINMENT EFFECT OF ASSUMPTIONS ON F1SSJON PRODUCT RELEASE TO CONTAINMENT

{

7 x 42' lines l

fully open (90').

l Given Q = Totti inventory of fission products in RCS at T*0.5 set af ter LOi If Q is released instantaneously to the total containment volume i

Fission product inventory discharged outside containrent over 7 secs = 4.301 Q I

If Q is released over tire with RCS inventory, and based en a uniform distribution within the inventory:

Fission product invertory discharged outside containrent 4

over 7 secs = 1.901 Q f}

'A '

m k

t I

I e

z t,

r y.:

s

?-

k-A l

E 25m g r.o' g2003

(

1,.

g I

1000

.g g

sw '

.W i

1 I

I ts 0'A 50 100 150 200 250 I

o TIME (SECONDS)

A AMNDix W CUD /r*/47W/ 77/*fd 1

1?lcl i

Fioure 14 F.219e Peek Clod Temperstvre - DECLG (CDM.0)

(Unit 1) 4

=

7 i

3.1.3.3 Thermal and bydraulic Limits The resctor core is designed to meet the following limiting therral and hydraulic criteria:

TheminimumallowableDNBRduring)normaloperation, including a.

anticipated transients is [3.30'.

b.

No fuel melting during any anticipated operating condition, To maintain fuel rod integrity and prevent fission product release, it is necessary to prevent clad overheating under all operating conditions, This is accomplished by preventing a departure from nucleate boiling (DNB).

DNB causes a large decrease in the heat transfer coefficient between the fuel rods ano the reactor coolant resulting in high clad temperatures.

5

'A N.

t t

h 9

8 1

The integrity of fuel rod cladding so as to retain fission products or fuel material is directly related to cladding stress and strain under normal operating and overpower conditions. Design limits and damage Ifmits (cladding j

1 perforation) in terms of stress and strain are as follows:

Damace timit Design Limit Stress Ultimate strength Yield strength-57,000 psi minimum 45,000 psi minimum Strain 1.7%

1.0%

t The damage limits given above are minimum values. Actual damage limits depend upon neutron exposure and normal variation of material properties and would generally be greater than these minimum damage Ifmits.

For most of the fuel rod life the actual stresses and strains are considerable below the design i

limits. Thus, significant margins exist between actual operating conditions and the damage limits.

The other parameters having an influence on cladding stress and stisin and the relationship of these parameters to the damage limits are as follows:

I-1.

Internal gas pressure:

The internal gas pressure required to produce cladding stresses equal to the damage limit under normal operating conditions is well in excess of

/

the maximum design pressure. The maximum design internal pressure under nominal conditions is 2250 psia which is equal to the coolant pressure.

(

The end of life internal ges pressure depends upon the initial pressure, void volume, and fuel rod power history, however it does not exceed the design limit of 2250 psia.

N 2.

Cladding temperature:

The strength of the fuel cladding is temperature dependent. The minimum ultimate strength reduces to the design yield strength at an average cladding temperature of approximately 850*F. The maximum average cladding temperature during normal operating conditions is given in Table 3.2.21[as720'F].

i i

3.

g i

(

Frevious experience with removaole roos has been attained at Saxton, Yankee anc 2crita; anc accitional excerien:e mill be a:cuired at the San Onofre Cycle 2 anc Surry Unit '1.

Over 300 fuel rocs mere removed anc re-lostrtec into assemoldes during the Saxton re-constitution without evioence of failure.

Leak cetection tests were performed on the assemblies after sT1 rocs were re-inserted, and no leakage was detected. An ecus 11y large owmoer of Saxton ro:s have been successfully rem:ved, examined and re-inserted into over 12 3x3 sucassemolles at Saxton.

In sedition, 28 full length Yankee rocs were re"eved, examined and re-inserted into Yankee Core V spe lal assemolles.

5!millar hancling of 22 removasle ro:s was su:essfully c:molete: ouring the first Zerjta refueling.

t.11 su:h fuel hanclings have teen cone rout!nely in:

.! now :!ffh ulty.

The same fuel roc de'si anc internal pressure,gn limits incicated in se: tion 3.2.) fuel ts : stat;.:e are maintainee for these re?ovable ro:s anc :nere is nt, It:v:tjen in cargin to DNS. Tnelt in:Jusion in the initial Zi:n U !: 1 et e hn:in-int

.t y :. t q s the saferunr:::v:es no accitic,a1 safety ::mskerati:.s ans in r: analysts anc relate 1

eviously submitte material in sup::rt of the license a:plicathn.

3.2.3.5 Evaluation of Core Comoonents reel Evaluation ine fission gas release and the associateo builcup of internal gas pressure in the fuel rocs is calculated by a coce based on experimentally ceterminec

stes. The in:rease of internal pressure in the fuel rod due to this

nen: mens is in:1veec in the cetermination of the maximum clad ing stresses at tme ens of ccre life when the fission proov:t gap inventory is a maximum.

The ensimum allewable st sin in the cladding, consicering the comoinec effects cf internal fission gas essure, external coolant pressure, fuel pellet s.elling and clac creep is limiteo to less than 1 per Cent throughout c:re life.

ine ass::Jate: stresses are cel== the yield strength of the material un:e: all n : mal c:erating conditions.

To assure that manufactured fuel rocs meet a high standare of excellence f cm the stancooint of functional recuirements, many inspections and tests are

erformed both on the raw material and the finished product. These tests and ins:e:tions in:Juce chemical analysis, elevated tencerature, tensile testing of fuel tubes, dimensional'inspe: tion, X-ray of both end plug welos, ultrasonic testing and hellum leak tests. See aeditional details in Se: tion 3.3.3.1.

In the event of cladding defe:ts, the high resistance of uranium di:xice fuel

ellets to atta:k by het mater prote:ts against fuel deterioration er ce::asse in fuel integrity. Thermal stress in the pellets, while causic; some fractu*e Of the evik esterial carin-tetcerature cycling, coes not resdt in i

viverization er gross voic formation in the fuel matrix. As ;shown by c:erating ex:erien:e and extensive exoerimental work in the industry, the g'

tre: mal :esi;n :arameters conservatively accoot for any chan;es in the

tal cerf::ran:e of the fuel element due t: :ellet f s:ture.

o!!);

3.2 38 OMOA

W It O

lJ l

l i-(,

• he conse:uen:es of a breach of claccin uranium clexice to retain fjssion proca:g are greatly reou:ed by the acility of ts in:1ucing these wn3th are gaseous or hipnly volatile.

This retentiveness cecreases with increasing tecerature anc fuel outnue, but remains a significant factor even at fulipomer ocerating temperature in the maximum burnup element.

A survey of high burnuo uranium dioxice fuel element behavior 1rdi:stes that for an initial uranium ciexiee void volume, which is a function of the L

fuel censity, it is possible to conservatively cefine the fuel swelling as a i

fun: tion of turnve.

The fuel swelling eccel eens!:ers the effe:t of ev:nup, r

t te=erature :!stricution, and internal voles.

It is an en:!r!:a1 n: el.nl:$

II

• as :een :.tt: e: with :sta from Entt!s, van <ee, CVTA, !arton en:.. t 5.

Tne it

ellet :ensities for the tnree re;1crs are listem in Ta:1e 3.2.3-1.

li ne inte; !ty of fuel roc : lac:ing so as to retain.f3ssion p::cu:ts :: fuel 1

material is cire:tly relatee to : lac ing stress and strain unn: r.::* 61

erat! ; an: :,e:::.et :sn:!tions. Ces!;n limits sn: :s a;e lir.its (:;s::!r;

er'::st!:0 in terms of stress an: st: sin are as foll:.s Dama;e Lielt Cesien Limit Stress Ultimate < t.ength Yield strength.

]

57,000 psi minimum 45,000 ;si minimum ll Strain 1.7%

1.05 U

j.

a i '

The camage limits given above are minmlum values. Actual cara e limits cepenc ll v:en neutren exe:svre and normal variation of material propertles anc would

e.e

ally be ; ester tnan tnese minimum cams;e limits. For m:st of the fuel roc life the a:tual stresses anc strains are censicersely belew the cesign

. limits. Tnus, significant margins exist between actual e:erating con:Atlons l]

anc the cama;e limits.

The other :artmeters havin; an infiven:e on elaccin; stress an0 strain ano the relat!cnstic of these :arameters to the car. age limits are as fell:ws:

1.

Internal gas cressure:

The internal gas pressure recuired to procu:e cladding stresses epval to the damage limit un0er normal ocerating conditions is well in excess of the ranimum cesign pressure. The maximum cesign internal cressure un er i

nominal conditions is 2250 psia which is epual to the coolant pressure.

The enc of life internel gas pressure cepenes upon the initial pressure, l

void volube, and fuel roa power history, h:vever it does not exceed the g

design limit of 2250 psia.

2.

Cla::!n;: te*:eratute:

ine stren;th of the fuel :laccin; is tem:erature ce;encenb The minimum ultimate strength reev:es to the cesign yiele strength at an average claccing tet:erature of accreximately a50*r.

Tne maximum average :2st:in ter:eratu:e curin; normal e:e:ating :cncitions is gisen in ia:1e 3.2.2-1.;

t C1157 3.2-39 0060A j

We l

1 l

l IM

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Ultimate Strength I

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<- vield $trength j

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709 400 600 000 1000 Temperature (*F) j

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MECHAfilCAl. STRENGTH OF,,

• ROD TUDING VERSUS TEMPERATURE iq f

l t

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TIME (SECONDS) 7ser i

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=

+

-..--.,~,-ma.--.r,.e

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,--.e.e,

..--w

13 TABLE 1

(

14ad Engineeri@ttess As MunctliiinMcNirn%s 4tshisiire aniilYmel#enderIbsign Neop Stress (psi) for a 600 psi Differential Design Across the cladding Wall P

4570 SW 15:15 4540 SW 17:17 4280 C.E 16:16 4910 d ----

W 15:15 4 90-W 17:17 4050 St'8s8 3940

.iC15:15e c

3880 (NC8aP*

D. C. Cook. Unit 1

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. Sorst Tempere'ure ve.+rsus Stress et Burst fos,.Zw4. 5 / aNy.e:79a, ' b

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hl 45 q

1 1

---..m;

,I i

j FIGURE 10

[##h7

[

HEATING RATE LOO'C/ SEC

~Isostrain and rupture curves

,g @

plotted as a~ function of hoop stress and. temperature

\\

E fD\\,

.SC - S'9 #

for tubes heated at LOO *C/sec.

x\\ N s

s I

g i

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l

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. SHE ATM l EXPANSION---

i ti.

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f if I

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' 00 700 000 900 *8000 t800 8200 1300 6

g i

TEMPERATURE *C i

i

{

p &M ' r ' //2 -0fY D.

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10 15 20 25 ENGINEERING llOOP STRESS (KPSI) j-l Fla. 3 flM.*terrelation of rupture-temperature as a hection of eUNIHi45HWE Nftvets and 1esperatom vete with dete-fmm Internally heated Z1rtaloy claddian in aqueous

!j _

atmospheres.

Dsk w:

,s.

...m..

- ~

,~

n-g.

i l

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4

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50 / 7S l

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)

28 C/S

~

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l O'

O C/S 14 C/5

$l Y

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1 i

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to 15

-20 25 ENGINEERING HOOP STRESS (KPSI) i i

F10. 17 win model and ORNL correlatinn of rupture temperatiere as a functfan of enoineering i

teop stress and ramp rate.

l 1

l LN 1

-. -. -... ~., -. -

S.

e m. 2,4/ ~<a.

CLASS

)

p.

10 J

Clad Swelline and kucture Mod.il

...t 3.5 l in 7

During a 1.0CA the clad is assumed to strain unif ormly and plastical y.

i d t he dif ferential the radial dire tion provided that both the teciperature anIf the strain exceed, (a,c )

e s

pressure across the clad are suf ficiently high.

d as a f unct ion or the clad tetsperature esceeds the 5 urst temperature (determine s

alW an additional stress) the clad 1. a.ssumed to burnt g

an=====

of the instantaneoun

=.

==.

local strain is added to the burst node.

d Three empirical models are employed to evaluate the clad swelling an rupture behavior.

3.5.1 Clad Swelling prior to Rupture Y

intern.11 pres-perf ormed a series of tests in which rods with constant Hardy sure wet e raeped to a serier temperaturch at'variuu*. cenutant ramp rate.4 The pree mures rep.,rted by Hardy were converted to hoop strerec* by the formula e

(3-695 I

cortclated as and the strain at a given temperature and ramp rate wer6 The equation developed which best function,. of the derived hocp stress.

d e scr ibea. t h e d.s t a 16

• )

( n. t-)

r-(3-70) wherei

< n.o I

1 Y

i u

u l

1 1,

y 3:..

• imaas 19

!e WES" E00S!

4 1

)

(a,c) 1 (a.c)'

.i 1

(a.c)

)

f (a.c) l

[

{

L~

' 3.5.2 Clad burst t

g v

I f it eesches(10tlhoop strain based on the swelling #8)

Clad is assumed to burst i m

model described above or if the clad temperature an the burst node'rearlsen

. = = = = = =

c l

i

.! emperature is ca'culated an a function of

~

Eurst temperature. Ibu t

( cut hoop etress based on correlation of the W st ingliouse singic rod burst The bent estimate curve free figure 3-1 in used and data shown in Figure 3-1.

pressure is converted to hoop stress by the relationship described in Equat ion 3-69 using original test specimen geumetry. This best estimate 1

curve is described by the equation (a.h.r)

(3-71A)

Tg,g 3.53 Local Hot i Strain Af ter Surst l'

The loca'lized,dist ettsi swelling that occurs very rapidly _ at the time of,

tent dat.t oL i

is calculated from a correlation of sinr,le red burnt burst I'lgure M sh_!ws the correlat ion esce* the rangch Wcutinghmse and othe rn.

Expressed in terms of Imop st ress the correlation given of the data used.

g

~

s-Sk g g Ad*

d,

.I *

(3-728)

J l

I

. +. '

...w-+~e---+---,..-..._-___________m___

s..

t

=

j I

1. ~

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i i

g I

s i

x g z g-i g<

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+.

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s. e-

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++

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+

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s.,.

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ln 1

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a 8

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a i

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n

.s.

n 6

kin nim 88 5 EweM7 TEMPERATURE (DEG. C)

. '..... - - e, et,

== u-

.i.e<.e.mn, detM i ciaddIr.ginaqueousstrespheresforallheett. grates.

{

4 i

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p-NUREG 76/077 i

J.

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(

t r

g.

THE ROLE OF FISSION GAS RELEASE IN REACTOR UCENSING -

i 3-C CORE PERFORMANCE BRANCH I

[

i U. S. NUCLEAR REGULATORY COMMISSION NOVEMBER 1975

~

l I

l r

r i

e al:

s W

9 4

e O

I w

~w~~----wn-

,,,,,,e

.. -.,. - -., -, ~ - - - -

b c-M..

22 28 4

I I

~l

}

v 1,

'i i

i 4

lb i

i i

I 6

A d

1 d

=

)

A 4

k e

3

~

ic! -

7 4

E 4

g UP5tlNGl)DE TD CALCULATION 5.

UMIT ON WOLTEN FUEL TEMPER.

ATURE 4

M 5

A l

i 556 5

B 4

E C

cc EXP03URE KEY e O Cl-0.1 1020g,sgian,gg 20 ssion cc ti A

0.1-1.0 a 10 g

20 ssion/cc h

A 1.0-2.0 10 10-1 3

2.0-3.0 1020 giggion.gc y

3.0-4.0 a 1020g,gg,on;gg A e PREyl0U$LY REPORTED DAT A g

ALL OTHER-RECENT DATA 2000 2400 2000 3200 3600 4

g-2 1600 MAXIWUW FUEL ROD VOLUWETRIC AVERACE TEMPERATURE 0F 12 %

i The Hoffman & Coplin correlation of fission gas release Fig. 2.

as a function of temperature (from Ref. 35).

.r,..,

t 23 ZION e

CORE TEMPERATURE DISTRIBUTION 7

Assumptions: Operation at 3391 NWt for 500 days i'of Core Fuel Velvec local Temperature. 'F Above the Given Terperature 4100 0.0 3700 0.2 3300 3.8 2900 7.0 2500 24.5 i

l t

l 1

• )5

%},,.

Attachment UNitto sTATas

' E NUCLEAR REQULATORY COMMISSION

'i j

wasmorow.o. c. nosse f.

/

May 11, 1989 MEMORANDUM FOR: Thomas E. Murley, Director i

Office of Nuclear Reactor Regulation FROM:

Robert B. A. Licciardo, Reactor Engineer (Nuclear) j Plant Systems Branch 1

Division of Engineering and Systems Technology

SUBJECT:

DIFFERING PROFESSIONAL VIEW CONCERNING a)

Issuance of SER to Zion 1/2 allowing full power operation with open 42" containment isolation

valves, b) Methodology useo for calculating related offr.ite doses.

l The writer submits a Differing Professional View (DPV) in accordance with the provisions of NRC Manual Chapter 4125.

This issue has arisen out of the Safety Evaluation Report (SER) undertaken for the Zion Units 1 and 2 as prepared by the writer; see Attachment.

The principal issue is the prudent and conservative calculation of the additions

{' (

to offsite dose which may result from a LOCA at a facility during the use of t

.open purge supply and exhaust valves at full power.

n The licensee for Zion 1/2 has proposed full power operation of the facility with the 42" purge supply and exhaust containment isolation valves open to a limited position of 50', and capable of isolation within seven (7) seconds of the commencement of a LOCA.

The writers SER concludes that the 42" valves at Zion should remain closed I.

in Modes 1, 2 3 and 4 because the consequence of the offsite dose to thyroid (from iodine),during a LOCA is unacceptably high; whole body has not been evaluated. The least value for the additional offsite dose which may be l

proposed within the licensing basis is 64,000 rem over the first seven (7) seconds of the LOCA. Management staff has disagreed with the writer's methodology and conclusion and plans issuance of a separate SER permitting the operation requested. The writer requests non-issuance of the related SER to the Ifeensee.

He also proposes probability of a generic action on other i

facilities which have been granted such licenses based on the staff's current methodology.

In general, t'he management staff has adopted a criterion described in SRP BTP CSB 6-4 w'hich is that providing the maximum time for closure of these containment isolation valves does not exceed 5 seconds (and by plant-specific

~

exception, up to 15 seconds), then the valves would be closed before the onset of fuel failure following a LOCA so that the only contribution to offsite dose is from RCS operational levels of fission product directly dischayged into

/

containment during this period, and then through the open containment isolation valves before closure.

---

,e

~e-

~n,--

i.

. Thomas E. Murley-1

// f In evaluating the consequence for Zion, the writer has used an alternata Criterion in BTP CSB 6-4 which states that:

"The following analyses should be performed to justify the cont,ainment purge system design:

An analysis of the radiological consequences of a loss-of-coolant The analysis should be done for a spectrum of break accident.

sizes, and the instrumentation and setpoints that will actuate the The source term used in purge valves closed should be identified.the radiological calculations s the terrrs of Appendix K to detennine the extent of fuel failure and the concomitant release of fission products, and the fission product A pre-existing fodine spike should activity in the primary coolant.

The volume be considered in determining primary coolant activity.

of containment in which fission products are mixed should be justified, and the fission products from the above sources should be assumed to be released through the open purge valves during the The radiological maximum interval required for valve closure.

consequences should be within 10 CFR Part 100 guideline values."

Using these related guidelines for Zion,(the fuel performance over th is detailed and shows that fuel failure by infringement of DNBR criteria) occurs within i seconds of the correncement of the LOCA, and together with other licensing basis responses including fission product release from the fuel gap

('

and the thermal hydraulic conditions in the core, containment and discharge

(

nozzle, result in a substantive discharge of fission products to the environment of far greater consequence than are calcutsted by the staff.

The relative consequences of these differing approaches are that whereas the staff methodology gives additions to offsite dose resulting in total doses within 10 CFR Part 100 limits, the alternate approach used by the writer shows a substantially increased offsite dose exceeding 10 CFR Part 10011mits, with completely unacceptable consequences to Public Health and Safety.

The writer requests review of the Differing Professional View in a timely manner in accordance with the provisions of NRC Manual Chapter 4125.

f MW l

Robert B. A. Licciardo Registered Professional Engineer California 001056 Nuclear Engineering License No. NU Mechanical Engineering License No. M 015380 cc:

J. Sniezek D. Nuller l

S. Varga C. Patel sa F. Miraglia L. Shao A. Thadani J. Wermiel J. Kudrick

-w-

. -, - - -.,.,,. - - - - -.. - - -. ~ - - -

pussy'o UNITED STATES a

4 e

~g NUCLEAR REGULATORY COMMISSION j

e

.5

.j

]

wAemwaTow.o. c. eones

,/

May 11, 1989 t

~

Attachment Docket Nos. 50-295 and 50-304 o

- MEMORANDUM FOR:

Daniel Muller, Director Project Directorate III-2 Division of Reactor Projects III, IV, V I

and Special Projects I

r

- FROM:

Jared S. Wermiel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology

]

SUBJECT:

OFFSITE RADIOLOGICAL CONSEQUENCES OF LOCA DURING j

CONTAINMENT PURGE PROPOSED IN TS CHANGES FOR ZION 1 AND 2

Reference:

LettertoH.R.Denton(NRC)FromP,C.Leonarddated February 2,1986,

Subject:

Zion Nuclear Power Station, Units 1 and 2 Proposed Amendment to Facility Operating l

License No. DPR-39 and DPR-48

. Plant Name:

Zion Nuclear Power Station, Units 1 and 2 Licensee:

Consnonwealth Edison Company (2

TAC Nos.:

55417 and 55418 Review Status:

Complete i

Y, Zion Units 1 and 2 (Ceco) has responded to an NRC request to proposa TS to primarily constrain operation of the large (42") containment purge supply and exhaust valves on these units; see reference 1.

The former Plant Systems Branch, Section A, of the Division of PWR Licensing A, requested Section B cf the same branch to review the offsite radiological consequences of this proposal.

?

The enclosed Safety Evaluation Report has been prepared by the technical reviewer initially assigned to this task, namely Robert B. A. Licciardo.

The licensee's proposal is to allow full power operation of the facility with the 42" purge supply and exhaust containment isolation valves op(en to a7)secondsof limited position of 50' and capable of isolation within seven the commencement of a LOCA.

The review concludes that the 42" valves at Zion should remain closed in Modes 1, 2, 3. and 4 because the consequence of the offsite dose to thyroid (from fodine),during a LOCA is unacceptable high; whole body dose has not been h

The least value for the additional offsite dose which may be proposed evaluated:

e.g within the licensing basis is 64,000 rem over the first seven (7) seconds.

\\

The conveiitional treatment of BTP CSB 6-4 which assumes that fuel failure does notoccuroverthefirst5-15secondsafteraLOCAandtherebytyt,,onlyRCS operating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thermal hydraulic analyses for containment response, and licensing basis requirements (including criteria) for the calculation for, and the occurrence of, fuel damage and the quantification and treatment of resulting source terms.

i

~,----..--.---,~~---,-,,n-------n--

....,-n.n-..,-,

n, -

[..

L4 Daniel Muller

-2 i.i i.

Our SALP input is provided in Enclosure 2.

We consider our efforts ce TAC hos. 55417 and 55418 to be complete.

.i r,

p Jared S. Wermiel, Acting Chief Plant Systems Branch o

Division of Engineering and Systems Technology

Enclosures:

As stated s

cc w/ enclosures:

e C. Patel CONTACT:. R. Licciardo -

L XP0876 (i

l' l

l i,

e/ ff

. ~ _,

-.,-,-m-..

.,y,.

.,. _..... - _ _,_ -,-~_,,,..,, -,.

+

,a~...

1 11' Daniel Muller a-

,, o,-

Our-SALP input is provided in Enclosure 2.

We consider our. efforts on TAC Nos. 55417 and 55418 to be complete.

4 Jared S; Werniel, Acting Chief Plant Systems Branch Division of Engineering and Systems Technology s

Er. closures:

As stated-

^

cc w/ enclosures:

C. Patel s

CONTACT: R. Licciardo X20876 DISTRIBUTION pocket r11es Plant File-JWermiel JKudrick RArchitzel AThadani LShao TGody (SALP only)

RLicciardo l

SPLB: DEST SPLS: DEST SPLB: DEST RLicciargo;cf JKudrick JWermiel 5/A//89 5/ /89 5/ /89

/

/9' 5520 NAME: Zion TACs 55417/8 Licciardo l

ll

. ~

. ~.

( '

t t

'i. 4 i

UNITED STATES A(

3

.r NUCLEAR FLEGULATORY COMMISSION a

W ASHINGToN. o. C. 20596 c

// /*

i lp SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATI,0N PLANT SYSTEMS BRANCH OFFSITE RADIOLOGICAL CONSiCUENCE OF LOCA DURING CONTAINMENT PURGE ZION NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-295~and 50-304

1.0 INTRODUCTION

-l Zion Units 1 and 2 (Ceco) has responded to an NRC request to propose TS to 6

primarily constrain operation of the large (42") containment purge supply and exhaust-valves on these units.

l l

The former Plant Systems Branch, Section A, of the Division of PWR Licensing A, requested Section B of the same branch to review the offsite radiological consequences of this proposal.

2.0 EVALUATION Background review shows that the facility was evaluated on the basis of b/

normally closed purge valves so that these consequences were never included e

Further, that a letter from Westinghouse (onthesubjectof"Ofi~W)t to Cossnonwealth in the Zion SER.

si Edison Company dated October 22, 1976 LOCA and Containment Purge" (Ref. 2) has never been evaluated by the NRC.

Subsequent to that TMI-2 event, the operability and automatic control of these valves was evaluated leading to the request for the required TS, but the Radiological Assessment was left as a "long(er) term issue" (Ref. 3) which was intended to be resolved in a subsequent probabilistic risk assessment which definitively excluded it from consideration without any justification (Ref. 4).

uses an RCS The W analyses undet taken under Consnonwealth Edison instruction,f the accident operational inventory of 60 uc/gm equivalent I 131 at the time o with a resulting site boundary thyroid dose due to iodine (during closure of I

I' the valves), of 52 rem, and which added to the containment leakage dose of 123 I

rem gives a total 175 rem which is within the 16 CFR 100 Ifmit of 300 rem.

The total iodine inventory of the RCS is assumed to be released into containment L

on initiation of the LOCA; a 505 plate out is assumed leaving the residual 501 f

as part of containment inventory for discharge out through both fully open containment purge lines for a total of seven (7 seconds).

p However, when reviewed against the BTP CSB 6-4, Item B.5.a requires that:

"Tha source term used in the radiological calculations should be based on a calculation under the terms of Appendix K to deterinine the extent of L

and the fuel failure and the concomitment release of fission produftsj ic fission product activity in the primary coolant."

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2 qr Further: SRP 4.2 identifies fuel failure with infringement of DNBR criteria, with the related requirement that gap activity be considered as part of ~

the source term, and Regulatory Guide 1.77 recommends that under similar circumstances, gap activity should be assumed at 10% of core activity.- Tuel damage criteria also includes the occurrence of center line melting'Vith measures of additional activity release also guided by Regulatory Guide 1.77, j

but the Zion SAR shows this does not occur.

l Revising the source term to Appendix K calculations [in which all fuel goes to DNBP in i second] with related release of all gap activity into containment, with limited blowdown to offsite during the related 7 seconds closure time and absent a 50% plate out of iodine as can be interpreted from the above

'i referenced item B.S a. increases offsite dose due to containment purge above by a f actor of 3400 to 176,000 rem and would thereby be completely unacceptable.

Limiting the purge line valves to an cpening of 50' could reduce offsite dose to 64,000 rem and represents the least value which may be proposed within the licensing basis.

5 Note: The BTP CSB 6-4 proposing that valve closure within 5 seconds will ensure purge valves are closed before the onset of fuel failures has since been extended by the s*sff on a plant-specific basis to 15 seconds.

Further, the writer cannot fino any safety evaluation report supporting these positions.

These positions cannot be sustained for Zion since a) DNBR infringement (from Appendix X calculations) and hence fuel failure and gap activity release [Ref.

('

SRP 4.2) of 10% of core inventory (Ref. Regulatory Guide 1.77) occur within 1 second of the initiation of the LOCA, b) related maximum clad temperatures of 1750*F occur insnediately and never reduce below 1400*F, c) RCS pressure in the regicn of the core rapidly reduces from 2250 psia to 900 psia in 7 seconds activity to the RCS inventory, d)p across the cladding for release of gap increasing potential pressure dro the massive bulk boiling and blowdown l

surrounding the failed fuel ultimately discharges 270,000 lbs of RCS inventory into the containtrent at 7 seconds into the event increasing containment pressure from 0.3 psig to 23.8 psig (in these 7 seconds), and e) causes 15,000 lbs of the resulting containment inventory to be discharged to the environment through 2x42" fully open lines, or 5400 lbs for the same lines with valve closed to 50'.

3.0 CONCLUSION

t 1

i The 42" valves at Zion should remain closed in Modes 1, 2, 3, and 4 because the consequences of the offsite dose to thyroid (from iodine) during a LOCA is unacceptably high; whole body dose has not been evaluated. The least value for offsite dose to the thyroid which may be proposed within the existing licensing basis is 64,000 rem.

The conventional treatment of BTP CSB 6-4 which assumes that fuel failure does not occur over the first 5-15 seconds after a LOCA and thereby that only RCS operating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thermal hydraulic analyses for containment response, and licensing basis requirements (including criteria) for the calculation for, 3rd the occurrence of, fuel dynap and the quantification and treatment of the resulting source terms l

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1.

Letter from P. C. Blond (CEto) to H. R. Denton (NRC);

Subject:

Zion, Units 1 and 2. Proposed Amendment to Facility Operating'L4 cense Nos. DPR-39 and DPR-48 cated February 21,1986.

2.

Letter from R. L. Kelley (W) to C, Reed (Ceco);

Subject:

Offstte Dose During LOCA and ContaTnment Purge, dated October 22, 1986.

3.

Letter to L. O. De1 George (Ceco) from S. A. Varga (NRC);

Subject:

Generic Concerns of Purging and Venting Containments, dated September 9, 1981.

4 Memo for F. H. Robinson from R. W. Houston,

Subject:

" Evaluation of the Risk at Zion " dated August 14. 1985.

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Plant Name:

Zion Nuclear Generating Stations, Units 1 and 2 SER

Subject:

- Containment Purge and Vent Yalve Operation TAC Nos.:

55417/8 Summary of Review / inspection Activities The licensee provided an evaluation of offsite doses undertaken in 1976. This was undertaken with a methodology and source term chosen by the licensee. The Ifeensee did not present results from alternative more detailed methodologies which could be considered enforceable under existing regulatory positions and the related circumstances.

Narrative Discussion of Licensee Performance - Functional Area l

The single only methodology used by the licensee is not an acceptable approach for estimating doses under the proposed circumstances and especially since alternate detailed evaluations required by the SRP give greatly increased values beyond 10 CFR Part 100 limits. A prudent approach would have recognized the deficiencies and risks in the single methodology adopted with gi resulting substantively different recommendations to ensure public health and safety.

Author: Robert B. A. Licciardo Date:

May 11, 1989 i

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July 14, 1989 l

MEMORANDUM FOR: Frank J. Miraglia, Associate Director i

for Inspection and Enforcement I

FROM:

Robert B. A. Licciardo, Reactor Engineer Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT ISOLATION VALVES AT ZION By memo dated June 30, 1989, the writer proposed to submit requested clarifications of his DPV by July 17, 1989. He would like to re-schedule this submittal to July 20. He is of course, prepared to agree to 6n extension of the required formal completion of t1e review of his DPV, by the same time period, t

Robert B. A. Licciardo Registered Professional Engineer California u

Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. H 015380 cc:

J. Snierek C. Rossi F. Congel H. Smith l

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July 14,1989 HEMORANDUM FOR: Frank J. Miraglia, Associate Director for Inspection and Enforcement FROM:

Robert 8. A. Licciardo, Reactor Engineer Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT ISOLATION VALVES AT ZION By memo dated June 30, 1989, the writer proposed to submit requested clarifications of his DPY by July 17, 1989. He would like to re-schedule this submittal to i n 00. He is of course, prepared to agree to an extensionoftherehliredformalcompletionofthereviewofhisDPV, by the same time per lod.

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Robert B. A. Licciardo j

Registered Professional Engineer California Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. M 015380 cc:

J. Sniezek C. Rossi F. Congel H. Smith 1'

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' MEMORANDUM FOR: Frank J. Miraglia, Associate Director -

for Inspection and Enforcement

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FROM:

Robert B. A. Licciardo, Reactor Engineer Plant Systems Branch i

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Division.of Engineering and Systeias Technology l

SUBJECT:

DIFFERINGPROFESSIONALVIEW(DPV)CONCERNINGCONTAINMENT l

ISOLATION VALVES AT ZION On June'16, 1989, the writer did elaborate for the Standing Review Panel upon the principal regulatory positions sumarily presented in his DPV of May 11, 1989 He shall be please to clarify further on the specific issues identified I.

in your memo to him of June 23, 1989, and will do so by July 17, 1989.

Robert B. A. Licciardo i

Registered Professional Engineer California Nuclear Engineering License No. NU 001056 a

Mechanical Engineering License No, M 015380 r

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H i-MEMORANDUM FOR: Robert Licciardo

- Reactor Engineer Plant Systems Branch j

Division of Engineering and Systems Technology FRON:

Frank J. Miraglia, Associate Director for Inspection and Technical Assessment j

SUBJECT:

DIFFERING PROFESSIONAL VIEW (DPV) CONCERNING CONTAINHENT ISOLATION VALVES AT ZION The Standing Review Panel of Frank Miraglia, Charles E. Rossi and Frank Congel

'l-reviewed the material submitted to Dr. Murley on the subject matter. The Panel met with you on Friday, June 16, 1989 to further discuss your views. At that meeting the Panel requested that you more clearly state your concern regarding the time to fuel failure used in LOCA analyses. The Panel also requested that you also clarify the mechanisms for transporting fission products from the primary to containment used in your analyses.

In addition, the Panel requested that you provide your view as to the safety significance of proceeding with the proposed Zion amendment and the safety significance of your concern regarding LOCA analyses.

Please let me know when you will provide the requested infomation. As we have indicated to you previously it is our intent to comply with the milestones in NRC Manual Chapter 4125 and NRR Office Letter 300.

Original signed by hemiaJ.Eins11a i

Frank J. Miraglia, Associate Director for Inspection and Technical Assessment i.

cc:

J. Sniezek l

J. Larkins l

C. E. Rossi F. Congel l

DISTRIBUTION l

Central File ADT/RF FMiraglia f$}gfgh'Y DIFFERING PROFESSIONAL VIEW

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UNITED STATES NUCLEAR REGULATORY COMMISSION 5

WASHING TON, D. C. 70655 June 2, 1989 MEMORANDUM FOR: Robert Licciardo i

Reactor Engineer Plent Systems Branch Division of Engineering and Systems Technology FROM:

Frank J. Mirag116, Associate Director for Inspection and Technical Assessment

SUBJECT:

DIFFERING PROFESSIONAL VIEW (DPV) CONCERNIN ISOLATION VALVES AT ZION i

In accordance with NRC Manual Chapter 4125 and NRR Office Letter 300 Standing Review Panel of Frank Miraglia, Charles E. Rossi and Frank

, the reviewed the mater 161 submitted to Dr. Murley on the subject matter.

Panel has determined that adequete information has been supplied to ini The a review of your DPV.

It is our intent to meet with you in the near future.

i Frank J. k ragl a, sociate Director for Inspection and Technical Assessn:ent cc:

J. Sniezek J. Larkins C. E. Rossi F. Congel t

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MEMORANDUM FOR:

Frank J. Miraglia, Associate Director y

for Inspection and Technical Assessment, NRR C. Ernie Rossi, Director Division of Operational Events Assessment, NRR p

Frank J. Conpel, Director Division of Radiation Protection and Emergency Preparedness, NRR FROM:

Thomas E. Murley, Director Office of Nuclear React,r Regulation

SUBJECT:

DIFFERING PROFESSIONAL VIEW OF ROBERT B.A. LICCIARDO CONCERNING CONTAINMENT ISOLATION VALVES AT ZION Enclosed is a memorandum from Mr. Liectardo to Dr. Murley, dated May 11, 1989 expressing a Differing Professional View.

In accordance with NRC Manual f

, Chapter 4125 ard NRR Office Letter No. 300 dated March 24, 1989, you are hereby designated as the Panel to review and recommend to the Director, NRR the appropriate disposition of Mr. Licciardo's Differing Professional View.

If you deem it necessary, you may solicit _ input from other NRR technical staff or contractors.

In carrying out your review and formulating your reconnendations to se, you chould ba guided by the Appendix to NRC Manual Chapter 4125 with special i

4 emphasis on Sections B 6 and B.7.

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Thomas E. Murley, Utrector

Enclosure:

As stated Office of Nuclear Reactor Regulation cc:

J. H. Sniezek J. Larkins R. Licciardo l

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MEMORANDUM FOR:.

Thomas E. Murley, Director Office 'of Nuclear Reactor Regulation

. F ROM:'

Robert B. A. Licciardo Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

DIFFERENT PROFESSIONAL VIEW (DPV) CONCERNING ZION.

PERSONS PROPOSED AS THIRD MEMBER OF STANDING REVIEW e

PANEL.

n On May'19 I received your request to submit a listing of persons to consider as.the third (and) alternate member of the Standing Review Panel for the purpose of reviewing the writers D.P.V dated May 11, 1989.

For this purpose I nominate:

Steven A. Varga, Director, Division of Reactor Projects

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Gary M. Holahan, Acting Associate Director for Regions III and V Frank J. Congel, Director, Division of Radiation Protection and Emergency Preparedness My understanding from NRC Appendix 4125, Section B.1 is, that the curront role of the panel is to determine if enough information has been supplied to undertake a detailed review of the issue.

And that given a favorable review, i

the necessary interdisciplinary expertise can be assembled to formulate a final disposition.

On this besis,.the above persons are nominated.

l W

Robert B. A. Licciardo l

Registered Professional Engineer, California i

Nuclear Engineering License No. NU001056 Mechanical Engineering License No. M015380 cc:

J. Sniezek F. Miraglia J. Partlow J

S. Varga l'

G. Holahan l

E. Rossi I

L. Shao C. Patel l

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MAY 181989 MEMORANDUti FOR: Robert Licciardo, Reactor l

i Engineer (Nuclear)

Plant Systems Bran:h l

Division of Er.gineering I

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and Systens Technology f

FROM:

Thomas E. Murley, Director p

Office of Nuclear Reactor Regulation SUPOECT:

DIFFERING PROFESSIONAL VIEW CONCERNINC (A) ZION 1/2 l

CONTAllEENT ISOLATION VALVES, AND (B) METHODOLOGY USED i

FOR CALCULATING RELATED OFFSITE DOSES This is to acknuwledge that on May 12, 19F.9 i received your Differire Professional View (DPV) concerning the captioned subject. Pleasesubmite listing of persons you would like me to consider as the third metber of the Standing Review Panel and as an alternate menber for the Standing Review Panel.

The Standing Review Panel will determine within 7 days if adeouate informatior l

har been supplied to initiate a review of your D V.

E. Murle b r tor i

ma Office of I;ucle Rea tor Regulation cce J. Snierek l

F. Miraglia J. Part10w S. Varga G. Holahan E. Rossi L. Shao C. Patel i..

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l CONTACT:

H. Smith, PMAS I

X21287 l

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MEMORANDUM TOR: Thomas E. Murley, Director Office of Nuclear Reactor Regulation RobertB.A.Licciardo,ReactorEngineer(Nuclear)-

TRON:

Plant Systems Branch Division of Engineering and tyttems Technology J

DIFFERING PROFES$10NAL VIEW CONCERNING

\\

SUBJECT:

Issuance of SER to Zion 1/2 allowing full power i

a) operation with open 42" containment isolation valves.

b) Methodology used for calculating related offsite doses.

The writer submits a Differing Professional Ytew (DPV) in accordance with the l

provisions of NRC Manual Chapter 4225.

This issue has arisen out of the Safety Evaluation Report (SER) undertahn for the Zion Units 1 and 2 as prepared by the writer; ste Attachment.

The principal issue is the prudent and conservative calculation of the additions i

to offsite dose which may result from a LOCA at a facility during the use of c

['

open purge supply and exhaust valves at full power.

~

The licensee for Zicn 1/2 has proposed full power operation of the facility

(

with the 42* purge supply and exhaust containment isolatio i

j of the comencement of a LOCA.

The writers SER concludes that the 42' valves at Zion sho I

l (from iodine) during a LOCA is unacceptably hight whole body has not been l

The least value for the additional offsite dose which may be l

proposedwithinthelicensingbasisis64,000removerthefirstseven(7) evaluated.

Management staff has disagreed with the writer's l

methodology and conclusion and plans issuance of a separate S seconds of the LOCA.

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the operation requested.He also proposes probability of a generic action on other facilities which have been granted such ifcenses based on the staff's current to the licensee.

I methodology.

In general, the management staff has adopted a critation described in SRP BTP CSB 6-4 which is that providing the maximum time fo exception, up to 15 seconds), then the valves would be c is from RCS operational levels of fission product directly valves befcre closure.

Me apwur W

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i Thomas E. Murley 2

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In evaluating the consequence for Zion, the writer has used an alternate Criterion in BTP C58 6 4 which states that:

'The following analyses should be performed to justify the tontainment purge system design:

An analysis of the radiological consequences of a loss-of-coolant accident. The analysis should be done for a spectrum of break stres, and the instrumentation and setpoints that will actuate the a

purDe valres closed should be identified. The source term used in the radiological calculations should be based on a calculation under the terms of Appendix K to determine the extent of fuel failure and the concomitant release of fission products, and the fission product activity in the primary coolant. A pre existing todine s sike should be considered in determining primary coolant activity. Tae volume of containment in which fission products are mixed should be justified, and the fission products from the above sources should be assumed to be released through the open purge valves during the saximum interval required for valve closure. The radiological l

consequences should be within 10 CFR Part 100 guideline values.'

i UsingtheserelatedguidelinesforZion,(byinfringementofDNBRcriteria) the fuel performance over the 0-7 seconds

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is detailed and shows that fuel failure occurs within i seconds of the corsnencement of the LOCA, and together with other licensing basis responses including fission product release from the fuel gap and the thermal hydraulic conditions in the core, containment and discharge norrle, result in a substantive discharge of fission products to the i

environment of far greater consequence than are calculated by the staff.

The relative consequences of these differing approaches are that whereas the staff methodology gives additions te offsite dose resulting in total doses within 10 CFR Part 100 limits, the alternate approach used by the writer j

shows a substantially increased offsite dose exceeding 10 CFR Part 100 limits, l

with completely unacceptable consequences to Public Health and Safety.

The writer requests review of the Differing Professional View in a timely manner in accordance with the provisions of NRC Manual Chapter 4125.

i e A&f M Robert B. A. Licciardo Registered Professional Engineer California t

Nuclear Engineering License No. NU 001056 Mechanical Engineering License No. M 015380 cc:

J. Snier'ek D. Nuller S. Varga C. Patel F. Miraglia L. Shao l

A. Thadant J. Werniel J. Kudrick l

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May 11, 1989 ooo,e Docket Nos. 50-295 Attachment and 50 304 m

MEMORANDUM FOR: Daniel Muller, Director Project Directorate !!!-2 Division of Reactor Projects !!!, IV, Y

)

and Special Projects l

FRON:

Jared 5. Wermiel, Acting Chief Plant $ystems Branch Division of Engineering and Systems Technology j

$UBJECT:

OFFSITE RADIOLOGICAL CONSEQUENCES OF LOCA DURING CONTAINMENT PURGE PROPOSED IN TS CHANSES FOR ZION 1 AND 2

Reference:

LettertoH.R.Denton(NRC)FromP.C.Leonarddated l

February 2,1986,

Subject:

Zion Nuclear Power Station, i

Units 1 and 2 Proposed Amendment to Facility Operating License No. DPR-39 and DPR-48 i

Plant Name:

Zion Nuclear Power Station, Units 1 and 2 i

Licensee:

Cosunonwealth Edison Company TAC Nos.:

55417 and 55418 Review Status:

Complete ZionUnits1and2(Ceco)hasrespondedtoanNRCrequesttoproposeTSto primarily constrain operation of the large (42') containment purge supply and exhaust valves on these units; see reference 1.

The former Plant Systes.s Branch, Section A, of the Division of PWR Licensing A, requested Section 8 of the same branch to review the offsite radiological consequences of this proposal.

l t

The enclosed Safety Ivaluation Report has been prepared by the technical reviewer e

i initially assigned to this task, namely Robert b. A. Licciardo.

1 l

The licensee's proposal is to allow full power operation of the facility with the 42" purge supply and exhaust containment isolation valves op(en to a7)secondsof limited position c,f 50', and capable of isolation within seven the commencement of a LOCA.

I The review concludes that the 42" valves at Zion should remain closed in p.

Modes 1, 2, 3.and 4 because the consequence of the offsite dose to thyroid i

(from iodine),during a LOCA is unacceptable hight whole body dose has not been evaluated: The least value for the additional offsite dose which may be proposed uithin the Itcensing basis is 64,000 rem over the first seven (7) seconds.

The conventional treatment of BTP CSB 6-4 which assumes that fuel failure does not occur over the first 5-15 seconds af ter a LOCA and thereby that only RCS cperating inventory of fission products is released to the containment, and then to the environment, cannot in general be sustained against thennel hydraulic analyses for containment response, and licensing basis requirements (including criteria)forthecalculationfor,andtheoccurrenceof,fueldamageandthe quantification and treatment cf resulting source terms.

GivP907xoniWM 2tY

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Daniel Muller !

ss s' Our $ ALP ipput is provided in Enclosure 2.

We consider our efforts og TAC Nos. 55417 and $5418 to be complete.

Jared S. Wermfel. Acting Chief Plant Systems Branch l

Division of Engineerir.g and Systems Technology l

Enclosures.

As stated j

cc w/ enclosures.

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CONTACT: R. Licciardo X20876 l

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Daniel Muller 2-j l

Our SALP input is provided in Enclosure 2.

We consider our efforts os TAC

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Nos. 55417 and 55418 to be complete.

i Jared S. Wermiel, Acting Chief i

Plant Systems Branch l

Division of Engineering and Systems Technology l

Enclosures:

As stated cc w/ enclosures:

C. Patel CONTACT: R. Licciardo X20876 1!STRIBUTION Jocket files Plant File JWermiel 2.

JKvdrick RArchittel AThadani LShao TGody($ALPonly)

RLicciardo SPLB: DEST SPLB: DEST SPLB: DEST RLicciardo;cf JKudrick JWermiel

$////89 5/ /89 5/ /89 5520 NAME:

Zion TACs 55417/8 Licciardo

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5AFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

PLANT SYSTEMS BRANCH 0FF$1TE RADIOLOGICAL CONSE0VENCE OF LOCA DURING j

CONTAINMENT PURGE ZION NUCLEAR POWER STATION, UNIT $ 1 AND 2 DOCKET NOS. 50 295 and 50 304

1.0 INTRODUCTION

ZionUnits1and2(Ceco)hasrespondedtoanNRCrequesttoproposeT5to i

primarily constrain operation of the large (42*) containment purge supply and exhaust valves on these units.

j The former Plant Systems Branch, Section A, of the Division of PWR Licensing A, requested Section B of the same branch to review the offsitt radiological 4

i consequences of this proposal.

p 2.0 EVALUATION Background review shows that the facility was evaluated on the basis of b

l normally closed purge valves so that these consequences were never included i

i i

in the Zion SER. Further, that a letter from WestinghouseiW) to Comonwealth LOCA and Containment Nrge' (Ref.1976 on the subject of *0friite Doses During Edison Company dated October 22

2) has never been evaluated by the hRC.

l l

Subsequent to the TMI-2 event, the operability and automatic control of these t

valves was evaluated leading to the request for the required 75RadiologicalA 3)whichwas I

intended to be resolved in a subsequent probabilistic risk assessment which definitively excluded it from consideration without any justification (R6f. 4).

The W analyses undertaken under Commonwealth Edison instruction, uses an RCS operational inventory of 60 uc/gm equivalent ! 131 at' the time of the accident j

with a resulting site boundary thyroid dose due to iodine (during closure of the valves), of 52 rem, and which added to the containment leakage dose of 123 i

rem gives a total 175 rem which is within the 10 CFR 100 limit o' 300 rem.

i The total iodine inventory of the RCS is assumed to be released into containment l

on initiation of the LOCA; a SOE plate out is assumed leaving the residual SOS as part of containment inventory for discharge out through both fully open containment purge lines for a total of seven (7 seconds).

L However,when;reviewedagainsttheBTPCSB6-4,Jtem8.5.arequiresthat l

"The source term used in the radiological calculations should be based on a calculation under the terms of Appendix K to determine the extent of i

fuel failure and ths concomitment release of fission products, and the l

fission product activity in the primary coolant.'

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$RP 4.2 identifies fuel failure with infringement of DNBR csiteria,

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c tsith the related requirement that gap activity be consider Further:

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circumstances, gap activity should be assumed at 105 of core Activity.

i damage criteria also includes the occurrence of center line melting with l

measures of additional activity release also guided by Regulatory Guide 1.7r i

but the Zion SAR shows this does not occur.

RevisingthesourcetermtoAppendixKcalculations[inwhichallfuelgoes l

toONBRinisecond)withrelatedreleaseofallgapactivityintocontainment, i

with limited blowdown to offsite during the related 7 seconds closure time and absent a 50% plate out of iodir.e as can be interpreted from the above l

referenced item B.5.a. Increases offsite dose due to containment pu l

l i

by a factor of 3400 to 176 000 Limiting the purge line valves to an opening of 50' co l

licensing basis.

The BTP CSB 6-4 proposing that valve closure within 5 seconds will ensure purge valves are closed before the onset of fuel failures has since Note:

Further, been extended by the staff on a plant-specific basis to 15 seconds.

i the writer cannot find any safety evaluation report supporting these pos These positions cannot be sustained for Zion sinc i

Ref. Regulatory Guide 1.77) occur within i i

SRP 4.2) of 10% of core inventory (A, b) reltted maximum clad temperatures of second of the initiation of the LOC c) RCS pressure in the l

1750'F occur imediately and never reduce below 1400*F regionofthecorerapidlyreducesfrom2250psiato9b0psiain7 seconds increasing potential pressure drop across the cladding for release of gap i

d) the massive bulk boiling and blowdown activity to the ACS inventoryltimately discharges 270,000 lbs of RCS inventory surrounding the failed fuel u into the containment at 7 seconds into the event increasing containment :

from 0.3 psig to 23.8 psig (in these 7 seconds), and e) causes 15,000 h

the resulting containment inventory to be discharged to the environme 2x42" fully open lines, or $400 lbs for the same lines with valve clo

3.0 CONCLUSION

e The 42" valves at Zion should remain closed in Modes 1, 2, 3} and 4 be i

during a LOCA n

the consequences of the offsite dose to thyroid (from iodine The least value is unacceptably hight whole body dose has not been ev Itcensing basis is 64,000 rem.

The conventio'nal treatment of BTP CSB 6 4 which as i

RCS not occur over the first 5-15 seconds after a LOCA and thereby that on operating inventory of fission products is released to the conta lic to the environment, cannot in general be sustained against thernal hy(d including analyses for containment response, and licensing bas quantification and treatment of the resulting source terns,

r-e l

References f,/ <-

1.

LetterfromP.C. Blond (CECO)toH.R.Denton(NRC);

Subject:

Zion, Units 1 and 2, Proposed Amendment to facility Operating License Nos. DPR 39 and DPR-48 Cated February 21,1986.

j 2.

Letter from R. L. Kelley LW) to C. Reed (Ceco);

Subject:

Offsite Dose During LOCA and ContUnment Purge, dated October 22, 1966.

i 3.

LettertoL.O.De1 George (Ceco)fromS.A.Yarga(NRC);

Subject:

Generic Concerns of Purging and Venting Containments, dated September 9, 1983.

4 Memo for F. H. Robinson from R. W. Houston,

Subject:

" Evaluation of the Risk at Zion," dated August 14, 1985.

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$PLB SALP INPUT Plant Name:

Zion Nuclear Generating Stations Units 1 and ?__.

SER

Subject:

Containment Purge and Vent Valve, Operation TAC Nos.:

55417/8 lumpary of Review / inspection Activities The licensee provided an evaluation of offsite doses undertaken in 1976.

was undertaken with a methodology and source term c This The which could be considered enforceable under existing regulatory positions and the related circumstances.

Narrative Discussion of Licensee Performance - Functional Area for estinating doses under the proposed circumstances alternate detailed evaluations required by the SRP give greatly increased values beyond 10 CFR Part 100 Ifmits.

A prudent approach would have recognized the deffetencies and risks in the single methodology adopted with resulting substantively different recommendations to ensure pubite health and safety.

Author: Robert 8. A. Licciardo Date:

May 31, 1989 4

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