ML20140H064

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Provides Info Re Necessity of Further Explication of ACRS Rept Concerning catch-all Paragraph on Problems About Lwrs.Related Info Encl
ML20140H064
Person / Time
Site: Midland, 05000000
Issue date: 08/06/1976
From: Libarkin M
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
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ML19255C661 List: ... further results
References
FOIA-85-602 NUDOCS 8510080429
Download: ML20140H064 (7)
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.g [ ,g" NUCLEAR REGULATORY COMMISSION 39  ? ,, ,. ADVISORY COMMITTEE ON HE ACTOR SAFEGUARDS

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August 6, 1976 l

ACRS Members HIDLAND "CATCl!41.1." pARAGRAPli On July 26, 1976, R. Muller provided yea with a copy of a U. S. Court of Appeals ruling on Midland which stated, in part, "...further explication of the ACRS report was necessary..." in connection with the so-called " catch-all" paragraph:

"Other prob 1cma related'to large water reactors have been identified by the Regulatory Staff cnd the ACRS and cited in previous ACRS reports.

l The Comittec belicycs that resolution of these items should apply

  • equally to Midland Plant Units 1 and 2.",

'The Court concluded that the Midland licaring Boned dould have returned, , , , ,

the ACRS' Report to the Conuittee for claboration ef the reference to "other probicms." -

This memo is intended to provide background infomation in anticipation l

of the need for such an elaboration.

The Comittee's report on the Midland CP application was distributed on June 18, 1970. It stated that the HSSS and ECCS proposed for Midland l were escontially identical with those for the previously reviewed Oconce and Rancho Seco Plants. Attachment 1 is a discussion of major similaritics and difference excerpts from the Midland PSAR. The Oconce and Rancho

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Seco reports were written in July 1967 and July 1968, respectively. The list of generic items referred to as "other probicms" in the Ccemittee's report on Midland should therefore, be those which were thought to apply to Oconce and Rancho Seco plus any that may have been identified as being generally applicabic between July 1968 and June 1970. P e

The Rancho Seco generic items paragraph referred to a previous ACRS report on Crystal Rivor l

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"This reactor is similar to others designed by this vendor and' reviewed y previously (sec, for exatr.ple, the ACRS report on the Crystal River Plant, May 15,1968). The Concittee cuntinues to call attention to matters that warrant careful consideration by the manufacturcrs of all largo, water-cooled, powcr reactors. These matters, referred to in the abovn-mentioned report, apply similarly to the Rancho Seco project." .

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August 6, 1976 ,

..o Hembers The Crystal River Report cited refers in turn to other, still carlier, ACRS reports en Oconce (7/11/67) and Three Mile' Island Unit 1 (1/17/6S).

Neither of those latter included a " generic items" paragraph. They did include reference to matters of general applicability to water-cooled reactors and to reactors designed by B&W:

6 Three Mile Island (1/17/68)

1. Diversity of ECCS initiation
2. Ieproved scram reliability
  • 3. Separation of protection and control instrumentation
4. Development of a f ailed fuel cicment monitor
5. R&D aimed at providing assurance that LOCA-rclated fuel failures will not interfere uith ECCS function '
6. Potential for axial xenon oscillations
7. The effects of blowdoun forces on core internals
8. The ef fect on Pressure Vessel integrity of ECCS-induced thermal shock
9. The behavior of core-barrel check valves in normal opera tion.

0conec (7/11/67)

1. The effect on Pressure Vessel integrity of ECCS-induced thermal shock
2. The ef fects of blowdown forces on core internals and other primary system components
3. Assurance that LOCA-related fuel nod failures will not interferc with ECCS function
4. Evaluation of the overall effect of the core-barrel check valves
  • 5. Diversity of ECCS initiation
6. Improved QA and in-scryice inspection of the primary system
7. Fuel integrity during end-of-life transients ,

B. The potential for xenon oscillations

,'. 9. Improved containment liner wcld inspection e

b l A copy of each of the ACRS reports cited above is attached for information Also attached (Attachments 7-15) are cxcerpts from

. ( At tachments 2-6).

several ACRS reports written during the period 7/68 to 6/70 which identify matters which might have been includcd in the Committee's reference in the Midland r,cport to "Other probicms... cited in previous ACRS reports."

Likely candidates include:

Continuous monitoring of t aron concentration (H. B. Robinson, 4/16/70).

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2. The potential consequerces of fuel handling accidents (Indian Point 3, 1/15/69,; Hutchinson Island, 3/12/70; etc.)
  • 3. Environmental qualffication of vital equipment in containment (Palisades, 1/7.7/70).

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August 6, 1976 *

4. Instrumentation to follow the course of an accident (Hutchinson Island, 3/12/70; Beaver Valley, 3/12/70; Point Beach, 4/16/70;
  • l H. B. Robinson, 4/16/70; etc.)
5. Vibration and loose parts monitoring (Point Beach, 4/16/70).
6. Failure to scram during anticipated transients - ATWS (Beaver Valley, 3/12/70; Duane Arnold, 12/18/69; etc.)

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Ther,e seems to be no other guidance available which would, help in the requested clarification.

M. W. "

M. W. Libarkin, Assist, ant i Executive Director for Project Review .

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13 TAnuudi cwuvsCTERISTICS h ter-Table 1-2 is a comparative list of important design and operating c arac Municipal istics of the Midland Units 1 and 2, Rancho Seco Unit 1 )(Sacra = ento d Turkey Utility Dictrict), Oconee Units 1, 2, and 3 (Duke Power The Company , an design and oper-Point Units 3 and 4 (Florida Power and Light Company). its are close ating parameters of the Rancho Seco, Oconce, and Turkey Point unRancho Seco i t in other to those of Midland Units 1 and 2.

saine rated core power as the Midland units, licensing.

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and are near-dupl station deceriptions and in Safety Analysis Reports submitted for Amendment No. 5 1-5 11/ 3/69 p[i a

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. o 4 du The design of each of these stations is based on information developed from operation of commercial and prototype pressurized water reactors over a number of years. The Midland unit desiEn is based on this existing power reactor technology and has not been extended beyond the boundaries of known informa-tion or operating ex;crience.

The similarities and differences of the features of the reactor units listed in Table 1-2 are discussed in the following paragraphs. In each case, the item number used refers to the item numbers used in the table.

Item 1. Hydraulic and Themal Design Parameters The rated lower of each Midland unit is the same as that for the Oconee and Rancho Seco units. The slight variation in other parameters between the Mid-land units and the other B&W units is due to the utilization of canless fuel assemblies in place of the canned fuel assemblies. The canless assembly allovs a slightly larger fuul rod which increases fuel loading and heat trans-fer surface area. Elimination of the can wall results in a slightly lower Icwer leaking factor and the more open lattice of the can'less assembly in-creases coolant flow area. The reactor coolant flow rate, operating pressure, and orcrating coolant temperature are the same for the Midland, Oconee, and Psncho Seco units. The conservatism of design of the Midland units is evi-denced by the DNp2 cf 1.71 (W-3) at the overpower condition compamd to essen-tially the same value for the other B&W units and to a lower value for the other reactor presented.

p Item 2. Core Mechanical Design Parameters The table presente com;nrable mechanical design data for the canless fuel assembly for the Midland units, the canned fuel assembly for the Oconee and Rancho Seco units, and the canless fuel assembly used for the Turkey Point units. The dimennions, t.sterials, and technology fcr each of these reactors are similar. Differences between the B&W units and the Turkey Point units are related to differences in power levels.

The small differences in fuel rod dimensions between the Midland units and the other B&W reactors result from the utilization of larger fuel rods and a 3arger fuel rod pitch to match fuel assembly pitch of the canless and canned fuel assemblics. The number of fuel rods per core is unchanged. The lesser number of control rod assemblies in the Midland units compared to the other B&W units results from a reduction in the requirements for inserted control rod assemblics for equilibrium xenon and transient xenon control. Burnable poison red assemblies described for the Midland units result from the higher first cycle burnup shown in the Preliminary Nuclear Design Data, Item 3 Item 3 Preliminary Nuclear Design Data The core size, number of fuel assemblies, and number of fuel rods are the same for all of the B&W units and differ from the Turkey Point units primarily due to the difference in }over level. Fuel enrichments differ between the B&W units primarily due to the different fuel cycle burnup requirements. Enrich-ment increase for the higher first cycle burnup of the Midland units is par-tially offset by the reduced amount of structural steel in the canless 1-6 w

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assembly. The excess resetivity requirements for each reactor also vary with fuel cycle burnup; the higher burnup of Midland units is reflected in the higher initial excess reactivity. The Midland units have fewer cont.rol rod asnomblies than do the Oconee and Rancho Seco units and more control rods than the Turkey Point units. The reduction in the number of control rod assemblics and control rod worth for the tiidland units is due to less control rod inser-tion in the core during operation for compensation of equilibrium and tran-sient xenon reactivity changes. The movable control rod worth for shutdown la not changed. The utilization of burnable poison as a part of the control talance allows for a reductica. of the soluble icison concentration to obtain moderator coefficients within a desired range. The Doppler coefficient for all cases shown is negative over the core life.

Item 4. Principal resign Parameters of the Reactor Coolant System Most of the features in this section are directly related to material Proler-ties and the amount of heat produced in the reactor core. Note that the B&W units are identical. The parameters are scaled in proportion to the power of the reactor. The major difference is the number of coolant loops required to remove the heat produced.

?or the B&W units, only two loops are required since once-through steam gen-erators are used instead of the U-tubes-in-shell design. The greater cooling capacity of thece steam Cencrators permits a reduction in the number of cool-l ing loops for an equivalent amount of heat removed.

Ite: 5 Reactor Coolant System - Ccdc Requirements The B&W units are identical. Ccde requirements for the shell side of the steam generator conform to the AS:C III Class A Spacification. This is con-sidered ,to be a contribution to the safety of the vessel. It enhances the integrity because of the core stringent ASME III Class A design, material, and quality-control requirements.

Iteu 6. Principal Desirn Parameters of the Reactor Vessel The B&W units are identical. These vessel designs are characterized by a thinner thermal shield and a relatively larger shell diameter. The larger diameter provides for additional water between the edge of the core and the vessel which leads to additional neutren attenuation.

Item 7 Principal Design Features of the Steam Generators The steam generators in the B&W units are the same. They are basically dif-ferent from the Turkey Point units since they are a once-through design incor-jorating an integral superheat section.

Item 8. Princirni Design Parameters of the Reactor Coolant Pumps The B&W designs are the same. In each crecific tabular parnmeter the rela-tive number or size is in proportion to the totil amount of beat removed fron 1

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the core. The B&W reactor pumps have higher head and horsepover requirements than the Turkey Point units have for approximately the same flow becauce of differences in system pressure drops.

Item 9 Principal Design Parameters of the Reactor Coolant Piping The B&W designs are .he same. They utilize enrbon steel clad with stainless steel.

Item 10. Fenetor Building Parameters All reactor buildings are basically of the same design and construction. The differences are physical dimensicns, amount of concrete chielding needed and design incident pressures, which are a direct result of plant layout, engi-neered safeguards, system capacities, and site location. The reactor building

, design and shielding effer satisfactory protection to the surrounding popula-tien in case of accident and during norir.al operation of the generating units.

Item 11. Engineered Safeguards Er.gineered safeguards are 6enerally similar.

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