ML20235D774

From kanterella
Jump to navigation Jump to search
Summarizes Understanding of Matters Requiring AEC Review Prior to Const of Related Sys,Per 701002 Meeting.Review of Structural Design Documents in Progress.Inservice Insp Info Submitted by Found Acceptable
ML20235D774
Person / Time
Site: Brunswick, 05000000
Issue date: 02/08/1971
From: Morris P
US ATOMIC ENERGY COMMISSION (AEC)
To: Jackie Jones
CAROLINA POWER & LIGHT CO.
Shared Package
ML20235B311 List: ... further results
References
FOIA-87-111 NUDOCS 8709250342
Download: ML20235D774 (7)
v • d • e


Text

- _ - _ - - , - - -

f 4 UNITED STATES ATOMIC ENERGY COMMISSION

. M< . ;:>,- WASHINGTON, D.C. 20545 77 im

  • February 8, 1971 Docket Nos. 50-324 and 3U!

Mr. J. A. Jones Senior Vice President Operating & Engineering Group Carolina Power & Light Company 336 Fayetteville Street Raleigh, North Carolina 27602

Dear Mr. Jones:

At the public hearing for the construction permit for the Brunswick plant we identified a number of matters that we would review prior to construction of the related systems. We discussed these matters with your representa-tives at a meeting held on October 2,1970. The following paragraphs summarize our understanding of the status of those items that are not part of an R&D effort, but are related specifically to the design of the Brunswick Steam Electric Plant (as described in Section 8.3 of the Regulatory Staff Safety Evaluation):

1. Structural Design This information has been filed and our review is in progress. The first reports (numbered 4, 5, and 6) were filed on November 9, 1970 and our questions on these submittals were transmitted to you on December 28, 1970.

Design Report No. 7," Containment Design Report," was submitted on January 8, 1971 and our review is in progress.

2. Inservice Inspection This information was transmitted by letter dated March 9, 1970. We com-pleted our review of your proposed inservice inspection and found it  ;

acceptable as discussed in our letter of April 27, 1970. l l

3. Interlock Logic for the Automatic Depressurization System and the i Rod Block Monitor System This information, which was to have been filed by March, 1970, has been l submitted in a topical report on August 15, 1970 (NEDO-10139). We )'

understand this information is relevant to the design of the Brunswick Steam Electric Plant and that the design details for the Brunswick Steam Electric Plant will be filed in the near future.

g!

8709250342 8705421 PDR FDIA MENZ87-lgl pggg ]

Mr4 J . A. Jones February 8, 1971

4. Spent Fuel Pool Cooling Svstem As indicated in our Safety Evaluation we reviewed your design provisions for replenishment of water in the spent fuel pool. The normal water make-up rate of 300 gpm, and your provisions for interconnection with the residual heat removal system are being evaluated in conjunction with our review of the potential for and consequences of accidental loss of pool water. At this time we do not have any recommendations for improvement of your system.
5. Design Basis for Engineered Safety Features Amendment No. 8, filed on June 26, 1969, contained a commitment to use TID-14844 source term assumptions in determining the design bases for engineered safety features. At the time of the hearing you had not completed the analyses for all the engineered safety features. As stated in our Safety Evaluation, we will review this information prior to con-struction of the subsystems. We expect to receive your submittal on the design basis of the engineered safety features on a timely basis to permit completion of our review prior to your procurement of equipment.

In addition to the above items that were deferred for resolution until after the issuance of a Construction Permit, there are two other items for which additional design features or design modifications are deemed necessary. These are the provisions for control of hydrogen concentration in the containment following a loss-of-coolant accident and isolation ,

capability for the one-inch diameter instrument lines. Our comments on l these matters are given below:

l. Control of Hydrogen Concentration within Containment In Amendment No. 8 you described your intentions regarding your study of methods of controlling the concentration of hydrogen that might he a-w.rab ..

by metal-water reactions and radiolysis following .a loss-of-coolant accid ent . You stated that your program would be complete and the results submitted to the AEC by mid-1970. We have reviewed the results submitted by others and have concluded that:

a. The Brunswick Steam Electric Plant should have the capability for measuring the hydrogen concentration and for mixing the atmosphere in the containment following a loss-of-coolant accident, and for controlling combustible gas concentrations I without reliance on purging of the containment atmosphere, The continuous presence of combustible gas control equipment f

1

4 February 8, 1971 {

Mr. J. A. Jones i'

~ .

at the site may notlbe necessary provided it is available on an appropriate time scale; however, appropriate design and procedural provisions should be made for its use.

b. Combustible gas control systems and the provisions for mixing, measuring, and sampling should meet the design,. quality assurance, redundancy, energy source, and instrumentation requirements for an engineered safety feature and the system itself should not introduce safety problems that may affect containment integrity,
c. You should also have the installed capability for a controlled purge of the containment atmosphere through appropriate fission product removal systems,
d. The parameter values listed below should be used for the. pur-pose. of calculating hydrogen and ' oxygen gas concentrations in containments and evaluating designs provided to control and to purge combustible gases evolved in the course of loss-of-coolant accidents. -If you plan on using parameters other than these values given below they should be listed and adequately justified.

(1) Fraction of fission product radiation energy absorbed by the coolant (a) Beta (1) Betas from fission-products in the fuel rods: 0 (2) Betas from fission products intimately mixed with coolant: 1.0 (b) Gamma (1) Gammas from fission products in the fuel rods, coolant in core region: 0.1 (2) Gammas from fission products intimately mixed with coolant, all coolant: 1.0 i

f .

i

-_=_--__-_______--___ l

Mr. J. A. Jones February 8, 1971 (2) G(H2) 0.5 molecules /100 ev (3) G(0 2) 0.25 molecules /100 ev (4) Extent of metal-water reaction (percentage of fuel cladding that reacts with water) 5 (5) Fission product distribution model (a) 50% of the halogens and 1% of the solids present in the core are intimately mixed with the coolant water.

(b) All noble gases are released to the con-tainment.

(c) All other fission products remain in fuel rods.

(6) (a) Hydrogen concentration 4 volume percent limit (This limit should not be exceeded if more than 5 volume percent oxygen is present).

(b) Oxygen concentration limit 5 volume percent (This limit should not be exceeded if more than 4 volume percent hydrogen is present).

2. Instrument Lines Penetrating the Containment In our review of plants subsequent to the Brunswick Steam Electric Plant we have concluded that the isolation features provided for the one-inch diameter lines should be improved. We believe that your design for one-inch sensing lines should be modified to meet the following criteria:

l i

Mr. J . A. Jones -5 Februa ry 8,1971 l

a. Should satisfy the requirements for redundancy, independence, and testability if used as part of the protection system.
b. Should be sized or orificed to assure that in the event of a postulated failure of the piping or of any co:ponent (including the postulated rupture of any valve body) in the line outside  !

primary reactor containment during normal reactor operation, (1) the leakage is reduced to the maximum extent practical con-sistent with other safety requirements, (2) the rate and extent of coolant loss is limited to within the capability of the reactor l coolant makeup system, (3) the integrity and functional performance of secondary containment, if provided, and associated safety systems (e.g. , filters, standby gas treatment system) will be maintained,  :

and (4) the potential offsite exposure will be substantially below the guidelines of 10 CFR 100.

c. Should be provided with an isolation valve capable of automatic operation1 or remote operation from the control room or from another appropriate locations and located in the line outside the containment as close to the containment as practical. There should be a high degree of assurance that this valve (1) will not close during normal reactor operation, (2) will close or be closed if the instrument line integrity outside containment is lost during normal reactor operation or under. accident con-ditions, and (3) will reopen or can be reopened under the conditions that would prevail when valve reopening is appropriate. Power-operated valves should remain as is upon loss of power. The status ;open or closed) of all such isolation valves should be indicated. If a remotely operable valve is provided, sufficient information should be available in the control room or another appropriate location to assure timely and proper actions by the operator,
d. Should be conservatively designed up to and including the isolation valve and of a quality at least equivalent to the containment.

These portions of the lines should be located and protected so as to minimize the likelihood of their being da: aged accidentally.

They should be separated or otherwise protected from any other line. Provisions should be included to permit periodic visual inservice inspection, particularly of those portions of the lines outside containment up to and including the isolation valve.

A self-actuated excess flow check valve is acceptable as an automatically operated valve provided it has all other features specified.

l L______

s

- Mr. J. : A. Jones - -6. February 8,1971

e. Should not be so restricted.by components in the' lines, such as valves and orifices, that the response time .of. the connected

, instrumentation will be increased to an unacceptable degree.

If.pou have any questions on the above, please contact me.

Sincerely, 0

Peter A. Morris, Director Division of Reactor Licensing I

cc: G. F. Trowbridge, Esquire- l Shaw, Pittman, Potts, Trowbridge & Madden-

' 910 17 th Street, N.W.

Washington,: D. C.

20006 I

i

r , s 6 S' NO 8M Mr. J. A'. Jones'

e. Should not be' so restricted by components in the lines, such as valves and orifices, that the response time of the connected insertsnentation will be increased to an unacceptable degree.

'If you have any' questions on the above, plasse contact me.

Sincerely, crl$ nt e b S' "

Pet 6r Aa Worfil Peter A. Morris, Director Division of Reactor Licensing ec: G. F. Trowbridge. Esquire Shaw, Pittman, Potts. Trowbridge & Madden 910 17th Street, N.W.

Washington, D. C. 20006 Distribution:/

Docket.(2) /

AEC PDR (2)-

DR Reading DRL Reading BWR-1 File CKBeck, DR' MMMann, DR SHanauer, DR I FSchroeder, DRL .

RSBoyd DRL RDeYoung, DRL DSkovholt, DRL TRWilson, DRL .i EGCase, DRS RRMaccary,.DRS PWHowe, DRL )

BGrimes, DRL {

DRL/DRS Branch Chiefs )

WRButler, DRL DKartalia, OGC Compliance (2)

ACRS (16) i WNyer, (2)

Newmark & Hall (2)

SMKari, DRL i DRL DRL I, .DRL/BWR- 1 DRL/ :BWR orriCE > ..........#../..... . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . .

RSB yd eder

$URNAME >

DFK

......FSh..

PAM is l

DATE > .  !. . .. .

l. Form AEC 318 (Rev. 943) u.s. GOVERNMENT PmMTING OmG .1gst 364-590 1

_ _ _ . _ _ _ _ _ _ _ _ _ _ _  !

Retrieved from "https://kanterella.com/w/index.php?title=ML20235D774&oldid=3226663"