ML20010E082
| ML20010E082 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/24/1981 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8109030047 | |
| Download: ML20010E082 (6) | |
Text
_ _
[
DISTRIBUTION:
Cocket File bcc:
Ld#1 Rdg.
TERA DEisenhut PDR AUG 2 412 JYoungblood LPDR JSnell NSIC KKiper TIC Docket Nos.: STN 50-454-
~
~ "~~MRushbrook (2) ACRS (16) and STN 50-455
........SHanauer m
RVollmer Mr. Louis 0. DelGeorge murley Director of Nuclear Licensing i _ [._ ',,
' l ',', aa f Id' MPA Commonwealth Edison Company P. O. Box 767
- '" DELD' Chicago, Illinois 60690
~ 1.
J3)
Dear Mr. DelGeorge:
Subject:
Request for Additional Inf.ormation for, the. Review of the Byron
~
Plant, Units 1 & 2 - Regarding Chemical Engineering As a result of our continuing review of. the. Byron _ Plant, Units.1 and 2 FSAR, we find that we need additional information.to,. complete our. evaluation. The specific information required is in.the area of. chemi. cal engineer Ing_ and is presented in the Enclosure.
To maintain our licensing review s.che.dule fo.r,the. Byron Plant FSAR, wc wil.1 need responses to the enclosed request.by Sept. ember 30, 1981.
If you cannot meet this date, please infcrm us within _seyen days after receipt of_ this..,
letter of the date you plan to.. submjt your. resppqses so that we,may revjew,_
our schedule for any necessary. changes.
Please contact J. C. Snell, Byron. Lice.nsing. Project Manager, if you desire any discussion or clarification.of. the,englose,d report.
Sincerely, Ori ginel signed by*
B. J. Youngblood B. J. Youngblood, Chief Licensing Branch No.1 Division of Licensi p
Enclosure:
R
\\
As stated
.,, T '. ' '
//
h]
q 6 Dj, g c iggI A -
cc: See next page
- u. %,,,P
\\
'% roar
,cl-b/
8109030047 810824 D
PDR ADOCK 03000454 g/
/HI i
'b DT-omer)
), 1/1 4
ne BJ
,g lood suaNQW)
'/5)[5E'"9,' 5/1 781"'
8 Oan p NRC FORM 318 410 BCI N RC M O2 4 0 OFFICIAL RECORD COPY
' uSC" ' *3d *2'
a
'[r iot.is c. W Geurae Dire tor of k.
>a r L icer.s i ng Comonweal th Edison Ccmpany Post Of ice Box /67 Chicago, Illinois 60690 ccs:
Mr. William Kortier Mr. Edward R. Crass Atomic Power Distribution Nuclear Safeguards and LiceFv(ng Division Westinghouse Electric Corporation Sargent & Lundy Eng'neers' P. O. Box 355 55 East Monroe Street Pittsburgh, Pennsylvania 15230 Chicago, Illinois 60603 Paul M. Murphy, Esq.
Nuclear Regulatory Commission, Region III Isham, Lincoln & Beale Office of Inspection and Enforcement One First National Plaza 799 Roosevelt Road 42nd Floor Glen Ellyn, Illinois 60137 Chicago, Illinois 60603 Myron Cherry, Esq.
Mrs. Phillip B. Johnson Cherry, Flyr.n and Kanter 1907 Stratford Lane 1 IBM Plaza, Suite 4501 Rockford, Illinois 61107 Chicago, Illinois 60611 Professor Axel Meyer Deoartment of Physics Northern Illinois University DeKalb, Illinois 60115 C. Allen Bock, Esq.
P. O. Bo >. 34 2 Urbanan, Illinois 61801 Thomas J. Gordon, Esq.
Waaler, Evans & Gordon 2503 S. Neil Champaign, Illinois 61820 Ms. Bridget Little Rorem Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935 Kenneth F. Levin, Esq.
Beatty, Levin, Holland, Casofin & Sarsany 11 South LaSalle Street Suite 2200 Chicago, Illinois 60603
~
f ENCLOSURE CHEMICAL ENGINEERI!4G BYRON NUCLEAR STATION UNITS N0. 1 AND 2 DOCKET N0. 50-454/455 281.1 Describe the ' containment sump design features to insure that adequate solution mixing of ESF fluids will occur and proper sump chemistry will be maintained. Allowance should be made for " dead" volumes, sumps, etc., in the determination of sump ph and the quantititas of NaOH used to verify that no areas accumulate spray solutions with pH less than 7.0.
281.2 (a) Verify that the coatings on the steel containment, concrete (6.1.2) wall, and concrete floors described in pages 6.1-4 and 6.1-5 of your FSAR meet the requirements of ANSI 101.2 and Regulatory Guide 1.54 or meet the requirements of ANSI 101.2 and the alternative QA acceptance criteria for protective coatings described on page 6.1-7 of your FSAR.
(b) State whether the equipment coating described on page 6.1-6 and listed in Table 6.1-2 ? your FSAR meets either one of the above criteria specified for the containment and floor coating.
In addition, pro /ide the thickness for the coatings listed in Table 6.1-2 of your FSAR.
(c)
In order for the staff to estimate the rate of conbustible gas generation vs time becasue of exposure of organic cable insula-tion to DBA condition inside containment, provide the following information:
(1) the quantity (weight and volume) of uncovered cable and cable in closed metal conduit or closed cable trays.
We will give credit for beta radiation sheilding for that portion of cable that is indicated to be in closed condu n or trays. (2) A breakdown of cable diameters and associated con-ductor cross sections, or an equivalent cable diameter and cor.ductor cross section that is representative of total cable surface area associated with the quantity of cable identified in 1) above.
(3) The major organic polyner or plastic material in the cables.
If this information is not provided, we will assume the cable insulation to be polyethylene and assume a G value for combustible gas of 3.
1 1
2-5.
Verify that valves which are not accessible for repair after an accident are environmentally qualified for the conditions in which they must operate.
6.
Provide a procedure for relating radionuclide gaseous and ionic species to estimated core damage.
7.
State the design or operational provisions to prevent high pressure carrier gas from entering the reactor coolant system from ca line gas analysis equipment, if it is used.
8.
Provide a method for verifying that reactor coolant d'ssolved oxygen is at <.0.1 ppm if reactor coclant chlorides are determiend to be 0.15 ppm.
9.
Provide information on (a) testing frequency and type of testing to ensure long term op,erability of the post accident sampling system and (b) operator
~
training requirements for post-accident sampling.
- 10. Demonstrak that the reactor coolant system and suppression chamber sample locations are representative of core conditions.
In addition to the a.bove licensing conditions the staff is conducting a generic review of accuracy and sensitivity for analytical procedures and on-line instrumentation to be used for post-accident analysis. We will require that the applicant submit data supporting the applicability of each selected analytical chemistry procedure or on-line instrument along with documentation demonstrating compliance with the licensing conditions four months prior. to exceeding 5% power operation, but review and approval of these
~
pcccedures will not be a condition for full power operation.
In the event our yeneric review determines a specific procedure is unacceptable, we will require -
- Ute, applicant to make modifications as determined by our generic review.
l e
9 l
l l
e e
- ~--,.
. 281.3 Describe the samples and instrument readings and the frequency (9.1.3)'
of measurement that will be performed to monitor the water purity and need for spent fuel pit cleanup system demineralizer resin and filter replacement. State the chemical and radio-chemical limits to be used in monitoring the spent fuel pool water and initiating corrective actions.
Provide the basis for establishing these limits. Your response should consider vari-ables such as: boron, gross gamma and iodine activity, de-mineralizer and/or filter differential pressure, demineralizer contamination factcr,pH, and crud level.
281.4 Your FSAR did not indicate that the refueling water storage tank, (9.3.2) the' boric acid mixing tank, the chemical additive tank, and the
~
~*
sump tank will be sampled. Confirm that these tanks will be sampled according to Standard Review Plan 9.? 2.
281.5 Verify that sample line purge flows and duration times are suff-(9.3.2) cient to flush out stagnant lines to assure that a representative sample is obtained.(The NSSS vendor recommends a flush of 6-10 sample line volumes, at a purge flush rate of about 2-3 times the sample flow rates).
281.6 Describe how discharge flows are limited under normal and anticipated (9. 2) fault conditions (malfunction or failure) to preclude any fission product release beyond the plant release limitation given in the Technical Specification.
Installatica of an orifice or a fully qualified solenoid valve which will provide a remote sampling system isolation capability is acceptable.
281.7 (TMI II.B.3) Provide information that satisfies the attached proposed license conditions for post-accident sampling.
281.8 Describe provisions for monitoring filter /demineralizer differential
( 9.,3.4) pressure to assure that pressure differential limits are not exceeded (Section II.8.b. of Standard Review Plan 9.3.4)
I b
I;UREG-0737, II.B.3 - Post Accident Samplina Capability RE0VIREl4ENT_
Provide a capability to obtain and quantitatively analyze react.cr coolant and containment atmosphere samples, without radiation exposure to any individual exceeding 5 rem to the whole body or 75 rem to the extremities (GDC-19) during and following an accident in which there is core degradation.
!4aterials to be analyzed and quantified include certain radionuclides that are indicators of severity of core damage (e.g., noble gases, iodines, cesiums and non volatile isotopes), hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron and chloride in reactor coolant
_. s samples in accordance with the requirements of f;UREG-0737.
To satisfy the requirements, the application should (1) review and modify his' sampling, chemical analysis and radionuclide determination capabilities as necessary to comply with fiUREG-0737, II.B.3, (2) provide the staff with information pertaining to system design, analytical capabilities and pro-cedures in sufficient detail to demonstrate that the requirements have been met.
EVALUATION AtlD fit 1DIf1GS The applicant has committed to a post-accident sampling system that meets the requirements of fiUREG-0737, Item II.B.3 in Amendment
, but has not provided the technical information required by tiUREG-0737 for our evaluation.
Implementation of the requirement is not nece.isary prior to low power operation
~
because. only small quantities of radionuclide inventory will exist in the reactor coolant system and therefore will not affect the health and safety of the public.
Prior to exceeding 5% power operation the applicant must demonstrate the capability to promptly obtain reactor coolant samples in the event of an accident in which there is core damage consistent with the conditions j
stated below.
Demonstrate compliance with all requirements of i;UREG-0737, II.B.3, for 1.
sampling, chemical and radionuclide analysis capability, under accident
(
conditions.
Provide sufficient shielding to meet the requirements of GDC-19, assuming 2.
Reg. Guide 1.4 source terms.
Comit to meet the sampling and analysis requirements of Reg. Guide 1.97, 3.
Rev. 2.
Verify that all electrically powered components associated with post 7
4.
accident sampling are capable of being supplied with power and operated, within thirty minutes of an accident in which there is core degradation, assuming loss of off site power.