ML20213D764
| ML20213D764 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/27/1981 |
| From: | Benaroya V Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0374, CON-WNP-374, TASK-2.B.3, TASK-TM NUDOCS 8108060183 | |
| Download: ML20213D764 (6) | |
Text
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JUL 2 71981 DocketNo.5d-397 s(!
"o MEMORANDUM FCR: Robert L. Tedesco, Assistant Director W
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Victor Benaroya, Chief "n o 91981 = fr-u.s.gj,7 " p.8 Chemical Engineering Branch
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SUBJECT:
REQUEST FOR ADDITIONAL IflFORMATI0fl Oil Wi:P-2 j
Plant Name: WNP-2 Suppliers: General Electric, Washington Public Power Supply System 1
Licensing Stage: OL Dccket Number: 50-397 Responsible Branch and Project Manager: LB #2, R. Auluck Reviewer:
F. Ilitt Description of Task: Operating License Review Status: Q Review Complete The Chemical Technology Section of the Chemical Engineering Branch has reviewed SRP Sections 5.4.8, 6.1.1, 6.1.2, 9.1.3, 9.2.3, 9.3.2.and 10.4.6.
tie.need the enclosed additional infornation to complete our review. We have also attached the prcposed license conditions to meet the requirements of TMI Action Plan Item II.B.3, Post-Accident Sampling Capability. He will recorrend the license conditions be imposed should the applicant's respcnse be inadequate. To maintain the current review schedule, the applicant should respond by September 18, 1931.
griginal Signed by
[,,Bonar0ya Victor Benaroya, Chief Chemical Engineering Branch Division of Engineering
Enclosure:
As stated
Contact:
F. Uitt X29613 cc:
R. Vollmer A. Schwencer D. Eisenhut R. Auluck
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ADDITI0fiAL IftFORMATIOff REQUIRED BY CHEMICAL EfiGIflEERIf4G BRAf1CH FROM WilP-2 WASHIfiGT0fi PUBLIC POWER SUPPLY SYSTEM Chemical Technology Section/ Chemical Engineering Branch (CMEB) 281.1 In accordance with Regulatory Position C.1 of Regulatory Guide 1.56 (10.4.6) revision 1, describe the sampling frequency, chemical analyses, and established limits for purified condensate dissolved and suspended solids that will be performed and the basis for these limits.
281.2 Establish and state the sequential resin replacement frequency in (10.4.6) order to maintain acequate capacity margin in the condensate treat-ment system (Regulatory Position C.2 of Regulatory Guide 1.56, revision 1).
Include the basis for the resin replacement frequency.
281.3 Verify that the initial total capacity of new demineralizer resins (5.4.8)
(condensate and primary coolant) will be measured and describe the (10.4.6) method to be used for this measurement (Regulatory Position C.3 of Regulatory Guide 1.56, revision 1).
281.4 Describe the method of determining the condition of the demineralizer (5.4.8 units so that' the ion exchange resin can be replaced before an (10.4.6) unacceptable level of depletion is reached (Regulatory Position C.4 of Regulatory Guide 1.56, revision 1). Describe the method by which (a) the conductivity meter readings for the condensate cleanup system will be calibrated, (b) the flow rates'through each demineral-izer will be measured, (c) the quantity of the principal ions likely to cause demineralizer breakthrough will be calculated, and (d) the accuracy of the calculation of resin capacity will be checked.
281.5 Indicate the control room alarm set points of the conductivity meters (5.4.8) at the inlet and outlet demineralizers in the condensate and reactor (10.4.6) water cleanup systems when either (Regulatory Position C.5 of Regula-tory Guide 1.56, revision 1):
a.
The conductivity indicates marginal performance of the demineralizer system; b.
The conductivity indicates noticeable breakthrough of one or more demineralizers.
281.6 The reactor coolant limits and corrective action to be taken if the (5.4.8) conductivity, pH, or chloride content is exceeded will be established (10.4.6) in the Technical Specifications. Describe the chemical analysis methods to be used for their determination (Regulatory Position C.6 of Regulatory Guide 1.56, revision 1).
281.7 Describe the waterchemistry control program to assure maintenance of (10.4.6) condensate demineralizer influent and effluent conductivity within the limits of Table 2 of Regulatory Guide 1.56, revision 1.
Include con-ductivity meter alarm set points and.the corrective action to be taken if the limits of Table 2 are exceeded.
]
. 281.8 Regarding the Spent Fuel Pool Cleanup System, provide the follow-(9.1.3) ing information:
Describe the samples and instrumentation and their frequency of measurement that will be performed to monitor the Spent Fuel Pool water purity and need for ion exchanger resin and filter replacement.
State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and for initiating corrective action.
Provide the basis for establishing these limits. Your response should consider variables such as: gross gamma and iodine activity, demineral-izer and/or filter differential pressure, demineralizer decontamination factor, pH and crud level.
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281.9 a.
Indicate the total amount of paint or protective coatings (area and (6.1.2) film thickness) used inside containment that do not meet the require-ments of ANSI N101.2 (1972) and Regulatory Guide 1.54. We will use the above information to estimate the rate of combustible gas generation vs. time and the amount (volume) of solid debris that can be formed from these unqualified organic materials under DBA conditions that can potentially reach the containment sump. A G value of 5 will be used unless a lower G value is justified
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technically.
b.
In order for the staff to estimate the rat'e of combustible gas generation vs. time due to exposure of organic cable insulation to DBA conditions inside containment, provide the following infor-mation:
- 1) The approximate total quantity (weight and volume) of organic cable insulation material used inside containment, including uncovered cable and cable in closed metal conduit or closed cable trays. We will give credit for beta radiation shielding for cable in closed conduit or trays if informati.on is provided as to the respective quantities of cable in closed conduits or trays vs. uncovered cable.
- 2) The approximate breakdown of cable diameters and conductor or cross section associated or an equivalent cable diameter and conductor cross section which is representative of the total cable surface. area consistent with the quantity of cable sur-face area consistent with the quantity of cable identified l
j in 1) above.
- 3) The major organic polymer or plastic material associated with the cable in 1) above.
If this information is not provided, we will assume the cable insulation to be polyethylene and a G value of 3.
i 281.10 Verify that sample line purge flows and duration times are sufficient (9.3.2) to flush out stagnant lines to assure that a representative sanple is
- obtained, l
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. 281.11 Acceptance Criterion 2 9 in Standard Review Plan Section 9.3.2 (9.3.2) states that passive flow restrictions to limit reactor coolant loss
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from a rupture of the sample line should be provided. You do not address this criterion in the FSAR. Describe how the requirement of maintaining radiation exposures to aslow as is reasonably achiev-able will be met in the event of a rupture of the sample line containing contaminated primary coolant, in accordance with Regulatory Position C.2.i(6) of Regulatory Guide 8.8, revision 3 (June 1978).
281.12 Acceptance criterion 1.a in Standard Review Plan Section 9.3.2 (9.3.2) indicates that sumps inside containment and the standby liquid control storage tank should be sampled. Describe provisions to sample sump water inside the containment in accordance with the requirements of General Design Criterion 64 in Appendix A to 10 CFR Part 50.
281.13 Verify that provisions have been made for draining and venting (5.4.8) reactor water cleanup system components through a closed system in accordance with GDC 60 and 61.
281.14 Provide information that satisfies the attached proposed license (TMI II.B.3) conditions for post-accident sampling.
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m, SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WNP-2 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET N0. 50-397 NUREG-0737, I1.8.3 - Post Accident Samoling Capability REQUIREMENT Provide a capability to obtain and quantitatively analyze reactor 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.
Materials 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, baron and chloride in reactor coolant samples in accordance with the requirements of NUREG-0737.
To satisfy the requirements, the applicaticn should (1) review and modify his sampling, chemical analysis and radionuclide determination capabilities as necessary to comply with NUREG-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 AND FINDINGS The applicant has committed to a post-accident sampling system that meets the requirements of NUREG-0737 Item II.B.3 in Amendment 49, but has not provided the technical information required by NUREG-0737 for our evaluation.
Implementation of the requirement is not necessary 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 i
stated below.
-1.
Demonstrate compliance with' all requirements of NUREG-0737, II.B.3, for i
l sampling, chemical and radionuclide analysis capability, under accident conditions.
2.
Provide sufficient shielding to meet the requirements of GDC-19, assuming I
Reg. Guide 1.3 source terms.
l 3.
Comit to meet the sampling and analysis requirements of Reg. Guide 1.97, Rev. 2.
3 I
3 4.
Verify that all electrically powered components associated with post
."f 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.
5 h.
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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 on line gas analysis equipment, if it is used.
8.
Provide a method for verifying that reactor coolant dissolved oxygen is at < 0.1 ppm if reactor coolant chlorides are determiend to be 0.15 ppm.
9.
Provide information on (a) testing frequency and type of testing to ensure long term operability of the post accident sampling system and (b) operator training requirements for post-accident sampling.
- 10. Demonstrate that the reactor coolant system and suppression chamber sample locations are representative of core conditions.
In addition to the above 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 cperation, but review and approval of these procedures will not be a condition for full power operation.
In the event our generic review determines a specific procedure is unacceptable, we will require the applicant to make modifications as determined by our generic review.-
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A. Schwencer, Chief
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.FROM:
R. Auluck, Project Manager Licensing Branch No. 2, DL
SUBJECT:
FORTHCOMING MANAGEMENT MEETING WITH WASHINGTON PUBLICPOWERSUPPLYSYSTE!!(WPPSS)
DATE & TIME:
Wednesday, August 26, 1981 8:30 AM LOCATION:
Room P-114 Phillips Building Bethesda, MD PURPOSE:
Presentation by WPPSS management of its organizational structure and technical resources for operation of WNP-2 PARTICIP/.NTS:
NRC D. B. Vassallo, D. l1. Beckham and G. W. Rivenbark WPPSS l
G. D. Bouchey, G. C. Sorensen and R. Nelson Original simed by R. Auluck, Project flanager Licensing Branch No. 2 Division of Licensing cc: See next page
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Nac ronu sia (io-soi nacu c.2e OFFICIAL RECORD COPY usom ini-m.ua l
JUL 31 1981 I
Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 ccs: Nicholas Reynolds, Esq.
Debevoise & Liberman 1200 Seventeenth Street, N.W.
Washington, D. C.
20036 Richard Q. Quigley, Esq.
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympis, Washington 98504 Mr. Albert D. Toth Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 Mr. G. C. Sprensen, Licensing Manager i
Washington Public Power Supply System l
P. O. Box 968 Richland, Washington 99352 Mr. O. K. Earle. Project Licensing Supervisor Burns and. Roe Incorporated 601 Williams Boulevard Richland, Washington 99352 r
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ENCLOSURE PROPOSED AGENDA FOR DISCUSSION BY WPPS50F ITS ORGANIZATIONAL STRUCTURE AND TECHNICAL RESOURCE' FOR OPERATION OF WNP-2 1.
Describe the current overall corporate and WNP-2 plant. organizations.
2.
For the overall corporate o'rganization, describe a) how the nuclear portion fits into the rest of the utility's operations, b) the relation of the senior corporate officer with nuclear responsibility to other corporate officials, c) the responsibility and authority of senior corporate officer in charge of nuclear operations, d) the corporate level support to be provided to support plant operation.
3.
For the corporate level support, describe: a)'the responsibilities of each offsite organizational element that is related to the management or support of operation of the units, b) where located, c) the lines of authority and comunication between these offsite organizational elements, d) if not under direct control of senior nuclear officer, how support is obtained, e) number and qualifications of technical support personnel, f) how organized, g) the relations and the lines of authority and comunication between these offsite organizational elements and the organizational elements at the site.
4 Describe how the proposed organization for operation compares with the organization that is and has been in effect during the plant construction, with emphasis on changes that have been or will be made to strengthen the i
organization.
5.
Describe the qualification and staffing levels of the offsite managers and technical staff (provide these in writing prior to meeting if possible).
Engineering Group that is located onsite but Discus's the Independent Safety (May 6.
reports to offsite management 13, 1980 Info. Report - SECY 80-242).
7.
Describe how the corporate official in overall charge of nuclear plant l
operations is acti'vely involved in plant operational activities - i.e.,
what oversight and management responsibility and activities he personally handles.
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8.
For the plant staff, describe: a) the operating organization and how it interacts to assure safe plant operation, b) the number and qualifications of technical support personnel and maintenance personnel, c) the provisions will be made for security and the fire brigade, d) the proposed shift staffing, e) numbers and qualifications per shift, f) how many shifts.
9.
Describe the provisions to be made for Shift Technical Advisors including qualifications, training, reporting channels, relations to shift staff.
l 10.
Describe the training provisions for both licensed and unlicensed personnel, l
how many now in training, future intentions.
s
- 11. Describe the proposed QA organization for operation, relation to plant staff, where does it report, access to corporate level personnel.
12.
Describt. arrangement's for safety review of plant operations, including onsite review and corporate review and audit.
- 13. Describe a) how operational experience from WNP-2, other utilities, the NRC, INPO, etc., are obtained, reviewed and disseminated to plant operators and to other appropriate personnel at plant and corporate levels, and b) the pro-cedures and checks used to assure that the information actually reaches the appropriate personnel and appropriate action is taken based on that informa-tion.
14 Describe how outside contractual assistance is relied on as technical or other support to operation of the nuclear plant:
- 15. Describe the national standards and NRC regulatory guides that you use as criteria for your current offsite management and technical support staff.
Throughout our review, the staff will be comparing information obtained from WPPSS with the draft guidelines set forth in NUREG-0731.
In order that there are no misunderstandings, it would be helpful if WPPSS could provide comments on how their proposed operating organization meets the intent of these. guidelines.
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'f4EETING NOTICE DISTRIBUTION.'
.. \\TED July 31, 1981 3-hketFile(s)
I&E Region I NRC POR I&E Region II Lo. cal PDR I&E Region III Branch Reading File I&E Region IV HSIC I&E Region V TERA h
NRC
Participants:
E. G. Case D. G. Eisenhut/R. Purple N. Hughes T. Novak S. Varga T. Ippolito R. A. Clark J. F. Stolz (ORBf4)
R. Tedesco B. J. Youngblood A. Schwencer F. Miraglia E. Adensam (LBf4)
J. R. Miller bec: Applicant G. Lainas Service List D. M. Crutchfield B. T. Russell Branch Licensing Branch No. 2 J. Olshinski M.D. Houston R. H. Vollmer Project Manger R. J. Mattson S. H. Hanauer
/
Licensing Assistant M. Service T. Murley.
J. P. Knight W. Johnston (AD/ Materials & Qualif. Engr)
D. R. ?!uller P. S. Check W. E. Kreger L. S. Rubenstein F. Schroeder I
M. L. Ernst ACRS (16)
OI&E (3) 1 050 (7)
Attorney, OELD J. LeDoux, I&E V. Moore B. Grimes I
f