ML17286A565
| ML17286A565 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/23/1991 |
| From: | Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GO2-91-012, GO2-91-12, NUDOCS 9101300137 | |
| Download: ML17286A565 (12) | |
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WASHIiVGT>>ON PUBLIC POWER SUPPLY SYSTEM P.O. Box 96B
~ 3000 George lVashlngton Way
~ Richland, WashlnglonM352 January 23, 1991 G02-91-012 Docket No. 50-39?
U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Mail Station Pl-137 Mashington, D. C.
20555 Gentlemen:
Subject:
NUCLEAR PLANT NO. 2, OPERATING LICENSE NO. NPF-21 NRC INSPECTION REPORT 90-29 RESPONSE To NOTICE OF VIOLATION The Washington Public Power Supply System hereby replies to the Notice of Violation contained in your letter dard December 24, 1990.
Our reply, pursuant to the provisions of Section 2.201, Title 10, Code of Federal Regulations, consists of this letter and Appendices A and B (attached).
In Appendix A, the violation is addressed with an explanation of our position regarding validity, corrective action and date of full compliance.
Appendix B
addresses the issue pertaining to the capability of the Reactor Building Post-LOCA Grab Sampler to'uantify post accident radioiodine concentrations in effluents.
As requested in the cover letter to the subject inspection report, a schedule for correcting system performance is also contained in Appendix B.
Very truly yours, G. D. Bouchey, Dir r
Licensing
- 8. Assurance JDA/bk Attachments cc:
JB Martin - NRC RV NS Reynolds - Winston/Strawn PL Eng -
NRR DL Williams - BPA/399 NRC Site Inspector - 901A ADOCK P~PPP3p7 PDF Plp Is
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'ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
E ESSION NBR: 9101300137 DOC. DATE: 91/01/23 NOTARIZED:
NO DOCKET ACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFILIATION BOUCHEY,G.D.
Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
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SUBJECT:
Responds to NRC 901224 ltr re violations noted in Insp Rept I
50-397/90-29 on 901113-16.Corrective actions:,separate procedure will be developed to provide method to document D
transfer of radioactive matl to another licensee.
DISTRIBUTION CODE:
IEOSD COPIES RECEIVED:LTR f ENCL Q SIZE:. 5 TITLE: Environ 6 Radiological (50 DKT)-Insp ReWp /Notice of Violate.on Respons
/
NOTES RECIPIENT ID CODE/NAME PD5 LA ENG,P.L.
COPIES LTTR ENCL 1
0 1
1 RECIPIENT ID CODE/NAME PD5 PD COPIES LTTR ENCL 1
1 A
D INTERNAL: ACRS AEOD/DSP NMSS/LLOB 5E4 NRR/DLPQ/LPEB10 NRR/PMAS/ILRB12 EG 02 RGN5 DRSS/RPB RGN4 MURRAY,B EXTERNAL EG&G S IMPSON i F NSIC 2
1 1
1 1
1 1
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AEOD/ANDERSON, R COLLINS g D RGN 2
NMSS/SGTB 4E4 NRR/DREP/PRPB11 NUDOCS-ABSTRACT OGC/HDS1 RES RGN5 FILE 01 NRC PDR 1
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D NOTE TO ALL"RIDS" RECIPIENTS:
I'LEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOK I PI-37 (EXT. 20079) TO EL!i%'IINATEYOUR NAih!E FROih] DISTRIBUTION I.ISTS FOR DOCUP,IENTS YOU DON'T NEED!
OTAL NUMBER OF COPIES REQUIRED:
LTTR 26 ENCL 25
WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968
~ 3000 George IVashington N'ay
~ Richland, lVashington 99352 January 23, 1991 G02-91-012 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Hail Station Pl-137 Washington, D.
C.
20555 Gentlemen:
Subject:
NUCLEAR PLANT NO. 2, OPERATING LICENSE NO. NPF-21 NRC INSPECTION REPORT 90-29
RESPONSE
TO NOTICE OF VIOLATION The Washington Public Power Supply System hereby replies to the Notice of Violation contained in your letter dated December 24, 1990.
Our reply, pursuant to 'the provisions of Section 2.201, Title 10, Code of Federal Regulations, consists of this letter and Appendices A and B (attached).
In Appendix A, the violation is addressed with an explanation of our position regarding validity, corrective action and date of full compliance.
Appendix B
addresses the issue pertaining to the capability of the Reactor Building Post-LOCA Grab Sampler to quantify post accident radioiodine concentrations in effluents.
As requested in the cover letter to the subject inspection report, a schedule for correcting system performance is also contained in Appendix B.
Very truly yours, G.
D. Bouchey, Dir r
Licensing
& Assurance JDA/bk Attachments CC:
JB Hartin -
NRC RV NS Reynolds
- Winston/Strawn PL Eng -
NRR DL Williams - BPA/399 NRC Site Inspector
- 901A
~~:I01300137 AD)0).>-.
F'DR ADOCK 05000397 PDR Zgpg $
4
APPENDIX A During an NRC inspection conducted on November 13-16,
- 1990, a violation of NRC requirements was identified.
In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C (1990),
the'iolation is listed below:
10 CFR Part 71.5 requires, in part, that each licensee who transports licensed material outside of the confines of the plant shall comply with the regulations appropriate to the mode of transport of U.S. Department of Transportation (DOT) in 49 CFR Parts 170 through 189.
49 CFR Part 173.415(a)
- requires, in part, that each shipper of a DOT Specification 7A (Spec.
7A) package must maintain on file for at least one year after the latest shipment, a completed documentation of tests and an engineering evaluation showing that the construction
- methods, package
- design, and materials of construction comply with that specification (49 CFR 178.350);
Contrary to the
- above, March 21,
- 1990, the licensee shipped (WNP-2 shipment No. 90-20-02) radioactive materials in a
DOT Spec.
7A package, and on November 13, 1990 a completed documentation of tests and engineer-ing evaluations were not on file at the licensee's facilities.'his is a Severity Level IV violation (Supplement V).
Validit of Violation The Supply System acknowledges the, validity of this violation.
The reason for the violation is less than adequate procedures in that there were no guidelines in Plant procedures for documenting a transfer of custody for radioactive material.
In this particular situation, Battelle Pacific Northwest Laboratories requested a
Reactor Water Cleanup (RWCU) resin sample from the Supply System to use in support of an EPRI Waste Management Study.
The Supply System Health Physics person involved determined that Battelle would provide the shipping container, the transfer of material would take place within the confines of the WNP-2 Plant and that Battelle would transport the material offsite.
For these reasons, the Health Physics individual involved correctly viewed this transaction as a material transfer and not a shipment on the part of the Supply System.
The individual then documented this transaction on an Offsite Radioactive Shipment Record.
However, there was nothing on the form to indicate that this was not a Supply System shipment
- and, as such, documentation needed to be on file that a Department, of Transportation, Specification 7A, container was used in the process.
Efforts to obtain container test documentation from Battelle were unsuccessful.
Appendix A-Page 2 of 2 This type of transaction with Battelle has occurred in the past to support research projects; however, the type of material was designated as Limited Quantity.
Limited Quantity material does not require container specifica-tion test documentation
- and, as in this
- case, the transactions were transfers of 'radioactive material and not shipments.
In those few instances where the Supply System has shipped Type A material to Battelle for research, a certified container has been supplied by WNP-2.
Corrective Ste s Taken Results Achieved Plant Procedure (PPH) 11.2.23.6, "Shipping Other Than LSA Radioactive H'aterials,"
was modified on November 19, 1990 to include a signo'ff on the Radioactive Haterial Shipment Checklist that test results are on file for DOT, Specification A, containers.
Corrective Action to be Taken A separate procedure will be developed to provide a method to document a
transfer of radioactive material on Supply System property to another licensee authorized to transport such material on the Department of Energy Hanford Reservation.
Date of Full Com liance Although full compliance was achieved on November 19, 1990 when PPH 11.2.23.6 was revised, the separate proc'edure for. documenting the transfer of radioactive materials will be developed by Harch 31, 1991.
APPENDIX B
In the subject inspection report, the NRC followed up on its previous concern pertaining to the ability of the Reactor Building Post-Accident Grab Sampling System to representatively sample effluent releases under accident conditions involving steam, with attendant variable temperatures along the sample line due to halogen plateout and excessive condensation.
As requested in the cover letter to the inspection report, this appendix provides our response to this concern and includes the current schedule for correcting system performance.
Reactor Building Post-Accident Grab Sampler REA-SR-48 implements NUREG-0737 Item II.F. 1, Attachment 2, which requires the ability to representatively sample and quantify the effluent release rate for gaseous/particulate radioiodine under accident conditions.
In the NUREG, the type and extent of the accident for which sampling must be provided is not defined.
Testing on the system was initiated and results indicated that a correction factor was needed to attain a representative sample due to sample flow rate and line size associated with REA-SR-48.
The correction factor of 500 is large and considered to be undesirable for the long term.
- However, the factor is based upon independent offsite laboratory testing of the sample system (by means of a representative mockup) and is considered to be a valid and acceptable near-term compensation in the consideration of operability for the sample rack.
Furthermore, the identification of accident conditions involving steam, laden with radioiodine, venting through the elevated release path are beyond the WNP-2 design basis. 'ven in consideration of the requirement to evaluate the system to beyond-design-basis
- events, several failures must occur before such a
condition would be present.
Although NUREG-0737, Table II.F. 1-2, does require entrained moisture to be considered in the context of adsorber degradation, it is also not clear that this represents a requirement to discount the moisture removal provisions already provided >such as the Standby Gas Treatment System.
When combined with the environmental requirements for the attainment of threshold conditions causing'the potential for condensate accumulation, the beyond-design-basis accident scenarios are too complex to allow for design protection against all possible event paths.
Sampler REA-SR-48, with the correction factor applied and under conditions not causing condensate accumulation in the sample lines, is capable of reasonably quantifying post-accident radioiodine concentrations in effluents, even for many scenarios beyond the design basis.
Although the system is considered operable with the correction factor applied, the decision to assume that the instrument is unreliable would be accident-scenario-based and would be evaluated under those conditions by the emergency response team.
In the event of those beyond-design-basis conditions which would render the grab sampler inoperable, the high-range noble gas effluent radiation monitors would be utilized to quantify the radioiodine release rate using default values to reflect the presence in the stack of 25% of the core halogens that could be released based on the estimated percent fuel damage.
This conservative approach is based on Regulatory Guide 1.3 source term guidance and existing WNP-2 fuel damage assessment procedures, and is consistent with current Protective Action Recommendation (PAR) determination guidelines.
Appendix B
Page 2 of 2 However; it is also recognized that improvement in the performance of REA-SR-48 is advisable for the long term.
An evaluation is currently in progress to determine the best option for improving system performance with regard.to the requirements of NUREG-0737.
Included in this evaluation, is consideration of the use" of on-line effluent monitors incorporating recent developments in. pure germanium, helium-cooled sensor technology.
On-line analysis of the spectrum of nuclides present in the stack, combined with existing flow monitoring from REA-SR-37, could.ex'ceed the guidance in NUREG-0737.
The evaluation is currently scheduled for completion during July, 1991.
At the completion of the evaluation, a supplemental response will be provided describing the action to be taken and its scheduled implementation date.
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