ML17309A226
| ML17309A226 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/19/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TASK-06-01, TASK-6-1, TASK-RR LSO5-82-02-080, LSO5-82-2-80, NUDOCS 8202240001 | |
| Download: ML17309A226 (16) | |
Text
0 February 19, 1982 Docket No. 50-244 LS05 02-",080 o
Mr. John E. Maier, Vice President
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Rochester. Gas and Electric Corporation
,89 East,-Avenue Rochester, New York 14649
Dear Mr. Maier:
SUBJECT:
SEP TOPIC VI-1, ORGANIC MATERIALS AND POST-ACCIDENT CHEMISTRY R. E.
GINNA r
Enclosed is our draft safety evaluation of SEP Topic VI-1 for the R. E. Ginna Nuclear Power Plant.
This evaluation compares your facility with the criteria currently used for licensing new facilities.
Our evaluation of the Post-Accident-Chemistry portion of the evaluation is based on the safety assessment provided in your letter dated November 16, 1981.
The Organic Materials portion was based on information provided in your letters dated November 6, 1981 and January
, 1982.
Enclosure:
As stated o
cc w/enclosure:
See next page f" 820224000l 8202i9 L
PDR ADOCK 05000244 P
PDR Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility.
This assessment may be revised in the futurh if your facility design is changed or if NRC criteria relating to ggaf this subject are modified before the integrated assessment is completed.
Sincerely,
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Dennis M. Crutchfield, Chief pz>r Operating Reactors Branch 85 Division of Licensing
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Ginna Docket No. 50-244 Rev. 2/8/82 Mr. John E. Maier CC Harry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N.
W.
Suite 1100 Washington, D. C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York 14519 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West
- Ontario, New York 14519 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 U. S. Environmental. Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406
ENCLOSURE 1
SYSTE LUATION PROGRAM TOPIC VI-1 R. E.
GI NNA TOPIC:
VI-l, Or anic Materials and Post Accident Chemistr I.
INTRODUCTION The design basis for selection of paints. and other organic materials is not documented for most operating reactors.
Topic VI-l,.is intended to review the plant design to assure that organic materials, such as organic paints and
- coatings, used inside containment do not behave adversely during accidents when 'they may be exposed to high radiation fields.
In particular the possi-bility of coatings clogging sump screens should be minimized.
Low pH solutions that may be recirculated within the containment after a Design Basis Accident (DBA) may accelerate chloride stress corrosion cracking and increase the volatility of dissolved iodines.
The objective of Topic VI-1 is to assure that appropriate methods are available to< raise or maintain the pH of solutions expected to be recirpulated within containment after a
DBA.
P Or anic Materials:
An assessment of the suitability of organic materials in t e containment includes the review of paints and other organic materials used inside the containment including the possible interactions of the de-composition products of organic materials with Engineered Safety Features (ESF),
such as filters.
Post Accident Chemistr An assessment of post accident chemistry includes a determination of proper water chemistry in the containment spray during the injection phase following a DBA and that appropriate methods are available to raise or maintain the pH of mixed solution in the containment sump.
REVIEW CRITERIA Or anic Materials:
The plant design was reviewed with regard to General Design Criterion 1, "guality Standards and Records" of Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants" which requires that structures and systems important to safety be designed and tested to quality standards commensurate with the importance of the safety function to be performed.
Also, contained in the review was Appendix B to 10 CFR 50, "guality Assurance Criteria for Nuclear Power Plan'ty and Fuel Reprocessing Plants."
This guide describes an acceptable method of complying with the Comnissions quality assurance requirements with regard to protective coatings.
Post Accident Chemistr The design was reviewed with regard to General,-Design Crster)on Reactor Coolant Pressure Boundary" of Appendix A to 10 CFR Part 50.
This requires that the reactor coolant pressure boundary be designed'and erected so as to have an extremely low probability of abnormal leakage and gross rupture.
Also, regarded in the review was General Design Criterion 41, "Containment Atmosphere Cleanup," of Appendix A to 10 CFR Part 50.
This requires that systems to control substances released in reactor containment be provided to reduce the concentration and quality of fission products released to the environment following a postulated accident.
0 0
III.
RELATED SAFETY TOPICS The effectiveness of the'.iddine removal system:=is evaluated as part of Topic XV-19, for a spectrum of loss-of-coolant accidents.
Topic VI-7.E reviews the ECCS in the recirculation mode to confirm the effectiveness of the ECCS.
IV.
REVIEW GUIDELINES Or anic Materials':
Current guidance for the", review of organic materials in containment is provided in Sections 6.'l..'1, "Engineered Safety Features
~Materials" and 6.1.2, "Org'anic Materials" of the Standard Review Plan and in Regulatory Guicfe 1.54,
'Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants"..
Regulatory Guide 1.54 endorses the requirements and guidelines described in detail in ANSI N101.4-1972, "Quality Assurance for Protective Coatings (Paints) for the Nuclear Industry" and ANSI N5.12-1974, "Protective Coatings (Paints) for the Nuclear Industry".
Post Accident Chemistr Guidance for the review of post accident chem-istry is provided in Sections 6.1.1 and 6.5.2 of the Standard Review Plan.
Section 6.1.1 is '.related to assuring that appropriate methods are avail-able to raise or maintain the pH of the mixture'of the containment
sion product removal in the containment sump during the recirculation phase and to preclude long term corrosion problems after the accident.
Section 6.5.2 is related to providing proper water chemistry in the con-tainment spray and sump during the injection phase following a Design Basis Accident.
V.
EVALUATION Or anic Materials:
By letters dated June 14, 1979, November 6, 1'981, and January 27, 1982, the licensee provided references to the types and amounts and the environmental testing of organic<coating materials used in the plant.
The foam containment liner and protective coating systems comprise the bulk of the organic materials (outside of electrical cable insulation) in the containment.
Accident effects on cable insulation are reviewed under NUREG-0458 (Reference 1).
The inner surface of the containment building is insulated with panels of foam encapsulated in stainless steel sheets.
The total weight of insulating foam in the containment building is about 15,000 lbs.
The licensee provided an estimate of the amounts of hydrogen, organic gases and hydrogen chloride which would be produced by radiation from the decom-position of the foam during a
DBA, as well as an analysis of the possible safety consequences of the degradation process.
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Because of the high rate for radiolytic HCl generation from the foam (vinyl chloride) the licensee conservatively assumed that all of the chloride content of the foam would be slowly released over a period of several weeks after a
DBA, resulting in about 4 metric tons of HC1 gas.
The calculation indicated that this much HCl would be effective-ly buffered to a
pH of about 8.5 by the alkaline NaOH additions to the containment spray system.
We concur with the conclusion o'f the -anal-yses that corrosive attack by the buffered chloride-containing spray solution will not significantly affect the operation of engineered safety features.
The licensee also provided an analysis of the possible corrosive effects of thermally and radiolytically generated HCl acting on the inner surf-aces of the stainless panels where they are in direct contact with the foam.
This is of concern because loss of panel structural integrity could al.low foam and metal pieces to be introduced into the sump causing possible flow restriction.
Although the location of the sump makes the possibility of the transport of metal and foam fragments into the sump are remote, we re-analyzed the time required to cause such
'toss of integrity using a more conservative corrosion rate for stainless steel (2, 3) than that used by the licensee.
As a result of our analysis, we concluded that neutralization of the corrosive environment by the containment spray would occur before significant damage to the materials integrity and therefore the sump screen would not clog.
No specific data are available for gas generation from the irradiated vinyl foam used in the containment.
For the generically similar pure polyvinyl chloride, the only gas produced in more than trace quantities is HC1.
Thermal, decomposition of po'lyvinyI chloride at high temperatures in air also produces small quantities of other chemicals that would not appreciably affect corrosion and f'lammability.
We conclude that the quantities of other gases produces by radiolytic or thermal decomposition of foam are too low to affect the flammability of the containment atmosphere.
We hive reviewed the plant design with re'gard to the effect of paints and coatings under accident conditions.
In general,.phenolic based paints are among the most radiation resistant, remMning serviceable after radiation dosage in excess of 10g rad )References 4,
5 and 6).
For a severe Design Basis Accident (DBA), 10~ rad would be a conservative dose estimate.
Most paint areas would receive less than 107 rad.
Phenolic resins are also stable to temperatures of the order of 300'F and to mildly acidic or basic aqueous solutions (Reference 7).
On the basis of the above information, we find that there is reasonable assurance that the radiation, thermal and chemical resistance of the organic coatings used inithe plant is sufficiently high that deterioration under DBA conditions would not interfere with the operation of engineered safety features.
Certain small surface areas of plant equipment were coated with industr'ial coatings whose radiation resistance has not been tested.
- However, because only small areas of these coatings are exposed in the containment, we conclude that their failure under accident conditions would not present a
significant safety hazard.
Very smal'1 amounts of gas are evolved when aromatic organic compounds of the types found in radiation-resistant plastic are irradiated.
For ex-
- ample, a phenolic plastic ireadiated of a dose of 109 rads produced 3 ml (STP) a gas per gram of plastic (Reference 6).
For the approximately 25 cubic feet of organic'icoatingnexisting in the containment; approximately 8 cubic feet of gas would be generated for the conservatively estimated DBA dose of 108 rads.
The gas is mostly hydrogen and less than a tenth of it is volatile organic compunds.
The pr esence of this small amount of organic gases in containment after a
DBA would not interfere with the absorption of organic iodides by the charcoal filters in the containment purge system.".
The amount of hydrogen from this source is small compared to that which could be produced in a DBA from the zirconium-water r'eaction, from the radiolysis of water, or from the reaction of the zinc in inorganic zinc coatings with high temperature borate solutions..
(Reference 8).
Hydrogen generation from 'the latter sources is being reviewed independently of the SEP program as part of the TMI Task Action Plan (Task II.B.7, NUREG-0660).
By letter of November.6, 1981, the licensee committed to make. a visual inspection of the exposed paint surfaces in containment approximately every three years.
We find the proposed inspection frequency",acceptable for monitoring the condition of coatings.
On the basis of our review we conclude that the organic materials 'used in the plant are acceptable and will not interfere with the operation of engineered safety features under accident conditions.
gualification tests demonstrate that the types of organic coating materials used in the con-tainment will maintain their integrity arid remain in serviceable condition after exposure to the severe environmental conditions of a DBA.
Insignifi-cant quantites of organic gases and of hydrogen would be generated under these conditions.
To provide further assurance that delamination, flaking or peeling of coating materials will not interfere with the operation of engineered safety features, the licensee has proposed an acceptable inspection program for coated surfaces in containment.
Post Accident Chemistr The Containment Spray System, used both to reduce post-DBA contai:nment pres-sure and to remove post-accident fission products from the containment atmosphere (especially radioactive iodine), is automatically actuated by a high-high containment pressure signal.
The containment spray system users borated water, with a concentration ranging from 2000 to 2300 ppm of boron, from the refueling water storage
- tank, and a sodium hydroxide solution of 30 percent by weight from the
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chemical spr ay.
The resulting solution drains.to the containment sump'..
During the'recirculation phase of containment spray or ECC systems, the containment sump solution will be recirculated.'n inert cover gas of nitrogen is provided for the sodium hydroxide storage tank which is housed in the heated auxiliary building.
The licensee has provided an evaluation of the post accident chemistry by letter dated November 6, 1981.
The licensee'has calculated the range of pH for the injection phase of the containment spray using the following range of parameters:
~HH l tl RWST:
1650 gpm at 2300 ppm boron NaOH Tank:
10 gpm at 30 weight percent The resulting solution had a
pH of 8.3.
~HHH H
RWST:
1200 gpm at 2000 ppm boron NaOH Tank:
51 gpm at 30 weight percent The resulting solution had a
pH of 9.1.
The effectiveness of the containment spray for iodine removal is presen-ted in SEP Topic XV-19.
The licensee also provided an analysis using the following range of parameters which:were used to evaluate the upper;and lower limits of pH in the sump following a design basis loss'of coolant accident:
H RWST:
300,000 gal at 2300 ppm boron Boric Acid Tanks:
6000 gal.at 20,000 ppm boron Reactor Coolant System:
46,713 gal at 1000 ppm boron Accumulators:
2216 ft3 at 1800 ppm boron NaOH Tank:
2700 gal at 30 weight percent The resulting mixed sump solution had a
pH of 9.0.
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H RWST:
215,000 gal at 2000 ppm boron Boric Acid Tanks:
3000 gal at 20,000 ppm boron Reactor Coolant System:
46,713 gal at'0 ppm boron Accumulators:
2268 ft3 at 1800 ppm boron
'aOH Tank:
4500 gal at 30 weight percent The resulting mixed sump solution had a
pH of 10.3.H
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The staff has independently evaluated the pH of the containment sump solution which results from mixing of the containment spray solution with the reactor coolant and ECCS fluids in the containment sump during recirculation.
We verified by independent calculations that sufficient sodium hydroxide is available to raise the pH of the containment sump
~so ution above
%he minimum-level of 7>0, consistent with the guidance of Branch Technical Position MTEB 6-1, to reduce the likelihood of stress corrosion cracking of stainless steel components.
We have also verified that the sump maximum pH will not exceed a value of 10.5 as specified in SRP 6.5.2.
Our calculations were based on the volumes and concentra-tions provided by the licensee.
Additional information relative to the use and effectiveness of the Ginna Containment Spray System is provided in.the G'irma FSAR; Section 6.4.3; Appendix 6A, " Iodine Removal Effective Evaluation of the Containment Spray System";
Appendix 6C,
!Design Intention Regarding the Selection of a Spray Additive"; and Appendix 6E, "Materials Compatability Review".
We determined that there is no need for heat tracing for the chemical additive storage tank since the tank is indoors.
The plant Technfcal Specifications of R.E.
Ginna Nuclear Power Plant provide for demonstration of operability of the containment spray system and monthly testing of the sodium hydroxide concentration in the chemical additive storage tank, consistent with the Standard Technical Specifications for Westinghouse pressurized water reactors.
In addition, the plant Technical Specifications contain limiting conditions for opera-tion which specify the minimum volume and boron concentration for the re-fueling water storage tank, accummulators and boric acid tanks and the
~,minimum volume and NaOH concentration for the spray additive tank.
We have reviewed the plant Technical Specifications and have determined that they provide sufficient assurance that the pH values will be within the range specified in SRP 6.5.2 and are acceptable.
VI.
CONCLUSIONS Or anic Materials:
Based on the above, we conclude that there is reasonable assurance that the integrity of paints and organic coatings within the containment will be maintained under normal operating conditions and those of a DBA, and that there will be no undue hazard to the health and safety of the public, and therefore, the paints and organic coating materials are acceptable.
Post Accident Chemistr On the basis of the above evaluation, we conclude that the R.E.
Ginna Iodine Removal System meets the post accident chemistry requirements of SRP 6.5.2, SRP 6.1.1, and GDC 14 and is, therefore, acceptable.
VII.
REFERENCES 1.
NUREG-,0485, "Shoi;t Term'Safety Assessment on the Environmental gualification of Safety-"Related Electrical Equipment of SEP Opera-ting Reactors",,
May 1978.
2.
Metals Handbook, Vol. 1, 8th Ed. American Society of Metals,
- 1966, Page 569.
3.
L.L. Shreir, "Corrosion", Vol. 1, Newnes - Butterworth London, 1977 Pages 3:55 and 3:56.
4.
ORNL-3589, "Gamma Radiation Damage and Decontamination'valuation of Protective Coatings and Other Material for Hot Laboratory and Fuel Processing Facilities", G.A. West and C.D. Watson,
- February, 1965.
5.
ORNL-3916, "Unit Operations Section quarterly Progress Report",
July - September,
- 1965, M.E. Whatley et al., March 1966, Pages 66-75.
6.
!'Radiation 'Effects on Organic Materials", edited by R.O. Bolt and J.G. Carroll, Academic Press, New York and London, 1963, Chapter 6,
Page 239.
7.
"Chemical, Engineers Handbook", J.H. Perry, Editor, 5th Edition, Pages 23-60 to 23-68.
8.
H.E. Zittel, "Post Accident Hydrogen Generation from Protective Coatings in Power Reactors",
Nuclear Technology 17, 143-6 (1973).
9.
Rochester Gas and Electric-,Corporation, Robert Emmet Ginna Nuclear Power Plant Unit No. 1, Final Facility Description and Safety Analysis Report (FSAR), Volume 2, Chapter 6.
10.
R.E.
Ginna Plant Technical Specifications.
11.
NUREG-0452, "Standardized 1'echnical Specifications for Westinghouse PWR's",,
June 15, 1978.
12, Letter from John E. Maier to Dennis M. Crutchfield, dated November 6, 1981.
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