ML19290F128
| ML19290F128 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/06/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Mary Johnson SOUTH CAROLINA ELECTRIC & GAS CO. |
| References | |
| NUDOCS 8003180156 | |
| Download: ML19290F128 (2) | |
Text
.
e
/
'o UNITED STATES
~,,
NUCLEAR REGULATORY COMMISSION n
.E REGION 11 6
o 101 MARIETTA ST., N.W., SulTE 3100 ATLANTA, GEORGIA 30303 o
MAR 0 61980 In Reply Refer To:
RII_;)[P0
-39 >
9 South Carolina Electric and Gas Company Attn:
M. C. Johnson, Vice President Special Services and Purchasing P. O. Box 764 Columbia, SC 29218 Gentlemen:
The enclosed IE Circular is forwarded to you for information. No written response to this IE Circular is required.
If you have any questions related to the subject, please contact this office.
Sincerely,
/
/
k James P. O'Reilly Director
Enclosures:
1.
IE Circular No. 80-03 2.
List of IE Circulars Recently Issued 8003180t56
~
" O O e 19e -
South Carolina Electric anu Gas Company,
cc w/ enc 1:
T. B. Conners, Jr.
Conners, Moore and Corber 1747 Pennsylvania Avenue, N.W.
Washington, D.C.
20006 A. A. Smith Quality Assurance Post Office Box 8 Jenkinsville, South Carolina 29065 O. S. Bradham, Manager Nuclear Operations Post Office Box 8 Jenkinsville, South Carolina 29065
b SSINS: 6830 Accession No.
UNITED STATES 7912190685 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 g-O{
March 6, 1980
=d IE Circular No. 80-03 PROTECTION FROM T0XIC GAS HAZARDS Chlorine gas releases have been reported at two different reactor facilities in the past two years.
At Millstone, in March 1978, a leak of about 100 standard cubic feet of chlorine (about a gallon of liquid) occurred over a ten minute period, resulting in the hospitalization of 15 people.
The ventilation system carried the chlorine into the plant buildings, where personnel distress was noted. No injuries occurred in the buildings due to the small size of the release.
At Browns Ferry, in June 1979, a small leak from a diaphragm on a chlorine reducing valve resulted in the hospitalization of five people, including a control room operator.
Chlorine is highly toxic, producing symptoms after several hours exposure in concentrations of only one ppm.
Concentrations of 50 ppm are dangerous for even short exposures and 1000 ppm is fatal for brief exposures.
- Chlorine, used at some power stations to control organisms in the circulating water, is normally supplied in one ton containers or in tank cars of up to 90 tons capacity.
Other potential sources of toxic gas that have been identified at nuclear power plants include:
Nearby industrial facilities. At Waterford, in July 1979, construction forces had to be evacuated for two and a half hours due to a chlorine gas release from a nearby chemical plant.
Chlorine transportation on adjacent highways, railways and rivers.
Large tanks of aqueous ammonia stored near plant uuildings.
Both acid and caustic storage tanks located in a common building near the control room. At the Dresden site, in August 1977, accidential mixing of acid and caustic solutions resulted in toxic fumes that entered the control room via the ventilation system.
Criterion 19 of Appendix A to 10 CFR 50 requires a control room from which action can be taken to maintain the reactor in a safe condition under accident conditions. The control room designs in current license applications are O
me m e--e
IE Circular No. 80-03 March 6, 1980 Page 2 of 2 reviewed for operator protection from toxic gases (as well as radiation), in accordance with Standard Review Plan (SRP) 6.4 (NUREG 75/087 dated 11/24/75).
Related information on the identification of potential hazards and the evalua-tion of potential accidents can be found in SRP sections 2.2.1-2.2.2 and 2.2.3 respectively.
The SRP references Regulatory Guide 1.78 (dated June 1974) on control room habitability during chemical releases.
It also references Regulatory Guide 1.95 on requirements for protection against chlorine releases specifically.
The majority of the plants currently operating, however, were built and licensed prior to the development and implementation of this guidance. A review of some older plants, with respect to toxic gas hazards indicates that they do not have the degree of protection that would be required for present day plants. Evaluation of the protection of control rooms from toxic gas releases is part of the systematic evaluation program currently being carried out on certain older plants. Also, as older facilities submit requests for significant license amendments, their design features and controls for protec-tion of control rooms are reviewed and, if appropriate, are required to be changed. However, the recent history of frequent toxic gas release incidents appears to warrant a more rapid implementation of the newer toxic gas protec-tion policies.
For the above reasons, it is strongly recommended that:
You evaluate your plant (s) against section 6.4 and applicable parts of sections 2.2.1-2.2.2 and 2.2.3 of the SRP with respect to toxic gas hazards.
Where the degree of protection against toxic gas hazards is found to be significantly less than that specified in the SRP, provide the controls or propose the design changes necessary to achieve an equivalent level of protection.
No written response to this circular is required.
If you desire additional information regarding this matter, contact the Director of the appropriate NRC Regional Office.
Attachments:
e
- tions 2.2.1-2.2.2; 2.2.3 and 6.4 of NUREG 75/087
t l
IE Circular No. 80-03 Enclosure March 6, 1980 RECENTLY ISSUED IE CIRCULARS Circular Subject Date of Issued to No.
Issue 80-03 Protection from Toxic Gas 2/6/80 All holders of Power Hazards Reactor OLs 80-02 Nuclear Power Plant Staff 2/1/80 All holddrs of Reactor Work Hours OLs, including research and test reactors, and cps 80-01 Service Advice for GE 1/17/80 All licensees 'of Induction Disc Relays nuclear power reactor operating facilities and holders of nuclear power reactor cps 79-25 Shcok Arrestor Strut 12/20/79 All licensees and Assembly Interference holders of power reactor cps 79-24 Proper Installation and 11/26/79 All Holders of a Calibration of Core Spray Power Reactor OL or Pipe Break Detection CP Equipment on BWRs.
79-23 Motor Starters and 11/26/79 All Power Reactor and Contactors Failed Operating Facilities to Operate and Holders of Reactor cps 79-22 Stroke Times for Power 11/16/79 All Power Reactor Operated Relief Valves Operating Facilities and all Utilities having a CP 79-21 Prevention of Unplanned 10/19/79 All holders of Power Releases of Radioactivity Reactor OLs and cps 79-20 Failure of GTE Sylvania 9/24/79 All holders of Power Relay, Type PM Bulletin Reactor OLs and cps 7305, Catalog SU12-11-AC with a 12V AC Coil 79-19 Loose Locking Devices 9/13/79 All Holders of Power on Ingersoll-Rand Pumps Reactor OLs and cps
NUREG 75/087 Sa ns fo o
U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTIONS 2.2.1 - 2.2.2 IDENTIFICATION OF POTENTIAL HAZARDS IN SITE VICINITY l
REVIEW RESPONSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - None I.
AREAS OF REVIEW
. ocations and separation distances from the site of industrial, military, apd transportation L
facilities and routes in the vicinity of the sf +a Such facilities and routes include air, ground, and water traffic, pipelines, and fined manufacturing, processing, and storage facilities. Potential external hazards or hazardous materials that are present or which may reasonably be expected to be present during the projected life time o,f the proposed plant. The purpose Of this review is to establish the infomation concer'ning the presence of potential external hazards which is to be used in further review in Sections 2.2.3, 3.5.1.5, and 3.5.1.6.
II. ACCEPTANCE CRITERIA 1.
Data in the SAR adequately describes the locations and distances of 1.x.ustrial,,
I military, and transportation facilities in the vicinity of the plant, and is in agreement with data obtained from other sources, when available.
2.
Descriptions of the nature and extent of activities conducted at nearby facilities, including the products and materials likely to be processed, stored, used, or trans-ported, are adequate to permit evaluations of possible hazards in Part 3 review sections dealing with' specific hazards.
3.
Where potentially hazardous materials may be processed, stored, used, or transported in the vicinity of the plant, sufficient statistical data on such materials are l
provided to establish a basis for egluating the potential hazard to the plant.
III. REVIEW PROCEDURES Selection and emphasis of various aspects of the areas covered by this review plan will be rEade by the reviewer on each case. The judgnent of the areas to be given attention during the review is to be based on an inspection of the mate:'cl presented, the similarity of the material to that recently reviewed on other plants, and whether items of special l
safety significance are involved. The following procedures are followed:
DUPLICATE DOCUMENT
, = e *
- o'a=
,,,--n-
= ~a -
e ***,dC;,,,,
af"M'""".!
Entire docutnent previously entered into system unde
~
s ANO
_u
-e.-**=
,w.g.
n.p.,een up e.a. D C. M No. of pages:
- * = =, --
u
a o
- p..to NUREG 75/087 f
o j,'
U.S. NUCLEAR REGULATORY COMMISSION
)
o i
STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS REVIEW RESPONSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - Applied Statis:ics Branch (ASS /MFA)
I.
AREAS OF PEVIEW The a;;plicant's ident fication of potential accident situations in the vicintty of tPt plant is reviewed to i etermine the corpleteness of and the bases upon which these potential accidents were or were not accorrodated in the design. (See Standard Review Plans 2.2.1 and 2.2.2.)
The applicant's probabi.ity analyses of potential accidents involving hazardous mater'ials or activities in the vicinity of the plant, if such analyses have been performed, are also reviewed by ASS /MPA on request by AAB to determine tha* appropriate data and analytical riodels have been utilized.
The analyses of the consequences of accidents involving nearby industrial, military, and transportation facilities which have been identified as design basis events are reviewed.
II. ACCEPTANCE CRITERIA The identification of design basis events resulting froa the presence of hazardous materials or activities in the vicinity of the plant is acceptable if the design basis events include each postulated type of accident for which the expected rate of occurrence of potential exposures in excess of the 10 CFR Part 100 guidelines is estimated to exceed the NRC staff objective of approximately 10 per year. Because of the difficulty of assigning accurate numerical values to the expected rate of unprecedented potential hazards generally con-sidered in this review plan, judgment must be used as to the acceptability of the overall risk presented.
s The probability of occurrence of the initiating events leading to potential consequences in excess of 10 CFR Part 100 exposure guidelines should be estimated using assumptions that are as representative of the specific site as is practicable. In addition, because of the low probabilities of the events under consideration, data are of ten not available to permit accurate calculation of probabilities. Accordingly, the expected rate of occur-rence of potential exposures in excess of the 10 CFR Part 100 guidelines of approximately 10 per year is acceptable if, when coribined with reasonable qualitative arguments, the realistic probability can be shown to be lower.
USNRC STANDARD REVIEW PLAN ei $"w.YN."I.Y.Nw. N7m E.
.e mY. _ N.Y. iNU N"~EU.U
~
~
- "*,:': '",,:',.~'. ':'O 0 ".TR ".* ::'
'.'*%' :=."". " "" "'"'.".T ',~ 00,.Ol'.". ** '"".T.~.,.,lCC
~.
(
~
m.u.
_.m k6W$60 6
d The effects of design basis events have been adequately considered if analyses of the l
effects of those accidents on the safety-related features of the plant have been performed and measures (e.g., hardening, fire protection) to mitigate the consequences of such f
events have been taken.
III. REVIEW PROCEDURES g
In some cases it may be necessary to consult with or obtain specific data from other branches, such as the Structural Engineering Branch (SEB) or Auxiliary Systems Branch l
(ASB), regarding possible effects of external events on plant structures or components.
The applicant's probability calculations are reviewed, and an independent probability analysis is performed by the staff if the potential hazard is considered significant enough to affect the licensability of the site or is important to the identification of design basis events.
All stochastic variables that affect the occurrence or severity of the postulated event are identified, and judged to be either independent or conditioned by other variables.
Probabilistic models should te tested, where possible, against all available information.
If the model or any portion of it, by simple extension, can be used to predict an observ-able accident rate, this test should be perfomed.
The design parameters (e.g., overpressure) and physical phenorrena (e.g., gas concentration) selected by the applicant for each design basis event are reviewed to ascertain that the values are comparable to the values used in previous analyses and found to be acceptable by the staff.
Each design basis event is reviewed to determine that the effects of the event on the safety features of the plant have been adequately accomodated in the design.
If accidents involving release of smoke, flammable or nonflamable gases, or chemical bearing clouds are considered to be design, basis events, an evaluation of the effects of these accidents on control room habitability should be made in SAR Section 6.4 and on the operation of diesels and other safety-related equipment in SAR Chapter 9.
Special attention should be given to the review of standardized designs which propose l
criteria involving individual numerical probability criteria for individual classes of external man-made hazards. In such instances the reviewer should establish that the envelope also includes an overall criterion that limits the aggregate probability of exceed-ing design criteria associated with all of the identified external man-made hazards.
Similarly, special attention should be given to the review of a site where several man-made hazards are identified, bt.t none of which, individually, has a probability exceeding the acceptance cr,1teria stated herein. The objective of this special review should be to assure that the aggregate probabilit*y of an outcome that may lead to unacceptable plant dainage meets the acceptance criteria of Part II of this SRP Section. (Ahypothetical example is a situation where the probability of shock wave overpressure greater than design
~h Rev. 1 2.2.3-2
~.
a eO
% e g
e.
.e
=
o overpressure is about 10 per reactor year from accidents at a nearby industrial facility, and approximately equal probabilities of exceeding design pressure from railway accidents, highway accidents and from shipping accidents. Individually each may be judged acceptably low; ine aggregate probability may be judged sufficiently great that additional features of design are wt T ied.)
IV.
EVALUATION FINDINGS If the reviewer verifies that sufficient information has been provided and that his evaluation is sufficiently complete and adequate to meet the acceptance criteria in Section II of this SRP, conclusions of the following type may be prepared for the staff's safety evaluation report:
"The applicant has identified potential accidents which could occur in the vicinity of the plant, and from these has selected those which should be considered as design l
basis events and has provided analyses of the effects of these accidents on the safety-related features of the plant. The applicant has derronstrated that the plant is adequately protected and can be operated with an acceptable degree of safety with regard to potential accidents which may occur as the result of activities at nearby industrial, military, and transportation facilities."
V.
REFERENCES Regulatory Guide 1.70, " Standard Forrat and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2.
Affidavit of Jacques B. J, Read before the Atomic Safety and Licensing Board in the matter of Skagit Nuclear Power Project, Units 1 and 2. July 15,1976. Docket Nos. STN 50-522, 523.
Atomic Safety and Licensing Board, Supplemental Initial Decision in the Matter of Hope Creek Generating Station, Units 1 and 2. March 28, 1977. Docket Nos. 50-354, 355.
Section 2, Supplement 2 to the Floating Nuclear Plant Safety Evaluation Report. Docket No. STN 50-437, September 1976.
2.2.3-3 Rev. 1
o NUREG 75/087 n arcg'o e
+
,! 1 3 2,8 i,.,
/
U.S. NUCLEAR REGULATORY COMMISSION l
STANDARD REVIEW PLAN N'%...p#
OFFICE OF NUCLEAR REACTOR REGULATION
+
e SECTION 6.4 HABITABILITY SYSTEMS REVIEW RESPONSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - Hydrology-Meteorology Branch (HMB)
Auxiliary Systems Branch (ASB)
Effluent Treatment Systems Branch (ETSB)
I.
AREAS OF REVIEW The control room ventilation system and control building layout and structures, as described in the applicant's safety analysis report (SAR), are reviewed with the objective of assuring that plant operators are adequately protected against the effects of accidental releases of toxic or radioactive gases. A further objective is to assure that the co'ntrol room can be maintained as the center from which emergency teams can safely operate in the case of a design basis radiological release. To assure that these objectives are accom-plished the following items are reviewed:
1.
The zone serviced by the control room emergency ventilation system is examined to ascertain that all critical areas requiring access in the event of an accident are included within the zone (cor trol room, kitchen, sanitary facilities, etc.) and to assure that those areas not requiring access are generally excluded from the zone.
2.
The capacity of the control room in terms of the number of people it can accommodate for an extended period of time is reviewed to confirm the adequacy of emergency food and medical supplies and self-contained breathing apparatus and to determine the length of time the control room can be isolated before CO levels become excessive 2
3.
The control room ventilation system layout and functional design is reviewed to determine flow rates and filter efficiengies for input into the AAB analyses of the buildup of radioactive or toxic gases inside the control room, assuming a design basis release. Basic deficiencies that might impair the effectiveness of the system are examined. In addition, the system operation and procedures are reviewed. The ASB has primary responsibility in the system review area under Standard Review Plan (SRP) 9.4.1.
The ASB is consulted when reviewing hardware and operating procedures, usNRC STAND DUPLICATE DOCUMENT W C O M " Of
.f E '""* Y U
Entire docunient previously
""f* A],'.'."""'"w entered into svstem under:
[
%--w.
ANO U '" Y E
~
~~
N o.
of pages:
'e
=
e.