ML18138A087
| ML18138A087 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 03/06/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Ferguson J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 8003180162 | |
| Download: ML18138A087 (31) | |
Text
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In Reply Refer To:
e UNITED STATES e
NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 MAR O 6 1980
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REGULATORY.DOClCET FILE COPY Virginia Electric and Power Company Attn:
J. H. Ferguson Executive Vice President-Power P. 0. Box 26666 Richmond, VA 23261 Gentlemen:
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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.
Enclosures:
- 1.
IE Circular No. 80-03
- 2.
List of IE Circulars Recently Issued
_Sincerely,
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,;:;o-J'- ;:mes P. O'Reilly Director.
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Virginia Electric and Power Company cc w/encl:
W.R. Cartwright, Station Manager Post Office Box 402 Mineral, Virginia 23117 P. G. Perry Senior Resident Engineer Post Office Box 38 Mineral, Virginia 23117 W. L. Stewart, Manager Post Office Box 315 Surry, Virginia 23883 MA°R O 6* 19£fC,..,,.
e UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 6, 1980 e
SSINS:
6830 Accession No.:
7912190685 IE Circular No. 80-03 PROTECTION FROM TOXIC GAS HAZARDS ft 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 cbntainers 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 buildings.
Both acid and caustic storage tanks located in a common building near the control room.
At the Dresden site, in August 1977, accidential m1x1ng 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
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e IE Circular No. 80-03 e
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 protect~on against chlorine releases specifically.
The majority of the plants currently operating, however, were buill and licensed prior to the development and implementation of this guidan~e.
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 evalu~tion 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:
Sections 2.2.1-2.2.2; 2.2.3 and 6.4 of NUREG 75/087
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IE Circular No. 80-03 March 6, 1980 e
RECENTLY ISSUED IE CIRCULARS Circular No.
80-03 80-02 80-01 79-25 79-24 79-23 79-22 79-21 79-20 79-19 Subject Protection from Toxic Gas Hazards Nuclear Power Plant Staff Work Hours Service Advice for GE Induction Disc Relays Shcok Arrestor Strut Assembly Interference Proper Installation and Calibration of Core Spray Pipe Break Detection Equipment on BWRs..
Motor Starters and and Contactors Failed to Operate Stroke Times for Power Operated Relief Valves Prevention of Unplanned Releases of Radioactivity Failure of GTE Sylvania Relay, Type PM Bulletin 7305, Catalog 5U12-11-AC with a 12V AC Coil Loose Locking Devices on Ingersoll-Rand Pumps Date of Issue 2/6/80 2/1/80 1/17/80 12/20/79 11/26/79 11/26/79 11/16/79 10/19/79 9/24/79 9/13/79
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Enclosure Issued to All holders of Power Reactor OLs Ii All holders of Reactor OLs, including research and test reactors, and CPs All licensees 'of nuclear power reactor operating facilities and holders of nuclear power reactor CPs All licensees and holders of power reactor CPs All Holders of a Power Reactor 01 or CP All Power Reactor Operating Facilities and Holders of Reactor CPs All Power Reactor Operating Facilities and all Utilities having a CP All holders of Power Reactor OLs and CPs All holders of Power Reactor OLs and CPs All Holders of Power Reactor OLs and CPs
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i SECTIONS 2.2.1 - 2.2.2 REVIEW RESPONSIBILITIES NUREG-75/087 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION IDENTIFICATION OF POTENTIAL HAZARDS IN SITE VICINITY Primary - Accident Analysis Branch (AAB)
Secondary - None I.
AREAS OF REVIEW
~ocations and separation distances from the site of industrial, military, ~d transportation facilities and routes in the vicinity of tbe site. Such facilities and routes *include air, ground, and water traffic, pipelines, and fixed 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 of the proposed plant. The purpose of this review is to establish the infonnation concerning 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 inciustria~,
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 act!yjties 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 hazar9ous materiaJs may be processed, stored, used, or transported in the vicinity of the'*:t,1ant, sufficient statistical data on such materials are provided to establish a basis for eYPluating 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 ~ade by the reviewer on each case.
The jud!Jllent of the areas to be given attention during the review is to be based on an inspection of the mate:-'ti.1 presented, the similarity of the material to that recently reviewed on other plants, and whether items of special safety significance are involved. The following procedures are followed:.
USNRC STANDARD REVIEW PLAN
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The reviewer should,, be especially alert, in the construction pennit(CP) stage review, f.or any potentially hazardous activities in close proximity of the plant, since the variety of activities having damage potential at ranges under about one kilometer can be very extensive. All identified facilities and activities within eight kilometers (5 miles) of the plant should be reviewed.
Facilities and activities at greater distances should be considered if they otherwise have the potential for affecting plant safety-related features.
At the operating license (OL) stage, most hazards will already have been identified.
Emphasis should be placed on any new information.
At the operating license stage, any analyses _pertaining to potential accidents involv-ing hazardous materials or activities in the vicinJty of the plant will be reviewed to ensure that results are appropriate in light of any new data or experience which is then available. Facilities which are likely tp either produce or coniume hazardous materials should be investigated as possible sources of traffic of haz'arl:lous materials past the site.
- 2.
Information should be obtained from sources other than the SAR wherever available, and should be used to check the accuracy and completeness of the infonnation sutxnitted in the SAR.
This independent information may be obtained from sources such as U.S.
Geological Survey (USGS) maps and aerial photos, published documents, contacts with state and federal agencies, and from other nuclear plant applications (especially if they are located in the same general area or on the same waterway.)
Information should also be obtained during the site visit and subsequent discussions with local officials. (See Standard Review Plan 2.1.l for further guidance with regard to site visits.) An attempt should be made to investigate future potential hazards over the proposed life of.the plant.
- 3.
The specific information relating to types of potentially hazardous material, includ-ing distance, quantity, and frequency of shipment, is reviewed to eliminate as many of the potential accident situations as possible by inspection, based on past review experience. At the operating license stage, nearby*industrial, military and trans-portation facilities and transportation routes will be reviewed for any changes or additions which may affect Jhe safe operation of the plant.. If these changes alter the data or assumptions used in previous hazards evaluations or demonstrate the need for new ones, appropriate evaluations w1'11 be performed.
Rev. l For pipeline hazards, Reference 7 may be used as an example of an acceptable risk assessment.
For cryogenic fuels, Reference.9 may be.used, and for tank barge,
risks, Reference 8.
For military aviation, Reference 10 may be used.
Safe separation distances for explosives are.identified in Reference 2, and for toxic *chemicals, References 3 and 4 should be consulted.
The distance frcm nearby railroad lines is checked to determine if the plant is within the range of a "rocketing" tank car which, from Reference 51 is taken to be
- 350 meters with the range for smaller pieces extending to 500 meters.
2.2.1-2
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Potential accidents which cannot be eliminated from cons.ideration as design basis events because the consequences of the accidents, if they should occur, could be serious enough to affect plant safety-related features, are identified. Potential accidents so identified are assessed in detail, using criteria in Standard Review Plan Sections 2.2.3, 3.5.1.5, or 3.5.l.6, as appropriate.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided, and that his evalu-ation is sufficiently complete and adequate to-support conclusions of the following type, to be used in the staff's safety evaluation r~port:_
"The nature and extent of activities involvin_g potentially hazardou\\materials which are conducted at nearby industrial, military, and transportat_ion fadlities have been evaluated to identify any such activities which have the potential for adversely affecting plant safety-related structures. Based on evaluation of information con-tained in the SAR, as well as information independently obtained by the staff, it is concluded that all potentially hazardous activities in the vicinity of the ~lant have been identified. The hazards associated with these activities have been reviewed and are discussed in Sections ~~~and of this SER."
If the activities are identified as being potentially hazardous, the evaluations described in Standard Review Plans 2.2.3; 3.5.l.5 and 3.5.l.6 are performed with respect to the inherent capability of the plant or special plant design measures to prev~nt radiological releases in excess of the 10 CFR Part 100 guidelines.
V.
REFERENCES
- l.
Department of the Army Technical Manual TM5,.;1"300, "Structures to Resist the Effects of Accidental Explosions," June 1969.
- 2.
Regulatory Guide 1.91, "Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plant Sites."
3; Regulatory Guide 1.1a: "AssLm1ptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a P'ostulated Hazardous Chemical Release."
- 4.
Regulatory Guide 1.95, "Protection of N'uclear Power Plant Control Room Operators Against an Accidental Chlorine Release."
- 5.
National Transportation Safety Board Railroad Accident Report, "Southern Railway Company, Train 154, Derailment with Fire and Explosion, Laurel, Mississippi, January 25, 1969," October 6, 1969.
- 6.
Regulatory Guide l. 70 1 "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2.
2.2.1-3 Rev. l
Rev. l
- 7.
NUREG-0014 Safety Evaluation Report, Hartsville Nuclear Plants Al, A2, Bl, and B2, April 1976, Docket STN 50-518.
- 8.
Safety Evaluation of the Beaver Valley Power Station, Unit No. 2 November 9, 1976 and supplements.
Docket 50-412.
- 9.
Safety Evaluation Report, Hope Creek Generating Station, Units l and 2, Supplement No. 5, March 1976, Docket 50-354 and 50-355.
- 10.
Project 485, Aircraft Considerations, Preapplication Site Review, Boardman Nuclear Plant. October 1973.
-,, U. S. GOVERNMENT PRINTING OFFICE:
1978.-720-387/277 2.2.1-4
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e NUREG-75/087 U.S. NUCLEAR REGULATORY COMMISSION 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 Statistics Branch (ASB/MPA)
I.
AREAS OF REVIEW The applicant's identification of potential accident situations in the vicinity of the plant is reviewed to determine the completeness of and the bases upon which these potential accidents were or were not accommodated in the design.
(See Standard Review Plans 2.2. l and 2.2.2.)
The applicant's probability analyses of potential accidents involving hazardous materials or activities in the vicinity of the plant, if such analyses have been performed, are also reviewed by ASB/MPA on request by AAB to determine that appropriate data and analytical models 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 f~~~ the presence of hazardous materials or activities in the v*icinity ~f 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 10-7 per year.
Because of the difficult; of assigning accurate numerical values to~he 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.
The probability of occurrence of the initiating events leaeing to potential consequences in excess of 10 CFR Part 100 exposure guidelines should be estimated using assumpti~ns that are as repfesentai:ive of the specific site as is practicable.
In *addition, because of the low probabilities of the events under consideration, data are often 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 10-6 per year is acceptable if, when combined with reasonable qualitative arguments, the realistic probability can be shown to be lower.
USNRC STANDARD REVIEW PLAN Standard review plana.,. prepared for the guidance of the Office of Nuclear Raactor Regulation ataff rnponelbla for th* review of application* ta con*truct and operate nuclur power plenta. Theaa documanta are mada avallabl1 to tha publlc a pan of tha Commlaalon'a policy to Inform tha nuclear industry and the general public of regulatory procedurn and pollcin. Standard review plan* are not aubatltutN for regulatory guldH, or tha Commiaaion'1 regulation* and compRanca with tham ii not required. The atandard review plan NCtlona ara keyed to Aeviaion 2 of the Sundard Format and Content of Safety Analyal1 Raponi for Nuelaar Power Planh. Not all aection, of th* Sundard Format have* corraponding rwtew plan.
PubUslled.-c1arc1 review plane will be N¥INd ~dically, u approptiale, to accommodate cammenta and to reflect new information and aaperlence.
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e The effects of design basis events have been adequately considered if analyses of the 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 events have been taken.
III. REVIEW PROCEDURES 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 (ASB), regarding possible effects of external events on plant structures or components.
The applicant's probability calculations are reviewed, and an independent ~robability fl, 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 be 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 performed.
The design parameters (e.g., overpressure) and physical phenomena (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 acco11111odated in the design.
If accidents involving release* of smoke, fla11111able or nonfla11111able gases, or chemical bearing clouds are considered to'be desig~asis events, an evaluation of the effects of these accidents on control roo~ habitability should be made in SAR Section 6.4 and on the operation of diesels and other safety-related e~uipment in SAR Chapter 9.
Special attention should be given to the review of standardized designs which propose*
criteria involving indivii"!tial 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-mad~ hazards.
Similarly, special attention should be given to the review of a site where several man-made hazards are identified, but none of which, individually, has a probability exceeding the acceptance crJteria stated herein. The objective of this special review should be to assure that the aggregate probabilii'Y of an outcome that may lead to unacceptable plant da~ge meets the acceptance criteria of Part II of this SRP Section.
(A. hypothetical example is a situation where the probability of shock wave overpressure greater than design Rev. 1 2.2.3-2
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-7 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; the aggregate probability may be judged sufficiently great that additional features of design are warranted.)
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:
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"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 basis events and has provided analyses of the effects of these accidents on ~he safety-related features of the plant. The applicant has demonstrated that tne 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 Format 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 land 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, Septenber 1976.
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- T NUREG-75/087 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION 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 huilding 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 acciderytal releases of toxic or radioactive gases_
A further objective is to assure that the control 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 ~ollowing items are reviewed:
- l.
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 (control 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 co2 levels become excessive.
- 3.
The control room ventilatior{system layout and functional design is reviewed to determine flow rates and filter efficien5ies 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 STANDARD REVIEW PLAN Standard review plan*.,. prepared for the guidance o! the Office of NudNr Ructor Regulation at*ff rnpon*ible for the review of application* to construct and operate nuclear power planta. Thfle documenta '"' made evallabla to the public a part of the Cammluion'a policy to Inform th* nud.. ir Industry and the eeneral public of regulatory procedurn and pollda. StanUrd review" plana are not aubathutN for regulatory gutdn or the Commluion'a reguladon* and compllance with them la not required. The atanda~d review plan NCtlona are keyed to Reviaion 2 of the Standard Format and Content of Safety Anatpla Report.a for Nuclur,-""'"* Not all MCtlon1 oftlM Standard Format-* co-ndlng -- plan.
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The flow rates and iodine removal efficiencies used in the analysis are obtained from*
the ETSB (see SRP 6.5. 1).
- 5.
The physical location of the control room with respect to potential release points of hazardous airborne materials (SAR Chapter 2 and other pertinent chapters) is reviewed to determine the location and source strength of radioactive, toxic, or noxious materials.
The layout of the control building is reviewed to assure that airborne materials will not.enter the control room from corridors or ventilation ducts, etc.
Estimat~s of dispersion of airborne contamination are made in conjunction with HMB.
- 6.
Radiation shielding provided by structural concrete is analyzed to determine the effectiveness of shielding and structure surrounding the control roorp.,~ The control building layouts are checked to see if radiation streaming through door's. (or other apertures) or from equipment might be a problem.
- 7.
Independent analyses are performed to determine whether dose values or toxic QGS con-centrations remain below recommended*levels.
The HMB provides meteorological input and checks the X/Q values for the control room location.
II.
ACCEPTANCE CRITERIA Rev.
- l.
Control Room Emergency Zone See Section III. l of this plan.
- 2.
Control Room Personnel Capacity Food, water, and medical supplies should be sufficient to maintain the emergency team (at least 5 men) for 5 days.
- 3.
Ventilation System Criteria (See III.3 of this plan)
The following criteria deal with the verification of acceptable system performance and assurance of system availability:
- a.
Isolation Dampers - DamP,ers used to isolate the,control zone from adjacent zones or the outside must be leaktight."'This may be accomplished by using low leakage dampers or valves.
The degree of leaktightness should be documented in the SAR.
- b.
Single Failure - A single failure of an active component should not result in loss of the system's functional performance.
All the components of the control room emergency filter train will be considered active components.
See Appendix A for criteria regarding valve or damper repair.
- c.
Pressurization Systems - Systems that will pressurize the control room during a radiation emergency should meet the following requirements:
(l) Those systems having pressurization rates of greater than or equal to 0.5 volume changes per hour will require periodic (every 18 aonths) verification that the makeup is! 10l of design value.
During plant construction or 6.4-2
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- 1 after any modifications to the control room that might significantly affect its capability to maintain a positive pressure, measurements should be taken to verify that the control room is pressurized to at least 1/8-inch water gauge relative to all surrounding air spaces (while applying the design makeup air rate).
(2) Those systems having pressurization rates of less than 0.5 and equal to or greater than 0.25 volume changes per hour will have identical testing
- requirements as indicated in (1), above.
In addition, at the CP stage an analysis must be provided (based on the p)anned leaktight design features) that ensures the feasibility of maintaining 1/8-inch water gauge differential with the design makeup air flow rate.
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(3) Those systems having pressurization rates of less than 0.25 volume changes should meet all the requirements for (2), above, except that periodic verification of control room pressurization (every 18 months) will be~
required.
- 4.
Toxic Gas Protection Self-contained breathing apparatus for the emergency team (at least 5 men) should be on hand.
A six-hour onsite bottled air supply should be available with unlimited offsite replenishment capability from nearby location(s).
Refer to References 3 through 6, and see Section III.3 of this plan.
- 5.
Emergency Standby Filters See Standard Review Plan 6.5. l for acceptance criteria for control room ESF systems.
Credit for iodine removal efficiencies will be given in accordance with Regulatory Guide 1.52.
Filter effici"encies for systems not covered by Regulatory Guide 1.52 will be determined on a case-by-case basis by ETSB.
- 6.
Relative Location of Source jind Control Room In general, the control room inlets must be so placed in relation to the location of potential release points as_to minimize 1ontrol room contamination in the event of a release. Specific criteria as to radiation and toxic gas sources are as follows:
- a.
Radiation Sources As a general rule the control room ventilation inlet should be separated from the 11ajor potential release points by at least 100 feet laterally and by 50 feet.
vertically.
However, the actual minimum distances must be based on the dose analyses.
Refer to Section III of this plan and Reference 7 for further information.
- b.
Toxic gases The minimum separation distance is dependent upon the gas in -question, the co~tainer size, and the available control room protection provisions.
Refer to Regulatory Guide 1.78 (Ref. 3) for general guidance and to Regulatory Guide 1.95 (Ref. 4) for specific acceptable design provisions related to chlorine.
6.4-3 Rev. l
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e
- 7.
Radiation Shielding
- 8.
See discussion of General Design Criterion 19 below.
Radioactive and Toxic Gas Hazards
- a.
Radiation Hazards The dose guidelines (see General Design Criterion 19, Appendix A of 10 CFR Part 50) used in approving emergency zone radiation protection provisions are as follows:
(1) Whole body gamma:
5 rem (2) Thyroid:
30 rem (3) Beta skin dose:
30 rem*
The whole body gamma dose consists of contributions from airborne radioactivity inside and outside the control room, as well as direct shine from fission products inside the reactor containment building.
- b.
Toxic Gases For acceptance purposes, three exposure categories are defined:
protective action exposure (2 minutes or less), short-term exposure (between 2 minutes and l hour), and long-term exposure (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or greater).
Because the physiological effects can vary widely from one toxic gas to another, the following general restrictions should be used as guidance:
there should be no chronic effects from exposure, and acute effects, if any, *should be reversible within a short period of time (several minutes) without benefit of medication other than the use of self-contained breathing apparatus.
The allowable limits should be established on the basis that the operators should be capable of carrying out their duties with a minimum of interference caused by the gas and *subsequent protective measures.
The limits for the three categories normally are set as follows:
(1) Long-term limit (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or greater):
use a limit assigned for occupational exposure (40-hour week).
(2) Short-term limit (2 minutes to l hour):
use a limit that will assure that the operator will not suffer incapacitating effects after a 1-hour exposure.
- Credit for the beta radiation shielding afforded by special protective clothing and eye pro-tection is allowed if the applicant commits to their use during severe radiation releases.
However, even though protective clothing is used, the calculated unprotected skin dose is not to exceed 75 rem.
The skin and thyroid dose levels are to be used only for judging the acceptability of the design provisions for protecting control room operators under postulated design basis accident conditions.
They are not to be interpreted as acceptable emergency doses.
The dose levels quoted here are derived for use in the controlled plant envirQnment and should not be confused with the conservative dose computation assumptions used in evaluating exposures to the general public for the purposes of comparison with the guideline values
~
of 10 CFR Part 100.
Re~. l 6.4-4
(3) Protective action limit (2 min. or less) use a limit that will assure that the operator will quickly recover after breathing apparatus is in place.
In determining this limit, it should be assumed that the concentration increases linearly with time from zero to two minutes and that the limit is attained at two minutes.
The protective action limit is used to determine the acceptability of emergency zone protection provisions during the time personnel are in the process of fitting themselves with self-contained breathing apparatus.
The other limits are used to detennine whether the concentrations with breathing apparatus in place are applicable.
(They are also used in those cases where the toxic levels are such that emergency zone isolation without use of protective geat is suffi-cient.) As an example of appropriate limits, the following are the tnree levels for chlorine gas:
Long-term:
l ppm by volume Short-term:
4 Protective action:
15 (See Reference 3 for protective action levels for other toxic gases.)
III. REVIEW PROCEDURES The reviewer selects and emphasizes aspects of the areas covered by this review plan as may be appropriate for a particular case.
The judgment-~n areas to be given attention and emphasis in the review is based on an inspection of the material presented to see whether it is similar to that recently reviewed for other plants and whether items of special safety significance are involved.
- 1.
Control Room Emergency Zone._.-~
The reviewer checks to see 'that. the zone includes the following:
- a.
Instrumentation and controls necessary for a safe shutdown of the plant, i.e.,
the control room, including the critical document reference file.
- b.
The computer room, if it is used as an integral part of the emergency response plan.
- c.
The shift supervisor's office.
- d.
The operators* wash room and the kitchen.
The emergency zone should be limited to those spaces requiring operator occupancy.
Spaces such as battery rooms, cable spreading rooms, or any other spaces not requiring continuous or frequent occupancy after a design basis accident (OBA) generally should 6.4-5 Rev. l
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e be excluded from the emergency zone.
Inclusion of these spaces may increase the probability of smoke or hazardous gases entering the emergency zone.
They may also increase the possibility of infiltration into the emergency zone, thus decreasing the effectiveness of the ventilation system in excluding contamination.
It is advantageous to have the emergency zone located on one floor, with the areas included in the zone being contiguous.
- 2.
Control Room Personnel Capacity The reviewer checks to see that emergency food and water are provided.
Normally, a five-day supply for five men would be sufficient for land-based plants.
A medical kit is also helpful.
Specific requirements for these items have not been formulated.
The air inside a 100,000 cubic foot control room would support five #ersons for at least six days.
Thus, CO2 buildup in an isolated emergency zone is no~ normally considered a limiting problem.
- 3.
Ventilation System Layout and Functional Design This area is a major portion of the review.
The procedures are as follows:
- a.
The type of system proposed is determined.
The following types of protection provisions are currently being employed for boiling water reactor (BWR) or pressurized water reactor (PWR) plants:
(1) Zone isolation, with the incoming air filtered and a positive pressure maintained by the ventilation system fans.
This arrangement is often provided for BWRs having high stacks.
Air flow rates are between 400 and 4000 cfm.
(2) Zone isolation, with filtered recirculated air. This arrangement is often provided for BWRs and PWRs with roof vents.
Recirculation rates range from 2,000 to 30,000 cfm.
(3) Zone isolation, with filtered recirculated air, and with a positive pressure maintained in the zone.
This arrangement is essentially the same as that in (2), with the addition of the positive pressure provision.
(4) Dual air inlets for the emergency zone.
In this arrangement, two widely spaced inlets are located outboard (on opposite sides) of potential toxic and radioactive gas sources.
The arrangement guarantees at least one inlet being free of contamination (except under extreme no-wind conditions). It can be used in all types of plants.
Makeup air supplied from the contamination-free inlet provides a positive pressure in the emergency zone and thus minimizes infiltration.
6.4-6
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- 'f (5) Bottled air supply for a limited time.
In this arrangement, a flow rate of 400 to 600 cfm is provided from compressed air containers for about one hour, to prevent inleakage.
It is used in systems having containments whose internal atmospheric pressure becomes negative within an hour after a OBA (subatmospheric containments).
- b.
The input parameters to the radiological dose model are determined (see Item 5).
The parameters are emergency zone volum~, filter efficiency, filtered makeup air flow rate, unfiltered inleakage (infiltration), and filtered recirculated air flow rate.
- c.
The ventilation system components and the system layout diagrams are examined.
The responsible reviewer in the ASB should be consulted if there are4questions pertaining to the system design.
He will determine if the system meets the single failure criterion as well as other safety requirements (see Standard Review Plan Section 9.4. 1).
Damper failure and fan failure are especially,
important.
The review should confirm that the failure of isolation dampers,on the upstream side of fans will not" result in too much unfiltered air entering the control room.
The AAB dose analysis results are used to determine how much unfiltered air can be tolerated.
- d.
The following information may be used in evaluating the specific system types (see Reference 7 for further discussion):
(1) Zone isolation, with filtered incoming air and positive pressure.
These systems may not be sufficiently effecti~~.in protecting against iodine.
The staff allows an iodine protection factor (IPF), which is defined as the
'time-integrated concentration of iodine outside over the time-integrated concentration within the emerge*ncy zone, of 20 to 100 for filters built, maintained, and operated according to Regulatory Guide 1.52 (an IPF of 100 requires deep bed f!lters).
Such ~ystems are likely to provide a sufficient reduction in iodine':i:o,ncentration only if the source is at some distance from the inlets. Thus, in most:::cases only plants with high stacks c~ 100 m) would meet Criterion 19 with this system.
Normally the staff suggests that these systems be modified to allow isolation and operation with recircu-lated air since only minor ducting changes are necessary.
(2) Zone isolation, with filtered recirculated air. These systems have a.
greater potential for controlling iodine than those having once-through filters.
IPFs ranging from 20 to over 150 can be achieved.
These are the usual designs for plants having vents located at containment roof level. A system having a recirculation rate of 5000 cfm and a filter efficiency of 95% would be rated as follows:
6.4-7 Rev. l
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~
Infiltration (cfm) 200 100 50 25 IPF*
25 49 96 191
- Within the range of interest, ~he iodine protection factor is directly proportional to recirculation flow rate times efficiency.
Infiltration should be determined conservatively. The calculated or measured gross leakage is used to determine the infiltration rate that will be applied in the evaluation of the radiological consequenc~s faf postulated accidents. This rate is determined as follows:
(i) The leakage from the control room when pressurized to 1/8-inch water gauge is calculated on the basis of the gross leakage data. 9ne-half of this value is used to represent the base infiltration rate.
Component leak rates may be used to calculate gross leakage (see, for example, References 8 and 9).
(ii) The base infiltration rate is augmented by adding to it the estimated contribution of opening and closing of doors associated with such activ-ities as the required emergency procedures external to the control room.
Normally, 10 cfm is used for this additional contribution.
(iii} An additional factor that is used to modify the base infiltration rate is the enhancement of the infiltration occurring at the dampers or valves µpstream of recirculation fans.
When closed, these dampers typically are exposed to a several-inch water gauge pressure differential. This is accounted for by an additional infiltration contribution over the base infiltration at l/8-inch water gauge.
The use of an infiltration1-ate that is based on calculation is acceptable except in the case where the applicant has assumed exceptionally low rates of infiltration. In these cases, more substantial verification or proof may be required. for instance, if an applicant submits an analysis that shows a gross leakage rate of less than 0.06 volume changes per hour, the reviewer would require that th~ gross leakage be verified by periodic tests as described in Regulatory Position C.5 of Regulatory Guide 1.95.
(3)
Zone isolation, with filtered recirculated air, and with a positive pressure.
This system is essentially the same as the preceding one. However, an additional operational mode is possible. Makeup air for pressurization is admitted. It is filtered before entering the emergency zone. Pressurization reduces the unfiltered inleakage that is assumed to occur-when the emergency 6.4-8
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- I*
zone is not pressurized.
Assuming a filter fan capacity of 5000 cfm and a filter efficiency of 95%, the following protection factors result (flows in cfm):
IPF (Assuming IPF (Assuming MakeuE Air Recirculated Air No Infiltration)
Infiltration*2 400 4600 238 159 750 4250 128 101 1000 4000 96 80 The makeup flow rate should have adequate margin to assure that the control woom will be maintained at a pressure.of at least 1/8-inch weter gauge.
The applicant should indicate that an acceptance test will be *performed to verify adequate pressurization. If the makeup rate is less than 0. 5*,volume changes per hour, supporting calculations are required to verify adequate air flow.
If the makeup rate is less than 0.25 volume changes per hour, periodic.
verification testing is required in addition to the calculations and the acceptance test.
A question that often arises is whether pressurization" or "isolation and recirculation" of the control room is to be preferred.
Which design gives the lowest doses depends upon the assumptions~ to unfiltered inleakage.
Isolation is generally preferred in that it wjll limit the entrance of noble gases (not filterable) and, in addition, it is a better approach when the accident involves a.short-term "puff release." If infiltration is 25 cfm or less, "isolation" would be best in any event.
A second question related to. the first involves the method of operation.
The following pos~ibilities have been considered:
(i) Automatic isolation with s.ubsequent manual control of pressurization.
.:,r (ii) Automatic isolation wit'!, i111nediate automatic pressurization.
The first is advantageous in the case of external puff releases. Simple isolation would minimize the buildup of the unfilterable noble gases. It would also protect the filters from excessive concentrations in the case of a chlorine release.
However, the second method does guarantee that infiltration (unfiltered) is reduced to near zero innediately upon accident detection.
This would be beneficial inthe case where the contamination transport path Normally 10 cfm infiltration is assumed for conservatism.
This flow could be reduced or eli-minated if the applicant provides assurance that backflow (primarily as a Tesult of ingress and egress) will not occur.
This may mean installing two-door vestibules or equivalent.
6.4-9 Rev. 1
- - ---=--**'.,-*-.-..
Rev. 1 to the emergency zone is mainly inside the building.
Method (i) should be used in the case of a toxic gas release and either method (i) or (ii) should be used in the case of a radiological release, provided Criterion 19 guidelines can be satisfied.
(A substantial time delay should be assumed where manual isolation is assumed, e.g., 20 minutes for the purposes of dose calculations.)
(4) Dual air inlets for the emergency zone.
Several plants have utilized this concept.
The viability of the dual inlet concept depends upon whether or not the placement of the inlets assures that one inlet will always be free from contamination.
The assurance of a contamination-free inlet depends in part upon building wake effects, terrain, and the possibility of wind stagnation or revers a 1.
For ~:.:ample, in a situation where the inlets are,,~ ocated at the extreme edges of the plant structures (e.g., one on the north si~e and one on the south side), it is possible under certain low probability conditions for both inlets to be contaminated from the same point source.
Reference 7 presents the interim position for dealing with the evaluation of X/Q's, for dual inlet systems.
With dual inlets placed on plant structures are on opposite sides of potential radiation release points (e.g, containment building), and are capable of functioning with an assumed single active failure in the inlet isolation system, the following considerations may be applied to the evaluation of the control room X/Q's:
(i) Dual inlet designs without manual or automatic selection control -
Equation (6) of Reference 7 may be used with respect to the least favorable inlet location to estimate X/Q's.
The estimated values can be reduced by a factor of two (2) to account for dilution effects associated with a dual inlet configuration. This is based upon. the dilution derived from drawing in equal amounts of clean and contaminated air through tw6 open inlets.
(ii) Dual inlet designs limited to manual selection control - Equation (6) of Reference 7 may be used with respect to the more favorable inlet location to estimate the X/Q's.
The estimated values can be reduced by a factor of four (4) to account for dilution effects associated with a dual inlet configuration and the relative probability that the operator will make the proper inlet selection. The reduction factor is contingent upon having redundant radiation detectors within each air inlet. The reduction factor is based on the judgment that trained control room operators, in conjunction with radiation alarm
- indication, will select and close the contaminated air inlet.
(iii) Dual inlet designs with automatic selection control features -
Equation (6) of Reference 7 may be used with respect to the more 6.4-10
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- '?.
- 4.
- 5.
e e
favorable inlet location to estimate the X/Q's.
The estimated values can be reduced by about a factor of ten (10) to account for the ability to select a "clean" air inlet. The actual factor may be somewhat higher if the inlet configuration begins to approach the remote air inlet concept such that the probability of having one clean air inlet is relatively high.
Plant configuration and meteorological conditions should be used as the principal basis for reduction factors greater than ten (10).
The reduction factor of ten (10) or more is contingent upon having redundant radiation detectors in each inlet and the provisions of acceptable control logic which would be used in the automatic selection of a clean air inlet.
Because damage to the ducting might seriously affect the system'capability to protect the operators, the ducting should be seismic Category I and should be protected against tornado missiles.
In addition, the number and placement of dampers must be such as to assure both flow and isolation,,in each inlet assuming one single active component failure.
(See Appendix A for information on the damper repair alternative.) The location of the intakes with respect to the plant security fence should also be reviewed.
(5) Bottled air supply for a limited time.
In some plant designs the containment pressure is reduced below atmospheric within one hour after a OBA.
This generally assures that after one hour significant radioactive material will not be released from the containment.
Such a design makes it feasible to maintain the control room above atmospheric pressure by use of bottled air.
Periodic pressurization tests are requiJ?ed to determine that the rated flow (normally about 300 to 600 cfm) is sufficient to pressurize the control room to at least 1/8-i_nch water gauge.
The system is also required to be composed of several separate circuits (one of which is assumed to be inoperative to account for a possible single failure).
At least one (nonredundant) once-through filter syst~m for pressurization as a stand>y for accidents of long duration is also desir.able.
~
Compressed air bottles should be protected from tornado missiles or internally-generated missiles and should be placed so as not to cause damage to vital equipment or interference with operation if they fail.
Emergency Standby Filters Refer to Standard Review Plan 6.5. l.
Relative Location of Source and Control Room This review area involves identification of all potential sources of toxic, radioactive, or*otherwise potentially hazardous gases and analysis of their transport to the control room.
There are three basic categories:
OBA radioactive sources, toxic gases such as 6.4-11 Rev. 1
~-.J:".
Rev. 1
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e chlorine, and gases with the potential for being released inside confined areas adjacent to the control room.
- a.
OBA Radioactive Sources The LOCA source terms determined in Appendix A to Standard Review Plan 15.6.5 review are referred to and routinely used to evaluate radiation levels external to the control room.
The disper&al from the containment or the standby gas treatment vent is determined with a building wake diffusion model.
This model is discussed in Reference 7.
Other DBAs are reviewed to determine whether they might constitute a more severe hazard than the LOCA.
If this is suspected, an additional analysis is performed for the suspect DBAs.
The HMB p~ovides the meteorological input and reviews the AAB calculation of X/Q value~.
- b.
Toxic Gases The applicant is asked to identify those toxic substances stored (or transported) on or in the vicinity of the site which may pose a threat to the reactor',oper-ators by producing toxic gases upon accidental release.
The method used to determine whether the quantity or location of the toxic material is such as to require closer study is described in Regulatory Guide 1.78 (Ref. 3).
This guide also discusses the methods for analyzing the degree of risk and states, in general terms, the various protective measures that could be instituted if the hazard is found to be too great.
In the case of chlorine, specific acceptable protective provisions have been determined; these are described in detail in Reference 4.
In summary, the following provisions or tbejr equivalent are required (pertaining to the emergency zone ventilation system):
(1) Quick-acting toxic gas detectors.
(2) Automatic emergency zone isolation.
(3) Emergency zone leaktightness~
(4) Limited fresh air makeup rates.
(5) Breathing apparatus and associ~ted bottled air supply.
(Note that the best solution for a particular case will depend on the toxic gas in question and on the specific ventilation system design.)
- c.
Confined Area Releases The reviewer studies the control building layout in relation to potential sources inside the control building or adjacent connected buildings. The following con-cerns are checked:
6.4-12
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I
- 6.
e
- r (1) Storage locations of CO2 or other firefighting materials should be such as to*
eliminate the possibility of significant quantities of the gases entering the emergency zone.
(The ASB has the primary responsibility in this area.)
(2) The ventilation zones adjac~nt to the emergency zone should be configured and balanced to preclude air flow toward the emergency zone.
(3) All*pressurized equipment and piping (e.g., main steam lines and turbines) t~at could cause significant pr~ssure gradient~ when failed inside buildings should be isolated from the emergency zon~ by multiple barriers such as multiple door vestibules or their equivalent.
Radiation Shielding Contro 1 room ~perators as we 11 as other p 1 ant personne 1 are protected froni **radiation sources associated with a normally operating plant by various combinations of shield-ing and distance.
The adequacy of this type of protection for normal operating conditions is reviewed and evaluated by_ the RAB.
To a large extent the same'radia-tion shielding (and missile b'arriers) also provides protection from design basis accident radiation sources.
This is especially true with respect to the control room walls which usually consist of at least 18 inches of concrete.
In most cases, the radiation coming from external design basis accident radiation sources is attenuated to negligible levels.
However, the following items should be considered qualitatively in assessing the adequacy of control room radiation shielding:
- a.
Control room structure boundary.
Wall, ceiling, and floor materials and thick-ness should be reviewed.
Eighteen inches to two feet of concrete or its equivale~t will be adequate in most cases.
- b.
Radiation streaming. The control room structure'boundary should be reviewed with respect to penetrations (e.g., doors, ducts, stairways, etc.). The potential for radiation streaming from acciden~ sources should be identified, and if deemed necess.ary, quantitatively evaluated. Support from the RAB may be required for some radiation streaming dose calculitions.
- c.
Radiation shielding from internal sources. If sources internal to the control room complex are identified, radiation shielding against them should be reviewed.
Typical sources in this category include contaminated filter trains, or airborne radioactivity in enclosures adjacent to the control r*oom.
Evaluations of radiation shielding effectiveness with respect to the above items should be performed using simplified analytical models for point, line, or volume sources such as those presented in References 10 and 11.
If more extended analysis is required, analytical support from the RAB should *be requested.
The applicant*s coverage of the above items should also be reviewed in terms of completeness, method of analysis, and assumptions.
6.4-13 Rev. l
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- 7.
e e
- r-Independent Analyses
- a.
Contro 1 Room Doses Although the applicant is required to calculate doses to the control room operators, independent analyses are made by the AAB because of the diversity of control room habitability system designs and the engineering judgment involved in their evaluation.
Using the approach indicated in Reference 7, the source terms and doses due to a OBA are calculated.
The source terms determined by the AAB's independent analysis of LPZ doses for a LOCA are used.
The methods and assumptions for this calculation are pr~sented in Appendix A to Standard Review Plan 15.6.5.
The control room doses are determined by estimating the X/Q from the source points t.o the emergency zone using meteorological input supplied by the HMB, by determining the credit for the emergency zone's protection features, and by calculating the dose.
Figure 6.4-1 shows a form which may 'be used to summarize the information that is needed for the control room dose 'c:*a1culation.
The effective X/Q's are used for calculating the doses.
The dose is then compared with the guidelines of General Design Criterion 19.
If the guid~lines are exceeded, the applicant is asked to improve the system.
In the event that other DBAs are expected to result in doses comparable to or higher than the LOCA, additional analyses are performed.
The limiting accidents are compared with Criterion 19.
- b.
Other Analyses Special case analyses are performed when questions are raised about certain poten-tial sources of toxic or radioactive gases.
The methods used in these analyses conform to current OBA methods concerning dispersion and dose calculations.
Regulatory Guide 1.78 should be consulted_~y-the site analyst to see if nearby facilities could present a potential hazard that requires detailed analysis.
IV.
EVALUAT!ON FINDINGS The reviewer verifies that sufficient information has been provided and that the review and calculations support conclu,sions of the following type, to be included in the staff's safety evaluation report:
- 1.
If the plant meets Criterion 19, the following statement or its equivalent is made:
Rev. 1 11The applicant proposes to meet General Design Criterion 19 of Appendix A to 10 CFR Part 50 by use of concrete shielding and by installing redundant~~ cfm recirculating charcoal filters in the control room ventilation system.
These fil\\ers will be automatically activated upon an accident signal, high radiation signal, or high chlorine signal.
Independent calculations of the potential radiation doses to control room personnel following a LOCA show the resultant doses to be within the guidelines of Criterion 19.
11
.. r. ".
- ).
)
l
- 2.
e If the design is not adequate, the fact is stated.
Alternatives such as an increase i~
the charcoal filter flow rate may be indicated as is given in the example below:
- .r. ;;..
The staff has calculated the potential radiation doses to control room personnel following a LOCA.
The resultant whole body doses are within the guidelines of Criterion 19.
The thyroid dose resulting from exposure to radioactive iodine exceeds the dose guidelines.
The applicant will be required to commit to increasing the filtration system size from 2000 cfm to 4000 cfm.
This increased filtration wil.l be sufficient to keep the estimated thyroid doses within the guidelines."
3.. If special protection provisions for toxic gases are not required, the /ollowing state-ment or its equivalent is made:
"The habi tabi 1 i ty of the contro 1 room was eva 1 uated using the procedures descri becj in Regulatory Guide 1.78.
As indicated in Section 2.2, no offsite storage or transport of chemicals is close enough to the plant to be considered a hazard.
There are no onsite chemicals that can be considered hazardous under Regulatory Guide 1.78.
A sodium hypochlorite biocide system will be used, thus eliminating an onsite chlorine hazard.
Therefore, special provisions for protection against toxic gases will not be required.
Self-contained breathing apparatus is provided for the emergency crew to provide assurance of control room habitability in the event of occurrences such as smoke hazards."
- 4.
If special protection provisions are required, compliance or noncompliance with the guidelines of Regulatory Guides 1.78 and l.95 should be stated.
V.
REFERENCES
- 1.
10 CFR Part 50, Appendix*A, General Design Criterion 19, "Control Room."
- 2.
Regulatory Guide 1. 52, "Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere C*leanup System Air Filtration and Adsorption Units of
- 3.
Light-Water-Cooled Nuclear'Power Plants."
Regulatory Guide 1. 78, 11Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."
- 4.
Draft Regulatory Guide 1. 95, "Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release."
- 5.
- 6.
Draft Regulatory Gui de 8. X, "Acceptable Programs for Respiratory Protect ion."
"Manual of Respiratory Protection Against Airborne Radioactive Material,"
WASH-1287, U.S. Atomic Energy Commission (1974).
6.4-15 Rev. 1
- 7.
e e
K. G. Murphy and K. M. Campe, "Nuclear Power Pl ant Control Room Ventilation System Design for Meeting General Design Criterion 19," 13th AEC Air Cleaning Conference, August 1974.
- 8.
"Leakage Characteristics of Openings for Reactor Housing Components,"
NAA-SR-MEM0-5137, Atomics International, Div. of North American Aviation, Inc.,
June 20, 1960.
- 9.
R. L. Koontz, et al., "Leakage Characteristics of Conventional Building Components for Reactor Housing Construction," Trans. Am. Nucl. Soc., November 1961.
- 10.
R.G. Jaeger, et al., eds., "Engineering Compendium on Radiation Shie.ld,fog, 11 Vol. 1, "Shielding Fundamentals and Methods," Springer-Verlag (1968).
- 11.
N.M. Schaeffer, "Reactor Shielding for Nuclear Engineers," TID-75951, U.S. Atomic Energy Commission.
Rev. 1
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e
~I SECTION 6.4 APPENDIX A ACCEPTANCE CRITERIA FOR VALVE OR DAMPER REPAIR ALTERNATIVE The control room. ventilation system must meet the *criterion that it work properly even with a single failure of an active component.
In certain cases; complex valve or damper configurations are required to meet the single failure criterion.
For example, assurance of the isolation and operability of each leg of a dual inlet system at various times after a po$'tulated accident could require a four-valve arrangement in which two pairs of series valves are c.?nnected in parallel.
The mechnical, power, and control components of such arrangements combine to form
- a rather complex system.
Credit will be allowed for an alternative system that allowed the failed valve to be manually repositioned so that it will not interfere with the opera~jon of the system.
For example, in the case of a dual inlet system, if credit for repair is given, then two valves_,in series in each leg of the dual inlet would be acceptable.
Where a valve fails closed but meets the criteria given below, credit would be allowed for the valve to be repositioned and locked in an open position.
The approval of the repair option is contingent upon the intrinsic reliability of the internal components of the valve or damper and also upon the ease and ability to overcome the failure of the external actuating components (electrical relays, motors, hydraulic pistons, etc.).
The following criteria or.their equivalent will be required:
- 1.
The valve or damper components must be listed as to which are considered internal (nonrepairable) and which external (repairable). These must be designed to meet the following criteria.
- a.
Internal valve components (components that are difficult to repair manually without, opening the ductwork) must.be judged to have a very l~w probability of failure.
The component design details will be revi'ewed and characteristics such as simplicity, ruggedness, and suspectibility to postulated failure mechanisms will be considered in arriving at an engineered judgment of the acceptability of the internal component design with respect to reliability.
For example, a butterfly valve welded or keyed onto a pivot shaft would be considered a high reliability internal component.
Conversely, multiple-blade dampers, actuated by multi-element linkages or*
pneumatically-operated components internal to the ducts, would be viewed as being subject to failure.
- b.
External valve components (components including motors and power supplies that are to be assumed repairable or removable) must be designed to ensure that the failed valve component can be bypassed easily and safely and'that the valve can be manipulated into an acceptable position.
The electronic components must be isolated from other equipment to assure that the repair operations do not result in further equipment failure.
6.4-17 Rev. l
- 2.
The location and positioning of the valve or damper must permit easy access from the control room for convenient repair, especially under applicable OBA conditions.
- 3.
Appropriate control room instrumentation should be provided for a clear indication and annunciation of valve or damper malfunction.
- 4.
Periodic manipulation of the valve or damper by control room operators should be required for training purposes and to verify proper marual operability of the valve or damper.
- 5.
The need for manual manipulations of the failed valve or damper should not be recurrent
~
during the course of the accident.
Manipulation should not occur more th~~.once during the accident.
Adjustment or realignment of other parts of the system should be possible from the control room with the failed valve in a fixed position.
- 6.
The time for repair used in the computation of control room exposures should be taken as the time necessary to repair the valve plus a one-half hour margin.
No manual corrEction will be credited during the first two hours of the accident.
- 7.
Compliance with the above criteria should be documented in the SAR whenever the repair option is used.
Rev. 1 6.4-18
FIGURE 6.4-1 Sufllllary Sheet for Control Room Dose Analysis MEMORANDUM TO:
cc:
L. Hulman R.W. Houston (Habitability File)
CONCERNING CONTROL ROOM DOSE ANALYSIS FOR (Site Analyst)
(Meteorologist)
The following summarizes the X/Q' s used in determining *the control room operator dose for the subject plant:
VENTILATION SYSTEM DESCRIPTION SKETCH OF SYSTEM (and inlets/sources if applicable)
SUMMATION OF X/Q ANALYSIS Source/Receptor Type and Distance S/D Ratio Number of 22 1/2° Wind Direction Sectors that Result in Exposure Central Wind Sector 5% Wind Speed (m/sec)
Projected Area of Wake (m2)
Time 0-8 hr 8-24 hr 1-4 day 4-30 day Wind Speed Factor Wind Direction Factor ACTION REQUESTED Site Analyst For your information only K Factor (sector wind is blowing from)
_, **
- 40% Wind Speed (m/sec) 5% X/Q (sec/m3)
Occupancy Factor l
l 0.6 0.4 Effective X/Q's Please use the effective X/Q's in TACT run and provide control room doses.
In addi-tion, please sufllllarize safety system assumptions and indicate their status (interim or final).
Meteorologist These are ;nterim X/Q's.
Please review to determine their reasonableness.
These are final X/Q's.
Please determine ;f they are accurate based on your analysis of site data.
Please Contact -----------
6.4-19 Rev. 1
. *.r.
.'