ML19320C065
| ML19320C065 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/26/1980 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-1.A.1.1, TASK-2.K.3.09, TASK-2.K.3.10, TASK-2.K.3.12, TASK-TM NUDOCS 8007150714 | |
| Download: ML19320C065 (29) | |
Text
e7 fAlab;ma Powar Comesny -
- 600 North latn Street -
- -. r * -
. Post Ottoce Box 2641 At Birminghstn. ALTm:3*291-
- Telephone 205 323-5341 -
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' F. L. CLAYTON, JR.
abama NWer Senior Vice President the soutwn eectrc svs:em l
June 26, 1980
-Docket
- o. 50-3h8
~Mr. D. G. Eisenhut
- - Director, Division of Operating Reactors -
U. S. Nuclear Regulatory Commission 7920 ::criclh. War.us -
Bethesda, Maryland 20014
Dear Mr. Eisenhut:
In accordance with.your letter dated May 7,1980, concernits
-additional D1I-2 related requirements, Alabama Power Company submits the enciesed response for the Joseph M. Farley Nuclesr Plant - Unit 1.
If you have any questions, please advise.
Ycurs very truly,
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F. L. Clayton,.Jr.
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. Enclosure ec:
Mr.
R.'A. Thomas l!r. G. F. Trowbridge 001 S I
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j I.A.1.3 SHIFT MAmlIIIG
-Recuirement The minimum shift crew for a unit shall include three operators, plus an additional three operators when the unit is operating.
Shift staffing may be adjusted at mult-unit stations to allow credit for operators holding licenses on more than one unit.
In each control room, including co= mon control rooms for multiple units, there shall be at all times a licensed reactor operator for each reactor loaded with fuel and a senior reactor operator licensed for each reactor that is operating. There shall also be onsite at all-times, an additional relief operator licensed for each reactor, a licensed senior reactor operator who is designated'as' shift supervisor, and any other licensed senior reactor operators required so that their total number is at least one more than the number of control rooms from which a reactor is being operated.
Administrative procedures shall be established to limit maximum work hours of all personnel performing a safety-related function to no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous duty with at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> between work periods, no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven-day pericd, and no more than 1h consecutive days of work without at least 2 consecutive days off.
These requirements shn11 be met by July 1,1982.
Overtime requirements shall be =et by August 1,19S0.
Response
The' control room will, by July l',1982, be staffed as indicated in Table 1 (attached). These staffing levels vill meet this shift' manning requirement.
Alabama Power. Company is in total agreement with and fully supports the Commission's requirement to limit maximum work hours of personnel performing safety-related functions. At this time the Company is able to commit to 'vorking personnel under normal conditions to no more Ethan 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in' any seven-day period and to working these personnel
- no more than 1h consecutive days without at least two consecutive days off.. At the present time the Company is not able to fully.
commit to limit the work of personnel to more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> under continuous duty with at.least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> off between work periods.
Our inability to make this cormitment is due primarily to a binding contract with-the International Brotherhood of Electrical Workers (IBEW) which does not provide for such scheduling of personnel.
The Company is initiating negotiations with the. IBEW to meet the 12
- hour provisions of your requirement.
In addition, newly constituted training and licensing requirements raise uncertainties about total manpower needs which make your 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provisions, difficult to meet at this time.
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TABLE 1 TOTAL MINIMUM CRE'J C0;&OSITION FOR SHARED CONTROL ROOM Condition' of Unit 1 - No Fuel in Unit 2 APPLICABLE MODES LICENSE CATEGORY 1, 2, 3-& 4 5&6 SOL 2
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1 Hon-Licensed 2
1 Conditien of Unit 1 - Unit 2 in MODES 1, 2, 3 or h APPLICABLE MODES
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LICE'!3E CATEGORY 1, 2, 3 & 4 5&6 SQL**
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Uon-Licensed 5
h Condition of Unit i - Unit 2 in MODES 5 or 6 APPLICABLS MODES LICENSE CATEGORY _
1, 2, 3 & 4 5&6 SQL**
2 1*****
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2 Non-Licensed-h 3
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TABLE 1 (Continued)'
- ' Does not include the licensed Senior Reactor Operator ~ cr Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIOIIS after the initial fuel' loading.
Assumes each individual is licensed on the unit to which he. is assigned.
- Shift crew conposition for shared control room (including an individual-qualffled in radiation protection procedures) may be less than the minimum
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requirements for a period of time not to-exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on-duty shift crew members provided ic: mediate action is taken to restore the shift crew composition to -
within the minimum requirements of Table 1.
- - One individual licensed-on both units.
'"he licensed individuals, one licensed on Unit 1 and one licensed on Unit 2, may be substituted in lieu of that individual.
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' IIA.3 1 EREVIs2 SCOPE A :D CRITERIA FOR LICE'!SI:!G 0:A !ILT"IC !S E,
POSITION
- All reactor. operator license applicants shall take a written exa=i-nation with a new categorf dealing with the principles of heat transfer and fluid mechanics, a time-limit of nine. hours, and a passing grade of 80 percent overall and 70 percent in each categorf.
All senior reactor operator license applicants shall take the reactor operator examination, an operating test, and a senior -reactor operator written examination with a new category dealing with the theory of fluids and thermodynamics, a time limit of.seven hours, and a passing
_ grade of 80 percent overall and 70. percent in each en.tegory.
These requirements shall be met conmencing May 1, 1980.
(See letter of March 28,1980.)-
Response
The t aining scope and criteria for reactor operator and senior reactor operator licenses examinations are.being changed to comply with require-
.ments' stipulated in the March 23, 1980, letter on this subject. All requirements vill be i=plemented by the schedule stipulated in the March 28, 1980, letter.
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I.C.5' PROCEDURES FOR FEEDSACK C? OPEF".02:0 E:GE?T:CE Td PT_i:2
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-POSITION
]di accordance with Task Action Plan I.C.5, Preceduras f:r 723ita:h of Operating Experience to Plant Staff, each licensee shall review its procedures and revise them as necessary to assure that operating infor=ation pertinent to plant safety originating both within and outside-the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. These procedures, which are to be implemented by January 1, 1981, shall:
Reauirement (1) Clearly identify organisational responsibilities for reviav cf operating experience, the feedback of pertinent infor=ation to operators 'and other personnel and the incorporation of such in-for=ation into training and retraining programs.
Response
The operating experience assess =ent function vill be perfor ed by the plant's systes performance group which is composed of super-visory, engineering and technical-personnel. This group is not functionally a part of the plant operations group. The systers performance group is a multi-disciplined group which has overviev of all plant systems including mechanical, electrical and instru-mentation and control. This group is dedicated to the operating experience assessment function which includes but is not limited to the following:
A.
Engineering evaluation of the operating history of the plant (equipment failures, design proble=s, operations errors, etc.)
and Licensee Event Reports from other plants of similar design, with suitable dissemination of the results of such evaluaticas to other members of the plant -staff; B.
Engineering evaluation of the adequacy of the policy for maintenance, testing, equipment procurement, etc.;
C.
' Engineering evaluation of continuing adequacy of plant operations quality assurance; and D.
Engineering evaluation of adequacy of plant emergency and operating procedures.
'Recuirement (2) Identify the administrative and technical review steps necessary in translating reco==endations by the operating experience assessment group into plant actions (e.g., changes to procedures; operating orders).
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Response
Steps for implementing recommended procedural changes and changes in plant design are outlined in plant procedures.
Reports of significant events frem other plants, vendor notifications, and regulatory notifications are reviewed and evaluated in accordance with appropriate plant procedures.
Requirement
.(3) Identify the recipients of various categories of information from operating experience (e.g., Supervisory personnel, STA's, operators, maintenance personnel, H.P. technicians) or other-vise provide means through which such information can be readily related to the job functions of the recipients.
Response
Conduct of operations procedures identifies plant superintendents /
supervisors an the recipients of pertinent information. The functional group superintendents / supervisors vill distribute pertinent information as required.
Requirement (h) Frovide mesns to assure that affected personnel beccme aware of and understand information of sufficient importance that should i ---
not wait fer emphasis through routine training and retraining progrims.
Response
The Conduct of Operations Procedures for appropriate functional groups vill be revised as necessary to assure, timely distribution of information.
Requirement (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency.
Response
Plant procedures provide for screening of extraneous and unimportant information. This screening function is performed by the Systems Performance and Planning Group.
Requirement (6) Provide suitable checks to assure that conflicting or contradictory information is not co veyed to operators and other personnel until resolution is reached.
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Ee$pon3e Oh.ecks are perfo'rmed by both the Sy t200/Perfor:1nce group and by the indiviinal group superintendent /superviser to ensure th : conflicting or contradictory inforcation la not conveyed to plant personnel.
Requirement (7) Provide periodic. internal audit - to assure that the feedbach
- rogram functions effectively at all levels.
Response -
s Plant procedures will be revised to provide for periodic internal
. audits to ensure that the feedback program is functioning effectively.
Farley nuclear Plant Licensee Event Reports (LERs) and significant IJ3hs of other plants are presently being routed for STA review and signature in accordance with existing plant procedures.
In addition, when personnel performing operational assessment conclude that informatica exists which may be relative 'to the function of the STA, such infor=ation will be issued to the STAS.
Procedures for feedback of operating axperience vill be fully
- implemented by January 1, 1961.
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'TI.K.3.1 INSTALLATION AND TESTING OF AUTCMATIC FORV ISOLATION SYSTEd' POSITION
- (a) All PWR licensees should provide a system which uses the PORY block valve to protect.against a small break LCCA. This system vill automatically cause the block valve to close when the.
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reactor coolant system pressure decays after the PCRY has opened,
.to relieve excess pressure.-- An override feature shculd be incorporated. Justification should be provided to assure that failure of this system would not decrease overall safety by intensifying plant transients and accidents.
The proposed design should be submitted-for approval by July l', 1981.
(b)' Each licenace should perform a confirmatory test of the auto =atic block valve closure system installed in response to (a) above by the end'of the first refueling cycle after installation of this system.
Resnonse The Westinghouse Owners Group has performed an analysis of the
. ultimate heat sink function and the decreased intensity of a number' of plant transients, assuming the proper operation of the PORV(S) (e.g.. opening). This analysis shovs that failure a---
of the proposed automatic PORV isolation cystes could impeir this function. In addition, the plant modifications, procedure chc.nges, and' operator training (e.g., UUEZ3-0573 requirements) provide assurance that the function of the tutomatic isolation system is provided by operator action.
Failure to isolate
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stuck open PORV(S) has been analyzed and results in no core uncovery; therefore, Alabama Power does not propose to make any design changes.
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II.. 3. 2 F.!R E'OO3 RE70'T :'" ?^3'*
FAILUF3 RCUOTIG:I PCSITIC'I (a) Each' PWR vendor should submit a report tj January 1,1981, for staff review documenting the various actions vnich have been taken to decrease the probability of a small break LOCA caused by a stuck-open PORY and show hov they constitute sufficient improvements in reactor safety. This report should be submitt ed for staff review.
(b) Safety valve failure rate based on past history of the vendor designed operating plants should be included in the report submitted in response to (a) above.
Response
A report on PORV failure reduction vill be submitted to the NEC by Jar.uary 1, 1981.
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-II.K.3.3 REPORTING SAFETY AND RELIEF
- VALVE FAILURES AND CHALLE*iGES POSITION (a) Future failures of a relief valve to close should be reported promptly to the NRC.
'(b) Future challenges to the. relief valves should be documented in the annual report.
(c). Future failures of a safety valve o close should be reported promptly.to the NRC.
(d) Future challenges to the safety valves should be docu ented in the annual report.
(e) Licensee to provide annual report on SRV and RV failures and challenges as of April 1, 1980.
Response
Alabama Po.rer Company will provide as a part of its Annual Report a list of all Steam Generator and Pressurize-SRV and RV failures and challenges, and ec=mits to notifyin6 the IGC promptly.upon any failure of a PORV or safety valve to close j
on the Steam Generators or Pressurizer.
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II.K.3.5 AUTOVATIC TRIP OF REACTOR COOLANT P'JMPS DURING LOCA PCSITICH Tripping of the reactor coolang pu=ps in case of a LOCA is not an ideal solution. The licensees should con:ider other colutions to the. c=all break LOCA probles (for example, an increase in safety injection flow rate).
In t meanti=e, until a better solution is
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found, the reactor coolant peeps should be tripped automatically in case of a small break LOCA. The signals designated to initiate the pump trip should be carefully selected in order to differentiate between a small break LOCA and other events which do not require reactor coolant pump trip as discussed in NURIG-0623 Licensee to provide results of evaluation of alternate solution to reactor coolant pump trips to staff by Janusry 1, 1981, with any resulting modifications to be implemented by January 1,1982.
Resoonse The Westinghouse Owners Group analy?is of delayed RCP trip during small break LOCAs is documented in WCAP-958h. This i
WCAP is the basis for the Westinghouse Cwners Group position l
on RCP trip (i.e., autecatic RCP trip is not necessary for a l
Westinghouse PWR since sufficient tinc is nynilible f r =anual i---
tripping of the RCPs). This philcsophy has been incorporated in the Westinghouse Energency Operating Instructions which were reviewed and approved by the NRC Sulle: Ins and Orders Task Force and subsequently incorporated in the Farley Unit 1 Emergency Operating Procedures.
In addition, the Westinghouse criteria (basically a RCS. pressure below the shutoff head of SI pumps) provides for continued RCP operation and, therefore, j
forced circulation and decreased reliance on operator action for non-LOCA events. As requested by the HRC in a letter dated l -
April 15, 1980, ard as discussed with the NRC during the May 22, l
1980, meeting on this subject, we anticipate that the -
Westinghouse Owners Group will provide predictions of the LOFT
. test L3-6 by January 1,1981. The NRC has indicated that s=all break tests at the Seniscale and LOFT facilities as well as Owners Group test predictions will aid in NRC resolution of this issue.. Therefore, we believe that it is not appropriate
.to take any additional actions on this issue until the results of the URC sponsored testing (in particular L3-5 and L3-6) and Owners Group prediccions are completed and the results evaluated.
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II.K.3 9 PROPORTIONAL INTEGRAL DERIVATIVE (PID) CONTROLLER MODIFICATION POSITICII The Westinghouse-recc nended modification to the Propertional Integral Derivative (PID) controller should be implemented by affected licensees.
The Licensee is to L=plement actions and submit docunentation of the method for staff review by July 1, 1980.
Response
Upon recot=endation of Vestinghouse, Alabsma Power Company has incorporated into the Farley Unit 1 design, " rate-tine constant" in the PID Controller of zero seconds. This, in effect, recovea the derivative action from the controller which decreased the likelihood of opening the PORV since the actuation (opening) signal vill not be sensitive to the rate of change of the pressurizar prescure.
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II.K.3.10 PROPOSED AUTICIPATORY TRIP '<0DIFICATICU
. POSITION The anticipatory trip modification proposed by some licensees to confine the range of use to high power letels should not be made until it' has been shown on a plant-by-plant basis.that the small, break LOCA probability resulting from a s uck-open power-operated l
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l relief valve (POR/)'is little affected by the modification.
Response
The licensing basis for Farley Nuclear Plant includes an antici-patory reactcr_ trip upcn turbine trip which i: ty;acced at pover levels of 50 percent or less. The Westinghouse design criterion
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is that load rejections up to 50 percent should not require a reactor trip if all other functions operate properly. The power micmatch is taken up by the h0 percent steam dump and automatic rod incertion (10 percent). The PORV's are provided to reduce the likelihood of tripping the reactor on the high pressuricer pressure cign:1 and cpaning the pres:urizer safety l
valves. A series of 50 percent loai rejections from full power I
performed by Westinghouse (WCAP-81h2) indicate an increase it reactor coolant pressure of le'ss than 100 psi. A turbine trip at half-power without immediate reactor trip results in a peak pressuricer pressure of 232k psia, or 26 psi below the PORY opening setpoint. Therefore, the p0RV's would not be expected to open and_.the probability of a s-a1' break LOCA due to a stuck-open PORY is not significantly changed by the modification.
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II.K.3.12 CONFIR*4 E7.ISTENCE OF ANTICIPATORY TRI? UFON TURBINE TRIP PO3ITION Licensees with U-de:igned operating plants should confirm that their plants have an anticipatory reactor trip on turbine trip.
The licensee of any plant where this trip is not present should provide a conceptual design and evaluation for the installation of this trip by July 1, 1980.
Re:ronse The Farley Nuclear Plant has an anticipatory reactor trip on turbine trip.
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,3 II.K.3 17 - PIPORT ON CUTAGE OF ECC Sh.e mICENSEE REPORT AUD PROPOSED TECHNICAL SPECIFICAC:i CHANGES POSITION Several components'of the ECC systems are permitted by Technical Specifications to have substantial outags simes (e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> fcr one diesel-generator; 14 days for the E?;; system). In addition, there are no cumulative outage time limitations for ECC systems.
Licensees should submit a report by January 1, 1981, detailing outage dates and lengths of outages for all ECC systems for the last five
-years of operatida.
IL: report should also include the causes of the outages fe.g., controller failure, spurious isolation).
Response
Alabama Power Conpany will provide a report by January 1,1931, detailing outage dates, lengths. ',2d causes for all ECC systems
- that occurred during modes of operation for which the equipment is necessary in accordance with the' Technical Specifications.
It should be noted that no accurate records exist for the duration of outages. during modes of operation in which this equipment is not necessary in accordance with the Technical
- Specifications; therefore, the data vould be misleading for any time base. study.
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T"50L ?.00:4 HABI'"A?!MT7 S.NJ:???*EUTS POS:OIOU In accordance with action tv--
III.D.3.L, Control Room Habitability,
~ licensees shall assure that control rcon cperators will be adequately protected against the effects of accidental release of toxic and l ' ~~
rsdioactive gases and that the nuclear power plant can be safely operated or shut dovn under design basis a:cident conditions (Criterion 19, " Control Room," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50).
All facilities that have not been reviewei for confor=ance with the following sections of the Standard 2eview Plan:
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- 2. 2.1-2. 2. 2 Identification of Potention Hazard in Site Vicinity; 2.2.3 Evaluation of Potential Accidents; 6.h Habitability Systens; shall perform the necessary evaluations and reco= mend appropriate modifications to meet control room habitability requirements. The following documents may be used for guidance in performing the required evaluations:
P 1.
Regulatory Guide 1.73, " Assumption: cf Evaluating the Habitability of a nuclear Power Plant Control Room During a Postulated Hazardous Chamical Release."
2.
Regulatory Guide 195, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release."
3 K. G. Murphy and K. M. Campe, " nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19," 13th AEC Air Cleaning Conference, August, 1974.
The licensee's submittal shall include the results of the analyses of control room concentrations from postulated accidental release of
. toxic' gases and control room operator radiation exposures from airborne radioactive material and direct radiation resulting from design basis
- accidents... The toxic gas accident analysis should be perofrmed for all potential hazardous chenical. releases occurring. either on' the ' site or-within five miles of plant site boundary. Regulatory Guide 1.78
' lists the chemicals most commonly encountered in the evaluation of control room habitability but is not all-inclusive.
The-DBA radiation source term should be for the LOCA containment'
/ leakage 'and ESP leakage contribution cutside contain=ent as described w
i'n Appendix A and 3 of Standard Reviev Plan Chapter 15.6.5.
In additien, BWR facility evaluaniens'should add any leakage from the main steam isolation valves (e.g., valve stem leakage, valve seat
-leakage, main steam isolation valve leakage control system release) to the containnent leakage and E5F leakage _following a LOCA. Other DBA's should be reviewed to determine whether they cight constitute a more severe control rcom hazard than the LOCA.
In addition to the accident analysis results which should either
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identify the possible need for control roon modifications or provide assurance that the habitability syste=s vill operate under all postulated conditions to permit the control room operators to remain in the control room to take appropriate actions as required by General Design Criteria 19, the licensee should' submit sufficient information needed for an independent evaluation of the adequacy of the habitability systems. Attach =ent 1 lists the information that should be provided along with the licensee's evaluation.
Licensees should submit their responses to this request on or before January 1, 1981. Modifications needed for compliance with the control room habitability requirements specified in this letter should be identified and a schedule for completion of the modifi-cations should be provided.
Implementation of such modifications should be started without awaiting for the results of the staff's review. Additional needed modifications, if any, identified by the staff during its review vill be specified to licensees by July 1931. All =cdifications must be scheduled for completion by
'r-January 1, 1983.
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Response
J The control room habitability systems are designed to provide maximun safety and co= fort for operating personnel during normal operations and during postulated accident conditions. These habitability systems for the control room include shielding, charcoal filter systems, heating, ventilation and air conditioning, storage capacity of food and water, kitchen, sanitary facilities, and fire protection.
The control room habitability systems are designed to meet ABC General Design Criterion 19 Sufficient shielding and ventilation are provided to permit occupancy of the control room for a period of 30 days following a design basis accident (DBA) without receiving '
more than 5 rem whole body dose or its equivalent to any part of the body.
Design Bases The following design bases were used to determine the functional design of the habitability systems:
1.-
The postulated accident that determines'the habitability-design requirements is the Design Basis LOCA.
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2.
The assumptions regarding the sources and amounts of radio-activity that could pose a hazard to the control rec = are
-discussed.in Section 12.1 3 of the FSAR.
3 In the event of an accident, the ventilation system in the control room vill be triggered.by the
"!" signal which will automatically isolate the normal air systens and start both trains of the control RM A/C system, pressurication system, and filteration eystem. The control roca air vill thus be filtered through a HEPA charcoal filter system which is 1 --
capable'of removing 99 9 percent of both the inorganic and organic iodine. On a long-term basis (eight-day continuous flow) the filter. efficiency for both organic and inorsanic iodine is 99.0 percent.
The filter system and the control roc shielding cro capable of keeping the dose to the operators less than 5 rem whole body dose or.its equivalent to any pary of the body for the duratior,of the accident.
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Following postulated accidents, the most severe being a LOCA, the licitations on control room pressure, temperature, radioactivity concentrationc, deses, and ::::2ntrations of carbon dioxide are as follows:
Parameter Allawable Control Room Pressure 1/6" w.g.'
Control Room Temperature 1200 F Radioactivity Concentrations As stated in 10CFR20, Appendix B Doses ~
5 rem Concentrations of carbon dioxide 1 wt percent 5. The fire protection system in the control' room consists of an early warning ionization-type detection system with hand portable water extinguishers located in the control room itself. The ionization detectors used vill detect products of combustion in approximately four seconds. Detectors are located on the false ceiling to detect s=oke in the control rooe itself. Each detector is equipped with a light to indicate which detector has operated. There is a subpanel
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that indicates which detector has operated in spaces not readily. visible. -All detectors vill operate the visible and
- audible alarm on the main fire protection annunciator panel located in the control' room.
In addition, a fixed carbon dioxide hose: reel with 100 feet of hose is Jocated in the L= mediate vicinity outside the control room and can be used to back.up the hand. extinguishers.
Non-combustible materials are used in construction and equipment as much as possible..The quantity of ccmbustible.
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i material such as paper and other fla=mable supplies are kept to a minimum. A persor. trained in fire fighting vill be on duty-in the control room at all times.
CONTROL ROOM HVAC Design Bases The control room air conditioning and filtration system is designed with sufficient redundancy and separation of components to provide reliable operation under normal conditions and to ensure operation under emergency conditions.
The system is designed to provide an environment with controlled temper-ature and humidity to ensure both the comfort and safety of the operators and the integrity of the control room components. Design ambient conditions are approximately Th F and 50 percent relative humidity, taken from ASHRAE Standard 55-66, " Thermal Comfort Conditions."
Provisions are made in the system to detect and limit the introduction of airborne radioactive material into the control room.
Provisions are also made in the system for the remov.
of radioactive and foreign material from the control room environment.
The system is designed to per=it periodic inspection of the principal system components.
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System Descriution During normal plant, operation, one (1) of the two (2) 10' percent capacity, Category I air conditioning packaged units recirculates 21,000 cfm of cooled filtered air through the control room.
Each packaged unit consists of a recirculation fan, prefilter, refrigeration unit with a water-cooled condenser, cooling coil, and the associated instrumentation and controls. Should the operating unit fail to function in seme way, the second 100-percent-capacity unit can be manually started by the operator.
Two full-capacity, redundant, seismic Category I air pressurization systems are provided to maintain the control rocm at a positive pressure post-LOCA. Esch train -is capable of supplying 300 cfm through electric heating coils, prefilter, HEPA filter, and 6-in.
- deep bed charcoal filters, designed in accordance with the requirements of Regulatory Guide 1 52.
Installed in parallel to the suction side of each control room main air conditioning unit are 1000 cfm filtration units consisting of prefilters, HEPA filters, and 2-in. charcoal filters, and a 2000 dfm filtration unit' incorporating the sa=e composite filter elements as the 1000 cfm units.
Therefore, the overall recirculation filtration capacity is 3000 ces for each packaged A/C unit.
During normal operation, an air supply system delivers fresh outside air to the control room and to the computer room. The air supplied A 4
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to tne contrca room by this system maintains the control room at a slignt positive pressure, thereby preventing the intrcduction of air into the control room frca sources other than the design fresh air make-up system.
A smoke detector near the return-air duct to each recirculation fan vill sound an' alarm in the control rocs on high smcke evel.
If necessary, the operator can exhaust air from the cantrol room by manually opening the motor-operated exhaust damper and starting one(l) sof the two (2) 100-percent-capacity exhaust fans. An area radiation monitoring sy1 tem and redundant control room charcoal filter re-circulation systems are provided to detect and reduce radiat'on levels in the control room. The filters are composite units and have been furnishe to the same specifications as the filters for the penetration rocm filtration system. An area radiation monitor located in the control room alarms on high radiation level and alerts the operator. to the need for filtration of recirculated air.
Redundant Category I chlorine detectors and process radiation monitors are provided at the control room normal fresh air intake. Redundant Category I smoke detectors are provided in the fresh air intake and return air duct of the computer room air handling unit. The air-operated isolc ion damper en the conputer room recirculation line is interlocked with the computer rocm fire detectors, and vill auto-natically cle:e in the event of smoke detection. Closing of this damper will a'.co initiate cicuure of the computer rocs outside air supply line 1,olation vnive and shutdown of the cenputer roce air conditiening system.
The cc puter roem fire detectors are also the means for actaating the h?ler rele2ce into this area. The halon storage tank is located c;:alaa he ccmputer roo=.
Inadvertent release of halon is discussed below. Tripping any one of the detectors vill cause an alarm to sound in the control room and the closing of all air-operated and motor-operated isolation valves in the non-engineered safety features hVAC penetratin,g the control room boundary, thereby isolating the control room. If required, and if contaminants are within safe levels, the operator can drav outside air by manually starting the emergency pressurization system, in which the air is filtered through deep bed filters.
During post-LOCA conditions, the normal =akeup air is cut off and c ?1 control rocs isolation valves are closed. In this event, the positi e pressure in tae control room is maintained by the autematic.stamm af one of the. emergency pressurization syste=s, each consisting of an air inlet, an isolation valve, and a 300-cfm, deep-bed, charcoal filtration unit. In like manner, the standby filtration units vill automatically recirculate 3000 cfm out of the 21,000 cfd total room recirculation flow rate through charcoal filters.
All penetrations into the control room are sealed to eltninate inleakage of outside. air. Mechanical penetrations az e sealed by use of silicone-rubber foam or by fiberglass impregnated boo'.s.
Electrical penetrations
are sealed with silicone-rubber foam.
EVAC system duct penetrations are provided with air-tight automatic butterfly valves.
Safety' Evaluation The redundant system has been designed to provide mini =um filtering
.and ventilation, and ensures that no single failure vill prevent the safe occup'ancy of the control room under any mode of plant operation.
Power for the fan and refrigerant compressors of each air conditioning packaged unit is supplied from redundant emergency power supplies. A separate, independent system of distribution ducts is installed for each unit.
Cooling water for the condensers of the air conditioning packaged units is supplied from redundant headers in the service water system.
The duct from the computer room air conditioning unit that normally supplies fresh makeup air to the control room contains two pneumatic-operated valves in series. A single pneumatic-operated isolation valve is provided for each duct connected to the smoke purge exhsust system.
Two pneumatic-operated valves in series are provided for the utility exhaust subsystem. These valves are povered from redundant power supplies and close automatically on a containment isolation signal to isolate the control room.*-on outside air.
The control room habitability is maintainea 'c:e continually monitoring radiation levels and smoke concentration inside the room, plus con-tinually =onitoring radiation levels, including monitoring of smoke concentration in the control room air intake duct and computer" room return duct, and chlorine concentration at the control room air intake. To prevent inleakage, the control room is provided with normal and emergency pressurization systems designed to maintain greater than 1/8 in. v.g. positive pressure.
Upon toxic gas (ch1crine or smoke) detection, an alar = is annuncieted in the control room and all failsafe, airtight isolation valves are closed auto =atically and remain closed until they are reopened manually.
The normal makeup air is cut off.
In the case of smoke detection, the operator can manually start the exhaust fan to purge smoke; in the case of chlorine detection, the operator can make use of self-contained breathing apparatus. After a safe level of toxic gas concentration is reached, the operator can draw outside air either from the normal makeup : air subsystem or from the emergency (pressurization systen) makeup air subsystem.
Upon a high radiation signal from the rakeup air inlet an alarm is sounded in the control room and all isvim ion valves are closed automatically and remain closed until they are reopened manually.
This will result in a loss of positive pressure in the control room.
In this event, the operator vill manually start one of the redundant pressurization systems 'and one of the redundant recirculation fil-tration systems (one 2000 cfm and one 1000 cfm system). The HEPA and 6-in. deep bed charcoal filter unit in the pressurization system and the HEPA and 2-in.. charcoal filters in the recirculation system provide additional assurance that the doso receiv$d by control roca personnel vill not exceed the guidelines of Criterion 19.....2.-...
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' Upon receipt of a containment isolation signal, the control room 10 autcastically isolated as described above. The air pressurication cystem and recirculation system are autcastically actuated to main-tain the pocitive pressure and to provide control roc cleanup, respectively. A flow control, automatic camper =cunted on the bleed-off les of the pressurization system will respond auto:stically to the pressure controller in the control room.
If the positive pressure drops-belov 0.25 in. v.g., the bleed-off damper will automatically restrict the bleed-off flow to divert taxi =um air supply into the room, in order to achieve a rapid pressure buildup. On a rising room pressure, the bleed-off damper vill function in the opposite
=anner. Therefore, the control room pressure vill be sensed and naintained, as well := exhibited for operator infceration.
Upon receipt o' a s=oke detection signal from the cc puter room scoke detecto::, the ccmputer roo= HVAC is automatically isolated.
In addition, redundant seismic Category I smoke detectors down-
. stream of ti e return air subsystem from the computer room vill automatically isolate redundant seismic Category I isolation valves in the computer room recirculation line, in the event of smoke re-circulation follcuing a ec=puter roo= fire.
Radiatica monitors are provided within the control rocm boundary.
Radiation menitors are also provided within each of the various ventilation syste=3 serving all radiation release points in the plant. Ihese cocitors provide indication in the control rocs, and alarm whenever predetermined radiation levels are exceeded. These I ~~~
HVAC systems discharge through the plant vent stack. Additional ralie:lan a;;itors are provided at the vent etack discharge which vill provide a back-up means of detecting abnormal plant releases.
These =0nitors are designed to detect releases in excess of the MFC Guidelines established under 10 CFR 20, Appendix B.
Based on the availability and sensitivity of the monitoring systens provided, the operator vill have adequate indication and information to evaluate the magnitude of any abnormal plant releases, and vill canually isolate the control roo= if required.
An analysis of dose levels in the control room under accident conditions is presented in the applicable sections of FSAR Chapter 15.0.
Redundant chlorine detectors vill be provided in the circulating water chlorination house designed to transmit an alarm in the control room if a chlorine spill should occur. These alarms will provide sufficient time for operators to put on self-contained breathing apparatus.
An analysis of chlorine release accident has been performed. Because of the proximity of the closest circulating water chlorination house the analysis was performed for the release of two tons of chlorine (the maximum amount _of chlor.ine headered together at one time).
~ Twenty-five percent of the chlorine was assumed to flash to gas.
This is analyzed as. a puff release.
The remainder is assumed to form a 200 sq.~ ft. pool where it evaporates due to the heat load from the sun and from ambient air and ground temperature.
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No credit is taken for the channeling of the dense chlorine gas around buildings and along ditches. No credit is taken for an elevated air intake, even though the intake is at an elevation of approxicately 177 ft.
To evaluate the control room habitability, considering the closest circulating chlorination house, the following cases have been analyzed using the techniques outlined in References 1, 2, and 3 to Section 9 1.h.5:
A. -'Two-ton chlorine spill, h50 ft. from control room air intake, 0 5 =/see vind, Pasquill Class F meteorology shared Unit.1 and Unit 2 control room, 70 cfm unfiltered inleakage and no credit for building vake effect.
b.
The same as Case A except a vind speed of 1.0 m/sec is used.
C.
The sa=e as Case A except building vake effect is considered.
D.
The same as Case B except building vake effect is considered.
The time (after chlorine accident) required to reach 15, 30, h5, 60 and maximum enlorine concentration in ppm by volume inside the control room is given in FSAR Table 9.4-9 This table also shows the peak chlorine concentration. The analysis assumed a 5-sec detector response time from 5 pps setpoint (chlorine detector at air intake), a 5-see transport' time of chlorinated air from air intake to the isolation L ---
valve, and a 6-sec closing time of the isolation valve.
In Case 3, the operators vill have over 2 minutes to put on self-contained breathing apparatus before a concentration of 15 ppm is reached in the control room, after allowing 5 seconds detection time at 5 ppm for the chlorine detectors at the chlorination house. This is in accordance with Regulatory Guide 1.78.
An inleakage rate of 70 cfm is used based on 0.06 air exchange per hour for Type A control room defined in Regulatory Guide 1.78.
Self-contained breathing apparatus with a minimum 30 minutes air supply
_'are stored in the control room. Sufficient apparatus is provided to allow one spare per three units required. An additional six hour air supply is provided in the auxiliary building as close as practical to the control room. An offsite source of air vill also be available.
An analysis of the Halon 1301 concentration in the control room following an accidental spill in the computer room, without fire, has been performed. It is. assumed that the total enount of Halon 1301' i
flooding the ce=puter room mixes with the air in the computer room.
This yields a maximum initial concentration of 6 percent by volume
~in'the co=puter roca. Using the maxi =um control room makeup air quantity of 1650 cfm, the maximum concentration of Halon 1301 in the control room is.h7 percent by volu=e after 15 minutes. This is' well below the 7 percent maximum reco== ended concentration for fnormally occupied areas of UFPA Standard 12A.
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k Instrumentation Cne train of the control' room air ecnditioning system is operating during normal conditions snd both traine are automatically started daring post-LOCA conditions. Both trains of the control rocm re-circulation filtration and emergency pressurination systems are
. automatically started upon receipt of a CIAS cignal.
SNEe following are displayed 'and/or located in the :ontrol room:
~
a.
Control-room air temperature
- b. ' Air. intake chlorine concentration indication c.
Air intake smoke concentration indication d.- Air intake radiation level e.
Room smoke detectors f.
Alarm for high differential pressure across each filter train g.
Dirrerential pressure between the control recs and atmosphere Positive indication of all isolation valves and dampers is locally displayed. Differential pressure across each filter train and fan is locally displayed in the control room mechanical equipment room.
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Control Room Shielding Control room shielding design is based on the requirements set forth in 10 CFR 50, Appendix A, Criterion 19, which requires occupancy of and access to tne control ~ room under accident conditions. The dose to personnel vill-be Ibnited to 5 rem vnole body or its equivalent
- to any part of the body for tne duration of the accident.
The accident anslysis in FSAR Chapter 15 indicated that this dose to personnel vill be less.that 5 rem.
Protection of control room personnel from the fission-product release in the containment is provided by the concrete valls between them.
Storage Capacity of Food and Water There vill be sufficient storage capacity (including a food freezer)
.for-two shifts of-operators for 30 days. Water is drawn from the Potable and Sanitary. Water. System.
Kitchen Kitchen facilities: are1available for refrigeration, cooking, and the cleaning' of cooking and' esting utensils.
Sanitary Facilities Bathroom = facilities' are provided adjoining the control rocm.
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Flich Radiation Level Alarm An area radiation monitor located in the ccatrol room alarms on high radiatiori level and alerts the operator to the need f or filtration of recirculated air and isolation of the frech air cupply du:t.
t The monitor is capable of reading in the range of 10-4 R/hr to 10F/hr.
I The control room habitability will be reviewed for conformance with the Standard Review Plan (Sections 2. 2.1-2. 2. 2, 2. 2. 3, 6. 4 ) by January 1, 1981.
If any modifications are required as a result of the above review, a schedule vill be provided by January 1,1981.
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y N.K.3.30 REVISED SMALL-EREAK LCCA '2THODS TO SHO'4 COMPLIAIICS '4ITH 10 CFR 50, APPE:iDIX K POSITION The analysis methods used by USSS vendors and/or fuel suppliers
'for small break LOCA analysis for compliv.ce w th A;pendix K.to i
10 CFR Part 50 should be revised, documented, and submitted for URC approval by January 1,1982. The revisions shculd account for comparisons with experimental data, including data from the LOFT and Semiscale facilities._
' Response The present Westinghouse small break evaluation codel used to analyze Farley Uuclear Plant is in conformance with 10 CFR Part 50, Appendix K.
However, Westinghouse has indicated that they vill, nevertheless, address the specific NRC items con-tained in NUREG-0611 in a model change scheduled for completion by January 1, 1982.
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II.K.3 31 PL AMT SPECIFIC CALCULATIONS TO SHOW CO:GLI!C'CE UITH 10 CFR 50.Mi POSITICU Plant-specific calculations using UEC-c; proved models for small break LCCAs as described in II.K.3.31 above, t: sh: cenpliance with 10 CFR 50.h6 should be sube.itted for ::RC approval by all licenceec by January 1,1983, or one year after staff approval of LOCA analysis model.
Bes;cnce The present Westinghouse small break evaluation model and s=all break LCCA anal sa f:r 7arley '? nit 1.rc '- - nformance with
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10 CTR Part 50, Appendix K and 10 CFR Part 50.h6. As stated in the response to ite: II..3 30, 'Jectinghouce plans to subnit a new small break evaluation model to the I;BC for review by January 1, 1982.
If the results of this new Westinghouse model (and subsequent ITRC reviev and approval) indicate that the present small break LOCA analysis for Farley Unit 1 are not in conformance with 10 CFR Part 50.k6, a new analysis utilizing the new and approved Westinghouse model vill be subnitted to the URC in accordance with the IIRC schedule.
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v v-ATTACH'T."P 1 Information Required for Control Roc = Habitability Ivaluation.
1.
Control Room Mode of Operation, i.e., pressuri:Stien and filter re-circulation for radiological accident isolation or chlorine release 2.
Control Room Characteristics a.
air volume of control room b.
control room emergency zone (control room, critical files, kitchen, washroom, computer room, etc.!
control room ventilation systc= schematic with normal and emergency c.
air flow rates d.
infiltration leakage rate e.
HEPA filter and charcoal adsorber efficiencies f.
closest distance between containment and air intake g.
layout of control room, air intakes, containment building, and chlorine or other chemical storage facility with dimensions h.
control room shielding including radiation streaming from penetrations, doors, ducts, stairways, etc.
- i. automatic isolation capability-damper closing time, damper icakago and area J.
chlorine detectors or toxic gas (local or remote) k.
self-containei tre,. thin; apparatu: avail?.bility (number) 1.
bottled air supply (hours supply) emergency food and potable water supply (how many days and hev many n.
people) control room personnel capacity (normal and emergency) n.
o.
potassium iodide drug supply 3.
On-site storage of chlorine and other hazardous chemicals a.
total a=ount and size of container b.
closest distance from control room air intake k.
Off-site manufacturing, storage or transportation facilities of hazardous chemicals a.
identify facilities within a five-nile radius b.
distance from control room c.
quantity of hazardous chemicals in one container d.
frequency of hazardous chemical transportation traffic (truck, rail, and barge) 5 Technical Specifications (refer to standard technical specifications) a.
chlorine detection system l
b.
control room emergency filtration system including gross leakage I
determination by control room pressurization at 1.8 inch water gage, verification of isolation by test signals and damper closure times and filter testing requirements -.
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