ML20004C841
| ML20004C841 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/29/1981 |
| From: | Colbert W DETROIT EDISON CO. |
| To: | Kintner L Office of Nuclear Reactor Regulation |
| References | |
| EF2-53-421, NUDOCS 8106050479 | |
| Download: ML20004C841 (28) | |
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cMigan 48226 b %,ea V (313) 237-9000 May 29, 1981 EF2 - 53,421 L
Mr. L. L. Kintner
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Division of Project Management c,i S 4
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U. S. Nuclear Regulatory Commission O ~ fj
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,s Decr Mr. Kintner:
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Referencc:
Enrico Fermi Atomic Pcwer Plant, Unit 2 NRC Docket No. 50-341
Subject:
Miscellaneous NRC Requests for Additional'Information Please find attached one copy each of Detroit Edison's responses to several NRC requests.
Item 1 LQB I.A.1.1 Shift Technical Advisors Detroit Edison's response is detailed in Attachment 1.
This information will be included in a forthcoming FSAR amendment.
Item 2 LQB I.A.1.3 Shift Manning Detroit Edison's response is detailed in Attacment 2.
This information-will be included in a forthcoming FSAR amendment.
Item 3 RSB II.K.1.5 Safety Related Valve Position Detroit Edison's response is detailed in Attachment 3.
This information will be included in a forthcoming k
FSAR amendment.
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THIS DOCUMENT CONTAINS tp P00R QUAL.lTY PAGES 8.108 os n y7q
Mr. L. L. Kintner May 29, 1981 EF2 - 53,421 Page 2 Item 4 RSB II.K.1.10 Procedures'for Removing Safety Related Equipment from Service Detroit Edison's response is detailed in Attachment 4.
This information will be included in a forthcoming FSAR amendment.
Item 5 RSB II.K.1.22 Procedures for Auxiliary Heat-Removal Detroit Edison's response is detailed in Attachment 5.
This information will be included in a forthcoming FSAR amendment.
Item 6 RSB II.K.3.3 Reporting of SRV Challenges Detroit Edison's. response is detailed in Attachment 6.
This information will.be included in a forthcoming FSAR amendment.
Item 7 r,B II.K.3.17 Reporting of ECC Outages Detroit Edison's response is detailed in Attachment 7.
This information will be included in a forthcoming FSAR Amendment.
Item 8 RAB III.D.3.3 Improved Inplant' Iodine Instru-mentation Under Accident Conditions Detroit Edison's response is detailed inAttachment 8.
This information will be included in a forthcoming FSAR Amendment.
Sincerely, f A~.,
h William F. Colbert Technical Director Fermi 2 Project WFC/AEW:jl Attachments (8) e-gy
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H.I.A.l.1 Shift Technical Advisor
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H.I.A.l.l.1 Statement of Concern For an off-normal event in reactor operations, the shift j
supervisor's primary responsibility is the command and control function.
4 The other control room operators assume a manipula-tive, reactionary role in response to commands from the shift supervisor and in response to the various alarms and other indi-cators of plant conditions caused by the event.
Having reviewed the facts available on the-accident at TMI 2 and the general state of training and qualifications for present operators and senior operators, the Task Force has concluded that additional technical and analytical capability, dedicated to concern for the safety of the plant, needs to be provided in the control room to support the dia' gnosis of off-normal events and to advise I
the shift supervisor on actions to terminate or mitigate the l
consequences of such events.
H. I. A.l. l. 2 NRC Position Each licensee shall provide an onshift technical advisor to the shift supervisor.
The shift technical advisor may serve more than one unit at a multiunit site, if qualified to perform the advisor function for the various units.
(
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The shift technical advisor shall have a bachelor's degree or equivalent in a scientific c: engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents.
The shift technical advisor l
shall also receive training in plant design and layout, includ'-
ing the capabilities of instrumentation and controls in the
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control room.
The licensee shall assign normal duties to the shift technical advisors that pertain to the engineering aspects of ensuring the safe operation of the plant, including the review and evaluation of operating experience (Reference 1).
H.I.A.l.l.3 Detroit Edison Position Detroit Edison agrees with the intent of the NRC to upgrade the accident-assessment and the operational experience assessment.
and will :ncure that =adi*innal *ach nical and en:1" tic 21 capabbl-itics m;e pccvided ir the centrol recr for the diagnecie cf abner-mal event
'nd
- n =d"iee the operstors on act4ane-tc terminste Or Titig&te Lhc CG"caq"*"ce? cf such esenis.-
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, This will be provided by the assignment of either qualified shif t educ? tion 21 qu ll'(ficatienE on each sdTt.
technical advisor cr nuc1::: rhift ru cr"icers with cpgraded--
Shift tcchnical advicers "ill rve ir er adviscry capac4ty-te.
th; nucles: chift supeuvisuc.
L shift t echa i ca l-.adeisor-sha -
b="* = ba-heler'e deg:cc in engip : ring ce * *l a t ed -4ctence s, H.I.A.l.1-1 Amendment 33 - March 1981
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including 50 ::::stc: hcur: ef ::11ege-level at:::: tics, ccovius j
-physics, zesctoi therevdynamiva, oud wicuL6 lual ensinccc h19 W
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Upgr2dOd uclect sh i f t. superviSOCS, in addiLIGG t0 GGintainini
'e C # " I ^ r--C C e u lus operator'S 11 Cense, Shall advc o wscu...m 2
t collega-laual aanc=* inn nf 2hnur s;n comaeter heur; in mo lh;;;t _
irs : reacter physics, uceutes thcr.edyn2=ics, end clevLuival ~
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U.I.A.l.l.4 Reference
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U.S. Nuclear Regulatory Commission, Clarification of TMI-2,
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Action Plan Requirements, NUREG-0737, p.
l.A.l.1-1, October 1980.
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H.I.A.1.1-2 Amendment 33 - March 1981 ee
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H.I.A.l.3 Shift Manning H.I.A.I.3.1 Statement of Concern Complex transients in nuclear power plants place high demands on the operators in the control room.
The objective of the actions described in this task is to increase the capability of the shif t crews in the control room to operate the facility in a safe and competent manner by assuring that a proper number of individuals with the proper qualifications and fitness are on shift at all times.
The work to improve the design of control rooms is de-scribed elsewhere in this plan.
Studies indicate that with fatigue, especially because of loss of sleep, an individual's detection of visual signals deterior-ates markedly, the time it takes for a person to make a decision increases and more errors are made, and reading rates decrease.
Other studies show that fatigue results in personnel ignoring some signals because they develop their own subjective standards as to what is important, and as they become more fatigued they ignore more signals.
Inspections of personnel performance and training since the acci-dent at Three Mile Island have shown that in certain situations facility personnel are either required or allowed to remain on duty for extended periods of time.
Also, complaints have been received from some licensed nuclear power plant operators con-O cerning the number of continuous hours they have been on duty.
H.I.A.l.3.2 NRC Position Pending completion of the long-term development of criteria for shift staffing and administrative controls, the NRC staff has
- I developed interim criteria for the licensees of operating plants and applicants for operating licenses.
Except for senior reactor operators, these interim criteria for shift staffing shall remain as described in the Standard Review Plan, Section 13.1.2, NUREG-75/087.
Special requirements regarding the utilization and quali-fications of an onshift technical advisor to the shift supervisor t
have been previously provided.
l It is required that there be one licensed senior reactor operator in the control room at all times, other than during cold shutiown conditions.
This will therefore require that,there be a minimum of two senior reactor operators at each site at all times, other than during cold shutdown conditi'ons, to ensure the availability of one' senior reactor operator in the control room without affect-
,,ing the freedom of the shif t supervisor to move.about the site as needed.
The criteria for reactor and auxiliary operators are
' stated below, and the required staf fing levels for selected sta-tion configurations and various plant operating. modes are summa-rized in the enclosed table.
H.I.A.1.3-1 Amendment 33 March 1981 e
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EF-2-FS AR At any time a licensed nuclear unit is being operated in Modes (g) 1-1 for a PWR (Power Operation, Startup, Hot Standby, or Hot Shutdown, res pecti vely) or in Modes 1-3 for a BWR (Power Opera-tion, Startup, or Hot Shutdown, respectivcly), the minimum shift crew shall include two licensed senior reactor operators (SRO),
one of whom shall be designated as the shift supervisor; two licensed reactor operators (RO); and two unlicensed auxiliary operators (AO).
For a multiunit station, depending upon the sta-tion configuration, shift staffing may be adjusted to allow credit for licensed senior reactor operators (SRO) and licensed reactor operators (RO) to serve as relief operators on more than one unit; however, these individuals must be properly licensed on each such unit.
At all other times, for a unit loaded with fuel, the mini-mum shift crew shall include one shift supervisor who shall be a licensed senior reactor operator (SRO), one licensed reactor operator (RO), and one unlicensed auxiliary operator.
Adjunct requirements to the shift staffing criteria stated above are as follows:
a.
A. shift supervisor with a senior reactor operator's license, who is also a member of the station super-visory staff, shall be onsite at all times when at least one unit is loaded with fuel, b.
A licensed senior reactor operator (SRO) shall, at all
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times, be in the control room from which a reactor is bein,g operated.
The shift supervisor may from time to time act as relief operator for the licensed senior reactor operator assigned to the control room.
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c..' For any station with more than one reactor containing fuel, the number of licensed senior reactor operators onsite shall, at til times, be at least one more than
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the number of control rooms from which the reactors are being operated.
i d.
I. addition to the licensed senior reactor operators pecified in a, b, and c above, for each reactor con-taining fuel, a licensed reactor operator (RO) shall be in the control room at all times.
i e.
In addition to the operators specified in a, b, c, and d above, for each control room from which a reac-tor is being operated, an additional licensed reactor operator (RO) shall be onsite at all times and avail-able to serve as relief operator for that control room.
As noted above, this individual may serve as relief operator for each unit being operated from that con-trol room, provided he holds a current license for 4
each unit.
f.
Auxiliary (nonlicensed) Operators shall be properly lh qualified to support the unit to which assigned.
H.I.A.l.3-2 Amendment 33 Marca 1981 aw.
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9 In addition to the staffing requirements stated above, g/
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shift crew assignments during periods of core altera-tions shall include a licensed senior reactor operator (SRO) to directly supervise the core alterations.
This' licensed senior reactor operator may have fuel-handling duties but shall not have other concurrent operational duties.
Licensees of operating-plants and applicants for operating li-censes shall include in their administrative procedures (required by license conditions) provisions governing required shift staff-ing and movement of key individuals about the plant.-
These pro--
visions are required to ensure that qualified plant personnel to man the operational shifts are readily available in the event of an abnormal or emergency situation.
These administrative procedures shall also set forth a policy, the objective of which is to operate the plant with the required staff and develop working schedules such that overtime is avoided, to the extent practicable, for the plant staff who perform safety-related functions (e.g., senior reactor operators, reactor operators, health physicists, auxiliary operatccs, I&C technicians, and key maintenance personnel) (Referance 1).
The staff recognizes that there are diverse opinions on the amount of overtime that would be considered permissible and that there is a lack of hard data on the effects of overtime beyond the generally recognized normal 8-hour working day, the effects of shift rota-tion, and other factors.
NRC has initiated studies in this area.
Until a firmer basis is developed on working hours, the administra-tive procedures shall include as an interim measure the follow'ing E
guidance, which genera 11r follows that of IE Circular No. 80-02 (Refere,nce 1).
't In the event that overtime must be used (excluding extended pe-riods of shutdown for refueling, major maintenance, or major plant modifications), the following overtime restrictions should be followed:
a, An individual should not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not including shif t turnover time).
b.
There should be a break of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (which can include shift turnover time) between all work periods.
j c.
An individual should'not wor'k more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.
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d.
An individual should not be required to work more than 14 consecutive days without having 2 consecutive days l
off.
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H.I.A.1.3-3 Amendment 33 March 1981 ww
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However, recognizing that circumstances may arise requiring devi-ation from the above restrictions, such deviation shall be author-()
ized by the plant manager or his deputy, or higher levels of priate docu.entation of the cause.in accordance with published procedures and management g
j If a reactor operator or senior reactor operator has been working more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during' periods of extended shutdown duties away from the control board),
(e. g., at be assigned shift duty in the conttol room without atsuch individuals shall not least a 12-hour break preceding such an assignment.
.The NRC encourages the development of a staffing policy that would permit the licensed reactor operators and senior reactor operators i
to be periodically assigned to other duties away from the control board during their normal tours of duty.
4 If a reactor operator is required to work in excess of 8 contin-uous hours, he shall be periodically relieved of primary duties at the control board, such that periods of duty at the board do not exceed about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at a time.
The guidelines on overtime do not apply to the shift technical advisor provided he or she is provided sleeping accommodations and a 10-minute availability is assureds H.I.A.1.3.3 Detroit Edison Position (h
Minimum Shift Manning l
Detroit Edison is in agreement with the NRC pocition on minimub.
l shift manning requirements.
The minimum shift manning will be l
included in the technical specifications and administrative pro-
/ sa:Ett)cedures as stated in the NRC position.
y s4 Movement _of Shift Personnel and Overtime Detroit Edison is in agreement with the NRC position on the move-ment of key shift personnel about the plant and on the overtime restrictions.
These requirements will be set forth in Enrico
~Permi Orders and in the plant administrative procedures.
Detroit Edison is also in agreement that circumstances may arise, such as resignations, promotions, extended illnesses, and other uncontrollable factors, that may. create situations requiring ex-tended overtime beyond these guidelines.
These situations will be reviewed and approved by the plant superintendent or designee 1njaccordance with the administrative procedures.
fpe<rc~r s y
3 57-alification of Nonlicensed Ocerators Detroit Edison is in agreement with the NRC position on the quali-fication of nonlicensed operators.
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H.I.A.l.3-4 Amendment 33 March 1981 l
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Ins;rt 1 reg 2 i or 1 (H.I.A.1.3.3a)
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SHIFT MAfiftIf;G i
The minimum shift crew as specified in the Technical Specifications for the various operating conditions is:
Applicable Conditions License Category 1,2 & 3 4&5 Senior Operator Licenses 2
1*
Operator Licenses 2
1 Non-Licensed 2
1
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising core alterations.
The operating conditions are:
Condition Mode Switch Average Reactor Position Coolant Temoerature
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1.
Power Operation Run Any Temperature 2.
Startup Startup/ Hot Standby Any Temperature 3.
Hot Shutdown Shutdown Greater than 212 F 0
4.
Cold Shutdoun Shutdown Less than or equal to 2120F 5.
Refueling Shutdown or Refuel less than or equal to 2120F 7
As specified in FSAR Chapter 13, Section 1 the shift crew composition of operators for normal routine operations consists of the following:
Number Personnel
. NRC License Position Per Shift Requirement Nuclear Shift Supervisor 1
SR0 Nuclear Assistant Shift Supervisor 1
SR0 Nuclear Supervising Operators 2
R0 Nuclear Power Plant Operators 3'
None e,
Nuclear Assistant Power Plant Operators 2
None
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Insert 2' Page 1 of'1 (H.I.A.1.3.3 b)
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OVERTIME l'.::T :T: ':L l
The overtime ~ '- kti;n for operations personnel are as follows:
a) Operations personnel shall not be permitted to work more than twelve (12) hours straight (not including
' shift turnover time).
I b) There shall be at least a twelve hour break between all work periods (which can include s'hift turnover time).
c) Operations personnel hall. not work more than 72-hours i{
in any 7-day period.
d) Operations personnel shall not work more than fourteen g
(14) consecutive days without having two (2) consecutive days off.
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EF-2-FSAR The purpose of the training program for nonlicensed operators is to provide the necessary knowledge and training for the nonli-censed operators to perform their jobs efficiently and to ensure the reliability of plant systems and equipment.
The training program consists of two phases, as described below.
Area Qualifications.
Checklists are developed and established by plant area and by job classification to familiarize the operators with the specific job tests expected to be performed as part of the normal shift functions.
Area qualifications are based on the following plant areas:
a.
Turbine building b.
Reactor building c.
RHR complex d.
Auxiliary and service building e.
Radwaste building f.
General service water pumphouse g.
Circulating water pumphouse h.
Auxiliary boiler house 1.
Electrical power distribution j.
Fire protection Satisfactory completion of each checklist item by each operator is documented by a nuclear supervising operator, a nuclear assistant shift supervisor, or a nuclear shift supervisor.
O System Training.
System training increases the nonlicensed opera-tor's knowledge of the function and operation of plant systems.
It ensures the safety and reliability of plant operation as a result of the integrated activities performed by licensed and i
nonlicensed operators.
The objective of system training, which is provided in addition to area qualifications, is to give the
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nonlicensed operator a concept of the overall operation of the system, the purpose of systems, the interrelationships of sys-tems, and the operator's responsibilities relative to each system.
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~ Emphasis is placed on systems that can affect the safe operation i
or the safe shutdown of the plant.
System training uses both system walk-throughs and examinations to verify qualifications in.each system.
Walk-throughs and examinations are used for re-l qualification purposes to ensure that an optimum level of profi-ciency is maintained by the nonlicensed operator.
H.I.A.l.3.4 Reference g
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IE Circular No. 80-02, Nuclear-Power Plant Staff Work Hours,
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February 1, 1980.
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'H.I.A.1.3-5 Amendment 33 March 1981
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H.II.K.l.5 Safety-Related Valve Position H.II.K.1.5.1 Statement of concern The THI-2 accident of March 28, 1979, resulted from a series of events that, either directly or indirectly, pertained to the relative position of several safety-related valves within safety-related systems.
One method to prevent a similar type of circum-stance from occurring at another nuclear power plant would be for the licensee to review and modify, where necessary, the appropri-ate valve-related requirements, controls, ated with safety-related systems.
and procedures associ-H.II.K.l.5.2 NRC Position Review all safety-related valve positions, positioning require-ments, and positive controls to ensure that valves remain posi-tioned (open or closed) of engineered safety features.in a manner to ensure the proper operation Also, review related procedures, such as those for maintenance, testing, plant and system startup, and supervisory periodic (e.g., daily / shift checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations'and are maintained in their proper positions during all operational modes (Reference 1).
H.II.K.1.5.3 Detroit Edison Position Detroit Edison agrees with the NRC position and will conduct a review of the safety-related valves at Ectmi 2.
The review process will involve the examination of design-based valve posi-tions for pre-and postaccident positions, the positioning of valves under normal and abnormal plant operating conditions, and an analysis of the physical hardware used for. determining relative saf ety-related valve positions. % rusev -,'// 6c C otr1 P cc Tdo M SG PT6-% e n.
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i In addition, all procedures will be analyzed, reviewed, and modi-t fled, as necessary, to ensure that all valves are returned to their correct positions following necessary manipulations and are main-tained in their proper positions during all operational modes.
cnly will administrative and control room turnover procedures be Not modified, but surveillance and maintenance procedures will be re-written to reflect both retest and post-test valve positions.
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73 f 5 H.II.K.1.5.4 Reference jj 1.
IE Bulletin No. 79-08, Events Relevant to Boiling Water r
Power Reactors Identified During Three Mile Island f
Inci-D dent, April 1979.
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H.II.K.l.10
.O Review and Modifv (As Required) Procedures for Removing t
Safety-Related Systems From Service
'j (and Rectoring to Service) f To Assure Ooerability Status Is Known H.II.K.l.10.1 Statement of Concern The regttirement to have the various safety-related systems avail-
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abic for operation, regardless of the redundant' systems status,.
was found to be a necessary requirement in light of the TMI-2 accident.
As a result, it would be advisable to ensure that I
nystem inspection, postmaintenance, and/or system indi-proper operator notification.
f H.II.K.l.10.2 NRC Position (Reference 1)
Review and modify, as necessary, maintenance and test procedures to ensure that they required the following:
of redundant safety-related systems before the a.
of any safety-related system from service b.
Verification of the operability of all safety-related maintenance or testingsystems when they are returned to servic
'I Explicit notification of involved reactor operational c.
personnel whenever a safety-relate'd system is removed from and returned to service H.II.K.l.10.3 Detroit Edison Position
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D,et itgEd* on-il undertake a r'eview of a'll th I
main gnance nd rv pplicabla a'te
' modi'f y the lla'nce hst p'rocedur,es 'hqd s
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[eq ri-e res ensure that'the op(ing p o( 3 v'idual sy' stem vil. be ra s
s ownwto the ope < rat sonnel.
H.II.K 0.4 Prheedure Modifications (
N A 1 of t e ap 'icable maintenance and s\\arveill etestbroce-duges, on\\an in - idual basis, will be a6al g d an revieGed iri accordan.ce with the NRC Of fice of Inspection IE Bulletin No 79-08, the on3y bullet'IEJelevant to BWRs.
s the 7p-08 Burletin are\\ re'sented to, ensure The requfrements of e plant ' ocedu't.es requit --
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Verif cation,fb. test or in Rection, of the o, era-bility\\ f dundant safety-rerated s tems bef re
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the rem al of an' safety-relatedss st'em from s rvic
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e ificati,n of the operabilitys f all \\
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tems when they are' returned fety'-rel ted o
s service-llowin mai tenance r testing II.K.l.10-1 Amendment 33 - Mar h 1981 9
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EF 2-FSAR 1
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c.' Expliciti notificaition ofg involved <
ctor operational 5
wsonnel whenever a safety-related system is removed ifroAand\\ returned ho serv' ice /
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The\\mainten$ N willarece \\
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_ -udied to reflect specific requirements \\ that will\\be set, forth in various a*dminis-trative procedures concerhing th'e pre-and-post-test verification of safety-related redundant / systems.
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4.i v e g owed te mvdi h _p coaccrr. Ohif -dpchetic..5,.sintc..ance, r.d t w i~ e Toy
. Explicit noti'fication of reactor operati g personnel when\\a safety-helated sys\\em is removed fr'om and on-duty nuclear,kice~ will be handled \\ administratively \\through t. e returned to ser ift supd,rvisor, and indihidually by\\periodica ly inspecting the'sa ty feat res status \\ display, system (rpFSDS)
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By implementat' ion.,of the \\ procedure \\.
located inythe co rol roo.
changes,and 'through nuclear hift supervisor \\ notification, in addition to the SFSO the i tent of Bulletin No. 7%8 will be satisfactor11'A met.
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4 E.II.K.l.10.9 Reference 1.
IE_ Bulletin No. 79-08, Events Relevant to Boilina Water i
Power Reactors Identified Durinc Three Mile Island Inci -
dent, April 1979.
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H.II.K.l.10-2 Amendment 33 - March 1981 l
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The Detroit Edison Administrative Procedures, 21.000.01, Shift Operations and Control Room and 21.000.02, Operations Logs and Records presently require the Nuclear Shift Supervisor (NSS),
the Nuclear Assistant Shift Supervisor (NASS), the Nuclear N
Supervising Operator (NSO), the Nuclear Power Plant Operator i
l (NPPO), and the Nuclear Assistant Power Plant Operator (NAPPO),
! l to be % 7. aware of the plant status and changes in the j
plant status. This includes the status of redundant safety-i) related systems before the re= oval of any safety-related system
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,i from service. Additionally, the Surveillance Procedures require the Nuclear Shift Supervisor's approval prior to performance of any surveillance, thus assuring that the NSS verifies redundant
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system operability.
The maintenance procedures require permission from the Nuclear Shift Supervisor prior to performance, thus assuring the NSS verifies the operability of redundant, safety-related systems prior to the removal of any safety-related system from service.
h.
The Surveillance Procedures specify operability verification of safety-related systems. After the operability tests are completed, the NSS will review these tests to verify that they have been successfully performed and that they meet the acceptance criteria.
The Maintenance Work Packages specify the post test requirements to be performed by Operations after completion of work.
c.
Explicit instructions are provided to operating personnel through procedural requirements which specify that the NSS must be notified before any surveillance or maintenance activity is performed.
i At present, the maintenance and surveillancs procedures are being written. These procedures are expected to be completed and approved prior to fuel load.
I Written by:
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EF-2-FSAR i
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L H.II.K.1.22 Describe Automatic and Manual Actions for Procer f~)/
Functioning of Auxiliarv Heat Removal Systems When g,
Feedwater System Is Not Ooerable H.II.K.l.22.1 Statement of Concern On March 28, 1979, the THI-2 experienced core damage that re-sulted from a series of events initiated by a loss of feedwater transient.
Several aspects of the incident may have general applicability to operating BWRs.
Of particular interest are the actions required by the operator in a BWR in placing the high-pressure core injection (HPCI) and reactor core isolation cooling (RCIC) systems in the correct operating mode during an accident situation resulting from a loss of the main feedwater system.
The HPCI and RCIC syster.s may demand too much of the operator's attention or may involve them in a very complicated procedure for verifying that these systems are operating properly in either the automatic or manual mode during an accident condition.
This can result observe the operations of other emergency systems and plantin jeopard parameters.
H.II.K.l.22.2 NRC Position t'
i Describe the actions, both automatic and manual, necessary for
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proper functioning of the auxiliary heat removal systems (e.g.,
RCIC) that are us.ed when the main feedwater system is not oper-able.
For any manual action necessary, describe, in summary form, the procedure by which this action is taken in a timely sense (Reference 1).
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H.II.K.l.22.3 Detroit Edison Position Detroit Edison agrees to provide the actions, both automatic heat removal systemsand manual, necessary for proper functioning of the auxiliary (e.g., RCIC).
These actions are as follows.
If the main feedwater system is not operable, a reactor scram
'will automatically be initiated when reactor wate:
to Level 3.
level falls boil-off until the low-low-level setpoint, Reactor water level will continue to decrease Level 2, is reached.
At this point, the main steam lines will be isolated automati-cally, and the HPCI system and the RCIC s" stem will be automati-cally initiated to supply makeup' water to the reactor pressure vessel.
These systems will continue automatic injection until e
the reactor water level reaches Level 8, at which time the HPCI and RCIC turbine is tripped.
These systems (HPCI/RCIC) will restart automatically once the high-level trip signal clears and a low-low-level (Level 2) signal is received.
H.II.K.1.22-1 Amendment 33 - March 1981
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The following actions occur dur-ing an automatic initiation of lt RCIC and HPCI:
1 a.
Automatic Operation of RCIC
!.i The RCIC system will start automatically upon receipt
'I of the initiation signal -from the reacto; vessel low,
water-level sensor.
During startup f ro.a standby, the following events o cur automaticaliy:
1.
Turbine speed control given to RCIC system flow indicator controller 2.
RCIC test bypass valve to condensate storage tank closes (if open) 3.
Steam supply valve to turbine opens 4.
Barometric condenser ?ondensate pump discharge isolation valve closes S.
Pump discharge valve to feedwater line opens f"
6.
Barometric condenser vacuum pump starts d
7.
Steam supply line isolation shutoff valves to the turbine open (if closed) ll 8.
Cooling water supply valve to lube oil cooler opens i
The turbine starts as soon as the steam supply valve
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opens, since the turbine trip throttle valve and con-l trol valve are open.
The minimum flow bypass valve to suppression pool opens when pump discharge pressure in-e creases.
System flow starts when pump discharge pres-sure exceeds fee &sater line pressure.
As pump discharge
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pressure and steam inlet pressure change, the control signal adjusts turbine speed to maintain constant pump flow.
When pump flow reaches a prescribed value, the minimum flow bypass valve closes.
Upon occurrence of rith:[ a low water level in the con-densate storage tank 7
1pmmhc the suction to the RCIC pump changes automati-cally from condensate storage tank to the suppression pool.
b.
Automatic Initiation of HPCI i
l The following sequence of events occurs in the case of l
an automatic initiation of the HPCI system:
l.
Steam supply outboard isolation valve opens H.II.K.l.22-2 Amendment 33 - Marchil981
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EF-2-FSAR 2.
Steam supply inboard isolation valve opens (if closed) f j"
3.
HPCI suctionIvalve from condensate storage opent (if closed) 4.
HPCI pump discharge inboard and outboard isolation opens 5.
HPCI test line isolation valves close (if open')
6.
HPCI steam inle.t valve opens 7.
HPCI lube oil cooling water supply valve opens 8.
HPCI oil pumps start i
l 9.
HPCI condenser vacuum pump starts 10.
HPCI test return valves close (if open) j With the turbine stop valve and control valves open, steam is admitted-to the-turbine, accelerating it quickly to speed.
1 Upon occurrence of either a low water level in the condensate storage tank or a high level in tha sup-l 1
pression pool, the suction valve to the HPC1 pump changes over from condensate storage tank to the sup-pression pool.
The operator can switch the flow controller to the manual position and decrease flow rate to stabilize the water level in the reactor vessel.
This would be done before reaching the high-water-level isola-t tion.
Even if the operator does not manually take i
control and the HPCI trips on high level, the HPCI will restart automatically once the high-water-level isolation signal clears and a Level 2 low-low-water-l level signal is received.
For the loss of feedwater transient, the HPCI/RCIC systems are used to automatically provide the required makeup flow.
No manual operations are required.
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H.II.K.l.22.4 Reference l
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IE Bulletin No. 79-08, Events Relevant to Boiling Water Power Reactors Identified During Three Mile Island Inci-dent, April 1974.
g H.II.K.l.22-3 Amendment 33 - March 1981 e
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With the MSIV's closed, reactor pressure may rise to the setpoint of the safety relief valves which will operate to reduce reactor pressure.
The heat added to the suppression pool from the operation of the safety-f relief valves and the RCIC and HPCI systems will cause the suppression i
pool to heat up. As the average suppression pool terperature rises, the operator will initiate the Suppression Pool Cooling Mode of RHR to reduce
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this temperature prior to reaching' the. Technical Specification Limit.
'l A se::vnary of the operator actions are given below:
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1.
Start a RHR Punp.
l; tt 2.
Start the associated RHRSN pump.
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3.
Close the associated RHR Heat Exchangte By-cass Valve.
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4.
Verify the associated RHR Heat Exchanger Inlet Insolation i
Valve is OP M.
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5.
Verify the associated RHR Heat Exchange Discharge Valve I
is OPM.
6.
Open the associated RHR Icop Containment Spray / Test Insolation valve.
t Throttle the associated RHR Icop Containment Cooling / rest Insolation 7.
Valve.
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Insert 2 - to Appendix H,' Item II.K.1.22-
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In. addition to the operation of the HPCI/RCIC system upon loss of the normal feedwater system,' reactor vessel beat removal may-be augmented into operation of the RHR system inLthe' steam condensing mode.
The operator: actions necessary to place the RHR in-the steam condensing mode i
. are as follows:
1.
Remove the desired RHR locp from service, if operating.
2.
Place the RHN steam' condensing mode select switch in
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the desired loop position.
f 3.
Establish RHR service water flow through the selected-RHR heat exchanger.
L 4.
Open the motor operated valve, routing the heat.,
exchanger condensate back to the suppression 1 pool f
through the RHR pump flow tes* line.
5.
Open the motor operated bypass valve around the steam supplied valve ahead of the steam pressure control valve to-pre-warm the steam line.--
L 6.
Verify that.the steam line drain valves are functioning l
to discharge the steam line; drains to the main condenser.
7.
Gradually open the steam supply valve ahead of the pressure control valve and. verify proper pressure control valve operation.
8.
Verify proper operation of the heat exchanger ccnden-f sate level control valve as the condensat.e in the heat-exchanger reaches'the normal operating level.
9.
Verify' proper RHR service water system operation and place the RHR reservoir cooling towers in service as necessary.
10.-. Increase the steam flow to the heat exchanger to rated flow and monitor heat exchanger condensate water quality.-
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- 11. When the condensate water quality approaches the acceptable limits, align the RCIC system in the steam condensing mode.
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When the condensate water quality reaches acceptable limits, close the steam supply valve ahead of the steam pressure control valve.
13.
Close the heat exchanger condensate motor-operated drain valve to the RHR test line and open the heat exchanger condensate drain valve to the RCIC pump
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suction line.
Start the RCIC syg,gy and bring the pump to the EN5hl'"*'
14.
recirculatertflow4 W to flush the system to the suppression pool.
15.
Gradually open the steam supply valve ahead of the steam pressure control valve.
16.
As the heat exchanger condensate level increases, increase the RCIC pump
. speed
_ and begin returning the heat exchanger condensate to the reactor vessel.
17.
Increase the heat exchanger steam flow to the rated flow while increasing the RCIC pump speed to,
maintain the required heat exchanger condensate level.
The steam condensate mode of the RHR is now in service with reactor vessel heat being transferred to the atmosphere vi,a the RHR service water system.
Reactor vessel heat removal may also be accomplished through manual ac4uation of any of the 15 safety relief valves.
In th8Yfeactor vessel pressure reduction and heat removal is required through SRV operation, remote actuatian of l*
SRV is available and would be used in conjunction with the l
suppression pool cooling mode of the RHR sy, stem.
The operator actions necessary to place the RHR system in the suppression cooling mode are as follows:
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l 1
Remove the desired RHR loop from service, if operating.
2.
Establish RHR service water flow through the RHR loq.
heat exchanger.
3.
Align a loop RHR pump in the suppression pool cooling mode with suction from the suppression pool and discharge through the heat exchanger and back to thu suppression pool via the RHR test line.
4.
Start the RHR pump and adjust system flow as requ' ired by throttling the RHR test line valve.
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The RHR s'istem is now in the suppression pool cooling mode
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maticaining the required suppression pool temperature and heat distribution limits.:
During the mode of operation, the automatic depressurization system remains fully operational.and will automatically initiate if the conditions necessary for automatic depres-surization should occur.
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'# H.II.K.3.3 Report Safety and Relief Valve Failures Promptiv
{}g and Challenges Annually H.II.K.3.3.1 Statement of Concern The record of relief valve failures to close for all BWRs in the past 3 years of plant operation is approximately 30 in 73 reactor years (0.41 failures / reactor year).
This has demonstrated tnat the failure of a relief valve to close would be the most likely cause of a small-dreak LOCA.
The high' failure rate is the result of high relief valve challenge rate and a relatively high failure rate per challenge (0.16 f ailures/ challenge).
Typically, five j
l valves are challenged in each event.
This results in an equiva-3 lent failure rate per challenge of 0.03.
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H.II.K.3.3.2 NRC Position All future safety and relief valve challenges and failures should 3
be reported to the NRC.
This should include the prompt reporting of failures through Unusual Event Reports and the reporting of l
-challenges in the annual report (Ref erence 1).
H.II.K.3.3.3 Detroit Edison' Position
@utuvic satava og e e.
with the NRC couvetu &cgecding thc f;ilarc
)'N cf 2 ealiaF v C ": Lv close being the most likely cause v:
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Fer-i 2 ed.iuicitati<c eu vveduc ms will prov.ib %
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=*hnA? and frequency of repvuoing sarety and re116t Tai.c c.'. 21
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-lenges and failures.
d.
H.II.K.3.3.4 R,eference i
1.
U.S. Nuclear Regulatory Commission, Generic Evaluation of Feed-water Transients and Small-Break T.oss-of-Coolant Accidents in GE-Desicned Goeratino Plants and Near-Term Ooeratino License
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Applications, NUREG-0626, pp. 1-16, item B.14, January 1960.
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H.II.K.3.3-1 Amendment 33 - March 1981 g
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Detroit Edison will report, a failure of a safety relief valve f
to OPEN or CLOSE when called upon, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by phone,
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confirmed the first working day following the event by tele-graph (or similar transmission) and followed up with a written report in two weeks. This written report will be in the form of a License Event Report in accordance with the Plant Reporting Re-quirements, Administrative Procedure, 12.000.10.
,j The Detroit Edison annual report to the NRC will list each safety
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,i relief valve which is challenged during the year and will include l.
Il the number of times each is challenged.
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EF-2-FSAR H.II.K.3.17 Report on Outage of Emergency Core Cooling Systems B.II.K 3.17.1 Statement of Concern While present technical specifications contain limits on allow-able outage times for emergency core cooling systems (ECCSs) and components, there are no cumulative outage time limitations.
It is possible, therefore, that ECCS equipment could have a high unavailability record due to frequent outages, yet still be within the allowable technical specifications.
H.II'.K.3.17.2 NRC Position Applicants for an operating license shall establish a plan that allows for the accountability of the plant's ECCSs outage times.
The report of the plan must include--
a.
,ECCS outage dates and outage durations b.
Cause of the ECCS outage c.
ECCSS or components involved in the outage d.
Corrective action taken H.II.K.3.17.3 Detroit Edison Position
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No historical record of ECCSs outage is available at Fermi 2 because the plant has not previously been operational.
- However, a reporting program will b; d;velop;d and implementcd acucbg
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th; r; quired "CCS cetage inferm tien i decumented 2nd ce;ded 00 that futurc cet A cal of--th; cut:g information i: :: ily avail:ble. in Tbc form oS h rdvdear Plant 'R eliab i lit y Do4.
S de% ha.s been esk=b \\ ; shed, at Fermi 2, Eccs y
O d o.y in-for ma.t;e, d. h b e. reco ed ed, en H PRb -4 fem
- h poA of ran ce.".
The NPRD program it h6"g i
compvievised.
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r ea a', \\3 refr;eva%\\e.. Th NPRb dha. us ll b e-reviewed o,nk com p d ed o.w n va.
a.nd sob tw N ed to The N RC.
in the a w n vo.\\ r ( p o d N
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H.II.K.3.17-1 Amendment 33 - March 1981 l
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H.III.D.3.3 Inplant Iodine Radiation Monitoring H.III.D.3.3.1 Statement of concern During the TMI-2 accident, it was observed by the NRC that when either air samples or process monitors indicated the presence of radioactive materials in the air the conservative assumption was made that the material was radioiodine.
Based upon this assumption, areas were evacuated and/or personnel were required to wear air-supplying respiratory protection.
Definitive iso-topic analysis of air samples was required to determine if the material present was radiciodine (which would have required the protective actions described) or noble gas (which would not).
If the airborne activity was later determined to be noble gas, the problem was no longer one of internal contamination, but of external radiation exposure from the noble gas cloud.
Such exposure rates were normally quite low even for relatively high air concentrations of noble gas, and areas could be reentered and/or respiratory protection equipment removed.
The problem was that the isotopic analysis (used to differenti-ate between radiciodine and noble gas) was taking too'long for a-variety of reasons, including the high levels of radiation in the counting room (where the analysis is normally performed), causing this area to be rendered useless.
The concern was that essential areas could be evacuated needlessly and that personnel were wear-ing respiratory protection (which hamper,s both work and communi-cation) when it was not required.
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The NRC also felt that these conditions were commonplace in most i
facilities even during normal operation and maintenance.
t H.III.D.3.3.2 HRC Position
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a.
Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.
b.
Each applicant for a fuel-loading license to be issued
.bef ore January 1, 1981, shall provide the equipment, training, and procedures necessary to accurately deter-mine the presence of airborne radiciodine in areas within the plant where plant personnel may be present during an accident.
After January 1, 1981, each applicant and licensee shall have the s e capability to remove the sampling cartridge to a low-background, low-contamination area for further analysis.
Normally, counting rooms in auxiliary buildings will not have sufficiently low back-grounds for such an analysis following an accident. 'In the low-())
background area, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases.
H.III.D.3.3-1 Amendment 33 - March 1981
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1 EF-2-FSAR The licensee shall have the, capability to measure accurately
([f the iodine concentrations present on these samples under accident conditions.
There should be sufficient samplers to sample all vital areas.
(See Reference 1.)
H.III.D.3.3.3 Detroit Edison Position a
Detroit Edison is in agreement with the requirements for improved inplant iodine monitoring.
In order to meet the requirements, some procedures will be developed and others will be modified,
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and equipment will be purchased t llow rani'd determination _o iodine / noble gas concentrations.
The effrit; analycic cep= hili *y
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will be encured by the use of a rcmcLe eneliaio facility and routinc an& lysis; HealLh Physics has : Cc'Li; detector and the mecociated multicharnel analy:c,: and c0=puter. 2NSERT A H.III.D.3.3.4 Modifications To perform rapid 1.1 plant determinations of the airborne looine concentration, a stabilized sodium iodide. detector coupled.to.
an analyzer will be used to continuously evaluate an iodine.
adsorbent cartridge.
This cartridge will be coupled to a flow-stabilized air sampler.
The* entire unitswill be cart-mounted and portable.
Procedures for the use and calibration of the unit will be wri*ren and approved by the On-Site Review Organization.
Personnal will.be trained in the use and calibration of the units.
To evaluate air samples, Health Physics routinely will use a Ge (Li) detector coupled to a mu-ltichannel analyzer / computer to identify and quantify the results.
This analyzer will be backed up by two units in the Chemistry laboratory.
In addition, if both the Chemistry and Health Physics counting rooms are rendered I"5f C" inoperative +(cuch es duiluy occident conditienc),
2" offeite 2n:1-ysic facility will bc availabic for air c =pic analytic.
This
-efhf acill Ly 111 basically uce the ;;=c analycic equip = cat end-procedurec as that nocasily usca by "calth Phycics and Chemictry, In addition, a supply of silver zeolite, or equivalent, adsorbent cartridges will be obtained to allow the determination of the airborne iodine concentrations in the presence of noble gas.
H.III.D.3.3.5 Reference 1.
U.S. Nuclear Regulatory Commission, Clarification of TMI Action Plan Requirements, NUREG-0737, pp. 3-195 and 3-196, t
October 1980.
Deicoit Edison will purchase and, h ave avai lam e.
adeg ude. womker an oC in plant iodine monited og indeemed3 to provide cov era $e fov-
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hp 4.r. B. 2.4. (. 4.
h H.III.D.3.3-2 Amendment 33 - March 1981
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Three separate laboratory facilities with multi channel analyzer (GeLi) capability will be available; one in the-Chemistry Laboratory Counting Room (located in the Radwaste Building), one in the Health Physics Laboratory
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(located in the plant Office Service Building), and one l'
remote laboratory facility (located within the Detroit Edison property surrounding the plant site).
t B.
(such as might occur during worst case accident conditions),
a remote laboratory facility located within the Detroit Edison property plant site will be available for air sample analysis.
This remote facility will basically use the same analysis equipment and procedures as that normally used by Health Physics and Chemistry.
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