ML20002E096

Forwards Response to NUREG-0737,in Response to NRC 801031 Ltr.Items Requiring Followup by Unit 2 & Corresponding SER Are Included in Action Item for Requirement

ML20002E096
Person / Time
Site:	Farley
Issue date:	01/14/1981
From:	Clayton F ALABAMA POWER CO.
To:	Schwencer A, Varga S Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.5, TASK-1.D.1, TASK-2.K.2.13, TASK-TM NUDOCS 8101260360
Download: ML20002E096 (99)

Text

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Alabama Power Company 600 North 18th street Post Office Box 2641 Birmingham, Alabama 35291 l

Telephone 205 250-1000 F. L CLAYTON, JR. m Senior vice President Alabama Power the southern electnc system January 14, 1981 Docket Nos. 50-348 50-364 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Phillips Building 7920 Norfold Avenue Bethesda, Maryland 20014 Attention:-,Mr. S. A. Varp ,

Mr. A. Schwencer FARLEY NUCLEAR PLANT UNITS 1 AND 2 NUREG-0737 RESPONSE Gentlemen:

In response to NRC letter dated October 31, 1980, Alabama Power Company submits our response to NUREG-0737, Clarification of TMI Action Plan Requirements. As requested by the NRC Staff, Alabama Power Company only addressed items containing new or additional requirements. Items not falling into these categories contain references to previous submittals which have been reviewed and found acceptable.

A significant engineering, construction, and testing effort has been expended by Alabama Power to meet the original TMI requirements issued by the NRC letters dated September 13, 1979 and October 30, 1979, and NUREG-0694. Even though implee ntation dates have been extended for l several items by NUREG-0737, Alabama Power Company has continued work l in all areas in an effort to meet previously committed dates and comply with the new clarification.

In addition, items requiring follow-up by FNP Unit 2 low power operating license NFP-8, and the corresponding Safety Evaluation Report, are included in the " Action Item" for that requirement. These items are as follows:

1. Reactor vessel level description

2. Degraded core training completion O

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3. Shift Technical Advisor training completion B10 2 26 0 M

i Mr. S. A. Varga Mr. A. Schwencer January 14, 1981 If there are any questions concerning this response, please contact us.

Yours very truly, c y sq,. ~:-

ra L L hw F. L. Clayton, Jr."

ODKjr/rt Enclosure ec: Mr. R. A. Thomas SWORN TO AND SUBSCRIBED BEFORE Mr. G. F. Trowbridge M( THIS S A;- DAY OF Mr. E. A. Reeves (w/ enclosure 1 ., s,-v , 1981.

Mr. L. L. Kintner w/ enclosure /~ '

Mr. J. P. O'Reilly w/ enclosure o Mr. W. H. Bradford w/ enclosure 3

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1 I.A.l.1' SHIFT TECHNICAL ADVISOR ,

P'revious Response

, An individual on shift has beert designated for Unit i to serve as Shift  !

Technical Advisor (STA) during emergency conditions since January 1,1980.

An STA will be available for Unit 2 whenever the unit is in Mode 4 or

. above. The STA's have received additional training as stated in APCo

_- letters dated June 20 and August 1,1980. By letters to the NRC dated June 20, 1980, August 1, 1980 and August 14,1980 for Unit 2 and October 24, 1979, November 21,1979, Decembe'- 31, 1979, and March 14, 1980 for Unit 1 -

Alabama Power Company has previously submitted comitments and documented action taken related to STA for the Farley Nuclear Plant. .

Clarification Response A. Shift Technical Advisor Training Program - Short Tenn

1. Training ,

l Prior to January 1,1981, Alabama Power Company shift technical

. advisors met the requirements listed below as specified in Enclosure 2 of NRC letter dated September 13, 1979, which was referenced by the October 30, 1979 letter.

a. General technical education equivalent to about 60 semester

• __ hours at the college level in basic subjects of engineering and science. Desirable basic subjects include mathematics, chemistry, metallurgy, reactor physics, heat transfer fluid i flow, thermodynamics and electricity.

b. Reactor operator training in the design function, arrangement and operation of plant systems including the capabilities of instrumentation and controls in the control room.

c. Transient and accident response training as an aid in accident assessment and diagnostics, recognition of multiple failures of inadequate core cooling.

2. Speci/ically, the above training and qualification have been completed as follows:

a. Each candidate completed a four week STA training course covering

., the following subjects; chemistry, metallurgy, reactor physics, j heat transfer, fluid flow, thermodynamics, instrumentation and  !

controls and transient and accident response.

l b. Selecting candidates with a minimum college level training equivalent to about 60 semester hours in basic subjects of engineering and science.

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I.A.l.1 (continued) 2

c. Each candidate completed simulator training covering:

1) Control board familiarization.

ii) Transient demonstrations.

iii) Instrument failure response.

iv) Accident diagnostics.

d. Candidates completed training on the design, function, arrangement and operation of plant systems in one of the following ways:

1) Completion of a six-week systems training course.

ii) Completion of an individually tailored self-study program including instructor assistance and periodic examinations.

This fonn of training is used for experienccJ personnel only.

iii) Completion of NRC license training.

3. Retraininn Shift technical advisors will attend licensed requalification training and annual simulator retraining to maintain their level

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of proficiency.

B. Shift Technical Advisor Training Program - Long Term

1. Qualification Alabama Power Campany will select personnel for the STA position that have one year nuclear power plant experience, of which at least two months must be at an operating plant and six months at the plant where the position is to be filled. Additionally, the STA must have a degree in an engineering or scientific discipline or 60 semester hou s of college level training in basic subjects in engineering and sciences as described in A.2.b above.

2. Selection Alabama Power Company will continue to select personnel for the STA position based on their education training, experience and interest.

3. Training ,

Completion of training as described in D below which includes:

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a. Classroom lectures.

b. On-the-job training.

c. Simulator training.

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I.A.l.1 (continued) 3 C. Shift Technical Advisor - Phase Out Plan The unique position of STA will be phased out as one shift super-visor per shift completes training to meet the long term require-ments of Item B above.

D. Comparison of the Shift Technical Advisor - Long Term Program to Appendix C of NUREG 0737 (INP0 STA Guideline)

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The comparison of the Long Term STA Program to the INP0 Standard is provided as follows:

ITEM INF0 REQUIREMENT APCo Program Experience 1. 18 months nuclear power 1. 12 mo.

• plant experience 2 months experience at 2. **

2.

1 an operating plant

3. 12 months at the plant 3. 6 mo.

• where the position is to be filled l Post High 1. Mathematics - 90 hrs. 1. **

School Education & 2. Inorganic chemistry - 2. **

t- Training 30 hrs

3. Physics - 150 hrs 3. **

College Level 1. Engineering Mathematics 1. * (Laplace Fundamental through Laplace trans- Transforms Education forms - 90 hrs not required)

2. Reactor Theory - 100 hrs 2. **

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3. Reactor Chemistry - 3. **

30 hrs

4. Thermal Sciences - l?O 4. **

l hrs

5. Nuclear Materials - 40 hrs 5. None 1

6. **

6. Electrical Sciences -

7. **

7. Nuclear Instrumentation

& Controls - 40 hrs

8. **

8. Health Physics ,40 hrs

l 1.A.1.1 (continued) 4 ITEM INP0 REQUIREMENT APCo PROGRAM Applied  !

Fundamentals 1. Reactor Technology 1. **

2. Plant Chemistry 2. **

3. Reactor Instrumentation 3. **

& Controls

4. Reactor Plant 4. **

Materials

5. Reactor Plant Thermal 5. **

Cycle - 120 hrs Total Program - 120 hrs Management & 1. Leadership 1. **

Supervisory Skills 2. Interpersonal Communica- 2. **

tion

3. Motivation of Personnel 3. **

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4. Problem & Decision 4. **

Analysis

5. Command Responsibilities 5. **

and Limits

6. Stress 6. **

7. Human Behavior 7. **

Total Program - 40 hrs

- Plant Systems 1. Plant Systems - 200 hrs 1. **

Administrative 1. Applicable Administrative 1. **

Controls Procedures, Technical Specifications, Emergency Plan - 80 hrs General 1. Startup, power operation 1. **

shutdown, xenon follow, ECP and shutdown margin -

30 hrs Transient and 1. Emergency Operating 1. **

Accident Procedures, Abnormal Analysis and Operating Procedures, Emergency Transient & Accident Procedures Analysis - 30 hrs

I.A.1.1 (continued) 5 ITEM INP0 REQUIREMENT APCo PROGRAM Simulator 1. Combination classraom 1. **

Training and simulator training -

103 h. s l

l Lecture - 40 hrs 1. **

Annual 1.

Requalifica-tion Training 2. Simulator - 40 hrs 2.

• 24 hr

.The qualifications and training of individuals currently designated -

as STA's are being submitted to the NRC Staff for review of compliance with the criteria stated above (A.l.).

FOOTNOTE:

• • - fully meets or exceeds INP0 guideline

• - partially meets INPO guideline t

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6 I.A.1.2 ' SHIFT SUPERVISOR ADMINISTRATIVE DUTIES Previous Response .

l In previous letters dated June 20,198'O and August 22, 1980 for Unit 2 and October 24,1979, December 31, 1979 and June 26, 1980 for Unit 1, Alabama Power Company previously submitted commitments and documentation of actions taken at the Farley Nuclear Plant to implement the Lesso'ns

_ Learned Reclirements of NUREG 0578.

[ Clarification Response No clarification was provided in NUREG 0737. Therefore, no additional clarification is submitted.

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I.A.I.3 Shift Manning Previous Response In previous letters dated June 26,1980,for Unit 1 and June 20, 1980, August 7, August 14, 1980 and September 8, 1980 for Unit 2 Alabama Power Company previously submitted comitments and documentation of actions taken at the Farley Nuclear Plant.

Clarification Response Alabama Power Company intends to implement three shift coverage in the Sumer of 1981 for classifications performing safety related support

activities (i.e., Health Physics technicians and Maintenance /I&C Journeymen). Staffing levels at Farley Nuclear Plant are currently sufficient to staff these shifts without routinely requiring employees to violate the overtime restraints of HUREG 0737 (excluding extended periods for refueling, major maintenance or major plant modifications.)

Due to these facts, the administrative burden that would be imposed by formally tracking each exception to these requirements and requiring Plant Manager approval of exceptions is not justified in Alabama Power Company's opinion.

A proposed revision to Farley Nuclear Plant Unit 1 Technical Specification Table 6.2-1 which was transmitted to the NRC staff in Mr. F. L. Clayton's letter to Mr. A. Schwencer, dated December 15, 1980 and Table 6.2-1 of the Farley Nuclear Plant Unit 2 Technical Specification addresses NUREG 0737

• ~ overtime restrictions (1), (3) and (4) of NUREG 0737 clarification to I.A.l.3. The Alabama Power Company contract with the International Brotherhood of Electrical Workers requires that a man have 8. hours of rest between work periods. If he does not receive eight hours, overtime is paid until he has had 8 consecutive hours of rest. ~ Restriction (pay2) would be violated by a significant percentage of the work force whenever personnel rotate to another shift (if on fixed shift schedule) or change their shift (if on rotating schedule). While this happens infrequently, it would.be a significant administrative burden. On a three shift, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> a shift, schedule, when the schedule is changed or when personnel rotate to another shift, time off from the end of the old shift to the start of the new shift is either 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or equal to or greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Further, restriction (2) would prevent an employee from being able to swap shifts for :t single day for personal reasons since this would result in less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> off either before or after the shift which he swapped to for the l single day. For the above reasons Alabama Power Company feels that i restriction (2) of NUREG 0737 item I.A.l.3 is excessively restrictive and would routinely create problems and that restriction (1) and (2) of Table 6.2-1 are sufficient to ensure adequate rest is provided between work periods.

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l I.A.2.1 IM1EDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR OPERATOR I TRAINING AND QUALIFICATIONS Previous Response By letters of August 1,1980, August 8,1980, and July 15,1980 for Unit 2 Alabama Power Company previously submitted commitments and documented actions taken for the Farley Nuclear Plant.

Clarification Response After December 1,1980, applicants for a Senior Reactor Operator (SRO)

License will have one (1) year of experience as a licensed operator. This experience may be obtained as follows:

(a) Military experience equivalent to a licensed reactor operator or Senior Operator, or .

(b) Experience as a licensed reactor operator or Senior Reactor Operator at another nuclear facility, or (c) One (1) year of experience as a licensed reactor operator at the Farley Plant, or (d) Possess a degree in er.gineering or science In addition to the experience requirements listed above, applicants for a Senior Reactor Operators license will meet additional experience & training requirements as follows:

(a) 4 years responsible power plant experience of which a maximum of 2 years

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is academic or technical training. At least 2 years experience shall be Nuclear Power Plant experience with 6 months at the Farley plant.

(b) Participation in the plant SRO training program which includes training in the areas of heat transfer, fluid flow, mitigating core damage and increased emphasis on plant transients.

(c) 3 months on shift training.

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, 9 I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS Previous Response By letters dated July:15,1980 and August 1,1980 for Unit 2, Alabama Power Company addressed this requirement for the Farley Nuclear Plant.

Clarification Response Current plant instructors involved in training programs for licensed operators are SR0 licensed. Instructors obtained from other sources or future plant ir.structors will be SRO licensed or certified as per NRC 3/28/80 letter from Harold Denton.

i Instructors are required to attend retraining programs. All licensed SRO instructors attend the SRO requalification program.

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10-I.A.3.1 REVISED SCOPE AND CRITERIA FOR LICENSING EXAMS Previous Response

- By letters to the NRC dated June 20,1980, July 15,1980, and August 1,1980 related to Unit 2 and the letter of June 26, 1980 for Unit 1. Alabama Power Company addressed this item for the Farley Nuclear Plant.

Clarification Response Alabama Power Company has purchased a plant specific simulator which is scheduled to be operational by July 1983. A plant specific simulator is not available at this time and it is the opinion of Alabama Power Company that utilizing a non-identical simulator would not provide a valid operator examination.

Alabama Power feels such an examination would probably serve as a detriment to operator proficiency by requiring him to learn in detail a control board other than the one on which he would be licensed. In addition, currently available l simulators are booked months in advance. Therefore, Alabama Power Company proposes to not examine operators on a simulator until its ow plant specific simulator becomes available. ,

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I.B.l.2 INDEPENDENT SAFETY ENGINEERING GROUP

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Previous Response By letters of August 6,1980, August 8,1980 and June 20, 1980 related to Unit 2. Alabama Power Company addressed this requirement for the Farley Nuclear Plant.

Clarification Response Alabama Power Company, in the process of implementing the series of reviews and evaluations which meet the requirements of the independent safety engineering group (Section I.B.1.2), detennined the need to clarify our response to this item. The responsibility outline for each group described in our August 8,1980 letter is generic in nature.with Table 1 providing the more detailed description of each group's review and evaluation functions. Three items describing the review function of the System Performance & Planning Group are clarified by the attached pages.

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. TABt,E 1.' APCO'S SYSTEM DF INDEPENDENT REVIEV5 PAGE 1 REVIEW ORGANIZATIONS 1 REVIEW PURPOSE' .

REVIEW System Perfomance & Plant Operations Review Operation [ valuation planningGroup(5 PPG) . Safety Atidit & Engineering REQUIRLHINT (Reporting thru . Coninittee (FORC) Tearns (CET) Review Croup (SACAG) organlaational Structural o[f Operations Review Crcup Independent (Reporting Directly to the (neporting Offsite to the Reporting to Vice President-kuclear Hanager-Safety Ardit,8

. Parallel to Operations thePlantManager) ~

to the Plant Manager) Generation) Engineering Review)

Revitw procedures and changes to U procedures important to the safe fo ur a Reviews administrative Reviews procedures 1. Reviews administrative operation of the FNP tions ma ntenance and procedures and changes to affecting areas under procedures and changes survelllanceproce- other procedures which evaluation for overall thereto and changes to -

dures affected by may involve an unreviewed plant safety. other procedures unich dest Sn. changes safety question (USQ) for -

r.ay involve an uso for advisablitty of the pro- .

advisab~ility of the pro.

z e p cedures or their chan es cedures and their changes as a jnove h a}f ohrh operations assessment - -

2. Verf fles that adequate role.

• evaluation of technical

' , accuracy and clarity o F all procedores is per.

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formed.

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Review proposed tests'and taperleents Reviews engineering

• Reviens proposed tests 1. Reviews proposed tests ah tests and experiments sad e8Periments which may espertments which,esy th.

t those r involve an U50 to verify > .

volve an U50 to verify es-t(excehephysicshlated o co for 88 U5Q actually exists. . U50 actually eshts, technical advisability

. 2. Vertfles proper develos-NOTE: Technical Group i and review process.

(independent of.0pera

. tionsI reviews' core ,

. physics 1 tests and ' ' - -

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exp.eriments. . .- -

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TA81.E 1 [APC0'S SYSTEM Of INDEPENDENT R'.VitV5 .

PAGE 2 REVitW ORGAH12AT!0N5 4 REYl[V PURPOSE REVitV -

• System Perfomance & Plant Operations Review

. PlanningGroup(5 PPG) Operation tvaluation Safety Audit 8 Engineering REQUIREMENT (Reporting thru Coirgitter(PUR0) '

icas(CET)

Organlistlonal Structural o(Review Group Independent (lleporting Directly to the Review f Operations Repor Group (SA[RG)

(Reporting Offsit! to the Parallel to Operations the Plant Manager) ting to Ggneration) Vice Prn ident Huclear flanager-Safety Audit 1 to the plant Manager) Engineering Review)

3. Review changes and modift' cations to Reviews changes and Unit systems and equipment Reviews changes and modifications to. modifications which may 1. Reviews changes and endl-systems and equipment involve an USQ to verif fications which may .

for technical no U5Q actually asists.y ' involve an USQ to verify advisability no U50 actvally exists.

completed under the - '

2. Verifles proper develop-routine design change ment and review process.

system.

4. Review Safety Evaluations for changes to procedures. equipment. ce systems, Reviews safety evaluations .

for changes to procedures.

1. Reviews safety evaluations or tests and experiments to verify .

for changes to procedarcs.

that ac',lons do not constitute

• 0r unreviewed safety.tssues equipment, tests and experiments or systems.to equipment or systces, or verify that USQ are not tests and esperiments to involved. verify that USQ are not Involved.

, . 2. Verifles proper develop-ment and review process.

' $. Review changes, tests, or experiments Reviews changes, tests, e which involve unreviewed safety issues Revtews changes, tests, or '.

l or experiments which may experimeriu sMca involve '). Reviews changes, tests, or involve USQ for technical USQ for disposition. experleents which involve adetsability. '

USQ for disposition,

2. Verffles proper review process.

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I.C.1 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS

'AND ACCIDENTS

Previous Response i

._ By letter to the NRC dated June 20, 1980 and August 1, 1980 for Unit 2 and 4

letters dated October 24, 1979 and December 31,1979 for Unit 1, Alabama Power Company addressed this subject related to the Farley Nuclear Plant.

i l Clarification Response

! Alabama Power Company as a member of the Westinghouse Owner's Group has 1

participated in a significant upgrade of emergency procedures with accompanying

! appropriate operator training as a result of this requirement. The Westinghouse l Vwner's Group submitted a detailed description of its program to comply with the requirements of Item I.C.1 on December 15,1980 (WOG 1etter 80-179).

' The program identified previous owners group submittals to the NRC, which we believe will comprise the bulk of the response. The additional effort required to

obtain full compliance with this item, as discussed with the NRC on November 12, l 1980, together with a schedule for completion, was also identified in the

December 15, 1980, submittal. If additional guidelines are develorad, the Farley

,_ Nuclear Plant will utilize these guidelines to further upgrade appropriate 4

procedures and provide the associated operator training by the first refueling outage for each unit after January 1,1982.

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13 I.C.2 SHIFT RELIEF AND TURNOVER PROCEDURES Previous Response Alabama Power Company has submitted its response to this item by letters of June 20,1980 for Unit 2 and October 24, 1979 and December 31,1979 for Unit 1.

Clarification Response .

NUREG 0737 required no clarification response for this item.

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14 I.C.3 SHIFT SUPERVISOR RESPONSIBILITIES Previous Response Alabama Power Company has submitted its response to this item by letters to the NRC dated June 20, 1980 and August 22, 1980 for Unit 2 and October 24, 1979 and December 31,1979 for Unit 1.

Clarification Response NUREG 0737 did not require a clarification response.

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15 I.C.4 CONTROL ROOM ACCESS Previous Response By letters dated June 20, 1980 for Unit 2 and October 24, 1979, and December 31,1979 for Unit 1 Alabama Power Company previously submitted connitments and documented actions taken for the Farley Nuclear Plant related to this item. .

Clarification Response NUREG 0737 requested no c1&rification for this item.

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16 I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF Previous Response By letters dated June 20,1980 for Unit 2 and June 26,1980 for Unit 1,

- Alabama Power Company submitted comitments and documented actions taken for the Farley Nuclear Plant related to this item.

Clarification Response In addition to previous submittals, Alabama Power will utilize the IMPO See-In Program and/or monthly NRC LER Summary with dissemination of the results of such evaluations to appropriate members of the plant staff. The procedure for disseminating such information and the procedure for auditing the associated feedback program have been revised.

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-I.C.6 GUIDANCE ON PROCEDURES FOR VERIFYING' CORRECT PERFORMANCE OF OPERATING ACTIVITIES i

Previous Response This item was femally issued as a part of NUREG 0737, therefore there has been no previous response.

I Clarification Response Since the Farley Nuclear Plant was placed in service, Alabama Power Company as a matter of good operating practice, has had policies and procedures to insure that the operational status of power plant equipment was controlled by shift supervision, plant operators were infomed of the equipment's status, and that positive means were employed to insure equipment would perform its intended function when being returned to service.

Shortly after the accident at TMI-2, Alabama Power Company strengthened its policies and procedures to require an independe~nt verification be i

perfomed when returning to service equipment important to safety.1 This policy also included independent verification following refueling or major maintenance outages. Specifically, these current procedures include the j following:

H Authority to release equipment important to safety for maintenance on surveillance testing or return to service is delegated to the on-shift SRO with the stipulation that the shift supervisor be kept fully l infomed of such status. -

Plant operators (control room work location) are required to log the removal and return to service

~o f equipment important to safety.

Upon return to service of equipment important to ~

safety, a fomal verification of the lineup is conducted. The lineup and the verification are performed by individuals qualified on the equipment or system. The lineup and verificationincludes valves, switches, and breakers as appropriate.

At this time, Alabama Power Company does not feel that independent verification when removing equipment important to safety from service is justified in all cases. In most cases, the removal of equipment from service is verified from the control room as part of routine shift operation.

As equipment important to safety is removed from service, alams are received, meter readings change, status lights change, and/or various light indications change. For removal of equipment important to safety, the shift supervisor is directed by procedure to predetermine whether such indications will give adequate indication of resultant system status and if they will not, he directs a properly qualified individual to verify the resultant lineup. This is a new commitment on Alabama Power Company's part and will be inco'rporated into appropriate procedures by April 15, 1981.

18 I.C.6 (Continued)

I Equipment important (pressure boundary to safety)is components anddefined as pressurizer associated the reactorand coolant system pressure relief system, the residual heat removal system, engineered safety features systems, engineered safety features electric power systems, and cooling water systems necessary to operate the above systems.

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19 I.C.7 NSSS VENDOR REVIEW OF PROCEDURES (UNIT 2 ONLY)

Previous Response l

-- Alabama Power Canpany by letters of . June 17, 1930, July 1.7, 1980, September 2, 1980, September 11, 1980, October 13, 1980, and November 6, 1980, has responded to the NRC on this issue.

Clarification Response The review of the low-power physics tests and applicable emergency operatin9 procedures has been completed by the NSS5 vendor (Westinghouse).

Their review of power ascension tests will be complete by l

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20 I.C.8 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPLICANTS (UNIT 2 ONLY)

Previous Response By letter of June 30, 1980 Alabama Power provided selected emergency operating procedures for NRC review as a part of the full-power licensing process.

Clarification Response-NUREG 0737 required no clarification response to this item.

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1 I.D.1 CONTROL ROOM DESIGN REVIEWS Previous Response ,

By letter of June 10,1980, June 20,1980 and July 17, 1980, Alabama Power Company provided response to the NRC on this item for Unit 2.

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Clarification Response Alabama Power Company intends to implement similar connitments described in the July 17, 1980 letter for Unit 1 prior to return to power after the current refueling outage except for denoting normal, alert, and alarm ranges on other significant main control room meters (other than those described in emergency procedures) which will be completed by the end of the third refueling outage. Alabama Power Company will address the long term control room design review after issuance of NUREG 0700. Alabama Power Company intends to implement the commitments described in the July 17, 1980 letter for Unit 2 prior to exceeding 5% power except for denoting normal, alert, and alarm ranges on other significant main control room meters (other than these described in emergency procedures) which will be completed by the end of the first refueling outage.

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O 22 I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE Prev- as Response Alabama Power Company has not previously formally responded to this item.

Clarification Response Alabama, Power Company will address this requirement and the implementation schedule when the final version of NUREG 696 is issued.

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23 I.G.1 TRAINING DURING LOW-POWER TESTING (UNIT 2 ONLY) 4 Previous Response

_. By lettersdated July 17, 1980, September 2,1980, and September 11,1980 Alabama Power Company address and submitted pro:edures to the NRC associated with this item.

Clarification Response NUREG 0737 does not request a clarification response.

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I l 24 l II.B.1 REACTOR COOLANT SYSTEM VENTS Previous Response l

By letters dated June-20, 1980 and August 1,1980 for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979, and March 14, 1980 for Unit 1 Alabama Power Company provided infomation and description of actions i applicable to the Farley Nuclear Plant.

l Clarification Response ,

The reactor vessel head vent system (RYHVS) is designed to remove non-condensible gases from the reactor vessel, head area, hot legs and cold legs. The RVHVS is designgd to vent in excess on one-half the reactor coolant system volume of hydrogen at 650 F in one hour from one of two available flow paths. The RVHVS is orificed to limit the blowdown from a break or inadvertently opened flow control valve downstream of the orifice to the capacity of one charging pump. Since a break in

the head vent line (upstream of the orifice) would behave similarly to the hot l

leg break case presented in WCAP-9600 (Section 3.2, Case F), the results presented therein are applicable to a RVHVS line break; therefore, a vent line break upstream of the orifice results in no calculated core uncovery.

i-The system is operated from the control room and has control room indication of valve position. The system is safety grade and meets the single failure criteria for venting initiation and tennination. The vent valves, block valves, and position indications meet the requirements of IEEE-323-1974, 344-1975, and 328-1972.

The Westinghouse Owners Group, of which Alabama Power Company is a member, is -

developing generic procedure guidelines for the use of the reactor vessel head

- vent system which will be incorporated into the Farley Plant procedures when the Reactor Coolant System Vent System is operational and approved by the NRC. These procedures will also include guidelines for testing the system during each refueling outage. In addition, the Westinghouse PWR system has demonstrated in the past through operating experience that decay heat can be effectively removed from the reactor coolant system with the absence of non-condensible gases in the steam generators. Moreover, the Westinghouse Owners Group has developed a procedure entitled " Exceptional Emergency Operating Instructions 2-Loss of Heat Sink" which provides operator procedures for an event where a steam generator is not available to remove decay heat.

The additional displays and controls added to the control room as a result of this requirement will be considered as part of the long-term human-factors analysis and will include:

(a) the use of this infomation by an operator during both normal and abnormal plant conditions

25 II.B.1 (continued)

(b) integration into emergency procedures (c) integration into operator training, and (d) other alarms during emergency and need for prioritization of alarms.

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This system has been installed in Unit 2 and initial testing has been completed. Final testing is planned to be complete prior to exceeding 5% power.

The Unit 1 valves and piping are installed and the entire system is planned to be completed prior to return to power from the present refueling outage or no later than July 1,1982. This system will not be placed in an operational status on either unit until accepted by the NRC.

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26 l II.B.2 DESIGN REVIEW 0F PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF l EQUIPMENT FOR SPACES / SYSTEMS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS Previous Response l

By letters of June 20, 1980 and August 1,1980, for Unit 2 and October 24, 1979,

- November 21, 1979, December 31,1979, March 14,1980, and May 5, 1980 for Unit 1 Alabama Power Company documented comitments and actions taken for the Farley Nuclear Plant related to this item. ,,

Clarification Response A design review for the Farley Plant - Units 1 and 2 was conducted by Bechtel Power Corporation, using the TID source terms and the 10 CFR 20 and GDC19, 60-64 of Appendix A to 10 CFR 50, dose criteria.

This shielding design review considered several classifications of systems l which included recirculation systems, systems which are extensions of the containment atmosphere, portions of the liquid sampling system, and portions of the letdown system.

The liquid and gaseous radwaste systems were not included in these analyses.

The gaseous system was eliminated since the reactor vessel head vent would be

! r- - used for degassing operations rather than the VCT. The leak reduction program instituted at Farley Nuclear Plant, and venting of the reactor by the reactor

- vessel head vent and/or PORVs rather than the letdown system and VCT, minimized the need for the liquid waste processing system and therefore it was not consiC.ed. The high activity radioactive lab and counting room for the affected unit was not included among..those areas' where access is considered vital after an accident since for'two unit operation'these areas in the unaffe.c ted unit will be utilized for post-accident analyses.

Access areas with their corresponding post-accident occupancy time for Units 1 and 2 are listed below: Unit 1 Area Occupancy Period Control Room 24 hr/ day

'lealth Physics Area 24 hr/ day Primary Access Point 24 hr/ day Passageway to Unit 2 1 hr/ day Hallway 409 1 hr/ day Electrical Penetration Rooms ** 1 hr (approximately 1 hr.

afteraccident)

• • Design change to eliminate occupancy requirement is being considered, Zone maps.will be updated as necessary.

II.B.2 (C:ntinu:d) Unit 1 27 Area Occupancy Period Hallway 322 (Outside Sample Room) *** 1 hr/ day Gas Analysis Room *** 1 br/ day Cable Spreading Room 1/2 hr.*

Filter Rooms 2 hr/ day

• 1/2 hr.*

Switchgear Rooms (Elev.121')

Hot Shutdown Panel 24 hr/ day

• CCW Pump Room 1/2 hr.*

Corridor 161 1/2 hr.*

RHR Heat Exchanger Room 1/2 hr.*

Stairway No. 1 Transit to elevations at

. west side of aux. b1dg.

Staimay No. 2 *** Transit to elevations 77'/83' Staimay No. 8 Transit to elevations at north and east sides of

" aux. bldg.

• Infrequent-brief access to these areas may be required post-accident.

l l *** Complete evaluation of these areas is based on resolution of shielding associated with the electrical penetration rooms.

Unit 2 Area . Occupancy Period Control Room '24 hr/ day Technical Support Center 24 hr/ day Health Physics Area 24 hr/ day Primary Access Point 24 hr/ day Passageway to Unit 1 (2402) I hr/ day Hallway 2409 1 hr/ day Electrical Penetration Rooms ** 1 hr (approximately 1 hr. after accident)

• • Design change to eliminate occupancy requirement is being considered.

Zone maps will be updated as necessary.

II.B.2(Continued) 28 Unit 2 Area Occupancy Period Hallway 2322 {0utside Sample Room for Liquid Sample) *** 1 hr/ day Spectro Photometer *** 1 hr/ day Cable Spreading Room 1/2 hr

• Filter Rooms 2 hr/ day

• f Switchgear Rooms (Elev.121') 1/2 hr* .

Hot Shutdown Panel 24 hr/ day

• CCW Pump Room 1/2 hr

• Corridor 2161 1/2 hr

• RHR Heat Exchanger Room . 1/2 hr

• Stairway No. 1 Transit to elevation at west side of aux. b1dg.-

Stairway No. 2 *** Transit to elevations 77'/83'.

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F-Stairway No. 8 Transit to elevations at north and east sides of aux. b1dg.

• Infrequent-brief access to these areas may be required post-accident.

• • • Complete evaluation of these areas is based on resolution of shielding associated with the electrical penetration rooms.

Each of these areas has been analyzed to determine the dose rates following an accident. .

l The Primary Access Point was not initially ceasidered but was later designated a I-A area for direct radiation. The security center will be included in the radiation zone maps. The main control room and the technical support center were considered as areas requiring continuous occupancy.

As a result of these studies the following shielding modifications are listed for Units 1 and 2:

1. Add shielding to the portion of line 3" SCC-12 which is exposed in the area of the Seal Injection Filter valve station to reduce the dose rate in this area.

29 II.B.2 (Crntinued)

2. Place temporary shielding at the containment radiation monitor to reduce dose rate in the corridor (RE-011, 012).

3. Re-route the RCS sample discharge line so that spent samples are returned by a more direct route to the VCT without entering the letdown line.

4. Add additional sUelding outside the auxiliary personnel

_. hatch to minimize potential effects at the Elevation 155' for the access control area and Technical Support Center.

The following is a discussion of the computer programs used for these analyses: .

1. Source term concentrations in uCi/cc for each isotope along with the volumetric source strengths in itev/cc/sec. were calculated using the NUCLYD computer program. This program calculates values of the specific activity and volumetric source strength at any given decay time for a given mixture or isotopes. The program also provides an integral energy release for that given mixture of isotopes from t = 0 to any specified time. This computer program is analogous to ORIGEN.

2. Dose rates were calculated with the CYLSO computer program.

This program uses the Rockwell point Kernel theory for one

,__ dimensional cylindrical volumetric sources. Self attenuation in the source as well as the shielding effects of various construction materials such as steel, lead, concrete and water are considered in the code. The code output is in tenns of dose rate vs. distance, for various piping diameters and shielding configurations.

3. This computer program is similar to the SDC code.

__ In addition to the above study, the effects of radiation on equipment are being considered as part of the IE Bulletin 79-01B and NUREG-0588 review.

Source tenns for LOCA events in which the primary system may not depressurize will be addressed during the above review.

The shielding design evaluation is a complex iterative process. All modifications listed above and modifications required to resolve the outstanding design issues will be completed for Units 1 and 2 by January 1,1982.

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30 II.B.3 POST-ACCIDENT SAf1PLING CAPABILITY l l

t Previous Response l By letters to the NRC of June 20, 1980 and August 1,1980 for Unit 2 and of October 24, 1979, November 21, 1979 and December 21, 1979 for Unit 1, Alabama Power Company provided information related to post-accident sampling capability.

I j Clarification Response As described in previous submittals, a post accident sample of reactor coolant is drawn from the same sample line used for the gross failed fuel detector. The gross failed fuel detector is normally on recirculation at i

all times. Any deposition or plating in these lines will be in equilibrium.

This system is isolated with the containment isolation signal but can be reinitiated without startup of the letdown system. Recirculation is accomplished by routing the sample return to the volume control tank. The sample line is 3/8" stainless steel tubing all the way from its origin in containment, therefore, loss of coolant would be limited to the flow through this 3/8" line in case of sample line rupture. The flow is l

normally limited to about 0.6 gallons per minute and the length of the sample line has been adjusted so t it a representative sample may be obtained approximately one minute hJter initiating sample flow. The sample return line has been rerouted to minimize exposure and line length. The I sample system was designed to produce a predetermined sample volume to prevent overflow and spillage. The sample system is completely closed and pressurized to the point where the sample is extracted. The sample is degassed and depressurtzed at this point. The gasses taken from the sample are used as the sample for H2 and 02 in the RCS as well as for The Noble gas analysis. The gasses excapino from the system during the sampling evolution are drawn into the radwaite HVAC system.

The post accident containment atmosphere sample is collected through the same sample lines that are used for the nonnal containment atmosphere monitors REll and RE12. These lines have been designed according to ANSI 13.1

. with no sharp bends in order to provide smooth laminar flow, to prevent impaction and to be as short as possible, according to the required location of sampling equipment, to minimize plateout. A plateout study, however, is in progress and a plateout factor will be included in the calculation of radioactivity content of the containment atmosphere if the results of that study indicate that plateout will be significant. The inlet line is located at about the 134' elevation in a protected area above a ventilation duct and below a steel grating with the actual inlet turned down so that no debris will be picked up which could block the sample line. The sample line is made of 1" I.D. rigid stainless steel tubing with isolation valves which can be closed in case of line rupture. The sample system includes a pump which returns the residue to containment. Any gases released during the sampling evolution will be collected in the radwaste HVAC system.

I 3l II.B.3(continued)

The post-accident sampling system and the containment atmosphere sampling

, _ system provide collection of small aliquots of the sampled media which will be shielded for transportation to the laboratory for analysis. Special procedures have been developed for analysis of highly radioactive samoles

__ which include the use of lead glass windows and manipulating apparatus that will insure that no analyst will receive exposures exceeding 3 and 18 3/4 rems to the whole body and extremities respectively. Both of the above systems will enable sampling and sam)1e analysis within three hours. The above analysis capability includes tie ability to detect 1) radionuclides in the reactor coolant system that ma 2) hydrogen gases, 3) chlorides

• and 4) y beconcentration boron indicators of core damage, of liquids. The ability of dilute samples is also provided. The onsite liquid sample analysis program has the capability to permit sensitivity measurements of nuclide concentrations in the range from approximately 1 pci/g to 10ci/g.

The sample analysis will provide results within a factor of two error.

Emergency implementing procedures describe how adequate infonnation will be provided to the operator to describe the post accident reactor coolant system radiological and chemical status.

The sampling system for reactor coolant and containment atmosphere for Unit 2 is installed, initial testing completed, and is anticipated to

, be fully operational by March 10, 1980 or no later than exceeding 5% power.

, ! ~- The sampling system for reactor coolant and containment atmosphere is installed and operational in Unit 1.

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• Due to the location of the Farley Nuclear Plant, arrangements have been made to perform offsite analyses within four days following sampling.

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l 32 II.B.4 TRAINING FOR MITIGATING CORE DAMAGE

\ Previous Response

-By letters dated July 29, 1980, August 6, 1980, September 24, 1980, September 25, 1980, and August 1,1980 for Unit 2 Alabama Power Company documented connitments and actions taken for the Farley Nuclear, Plant, l

3 Clarification Response This program will be completed for Unit 1 by October 1,1981 and prior to Unit 2 operation above 5 percent power.

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33 II.D.1 PERFORMANCE TESTING OF BOILING WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES Previous Response  ;

__ By previous response dated July 17,1980, July 23,1980, and August 1,1980, for Unit 2 and October 24, 1979 and December-31, 1979, for Unit 1 Alabama Power Company's described commitments and actions taken for the Farley Nuclear Plant.

Clarification Response As indicated in the December 15, 1980, le.cter from R. C. Youngdahl (EPRI) to D. G. Eisenhut (NRC), the present EPRI program does not formally include the testing of block valves. However, a small number of block valves have been tested at the Marshall Steam Station Test Facility, and a preliminary scope and cost estimate study for a block-valve test progrrm has been completed by the EPRI staff. A detailed block valve test

program will not be resolved until after July 1,1981. Alabama Power l Company will supply further details of this program as they become available.

While Alabama Power Company does .ot support additional ATWS valve

, testing until regulatory issues are res)1ved, the major test facility for I~ the EPRI program was designed to provide the potential for additional valve testing at higher pressures for ATWS conditions.

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i 34 II.D.3 DIRECT INDICATION OF RELIEF AND SAFETY VALVE POSITION i

Previous Response By letters dated June 20, 1980, and August 6,1980, for Unit 2; and October 24,1979, Novembar 21, 1979, December 31, 1979 and March 14, 1980 for Unit 1. Alabama Power Company described commitments and actions taken for the Farley Nuclear Plant.

Clarification Response The described system has been installed for Units 1 and 2. In addition, backup methods of determining valve position are available and are discussed in plant emergency procedures. The additional displays and controls added to the control room as a result of this NRC position were considered as part of the human-factor analysis conducted in response to NUREG-0737, Item D.1.1.

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II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION l

Previous Response Alabama Power Company by letters of November 10,-1979, November 14, 1979, April 1, 1980 and May 27~, 1980, for Unit I and June 20, 1980, for Unit 2 provided response for this item for the Farley Nuclear Plant.

Clarification Response The staff has recently completed an extensive review of the Farley Unit 2 AFW system. Alabama Power's letter of December 14, 1980, indicated concurrence with the staff's review and docketed commitments for both Farley units.

Alabama Power Company's letters to the NRC dated July 29, 1980, and September 8, 1980, for Units 1 and 2, respectively, document our commitment to provide train separation for the auxiliary feedwater flow control valve solenoid valves. This modification will be completed prior to operation above zero power for Unit 2 and prior to return to power from the present refueling outage for Unit 1.

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36 II.E.1.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION By letters of August 1,1980, April 1,1980, March 3,1980, April 21,1980, May 27, 1980, June 20, 1980 and August 6,1980, for Unit 2 and July 29, 1980, October 24, 1970, December 14, 1979, December 31, 1979, November 20, 1979, and December 4,1980, for Unit 1 Alabama Power Company provided response to this item related to the Farley Nuclear Plant.

i Clarification Response

Initiation The automatic initiation signals and circuits associated with the auxiliary feedwater system meet safety grade requirements (i.e., IEEE 279-1971, seismic and environmental qualification) with the exception of the isolation circuitry associated with motor-driven auxiliary feedwater pump auto-start on main

. feedwater purr.p trip. The non-safety grade portion of these circuits will be modified to include isolation from the safety grade portions through isolation devices in accordance with the requirements of IEEE'279-1971. It is intended that these modifications ~will be completed prior to startup following the current refueling outage but no later than July 1,1981, on Unit 1 and prior to exceeding 5 percent power on Unit 2. -

Indication Auxiliary feedwater injection lines to each steam generator are provided with safety-grade flow indication for both units. This flow indication is on the main control board and is powered from the plant emergency power. These flow instrument loops are testable. Redundancy requirements are met by qualified steam generator level instrumentation (safety-grade). A description of the presently installed equipment is provided below.

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Local and control room indication of auxiliary feedwater flow to each of the steam generators is provided by flow orifices in each auxiliary feedwater supply line, located just upstream of the auxiliary feedwater stop check valves.

The auxiliary feedwater flow indication is backed up by three redundant safety-grade wide range steam generator level channels, per steam generator which have control room readouts.

l Testing of this equipment is conducted in accordance with the Farley Nuclear Plant Technical Specifications. The auxiliary feedwater flow :ndication channels and steam generator wide range level channels are calit. rated every 18 months. The steam generator narrow range level channels are functionally tested every 31 days and calibrated every 18 months.

The displays and controls associated with auxiliary feedwater system flowrate indication were considered as part of the human-factor analysis conducted in response to NUREG-0737, Item I.D.1.

37 II.E.1.2(Continued)

The auxiliary feedwater flow instrumentation channels and the steam generator narrow range channels receive their power from the Class lE vital instrument buses. The steam generater wide range channels also receive their power from vital instrument Loses.

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l 38 II.E.3.1 ENERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS Previous Rasponse By letters of June 20, 1980 for Unit 2, and October 24, 1979, December 31, 1979 and March 14, 1980 for Unit 1. Alabama Power Company documented comitments and actions taken for the Farley Nuclear Plant.

Clarification Response These heater groups for both units have the capability of being powered from the emergency section of 600V load centers 1A (2A) and 1C (2C), respectively, which constitute one C1 cts 1E division power supply each. The pressurizer heaters are normally supplied by offsite power. Upon loss of offsite power.

the pressurizer heaters require manual loading onto the diesel generators by operation of breakers controlled from the control room. Reset of a safety injection signal, if present, is not necessary to permit operation of the heaters since load shedding of the heaters is not initiated from the presence of such a signal. The connection of the preselected pressurizer heater to the emergency buses is accomplished by alignment of backup heater group A or B within 60 minutes of the loss of offsite power.

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i 39 II.E.4.1 DEDICATED HYDR 0 GEN PENETRATIONS Previous Response By letters of June 20, 1980 and August 1, 1980 for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979 and March 14, 1980 for Unit 1, Alabama Power Company has provided response on this for the Farley Nuclear Plant.

Clarification Response Since the Farley Nuclear Plant design does not utilize extern.sl recombiners, nor does it primaril rely

, on a purge system for combustible gas control, no further clarification is required.

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II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY 40 l 4 Previous Response By letters of June 30, 1980, for Unit 2 and October 24, 1979, December 31, 1979, November 21, 1979, and March 14, 1980, for Unit 1, Alabama Power Company responded to this item for the Farley Nuclear Plant.

Clarification Response The containment pressure-high (CP-H) setpoint is selected to limit the maximum pressure inside containment following a design basis accident, and to provide early isolation in the event of an accident but must be set high enough to avoid spurious isolation. Technical Specification 3.6.1.4 requires that the con-tainment internal pressure be limited to a maximnn of 3.0 psig under normal

, operating cor.11tions. The 3.0 psig limit is required to assure that the containment peak pressure does not exceed the design condition pressure of 54 psig during the LOCA condition.

All safety analyses affected by the CP-F setpoint have been reviewed. Based on this review, it has been determined that a decrease in the CP-H setpoint would have no adverse effect on any safety analyses. Thus, the analyses presented in the Farley Nuclear Plant FSAR remain conservative.

For large LOCAs, SI and containment isolation signals are generated almost

• ;- immediately from both low pressurizer pressure and CP-H. Lowering the CP-H setpoint would result in an insignifica~nt difference in signal generation time of only a few tenths of a second.

. Several categories of small break LOCA have been considered. For LOCAs within the capability of the charging pump there would be no core uncovery. For small break LOCAs beyond this range, SI and containment i W tion would occur before

! core uncovery begins. Since core damage and the subuquent release of large amounts of activity are predicated on core uncovery, the activity released

.- to the containment prior to core uncovery would be the activity normally present t in the reactor coolant system. Activity in t . amount would not significantly affect site boundary limits or plant habitability. If high activity in con-tainment resulted from an accident of this nature, purge and exhaust would isolate independent of any other protective signal closing off the containment to atmosphere path. SI and containment isolation occur much sooner than any significant fuel damage that would be caused following core uncovery. Reducing the CP-H setpoint may, for certain breaks, cause SI and containment isolation to occur sooner but the benefit to the plant, the operating staff, and the public would be questionable.

l The minimum obtainable CP-H setpoint would be the sum of the maximum allowable L normal operation containment pressure (3.0 psig) plus the instrument error of the l

containment pressure transmitter. The containment pressure transmitter error is

! equal to 2.5 percent of the instrument span based on instrument and rack drift, l

calibration' accuracy and temperature effects (-5 to +65 psig) or 1.8 psig. Thus, it is possible to reduce the CP-H setpoint to 4.8 psig (5.4 psig is current setpoint).

4I II.E.4.2 (Continued)

If the setpoint was reduced to 4.8 psig, a setpoint reduction of only 0.6 psig, there would be no margin for minor containment pressure increases associated with ,

plant operation and the probability of spurious safety injections would be significantly increased.

Based on the above information, along with the corresponding higher probability of safety injection initiation, Alabama Power Company has concluded that reducing the current CP-H setpoint is neither advantageous nor advisable.

Alabama Power Company's proposed actions concerning recent NRC Staff positions . .

related to the containment purge system were transmittad to the NRC Staff by a letter from Mr. F. L. Clayton, Jr. , to Mr. A. Schwencer, dated September 30, 1980. These proposed actions are currently under review by the NRC Staff.

This matter will be resolved with respect to Farley - Units 1 and 2 upon completion of the NRC Staff's review.

Farley Nuclear Plant - Units 1 and 2 containment purge and exhaust isolation valves were originally designed to close on receipt of a high radiation signal.

In response to I & E Bulletin 80-06, Engineered Safety Feature (ESF) Reset Controls, Alabama Power Company addressed valve position associated with reset of ESF equipment in our letters of September 29, 1980, and September 12, 1980, for Unit 2 and June 12, 1980, for both units.

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II.F.1 ADDITIONAL ACCIDENT MONITORING INSTRdMENTATION Previous Response By letters dated August 1, 1980, August 19 1980, June 20, 1980 and July 24, 1980, for Unit 2 and October 24, 1979, November 21, 1979, December 31, 1979

and March 14, 1980, Alabama Power Company described commitments and actions taken for the Farley Nuclear Plant.

i l Clarification Resp- .se Noble Gas Effluint Monitor i A. Vent Stack Monitor

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Alabama Power Company will install for both units an Eberline Sping 4 sampler to monitor noble gases in the plant vent stack. This sampler has a range of 10-7 to 105uCi/cc using multiple detectors. The. monitor draws a sample from

} the vent stack to a monitor unit located in the mechanical equipment room at

. elevation 175 of the auxiliary building. The readout for this unit is located

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in the main control room. An auxiliary readout is located in the low activity counting laboratory.

The noble gas measurement is performed by several detectors viewing a sample H volume. The low and medium range detectors view the same sample volume located in the SA-13 sampler assembly. The high range detector views the sample volume located in the SA-9 sampler assembly.

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(1) LOW RANGE NOBLE GAS: The gas chamber is monitored by a BETA scintillation detector (Eberline Model RDA-3A). Background correction for this channel is derived from the gama background detector, an energy-compensated GM detector (Eberline Model

.10450-B28). Since the external-(ambient) gamma radiation has a measurable effect on the BETA measurement (particulate and gas),

the gamma background channel is used as a source of subtraction for both the gas measurement and the particulate measurement.

(2) MEDIUM RANGE NOBLE GAS: An energy-compensated GM detector monitors the gas volume for the medium range noble gas ceasurement, with its output proportional to the gamma content of the sample. An additional identical detector is provided in the sampler shield as a measure of the external background at the sampler; this is the background detector. Thus the effects of a fluctuating external background on the medium range gas channel are nullified by measuring and subtracting the background.

43 II.F.1(Continued)

Noble Gas Effluent Monitor (Continued)

(3) HIGH RANGE NOBLE GAS: An energy-compensated GM detector monitc:s the gas volume of a section of 1" stainless steel tubing for the high range noble gas measurement. Its output is proportional to the gama content of the sample. -

An area monitor radiation detector assembly (Eberline Model DAl-1-CC) is mounted on the Sping 4 and provides a measure of the gama field at the instrument. This detector is an energy-compensated GM tube and is calibrated in radiation dose rate. Calibration is by use of an external calibration source and is performed upon installation and at intervals not exceeding each refueling outage. The Eberline Sping 4 monitor is capable of functioning both during and following an accident. Frequent filter replacement will ensure operability of the monitor's electronics after an accident. The monitor's accuracy is 2% of span.

The following list sumarized by channel number and type which calibration sources are provided.

CHANNEL CHECK SOURCE I

Number Type Content Isotope 1 Beta Particulate 30 microcuries 137Cs s 2 Alpha Particulate 3 Iodine (Gamma) 0.5 micurcurie 133Ba 4 Iodine Subtraction (Gama) 5 Beta Gas (Low Range Noble Gas) 30 microcuries 137Cs 6 Gama Area 0.5 microcurie 903 90y 7 Gama Gas (Medium Range Noble Gas) 8 Gama Background 9 Gama Gas (High Range Noble Gas) .05 microcurie 903 90y The plant vent noble gas concentration in pCi/ml is determined by sampling and/or by obtaining a value from the plant vent stack high range monitor.

The plant vent flow rate is decemined by the r. ..L2r of operating auxiliary building exhaust fans. The release rate in curies per second is detemined by the following equation:

Release rate (Ci/sec) = Concentration (pCi/ml) X flow rate (cfm) X conversion factor

44 II.F.1(Continued)

Noble Gas Effluent Monitor-(Continued)

The~above method to detemine noble gas release rate is described in emergency implementating procedures. During emergencies the release rate is calculated periodically as directed by the Emergency Director to determine if the accident classification should be upgraded.

The monitors have been environmentally qualified by the vendor.for the environment in which it is located.

B. Main Condenser Air Removal Monitor (SJAE)

The main condenser air removal exhause systems for Units 1 and 2 are monitored using the existing monitor (described in the FSAR) on the steam

~ jet air ejector exhaust for the nomal range of radioactivity. The accident range of radioactivity will be monitored for Units 1 and~2 by intermediate and high range detectors with overlapping ranges and located at the common vent duct for the turbine building. The accident monitor consist of 2 Eberline detectors and readouts. The intermediate range detector will be

- model Dal-1CS with an ED1-1 readout module with a range of indication of 0.1 to 100 mR/hr. The high range detector is a 'model Dal-4CS with an ECl-20 readout module with a range of 10 mR/hr. to 1,000 R/hr. The

relationship between mR/hr. and uCi/cc will be established for the ncole gas isotopes present during an accident. The range of the accident monitors in uCi/cc is from 10-5 to 103 with the normal range monitor measuring

!' concentrations down to 10-6 uCi/cc. This is the required range for the case where the SJAE exhaust is combined with turbine building ventilation exhaust.

F- The readout modules will be located in the control room and will provide l' continuous indication. The accident Atectors will be shielded from back-ground radiation with 6 inches of lead. Calibr:a. ion is by use of an external j calibration source and is performed upon installation and at intervals not exceeding each refueling outage.

C. Steam Generator Atmospheric Relief and Safety Valve Monitors The discharge from steam generator safety relief valves and atmospheric

- dump valves for Units 1 and 2 will be monitored by measuring the radiation levels from these steam plumes. There will be four Eberline model DAl-4CS detectors per unit mounted on the main steam roof with a range of 10 mR/hr.

to 1,000 R/hr. The relationship between mR/hr. and pCi/cc has been established for the noble gas isotopes present during an accident. The range of the monitors in uCi/cc will more than cover the r.equired range from 10-1 to 103 for cases with just the PORC open to cases with the PORV and all safties open. Each detector will be connected to an Eberline ECl-20 readout module in the control room, providing continuous i indication. Since the safety relief valve and atmospheric dump W ye discharges are grouped together for each of the three steam generators, ,

one detector will be used to monitor the combined effluent steam plume from each steam generator. The fourth detector is used to monitor the plume from the steam driven auxiliary feedwater pump turbine exhaust. Each detector is collimated and background shielded with 7.5 inches of lead. Calibration is by use of an external cal,vation source and is performed upon ir.s allation and at intervals not exceeding each refueling outage.

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. 45 II.F.1 (Continued)

(Noble Gas Effluent Monitor (Continued)

D. Design and Installation Schedule for Nobles Gas Effluent Monitors The noble gas effluent monitors will be powered from a vital instrument bus. Procedures will be developed for use, calibration of the system, and dissemination of release rate information. The Sping-4 for both units is onsite hardware to support installation of the main condenser air removal monitors and the steam generator atmospheric relief and safety valve monitors are onsite or in the process of being shipped.

The installation of the vent stack monitor for Unit 2 is scheduled for March 10, 1981, or prior to exceeding 5 percent power. This instrument is currently scheduled for installation in Unit 1 for prior to the end of the current refueling outage but not later than January 1,1982.

The original Alabama Power Company position was to monitor the main condenser air removal exhaust and the discharge from the steam generator safety relief valves and atmospheric relief valves with a portable gamma survey instrument. Alabama Power Company, however, finalized the above position based on NRC questions during the latter part of 1980. Based on the current material availability and status of the complex shielding design required, installation for both units is scheduled for completion by January 1, 1982. Alabama Power Company purchased the best available monitors upon finalization of this position. In order to ensure accurate reading of each of these monitors, a complex shielding design is required to discriminate actual readings from background including containment shine.

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II . F.'1 (Continued) 46 Sampling and Analysis of plant Effluents Alabama Power Company has the capability to provide continuous sampling of plant gaseous effluent for post accident releases of radioactive iodine and particulates at the plant vent and the condenser air removal system. The sampling method involves passing the effluent gases through a filter assembly

_ and transporting the filter to a counting room for analysis. The sampling system has the following capabilities:

(1) Effective iodine absorption of greater than 90% for all forms of gaseous iodine.

(2) Greater than 90% retention of particulates for 0.3 micron diameter particulates.

(3) Design intent meets sampling requirements of ANSI N 13.1-1969.

(4) Continuous collection whenever exhaust flow occurs.

(5) Analytical facilities and procedures considered the design basis sample.

l (6) Shielding factors were considered in the design.

!~

On-site laboratory capability exists to analyze or measure these samples. The sampling system design is such that plant personnel can remove samples, replace sampling media, and transport the samples to the on-site analysis facility with radiation exposures that are not in excess of the GDC 19 criteria

of 5 rem whole body and 75 rem to the extremities during the duration of the accident assuming the' design basis shielding envelope.of NUREG-0737.

The Eberline Sping 4, which samples vent stack effluents, uses an isokinetic nozzle in the stack to draw its sample into its filter system and the flow rate can be adjusted at the pumping unit to attain a sample velocity that will match stack flow rates. There are presently two exhaust fans that detemine effluent velocities. In addition, tWe will be a Victoreen vacuum pump with charcoal filters that will allow th .emistry and Health Physics Group to draw 15 minute iodine and particulate sa. >les to be analyzed in the laboratory.

This pump has bypass lines that allow drawing an isokinetic sample by passing portions of the sample back to the stack.

The steam jet air ejector sample point is located on the vertical section of the turbine building exhaust ventilation duct. Locating the sample point on the vertical section of the'cxhaust duct ensures that the absorber material is not degraded with entrapped water.

.- J A .-

l 47

, II.F.1 (Continued) l l

Sampling and Analys.. of Plant Effluents (Continued) l The primary sampling system for the vent stack (Sping-4) is scheduled to be installed by March 10, 1981, but prior to exceeding 5 percent power for Unit 2 and prior to return to power in Unit 1 following the current

-. refueling outage but no later than January 1, 1982, to provide indication (uCi/ml) in the main control rooin and the counting room. ,

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II.F.1 (Continued) 48 Containment Pressure Monitor The present containment pressure indication provides continuous redundant indication in the main control room and has an indication range of -5 psig to 60 psig. Additional monitoring capability with control room indication having a range of 0 to 210 psig is scheduled to be installed for Unit 1 by return to power after the current refueling outage but no later than January 1, 1982, and for Unit 2 by March 10, 1981, but no later than exceeding 5 percent power. Continuous display and recording of the containment pressure is provided in the control room. The indication accuracy of both the wide and narrow range instruments is13.5% with a response tima of less than 180 milliseconds for a 10% to 90% step function change in pressure.

The environmental qualfication for these items are being addressed as a part of Alabama Power Company's response to I.E. Bulletin 79-01B and NUREG-0588.

i i

I l _

l l

e

49-II.F.1 (Continued)

Containment Water Level Monitor The Farley Nuclear Plant present design has two wide range containment (ECCS sump)waterleveldetectors. These detectors provide indication in the main control room that meets the wide range requirements as specified in the

- various clarification letters. These level transmitters and associated readout are safety grade and measure volumes up to and above 600,000 gallons.

In addition, a narrow range containment (reactor vessel cavity sump) level

, system meeting the various clarification letters is scheduled to be installed for Unit 1 by return to power after completion of the current refueling outage but no later than January 1,1982. The schedule for Unit 2 installation is March 10,1981, or no later than exceeding 5 percent power. The accuracy of the narrow range level instrumentation is + 1/2 inch with an instantaneous response time. Qualification will be addressed in Alabama Power Company's response to IE Bulletin 79-01B and NUREG-0588.

i 1

s O

P 50 II.F.1 (Continued) -

Containment Hydrogen Monitor Two independent, redundant systems for containment hydrogen monitoring are

-~

provided for Units 1 and 2. The design of these systems meets the requirements for safety-related protective systems as defined by IEEE 279-1971. The output signal of the analyzers are indicated at the analyzer panel location and are alarmed and recorded in the main control room. Each system is supplied

, electrical power from an independent and redundant Class 1E Power Supply.

The system meets the single failure criteria and remains operable under the postulated accident. Any single failure in one hydrogen monitoring system does not affect its redundant and independent counterpart. The accuracy of the hydrogen monitor is +2% of span with a response time of 0.45 minutes. The range of indication Ts 0-10%. Qualificaton requirements are being addressed in Alabama Power Company's response to I.E. Bulletin 79-01B and NUREG-0588.

The indication and recording of hydrogen concentration will be initiated as required by emergency procedures in less than one hour after a safety injection initiation.

m l

I

51 II.F.1 (Continued) i Containment High-Range Radiation Monitor l Alabama Power Company has ordered redundant Victoreen Model 875 Radiation l Detection Systems to meet the requirements for a high containment radiation monitor. Each system consists of an ion chamber detector, readout panel, and interconnecting cables. The monitors will be located inside containment about six feet above the operating deck and approximately 900 apart. These locations ensure the monitors are not protected by massive shielding and that they will provide a reasonable assessment of area radiation conditions inside the containment during and following an accident.

(a) Each detector is designed to measure gamma radiation.

(b) The range of each detector is 1 R/hr. to 107 R/hr. for photon radiation.

(c) The energy response is -15% to 80 key and 8% from 100 key to 3 Mev.

l l (d) The calibration frequency will be at a maximum interval of 18 months. Presently it will be necessary to return the monitors to the vendor for calibration.

(e) The containment high radiation monitors are being installed and should i- be operational for Unit 2 by March 10, 1981, or prior to exceeding 5 percent power. Such monitors are being installed and should be operational prior to return to power following the current refueling outage for Unit 1 but no later than January 1,1982.

Victoreen has completed the preliminary qualification review and is near completion for the final qualification program. The radiation monitors satisfy the requirements of the vendor qualification program. The only remaining component to be qualified is the electrical connection between the power cable and the radiation monitor. Victoreen is currently in the process of qualifying this component. Alabama Power Company will update the NRC on the qualification program as information becomes available.

Capability exists for on-site calibration of the radiation monitor to 10R/hr.

Calibration above 10R/hr. will be completed by utilizing an electronic signal.

As part of the vendor testing program, Victoreen has stated that at least one point per decade of the range between 1 R/hr..and 103 R/hr. had the calibration certified.

l l

i 52 II.F.2 INADEQUATE CORE COOLING INSTRUMENTATION Previous Response By letters of December 31, 1979, for Unit 1 and of July 17, 1980, July 24, 1980, June 20, 1980, August 19, 1980, August 6, 1980 and August 1, 1980, for Unit 2, Alabama Power Company provided response to this item for the Farley Nuclear Plant.

Clarification Response The Westinghouse Owners Group also provided the NRC with additional procedural guidelines via letter of November 10, 1980 (0G-44) relating to Inadequate Core Cooling. The target date for plant specific implementing procedures is July 1,1981 for the Farley Nuclear Plant.

A. Vessel Level Alabama Power Company, in coordination with EPRI, has undertaken a vessel level measurement program for the Farley Nuclear Plant to demonstrate the capability of a non-invasive prototype system on an experimental basis.

EPRI wil', continue to fund the analytical investigations associated with the prototyne demonstration program and will assist APCo in establishing the necessary test programs and evaluation of the collected date to determine the feasibility of such a system.

e i

Initial testing on a temporary special test setup for Farley Unit 1 was performed during the period of November 10-14, 1980. The objective of the test was to obtain definitive measurements of the relationship between neutron count rate above the reactor vessel water level. A second objective was to discover which variables such as core reactivity, shielding, and neutron background affected this relationship.

~

GENERAL DESCRIPTION OF THE NON-INVASIVE WATER LEVEL MEASUREMENT SYSTEM A description of the non-invasive water level measurement system was submitted to the NRC for its review and approval by Alabama Power Company letter dated August 6, 1980. This is a prototype system unique to the Farley Nuclear Plant, and its use has been supported by the NRC in issuance of Supplement 4 to the Safety Evaluation Report associated with the low power l operating license for Farley Unit 2.

The non-invasi water level measurement system consists of a set of externallymountedgBF3 neutron detectors above and below the reactor vessel.

The principle used in the detection of photoneutrons from the reaction of high energy gammas with the deuterium impurity present ir, the reactor coolant system. A simplified diagram of the system is shown in Figure 1. Each detector setaboveorbelowthe5eetrvesse c nsists f eight (8, 2-inch diameter, 24-inch active length I BF3 filled thermal neutron counters. These detectors are made from stainless steel and are filled to 70 cm. Hg. pressure. They are shielded by a 1/2-inch thick lead sleeve, and they are surrounded by a plastic moderator.

II.F.2 (continued) 53 Additional 1/2-inch thick lead shielding is provided between the vessel and the detectors. The ratio of the count rates from these two sets of detectors is used to determine the water level in the vessel above the core.

The detector assembly above the reactor vessel consists of two counters sheathed in 0.065-inch thick steel. Figure 2 shows the preliminary sketch of the cross-section of the top detector assembly. Four detector assemblies are then mounted in position above the vessel head as shown in Figure 3. The top detectors are expected to have a combined sensitivity of approximately 500 counts per NV. The bottom detectors are mounted in the area below the reactor vessel. All eight of these detectors are consolidated into a single assembly as shown in Figure 4. tive for use during operating power levels, a pair of small, less sensitiveSincetheselarge lined detectors will be used to measure level during power operatian, if necessary. While details of electrical wiring diagrams for the sysem have not yet been finalized, Figure 5 shows the features that will be incorporated in this sytem.

The 10 BF3 detector assemblies are divided into pairs with their associated preamplifiers so that the redundancy requirements will be satisfied. The amplifiers drive two counter timers. One counter operates on the top detectors, set-up for a preset number of counts. Operation of this counter gates the j second counter which accumulates counts from the bottom detector during the period when the first counter is counting a preset number of counts. Consequently, the counts from the second scaler is proportioned to:

Bottom Count Rate Top Count Rate The reduction in the amount of water above the reactor core increases the top count rate, thereby reducing ';he ratio. Thus, the second counter display counts increase with increased water level.

Background:

ansorship, the National Nuclear t Corporation (NNC) conducted tests 2using aDuring the Sumer of 1979, under mockup to determine whether neutron measurements outside a reactor vessel would provide an unambiguous measure of the water level inside the vessel. The method chosen is shown schematically on Figure 1. An array of neutron detectors were placed above and below the vessel, and the ratio of the counts from these arrays were related to the water level. Figure 6 shows count rates measured l above the vessel as a function of water level. The initial rapid drop off for about four feet was due to shielding, by the water, of fission neutrons from the source (or reactor). Beyond four feet, neutrons were principally produced from the action of high energy (over 2.2 MeV) gamma rays on the deuterium within the water. Since these gammas travel further than neutrons in water, these photoneutrons predominate when the water in the tank was deeper than four feet over the core. When the counts from the lower detector are divideri by counts from the upper detector (as in Figure 7), a relationship roughly proportional to water depth is obtained, except just before core uncovery when a much greater effect is observed.

4 II.F.2(continued) -

. 54 l

l As shown by this data from tanks tests at NNC, for water levels over four .

feet above the core, most of the neutrons detected arise from interaction of '

high energy feet (where gamas there with the is a danger of deuterium impurity)in core uncovery the neutron thelevel water, while above the below five reactor rises very rapidly due to neutrons produced by fission in the reactor ,

core. Thus, this system provides a vivid warning well before core uncovery.

Less dramatic indications are provided to gauge water level in the range

. - between five feet and full. This is shown in the correlation on Figure 7. In this test, top counts were referenced against a side d2tector, instead of the bottom detectors used in the actual installation.

Following these tests and additional tests at Prairie Island and Rancho "

Seco, equipment was built and used to demonstrate successfully the operation of this system at Trojan during initial drain-down. Data from the Trojan test is shown on Figure 8. Based on this data, and on Trojan side counter data, the curve on Figure 9 has been projected to indicate the perfonnance.of the i actual system one day after shutdown.

Following the Trojan test, analytical studies have been made to further investigate the system's capabilities. It has bem brought out that the system reads weight of water above the core, thus giving a valid indication indication before core uncovery.

The preliminary system discussed above will be installed in Farley Unit 2 prior to exceeding 5 percent power. In order to provide are indication of

'- lowering reactor vessel water level on Unit 1, an abbreviated version of the system installed on Unit 2 is being installed on Unit 1 during the current refueling outage. This system will consist of one detector assembly installed above the Unit i reactor vessel head and an alarm set at a predetermined count rate to give an indication of decreasing water level. Table 1 is a schedule of development, installation, and testing of the vessel level systems for Units 1 and 2.

The detector installation fg the test consisted of four (4) detector assemblies, each containing two BF3 counters, distributed around the reactor vessel head area on top of trie vessel head insulation.: After the detectors were installed and calibrated, the reactor vessel water level was lowered in two (2) foot intervals until reactor vessel water level was

at the centerline of the vessel nozzles. The water level was then raised j in two (2) foot intervals until the vessel was full. At each level a 1,000 second count was taken. The test was repeated with the detectors shielded for a total of three test runs. ,

The results of the tests on Farley Unit 1 demonstrate that neutron detectors mounted above the reactor vessel respond to changes in water level within the vessel. Improvements in the system may be made through better threshold adjustments of the detectors and detector shielding. Areas that require further investigation include a clearer understanding of the mechanism whereby neutrons reach the detectors when the core is covered by 10-20 feet of water. Tests such as those planned at LOFT should yield a better understanding of behavior at lower water levels, shorter times after shutdown, and under transient conditions.

. --,.-.- -- - --a

II.F.2(continued) 5d Following the installation of the complete system on Farley Unit 2, the effects of density change with temperature will be investigated during plant startup. Also, during a full-power operation period, the detectors will be evaluated for their performance. Additional necessary water level measurement tests will be performed during unplanned outages, but no later than the next refueling outage currently planned to commence during the fall of 1982. During this entire period continued engineering evaluations I of the system will be performed towards improvement of the design, installation, I and qualification of the system. If the system proves to be viable and reliable during the Unit 2 tests, then the same system will be installed in Unit 1 during the refueling outage currently planned to cortnence in the fall of 1982. However, if the system fails to meet performance expectations, then Alabama Power Company will install, in both units, the best system available as rapidly as practicable.

It is recognized that the non-invasive water level system is still in its developmental stage and the effort undertaken by Alabama Power Company could be construed as the first prototype field testing program.

As further information becomes available, Alabama Power Company will keep the NRC fully informed. If the test program shows that this reactor vessel water level system is a viable and reliable system, Alabama Power Company will make additional submittals to the NRC to provide a suninary of the key operator action instructions in the emergency procedures for inadequate core cooling and to demonstrate the ability of the system to provide unambiguous, easy-to-interpret indication of inadequate core cooling. These additional submittals will also describe the program to qualify the reactor vessel water level

- system in accordance with Appendix B of NUREG-0737 and the human factors considerations in the design of the vessel level displays. Such qualification will be " cormed in conjunction with EPRI.

M e

O

l II.F.2(continued) 56 TABLE 1 ,

I REACTOR VESSEL LEVEL SYSTEM DEVELOPMENT AND PLANNED INSTALLATION SCHEDULE FARLEY NUCLEAR PLANT FARLEY NUCLEAR PLANT UNIT ACTIVITY SCHEDULE DATE Unit 1 REACTOR VESSEL DRAIN DOWN November, 1980 Unit 2 INSTALL PROTOTYPE SYSTEM Prior to Exceeding 5% Power Unit 1 REACTOR COOLANT SYSTEM HEATUP End of second refueling outage TESTS Unit 1 COMPLETE INSTALLATION OF End of second refueling outage ABBREVIATED PROTOTYPE SYSTEM I

Units 1 COMMENCE DEVELOPMENT OF March, 1981 and 2 PERMANENT SYSTEM Unit 1 ADDITIONAL TESTING AND DATA Forced Outages (between end of r-COLLECTION WHEN FEASIBLE second refueling and beginning of third refueling)

Unit 2 PROTOTYPE TESTING WHEN Forced Outages (between initial i FEASIBLE criticality and beginning of first refueling when decay heat l is available)

Unit 2 FINAL TESTING OF PROTOTYPE Refueling Outage 1982 SYSTEM Units 1 COMPLETE DEVELOPMENT OF January 1, 1982 and 2 PERMANENT SYSTEM Unit 1 INSTALL PERMANENT SYSTEM Refueling Outage 1982 l

57 II.F.2(continued)

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1 h II.F.2 (continu:d)

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l Figure 9 Perfo' rman'ce Of Top . Detectors In Final System 8 Stainless. ' l

                                                                                      ',Dete'ctors 1 Day Af.ter Shutdown Hot Coolant
                             *~ *=: ua. . . . . .               +
                                                                                                    ..f...~
                                                                                                                       ....-------u-                                                                                                   ~
                                                                                                                           """"?!'FF"r""N"""?!"!!!!"/E'?"*!!!I ooee e q _a 2.o   v.__L .. .. .u_t_umumu nu:nu::=mturnmpwynr                                                                                               ""/-        #E""# MW-                  9  o ME"DY " #9

• i II.F.2 (continued) l 66 B. Core Subcooling Meter The installed core subcooling monitors have been tested by a calibration l

, and functional test procedure for both units. These tests are a comprehensive l software and hardware performance verification which includes initial calibration of inputs and functional testing of microprocessor self test features, calculation outputs, display capabilities and alarm outputs. The maximum thermocouple indication error found during testing of the Unit 1 Subcooling lionitor was 7oF. at 20000F in a conservative (high) direction. The maximum error found during the Unit 2 test was 60F. C. Incore Thermocouples As a result of Alabama Power Company's response to I&E Circular 80-15, a modification was implemented for reallignment of certain thermocouples to the upper head region as a means of indicating upper head voiding during natural circulation. There are 16 themocouples inputs for the core subcooling meter (eight per channel )two of which are upper head themo-couples. Westinghouse will initiate a generic program in January,1981, to qualify the present thermocouples to the requirements of IEEE 323-1974. Alabama Power Company will notify the flRC as more information becomes available regarding a schedule for completing this program. All of the control equipment for the Thermocouple System is located e cn a rack in the control room. A multipoint precision indicator has been provided to indicate the temperature sensed by the themocouples. Only one thermocouple at a time can be connected to the indicator. Switches have been provided on the front of, and above, the indicator to select that thermocouple desired to be read. An additional selector switch located on the front of the panel allows either the low (100-4000F) or high (400-7000F) range measuring circuit to be used. Besides being directed to the indicator, the thermocouple outputs are also applied to tne plant computer (up to 19000F.). D. Incore Thermocouple Readout Panel Initial Testing i Testing of the Incore Thermocouple Readout Panel was performed for Unit 2 by functional test procedure. Thermocouple inputs and readout panel performante was verified to be within expected tolerances. Input verification was performed on each thermocouple input. Maximum thermocouple indication error over the temperature range tested was l 4.50F in the conservttive (high) direction on Unit 2. Similar tests l will be performed for Unit 1 at the next refueling outage. l l l l

I II.F.2 (continued) 67 Periodic Retesting Retesting is to be accomplished under the Preventive Maintenance program and will be perforned at refueling intervals. Diagnostic Capabilities There are no specific diagnostics internal to the Incore Thermocouple Readout Panel. Good industry practice will be utilized to ensure operability of the readout panel. E. Plant Process Computer (P2500) Initial Testing The plant process computers for both units have been tested using a combined input verification and acceptance test procedure. These tests are a comprehensive software and hardware performance verification which includes calibration of inputs, individual program performance testing, and calculated value, display, and alarm outputs verification. The thermocouple program package and incore therinocouple inputs are verified and calibrated at this time. Unit 1 and Unit 2 preoperational test results indicated errors of less than 40F over the range tested. Periodic Retesting

            .              Incore thermocouple input verification to the process computer will be

, t' performed at refueling intervals beginning at the next refueling outage. Diagnostic Capabilities Incore thermocouple inputs are monitored and alarmed by the process

computer for sensor input failure and temperature alarm limits, i

Thermocouple program outputs are also monitored by the computer f,or alarm limits. Preventive maintenance and good industry practice is utilized to ensure process computer system operability. l

l l l 68

II.G.1 EMERGENCY POWER FOR PRESSURIZER EQUIPMENI Previous Response By previous letters dated June 20, 1980 for Unit 2 and October 24,

         ~ December 31, 1979 and March 14, 1980 for Unit 1, documented commitments
  . _ _    and actions taken for the Farley Nuclear Plant.

Clarification Response Power Supply for Power-Operated Relief Valves (PORV) i Each power-operated relief valve (PORV) is equipped with two solenoid I valves. Solenoid valves SV 0445AA-A and SV 0445AB-A associated with

.          PORY PCV 445A are powered by Train A 125V D.C. system.lA (2A) which is

! supplied by Train A 125V D.C. bactery LA (2A) or from offsite power through l a Train A battery charger. On LOSP the Train A battery charger will be

;          automatically transferred to a Train A diesel generator. Similarly, solenoid valves SV 0444BA-B and SV 0444BB-B associated with PORV PCV 444B are powered by Train B 125V D.C. system 1B (2B) which is supplied by Train B 125V D.C. battery IB (2B) or from offsite power through a Train B battery charger. On LOSP the Train B battery charger will be automatically transferred to a Train B diesel generator. Therefore, the PORV'S ass 0Ciated circuits meet the separation requirements for redundant systems. In addition, the motive had control power interfaces with the emergency buses are accomplished through safety grada devices. Accordingly, this design

provides the capability to either open or close the PORVs.

Power Supply for Block Valves l

Each PORV has a motor-operated block valve located in its respective piping.

l MOV 8000A-A receives its power supply-from Train A emergency 600V NEC IU (2U) l_ which is energized by Train A emergency 600V Load Center ID (2D). Similarly, MOV 8000B-B is powered by Train B emergency 600V MCC IV (2V) which is energized i by Train B emergency 600V Load Center lE (2E). On LOSP automatic transfer is made to the Train A and B diesel generators, respectively. Therefore, the block valve associated circuits meet the separation requirements for redundant systems. In addition, the motive and control power interfaces with the emergency buses are accomplished through safety grade devices. Power Supply for Pressurizer Level Indication Regarding the level indicators, there are three channelized pressurizer level indicators that receive signals from the corresponding channelized level transmitters. LI 459 and LI 460 receive their power supply from 120V A.C. Vital Buses lA (2A) and IB (2B), respectively. Accordingly, these buses are supplied by Train A emergency MCC 1A (2A) which is energized by Train A emergency Load Center ID (2D). Similarly, level indicator LI 461 receives its power from 120V A.C. Vital Bus 1C (2C). Accordingly, this bus is supplied by Train B emergency MCC IB (2B) which is energized by Train B emergency Load Center lE-(2E). On LOSP automatic transfer is made to the Train A and B diesel generators, respectively. l

69 II.K.1 IE BULLETINS ON MEASURES TO MITIGATE SMALL-BREAK LOCAs AND I,055 0F FEEDWATER ACCIDENTS (II.K.1.5, II.K.l.10. II.K.l.17) Previous Response By letters of June 20,1980 for Unit 2 and April 24,1979 and August 28, 1979 for Unit 1, Alabama Power Company has responded to this item related to the Farley Nuclear Plant. Clarification Response

• NUREG 0737 does not require clarification to this item. Administrative procedures require that the reactor operators walkdown the main control board as a part of normal shift turnover. Double valve verification of equipment is discussed in section I.C.6.

 ~

i 9

70 II.K.2.13 THERMAL MECHANICAL REPORT - EFFECT OF HIGH PRESSURE INJECTION ON i 4 VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENI WITH NO AUXILIARY FEEDWATER l i Previous Response Orders on B&W plants have not previously been required to be addressed for Westinghouse plants. Clarification Response . I To completely address the NRC requirements for a detailed analysis of the thermal-mechanical conditions existing in the reactor vessel during recovery from small breaks with an extended loss of all feedwater, Alabama Power Company, as a member of the Westinghouse Owners Group, is participating in a program consisting of analysis for generic Westinghouse PWR plant groupings. The program will be completed and documented to the NRC by January 1,1982. Following completion of this generic program, additional plant specific analyses, if required, will be provided. A schedule for the plant specific analysis will be determined based on the results of the generic analysis. e-i

71 II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS l Previous Response

     --        Orders on B & W plants have not previously been required to be addressed on Westinghouse plants.

Clarification Response - The Westinghouse Owners Group, of which Alabama Power Company is a member, i is currently addressing the potential for void formation in the Reactor Coolant System (RCS) during natural circulation cooldown conditions, as described in Westinghouse letter NS-TMA-2298 (T. M. Anderson of Westinghouse to P. S. Check of the NRC). We believe the results of this effort will fully address the NRC requirement for analysis to determine the potential for' voiding in the RCS during anticipated transients. A report describing the results of this effort will be provided to the NRC by January 1, 1982. 9 m 'l G

I 72 II . K. 2.19 SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS i Previous Response Orders on B & W plants have not been required to be addressed for Westinghouse plants. . Clarification Response The transient analysis code, LOFTRAN, and the present small break evaluations analysis code, WFLASH, have both undergone benchmarking against plant information or experimental test facilities. Inese codes, under appropriate conditions, have also been compared with each other. Alabama Power Company, as a me 9:r of the Westinghouse Owners Group, will provide, on a schedule consistent with the requirements of Item II.K.2.19, a report addressing the benchmarking of these codes. e G l l l

                                                                           - - - . - - _ - - -    -D_----

73 II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM Previous Response

             ~

This requirement was formally issued by NUREG 0737. This item however was responded to for Unit 1 by Alabama Power Company letter dated June 26, 1980, Clarification Response The Westinghouse Owners Group, of which Alabama Power Company is a member, is in the process of developing a report (including historical POR.V,' block, and safety valve failure rate data and documentation of actions taken since the TMI-2 event to decrease the probability of a stuck-open PORV) to address the NRC concerns of Item II.K.3.2. However, due to the time-consuming' process of data gathering, breakdown, and evaluation, the report is scheduled for submittal to the NRC on March 1, 1981. As required by the NRC, this report will be used to support a decision on the necessity of an automatic PORV isolation system as specified in Task Action Item II.K.3.1. l l_

74 II.K.3.2 REPORT ON PORV FAILURES Previous Response

  ~

By letter dated June 26, 1980 for Unit 1 Alabama Power Company document our cannitments for the Farley Nuclear Plant. This iten was included in the Unit 2 Technical Specifications. Clarification Response The Westinghouse Owners Group, of which Alabama Power Company is a member, is in the process of . developing a report (including historical PORV, block, and safety valve failure rate data and documentation of actions taken since the TMI-2 event to decrease the probability of a stuck-open PORV).to address 'the NRC concerns of Item II.K.3.2. However, due to the time-consuming process of data gathering, breakdown, and evaluation, the report is scheduled for submittal to the NRC on March 1, 1981. As required by the NRC, this report will be used to support a decision on the necessity of an automatic PORV isolation system ~as specified in Task Action Item II.K.3.1. l

76 . I II.K.3.5 AUTOMATIC TRIP OF RF. ACTOR PUMP DURING LOSS-OF-COOLANT Previous Response By letter of June 26, 1980 for Unit 1 Alabama Power Company addressed this item for.the Farley Nuclear Plant. This is a new item for Unit 2. Clarification Response The Westinghouse Owners Group, of which Alabama Power Company is a member, has performed analyses using the Westinghouse small-break evaluation model (WFLASH) to show ample time is available for the operator to trip the reactor ' coolant pumps following certain size small breaks (see WCAP-9584). In addition, the owners group is supporting a best-estimate study using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time af ter a small-break will lead to acceptable results. For both of these analysis efforts, the Westinghouse Owners Group is performing blind post-test predictions of LOFT erperiment L3-6. The input data and model to be used with WFLASH on LOFT L3-6 was submitted to the staff on December 1, 1980. The information to be used with NOTRUMP on LOFT L3-6 will be submitted prior to performance of the L3-6 test, as stated in owners

 , . . group letter OG-45, dated December 3, 1980.

The LOFT prediction from both models will be submitted to the NRC on February 15, 1981, given that the test is performed on schedule. The best estimate study is scheduled for completion by April 1, 1981. Bared upon these studies, Alabama Power Company believes that resolution of this issue will be achieved without any design modif-ications. In the event that automatic trip of the reactor coolant pumps is required after the

  -    NRC determination of model acceptability, a schedule will be provided for

potential modifications.

l l l

1 75 II.K.3.3 REPORTING SAFETY AND RELIEF VALVE FAILURES'AND CHALLENGES Previous Response By letter of June 26, 1980 for Unit 1 and June 20, 1980 for Unit 2, Alabama Power Company has responded to this item for the Farley Nuclear Plant. Clarification Response Alabama Power Company will provide as part of its Annual Report a list of all Steam Generator and Pressurizer SRV and RV failures and challenges, and connits to notifying the NRC promptly upon any failure of a PORV or safety valve to close on the Steam Generators or Pressurizer for Units 1 and 2. 1

l 1 77 II.K.3.9 PROPORTIONAL INTERGRAL DERIVATIVE (PID) CONTROLLER MODIFICATION Previous Response By letters dated June 20, 1980 for Unit 2 and June 26, 1980 for Unit 1 Alabama Power Company responded to this item for Farley Nuclear Plant. Clarification Response The clarification for item has been addressed in previous responses. y.. I

78 II.K.3.10 PROPOSED ANTICIPATORY TRIP MODIFICATION Previous Response

   ~~

By letters dated June 20, 1980 and September 16, 1980 for Unit 2 and June 26, 1980 for Unit 1 documented Alabama Power Company's commitments and actions for the Farley Nuclear Plant. Clarification Response The clerification has been addressed by previous responses. t ( l l

79 II.K.3.ll JUSTIFY USE OF CERTAIN PORV'S  ; Previous Response This item was formally issued by NUREG 0737. _. Clarification' Resoonse The Farley fluclear Plant Units 1 and 2 power operated relief valves are manufactured by Copes Vulcan and, therefore, the requirements of NUREG 0611, Section 3.2.4.d are not applicable for Farley Nuclear Plant. e

l 80 II.K.3.12 CONFIRM EXISTENCE OF ANTICIPATORY TRIP UPON TURBINE TRIP Previous Response

  ~

By letters dated June 26, 1980 for Unit 1 and June 20, 1980 for Unit 2, Alabama Power Company provided response to this item for the Farley Nuclear Plant. Clarification Response NUREG 0737 requires no clarification response. This anticipatory trip was addressed in previous responses, i k

81 II.K.3.17 REPORT ON OUTAGE OF ECC SYSTEMS - LICENSEE REPORT AND PROPOSED TECHNICAL SPECIFICATION CHANGES Previous Response Alabama Power Company's letter dated June 26, 1980 responded to this

       ,     item for Farley Nuclear Plant Unit 1.

Clarification Response Table 1 is a listing of ECC systems outages for Farley Nuclear Plant - Unit 1 since December 1,1977 during modes in which the systems are required to be operable in accordance with technical specifications. Included in this listing is the duration of the outage, the affected ECCS component, the cause of the outage, and corrective action taken to return the component to operable status. ECC system outages routinely required for surveillance testing are not included in the listing. It should be noted that no accurate records i exist for the duration of outages of ECCS equipment during modes of operation

 ,           in which this equipment is not required by technical specifications.

Unit 2 has experienced no ECCS outages due to the fact that no fuel has been loaded into Unit 2. i x

II.K 3.17 (continued) 82 TABLE 1 LISTING 0F ECCS QUTAGES Date Begun Date Finished . l Time Date Time Date System Duration Cause & Corrective Action 0145 12/6/77 0323 12/6/77 BIT 1.6 Boron concentration .below 20,000 ppm. Boric a'cid batched to raise concentra-l 1

                                                                              ' tion..

i 1145 12/28/77 0245 12/29/77 BIT 15 Inadvertent S.I. caused low boron concentration. Boric Acid batched to raise con-centration. - 1000 1/21/78 2200 1/23/78 RHR 60.0 RHR Pump 1A discharge valve _ . controller output re-adjusted and returned to service. 1020 4/25/78 1245 4/25/78 BIT 2.4 Baron concentration below

                                .                                              20,000 ppm. Boric Acid batched to raise concentra-tion.
 '1140       5/29/78          2010    5/29/78          BIT           8.5       Boron concentration below 20,000 ppa. Boric Acid batched to raise concentra-tion.

1355 6/7/78 1650 ,6/7/78 RWST 2.9 RWST volume below minimum value. Level increased with - blended makeup from CVCS blender. , 1250 6/10/78 2120 6/10/78 RHR 8.5 RHR Pump 1A tagged out to . perform maintenance on  ! leaking CCW relief valve. ' The valve was repaired, i tested satisfactorily, and returned to service. 0330 7/21/78 2215 7/24/78 BIT 85.8 BIT heat tracing de-energized to perform modification to j flow transmitter. 1 0900 7/25/78 1455 7/25/78 BIT 5.9 BIT heat tracing de-energized to perform modification to flow transmitter.  ![' l 1 l

                     .                                                                                          .)

p d _ - 0

II.K.3.17 (continued) TABLE 1 83 LISTINGOFECCSOUTAGES(CONTINUED) Date Begun .Date Finished l Time Date Time Date System Duration Cause &. Corrective Action 0745 8/4/78 1705 8/4/78 RHR 9.3 Modification to level switch . _.- required tag out of RHR suction valves. l 0840. 8/7/78 1500 8/7/78 - RHR 7.3 Modification to level switch l required tag out of RHR , suction valves. , 0800 8/9/78 1125 8/10/78 RHR 27.4 RHR suction valves tagged '

                                                                         . out for work on RWST level switches.

1040 8/14/78 1000 8/15/78 RHR 23.3 , RHR suction valves tagged out' for work on RWST level sw' itches. 1415 8/ 21/78

               ~

0312 8/22/78 ' ~ACC 13.0 Accumulator 1B nitrogen vented to allow for main-

                                                 '                         tenance of outlet valve.

Returned to service. 0245 12/21/78 0400 12/21/78 BIT 1.3 Lost BIT recirculation flow due to performance of I surveillance test. . . 1050 1/11/79 1120 1/11/79 RHR 0.5 RHR Pump 1B breaker racked out to perform preventive maintenance.

 '0530      1/23/79        2340    1/23/79         ~ BIT       18.2 BIT inlet valve failed to               j operate. Motor actuator repaired and BIT returned                 .I to service.                    .

0730 1/23/79 1030  ; l 1/23/79 BIT 3.0 Inl.et sample low boron concentration. Borated and  ! returned to service. , 1635 2/23/79 1707 2/23/79 BIT 0.8 BIT declared inoperable 8 L voluntarily in order to 'l.; repair air line to recire valve. Air line repaired .I and returned to service. l! Ih

                                                                                                                *I a -

l

II.K.3.17(' continued) - 84 TABLE 1

                                                                                                                                       ~
                                                  'LISTINGOFECCSOUTAGED(CONTINUED)                                                                   i Date Begun               Date Finished
  - Time              Date'         Time        Date           System    Duration                        Cause & Corrective Action
   -1045          .2/27/79          1145     2/27/79             RHR                1.0                  RHR pump 18 tagged out to         -

change oil. .

    '0100          2/28/79          0145     2/28/79             RHR               0.8                   RHR pump 1A tagged out.to
                                                                                        .               change oil.

1427 10/24/79 1432 10/25/79 S.I. 24.0 ' Spurious SI on low pressurizer pressure. Due to defective block card. - 1325 11/6/79 2200 11/6/79' 8.5 RWST RWST Level i[1dicators out of calibration calibrated and returned to service. 0532 11/21/79 '1745 11/21/79 BIT 12.2 -Boron concentration below 20,000 ppm. Boric Acid

                                                            '                                         batched to raise concentra-
                                                        ,                                             tion.
   '0105         .12/4/79          1920     12/5/79           . RHR       .23.0                      RHR pump 1B declared inoperible due to tag out
                                                                                                  .for repair of RHR miniflow valve.                                  -

1905 12/21/79 2105 12/21/79 CHG PMP 2.0

                                       -                                                             Charging pump 1B tagged out for oil change.

2125 12/26/79 0500 12/27/79 CHG PMP 7.5 10 nly one train of operable charging pumps. Pump 1C I racked out while IB was ( racked in to train A. Due

                         ;                                                                           to administrative error by i

shift foreman. . 1035 2/2/80 2015 2/2/80 BIT 10.0 Boron concentration below 20,000 ppm. Boric Acid batched to raise concentra-tion. I

                                                           ..h.

u=> >- _ _ _ _ _ - - - - - -- ^ ^ - -

F

                                                                                               .                             l
                                                                   ~

l II.K.3.17(continued) i TABLE 1 l l LISTINGOFECCSOUTAGES(CONTINUED) D:te Begun Date Finished Time Date Time Date System Duration Cause & Corrective Action 2020 3/4/80 2100 3/4/80 RHR 0.7 Power supply breaker to RHR pump 1A racked out for- -.

      ~

performance of general maintenance procedure. . D000 3/5/80 0420 3/5/80 RHR . 4.4 RHR pump 18 tagged out for ' ' oil change. , ' 1600 3/20/80 1620 3/20/80 RHR 0.3 RHR pump 1B tagged out for routine preventive main-

                                                                           . tenance.                         '
                                    ~

i400 3/29/80 0242 3/30/80 BIT 12.8 No recire flow through BIT. ' Due to clogged drain line

                                                                       . and seal leakage.

l ' !200 5/28/80 0950 5/29/80 - RHR 12.0 RHR heat exchanger discharge

  '                                                                          valve actuator failed.

Following maintenance valve was returned to service. l . - 125 10/20/80 2100 10/20/80 CHG PMP 9.5 - RWST supply to charging pump - suction valve would not open - during surveillance testing. . Repaired contact arm on valve I limit switch. Test completed satisfactorily. , . O e e 4

86 II.K.3.25 EFFECT OF LOSS OF ALTERflATING-CURRENT POWER ON PUMP SEALS Previous Response This item is new for Westinghouse plants. Clarification Response l i As stated in NUREG-0737, the loss of AC power is construed to be loss of offsite power. One acceptable solution is to provide an onsite emergency power supply to permit continued cooling of the RCP seal system, following loss of offsite power. The Farley Units 1 and 2 design is such that component cooling water to the RCP thernal barrier and seal injection to the RCP seals are automatically provided following a loss of offsite power. This is accomplished by automatically supplying the emergency electrical buses from onsite diesel generators and automatically sequencing the necessary RCP support systems on the emergency electrical buses. The RCP support systems consist of the component cooling water pumps (thermal barrier flow) and centrifugal charging pumps (seal injection flow). Neither the flow path associated with thermal barrier flow nor the flow path associated with seal injection flow isolate during an accident situation requiring containment I isolation. The Farley Units 1 and 2 RCP seals will not be adversely affected after the loss of offsite power if either one CCW pump or one centrifugal charging pump is operating. Both of these redundant (train A and B) cooling water and seal injection sources are provided even though only one source is required to assure seal integrity for an extended period of time. l t l

87 II.K.3.30 REVISED SMALL-BREAK LOSS-OF-COOLANT ACCIDENT METHODS TO SHOW COMPLIANCE WITH 10 CFR PART 50, APPENDIX X - Previous Response Alabama Power Company's letter dated June 26, 1980 responded to this item for Farley Nuclear Plant Unit 1. This is a new requires:2ent for 4 Westinghouse plants applying f e an operating license. Clarification Response - The present Westinghouse small-break model used to' analyze Farley Nuclear Plant Units 1 and 2 is in conformance with 10CFR Part 50, Appendix K. However, Westinghouse has committed to review their code and address the concerns of the Bulletins and Orders Task Force. The scope and schedule for the Westinghouse effort has been submitted to the NRC (see Westinghouse letter NS-TMA-2318, T. ft. j Anderson of Westinghouse to D. G. Eisenhut of the NRC, dated September 26,1980). 1 I I I:

                                         , . . - - - . - _ _              - - - - . .     - - - - . - - -                          - i

88 II.K.3.31 PLANT SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR PART 50.46 i Previous Response 1 Alabama Power Company's letter dated June 26, 1980 responded to this

 .              item for Farley Nuclear Plant Unit 1. This is a new requirement for Westinghouse plants applying for an operating license.                    -

Clarification Response - Westinghouse is pre'sently reviewing the model used to analyze small-break LOCAs on a schedule consistent with Iten II.K.3.30. Based upon the results of ' this review and the extent of changes made to the model used to analyze Farley-Units 1 and 2, Alabama Power Company will submit new calculations, if required, to show compliance with 10 CFR Part 50.46 for Farley-Units 1 and 2 by January 1,1983, or one year after staff approval of LOCA analysis models, whichever is later. ' l l

89 III.A.l.2 UPGRADE EMERGENCY SUPPORT FACILITIES Previous Response By letters of August 1,1980 and June 20, 1980 for Unit 2 and October 24, 1979, December 31, 1979, and-March 14, 1980 for Unit 1, Alabama Power Company responded to this item for the Farley Nuclear Plant. Clarification Response NUREG 0737 requires no clarification for this item. The Technical Support Center is currently scheduled to be completed prior to Unit 2 exceeding 5 percent power. s-1 1

90 III.A.I.1 EMERGENCY PREPAREDNESS -- SHORT-TERM and III.A.2 IMPROVING LICENSEE EMERGENCY PREPAREDNESS -- LONG-TERM 0 Previous Response By letters dated October 24, 1979, June 20, 1980, October 28, 1980, November 7,- 1980, and December 16, 1980, Alabama Power Company submitted the Radiological Emergency Response Plan and implementing procedures which describe our program regarding emerger.cy preparedness. Clarification Response Alabama Power Company submitted the Farley Nuclear Plant Radiological Emergency Response Plan and implementing procedures via letters dated October 28, 1980 and November 7, 1980, respectively. The compensating actions provided in the alternate to milestone (3) will be utilized by Alabama Power Company. In no case, however, will the alternative be exercised after July 1,1982, without prior approval by the NRC. Milestones (4) through (8) are scheduled for

 *.                 completion in accordance with the dates specified in NUREG-0737. In the event that the above schedule cannot be met the NRC Staff will be notified.

Attachment 1 provides a description of the compensating actions taken as an alternate to milestone (3). - _ - -__- __-__ m-_ __ _m- _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ -

• _ _ _ _ _ _ _ _ _ - - _ _ . _ _ _ . _ _ _- a

Attrchm:nt 1 (i) if only element 1 cy; ele =ent 2 is in use: 9I O The licensee (the person who will be responsible for making offsite dose projections) shall check communications with the cognizant National Weather Service (NWS) first order station and NWS forecasting ststion on a monthly basis to ensure that routine meteorological observations and forecasts can be accessed.

RESPONSE

The cognizant National Weather Service (NWS) first order station and forecasting station are identified in the appropriate emergency plan implementing procedure. A change to the emergency plan implementing procedure which pro-vides instructions and describes responsibilities for testing communications networks is being processed to specify the monthly communications check with the appropriate NWS stations. NOTE: NRC is on controlled distribution for EIPs. O The licensee shall calibrate the meteorological measurements pro-gram at a frequency no less than quarterly and identify a readily available source of meteorological data (characteristic of site conditions) to which they can gain access during calibration periods.

RESPONSE

Past calibration frequency was established as semi-annual. This calibration frequency will be increased to meet the quarterly calibration requirement. A source of meteorological data (characteristic of site conditions) is identified in the appropriate emergency plan implementing procedure. O During conditions of measurements system unavailability, an alternate source of meteorological data which is characteristic of site conditions shall be identified to which the licensee can gain access.

RESPONSE

i See response immediately above. l 0 The licensee shall maintain a site inspection sheedule for evaluation of the meteorological measurements program at a frequency no less than weekly.

RESPONSE

The current inspection schedule specified in approved plant procedures meets this requirement. Specific instructions are included in the appropriate environmental procedure which verifies operability of wind speed, wind direction and temperature gradient instrumentation. i ___________1

l' l 92 0 It shall be a reportable occurrence if the meteorological data unavailability exceeds the goals outlined in Proposed Revision l 1 to Regulatory Guide 1.23 on a quarterly basis.

RESPONSE

l A change is being processed to include this requirement in the appropriate environmental monitoring procedure to ensure that the requirement is promptly recognized. (ii) The portion of the DC5 relating to the transport and diffusion of gaseous effluents shall be consistent with the characteristics of the Class A model outlined in element 3 of Appendix 2 to NUREG-0654.

RESPONSE

Farley Nuclear Plant will implement a Class A Dose Calculation Method by April 1, 1981. (iii) Direct telephone access to the individual responsible for caking off-site dose pr.jections (Appendix E to 10 CFR Part 50(IV)(A)(4) shall be available to the NRC in the event of a radiological emergency. Procedures for establishing centact and identification of contact individuals shall be provided as part of the implementing procedures.

RESPONSE

The NRC ringdown telephone provides this capability. The shift supervisor and the Emergency Director are identified as contact individuals. l t l 1

93 III.D.l.1 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN

               . RADI0 ACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND BOILING WATER REACTORS Previous Response
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By letters dated August 20, 1980 and June 20, 1980 for Unit 2, and October 24, 1979, November 21, 1979, December 31, 1979, and March 14, 1980 for Unit 1, Alabama Power Company has addressed this item for the Farley Nuclear Plant. , Clarification Response Alabama Power Company has instituted a program for Units 1 and 2 to maintain leakage rates of systems outside containment to as low as practical as described in the referenced letters. Leakage measurements results for Unit 1 were submitted in our December 31, 1979 letter. Such leakage rates for Unit 2 will be determined prior to exceeding 5 percent power. A walkdown of scoped systens described in I&E Circular 79-21 has been completed for Unit 1. A similar walkdown will be completed on Unit 2 as soon as plant conditions permit. i l e - , - - -

94 III.D.3.3 IMPROVED INPLANT I0 DINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS Previous Response By letters of July 24, 1980 and August 1,1980 for Unit 2 and October 24, 1979, November 21, 1979 and December 31, 1979, Alabama Power Company addressed this item for the Farley Nuclear Plant. Clarification Response Alabama Power Company has a portable monitoring system available for both units which uses an iodine silver xeolite sampler and single channel analyzer. Emergency procedures have been revised to address the use of this portable monitor. Appropriate shift personnel have bean trained on the use of this analyzer. Alabama Power Company presently has the capability of purging these samples of entrapped noble gases by the use of nitrogen gas and analysis by Ge(Li) gamma ray spectroscopy in a low background counting facility. Alabana Power Company has the capability to remove the iodine sampling cartridge to a low-background, low-contamination area, if required, for additional analysis. Farley Nuclear Plant will have two counting rooms (one per unit). In the event of an accident, it is anticipated that the background radioactivity level in the non-cffected unit's counting room

 . will be low enough to perform the measurements. However, if the background radioactivity level is too high in the non-affected unit, the necessary measuring equipment will be relocated within one hour to the water treat-ment plant or to the emergency operations facility upon its completion.

The iodine concentrations will be measured accurately, as required, under accident conditions. Two Eberline IM-2 single channel analyzer (SCA) Iodine monitors have been installed as in-plant iodine instrumentation. These units constitute a sufficient quantity of instrumentation for all vital areas where continuous occupancy is required. l l 9

95 III.D.3.4 CONTROL ROOM HABITABILITY Previous Response i By letters of June 26, 1980 for Unit 1 and of June 20, 1980 for Unit 2, Alabama Power Company addressed this item for the Farley Nuclear Plant. Clarification Response Previous responses address all new clarification items. l {

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