ML19351G246
| ML19351G246 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19351G242 | List: |
| References | |
| NUDOCS 8102230346 | |
| Download: ML19351G246 (15) | |
Text
' O ENCLOSURE 2 FEB 101981 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 2 TO FACILITY LICENSE NO. NPF-8 ALABAMA POWER COMPANY
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JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 DOCKET N0. 50-364
Background
License condition 2.C.(13)b requires staff approval of licensee's program for I.G.1, " Training During Low Power Testing".
This program is one of the require-ments for fuel loading and low power testing identified in NUREG-0694, "TMI Related Requirements for New Operating Licenses," June 1980.
" Clarification of TMI Action Plan Requirements", November 1980, superseded and provided changes to the requirements in NUREG-0694.
The staff has evaluated licensee's safety analysis and test procedures for Item I.G.I.
In addition staff has reviewed licensee's response to those items in NUREG-0737 which change fuel loading and low power testing requirements identified in NUREG-0694.
Our evaluation and conclusinn regarding these items is provided herein.
I.G.1 _ Training During Low Power Testing Requirement Section 22.2, Item I.G.1 of ' Supplement 4 to the Safety Evaluation Report related to the Operation of Joseph M._ Farley Nuclear Plant, Unit 2, September 1980, required that augmented low power tests be performed during initial plant startup prior to exceeding 5 percent power to provide data and operator training for anticipated abnormal conditions.
The specific tests required by Supplement 4 were:
Test 1 Cooldown capability of the charging and letdown system (6)*
Test 2a Natural circulation test (1).
Test 2b Natural circulation with loss of pressurizer heaters (3)
- Numbers in parentheses are those used to designate the tests in Supplement 4; (8) was not required and (98) will be run after the Westinghouse full power acceptance run.-
810 2280 W
5 0 Test 2c Natural circulation at reduced pressure (5)
Test 3 Natural circulation with simulated loss of offsite power (2)
Test 4 Effect of steam generator secondary side isolation on natural circulation (4)
Test 5 Forced circulation cooldown (9A)
Test 6 Simulated loss of all onsite and offsite AC power (7)=
Evaluation By letter dated September 2, 1980, licensee transmitted its safety analysis and procedures for Tests 1, 2, 3, 4 and 6.
The safety analysis for Test 5 (9A)* was not included because it was a prerequisite test for the tests of boron mixing and cooldown (98)* if they were to be run using r.uclear heat; however, the licensee proposed, and Staff agreed, that the baron mixing and cooldown tests could be run following the plant power escalation and full power acceptance run using decay heat.
Subsequently, Test 5 (9A) was incorporated in Test 4 (4).
By letter dated September-ll,1980, licensee transmitted revised procedures for Tests 3 and 6, using nuclear heat for Test 3 and reactor coolant pump heat for Test 6.
By letters dated November 18, 1980 and January 16, 1981, licensee transmitted its revised safety analysis, which is the basis for our approval of the tests.
The draft test procedures which were reviewed and accepted by the staff are:
FNP Test No.
Date of Draft Test 1 501-7-C01 September 17, 1980 Test 2 501-7-002 September 17, 1980 Test 3 501-7-003 September 13, 1980 Test 4 501-7-004 September 18, 1980 Test 6 501-7-006 September 17, 1980 The purpose of this safety evaluation is to present the results of the NRC staff review of Tests 1, 2,'3, 4 and 6 which constitute the licensee's augmented low power test program.
Staff approval of this test program satisfies NPF-8 License Condition 2.C.(13)b.
As identified above, Alabama Power Company (licensee) submitted the results of an analysis of the safety effects of the special conditions of the augmented low power test program, including the exceptions to the Technical Specifications, which lead to operating conditions that are outside the boutds of conditions assuined in the Final Safety Analysis Report (FSAR).
The effects of these special conditions on the Condition II, III, and IV events treated in Chapter 15 of the FSAR-were evaluated.
- Numbers in parentheses are those used to designate the tests in Supplement 4; (8) was not required and (98) will be run after the Westinghouse full power acceptance run.
. As the result of licensee's safety analysis of the augmented low power test program, a set of operational safety criteria have been specified for test conditions and for conditions requiring prompt operator initiation of reactor trip or safety injection or termination of test.
The operational safety criteria which are provided in Section 3.2 of licensee's safety analysis, November 18, 1980, include:
Limits on maximum core exit temperature, maximum loop AT for any loop, a.
maximum coolant cold leg and average temperature, and minimum subcooling.
These limits and operator actions are provided to ensure adequate margin to the saturation temperature and adequate core cooling.
b.
Limits on the minimum steam generator water level to provide a sufficient secondary side heat sink.
Limits on the minimum pressurizer water level for heater coverage and c.
pressure control.
d.
Limits on maximum insertion of control bank D to minimize consequences of inadvertent rod withdrawal and maintain a small moderator temperature coefficient while providing sufficient margin for shutdown.
Limits on the Power Range Neutron Flux low setpoint and Intermediate e.
Range Neutron Flux reactor trip setpoint to limit maximum power to low values following possible uncontrolled power increases.
f.
Limits on containment pressure and unplanned or unexplained changes in pressurizer water level and pressure.
Exceptions to a number of Farley Unit 2 Technical Specification requirements are needed to cor. duct the augmented low power test program. Some exceptions are needed because.of operation with a critical reactor under conditions outside of the range allowed in the Technical Specifications (e.g., natural circulation conditions and low coolant temperatures and pressure).
Other exceptions are required because some systems normally required to be operable will be rendered temporarily inoperable as part of the test program (e.g.,
simulated loss of offsite power and simulated loss of all AC power). The exceptions required are provided in Table 3-1 of Licensee's safety analysis, January 16, 1981, and listed in Table A of this Safety Evaluation for each of the tests in the augmented low power test program.
The Licensee presented results of offsite dose analyses for a hypothetical accident during the augmented low power test program, using conservative assumptions. The analysis was made for an accident with a coincident loss of 4
main condenser vacuum which did not involve a break in the reactor coolant f
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. pressure boundary. This accident bounds the consequences of Condition II type transients analyzed in the FSAR. The results of the analysis show that the two hour site boundary doses would be 5 rem thyroid, 0.9 rem whole body, and 0.4 rem to the skin.
The test procedures for the augmented low power test program as ident.ified in the Background of this Safety Evaluation have been reviewed by the staff.
The procedures have also been reviewed by the reactor system vendor, Westinghouse.
The reactor system vendor's safety analysis stated that tre program can be safely performed.
Independent staff review also concludes that the tests can be safely performed.
In order to perform the tests certain Technical Speci-fications must be excepted for the period of the tests as described above.
The low power levels, low core fission product inventory, and operational safety criteria described above permit the exceptions to be made and still retain adeq'; ate safety margins.
On the basis of our review of the licensee's safety analysis and procedures for the tests which include the operational safety criteria, effects of the exceptions to the Technical Specifications, offsite dose analyses, and test procedures, the staff concludes that the augmented low power test program at Farley Unit 2 is acceptable.
I.A.l.1 Shift Technical Advisor Requirement Each licensee shall provide an on-shift technical advisor to the shift super-visor. The shift technical advisor (STA) may serve more than one unit at a cultiunit site if qualified to perform the advisor function for the various units.
The STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the piant for transients and accidents. The STA shall also receive training in plant design and layout, including the capabilities of instrumenta-tion and controls in the control room. The licensee shall assign normal duties to the STAS that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.
Training shall be completed by January 1,1981 or by the time the fuel loading license is issued. See NUREG-0578, Section 2.2.1.b, and letters of Septerber 27 and November 9,1979 and October 31, 1980 (NUREG-0?37).
. Clarification The letter of October 30, 1979 clarified the short-tcrm STA requirerents.
That letter indicated that the STAS must have completed all training by January 1,1981.
This paper confirms these requirements and requests additional information.
The need for the STA position may be eliminated when the qualifications of the shif t supervisors and senior operators have been upgraded and the man-machine interface in the control room has been acceptably upgraded. However, until those long-term improvements are attained, the need for an STA program will continue.
The staff has not yet established the detailed elements of the academic and training reg'lirements of the STA beyond the guidance given in its October 30, 1979 letter. Nor has the staff made a decision on the level of upgrading required for licensed operating personnel and the man-machine interface in the control room that would be acceptable for eliminating the need of an STA.
Until these requirements for eliminating the STA position have been established, tne staff continues to require that, in addition to the staffing requirements -
specified in its July 31, 1980 letter (as revised by item I.A.l.3 of this enclosure), an STA he available for duty on each operating shift when a plant is being operated in Modes 14 for a PWR and Modes 1-3 for a BWR. At other times, an STA is not required to be on duty.
Since the October 30, 1979 letter was issued, several efforts have been made to er+ ;1ish, for the longer term, the minimum level of experience, education, and training for STAS. These efforts include work on the revision to ANS-3.1, work by the Institute of Nuclear Power Operations (INP0), and internal staff efforts.
INPO recently made availa' ale. a document entitled " Nuclear Power Plant Shift Technical Advisor--Recommendations for Position Description, Qualifications, Education and Training." A enpy of Revision 0 of this document, dated April 30, 1980, is attached as Appendix C.
Sections 5 and 6 of the INPO document describe the education, training, and experience reouirements for STAS. The NRC staff finds that the descriptions as set forth in Sa ns 5 and 6 of Revision 0 to the INP0 document are an acceptable approach 1 e selection and training of personnel tc staff the STA positions.
(Note:
.s should not be interpreted to mean that this is an NRC requirement at this time. The intent is to refer to the INP0 document as acceptable for interim guidance for a utility.in planning its STA program over the long term (i.e., bevond the January 1,1981 requirement to have Sl As in place in accordance with the qualification requirements specified in the staff's October 30, 1979 letter).)
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. No later than January 1,1981, all licensees of operating reactors shall provide this office with a description of their STA training program and their plans for requalification training. This description shall indicate the level of training attained by STAS by January 1,1981 and demonstrate conformance with the qualification and training requirements in the October 30, 1979 letter. Applicants for operating licenses shall provide-the same information in their application, or aEendments thereto, on a schedule consistent with the NRC licensing review schedule.
No later than January 1,1981, all licensees of operating reactors shall provide this office with a description of their long-term STA program, including qualification, selection criteria, training plans, and plans, if any, for the eventual phasecut of the STA program.
(Note:
The description shall include a comparison of the licensee / applicant program with the above-mentioned IMP 0 document. This request solicits industry views to assist NRC in establishing long-term improvements in the STA program. Applicants for operating licenses shall provide the same information in their applicatinn, or amendments thereto, on a schedule consistent with the NRC licensing. review schedule.)
Evaluation The NRC letter of October 30, 1979 refers to Sections A.1, A.2 and A.3 of to a September 13, 1979 NRC letter to licensees for detailed guidance concerning the ter'.nical education and training qualifications of STAS that are to be met by January 1,1981 for operating plants and prior to fuel loading for operating license applicants. This guidance is as follows.
A.
Accident Assessment Function 1.
General Technical Education I
j The technical education of at least one person in the control room
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under off normal conditions should include basic subjects in engineer-ing and science. ~The purpose of this education is to aid the operator in assessing unusual situations not explicitly covered in the current operator training. The following is a tentative list of areas of knowledge that are considered to be desirable:
Mathematics, including elementary calculus Reactor physics, chemistry and materials 1
2
o Reactor thermodynamics, fluid mechanics, and heat transfer Electrical Engineering, including reactor control theory These areas of knowledge should be taught at the college level and would be equivalent to about 60 semester hours. Although a college graduate engineer would have many of these subjects and more that would not be essential, some engineers might be deficient in a few of these specific aieas, e.g., reactor physics. Although the time to teach these subjects to a licensed senior reactor operator could be as short as two years, depending on the scope and content of the subjects, the selection of a graduate engineer would likely be a more rapid means of fulfilling this characteristic.
2.
All persons assigned to duties in the control room should be trained in the details of the design, function, arrangement and operation of the plant systems. This training is necessary to assure that the meaning and significance of instrument readings and the effect of control actions are known. A licensed operator or supervisor of an operator would not be required to have further training in order to fulfill this characteristic. A graduate engineer not previously licensed or trained as an operator or senior operator would require additional trainirig in order to fulfill this characteristic.
3.
Transient and Accident Response Training In addition to the training in normal operations, anticipated transients, and accidents presently required of operators and senior operators, one person in the control room under off normal conditions should be trained to recognize and react to a wide range of unusual situations including multiple equipment failures and operator errors. This training should not be limited to written procedures or specific accident scenarios, but should include the recognition of symptoms of accident conditions such as complex transient responses or inadequate core cooling and possible corrective actions.
The purpose of this training is to broaden the ability for prompt recognition of and response to unusual events, not to modify the instinctive, rapid procedural response to transients and accidents provided by reactor operators. The training is required in recognition of the fact that real accidents inherently are initiated and accompanied by unusual and unexpected events. The training 1.s also to emphasize need to focus on the essential parameters that indicate the status -
of the core and the primary coolant boundary. This additional training
. would take up to a year to accomplish for a person not already ex-perienced in nuclear plant transient and accident analysis or evaluation.
Both inexperienced graduate engineers and currently licensed operators would require additional training to fulfill this characteristic."
By letter dated February 5,1981, APCo has info _rmed us that its STAS meet the technical education and training requirements as specified in the NRC September 13, 1979 letter to licensees.
APCo also submitted information listing the education and training qualifications of each of its five STAS.
It has assumed and attributed equivalent college semaster hour credits for some of the training received by these STAS in the Farley STA and SR0 training programs and in U.S. Navy training programs.
APCo relies on these equivalent credit hours to reach the 60 semester credit hours in technical subjects specified by the September 13, 1979 letter for two of the Farley STAS.
The other three Farley STAS have Bachelor of Science Degrees in either Nuclear or Mechanical Engineering and appear to have well over the specified 60 semester credit hours without relying on an equivalence between the APCo-provided STA and SRO training program and college semester hour credits for technical education.
We have reviewed licensee's submittal regarding the technical education and training of its five shift technical advisors.
Based on our review, we conclude that the technical education and training of the Farley Nuclear Plant shift-technical advisors meet the requirements specified in Item I.A.1.1, and are acceptable.
I.A.l.3 Shift Manning Requirement Shift staffing and overtime restrictions for normal operation shall. be in accordance with Mr. D. G. Eisenhut's letter of July 31, 1980, as revised by NUREG-0737.
This requirement shall be met before fuel loading. See letters of July 31, 1980 and October 31, 1980 (NUREG-0737).
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C_larification NUREG-0737 supersedes page 3 of the July 31, 1980 letter in its entirety with the following:
Licensees of operating plants and applicants for_ operating licenses shall-include in their administrative procedures (required by license conditions) provisions governing required shift staffing and movement of key individuals about the plant.
These provisions are required to assure that qualified plant personnel to man the operational shifts are readily available in the event of an abnormal or emergency situation.
These administrative procedures shall also set forth a polity, the objective of which is to operate the plant with the required staff and develop working schedules such that use of overtime is avoided, to the extent practicable, for the plant staff who perform safety-related functions (e.g., senior reactor operators, reactor operators, health pnysicists, auxiliary operators, IAC technicians and key maintenance personnel).
IE Circular No. 80-02, " Nuclear Power Plant Staff Work Hours," dated February 1, 1980 discusses the concern of overtime work for members of the plant staff who perform safety-related functions.
1 The staff recognizes that there are diverse opinions on the amount of overtime that would be considered permissible and that there is a lack of hard data on i
the effects of overtime beyond the generally recognized normal 8-hour working day, the effects of shift rotation, and other factors.
NRC has initiated studies in this area.
Until a firmer basis is developed on working hours, the administrative procedures shall include as an interim measure the following guidance, which generally follows that of IE Circular No. 80-02.
In the event that overtime must be used (excluding extended period of shutdown for refueling, major maintenance or major plant modifications), the following overtime restrictions should be followed:
4 (1)
An individual should not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not including shif t turnover time).
(2)
There should be a break of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (which can include shift turnover time) between all work periods.
(3)
An individual should not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.
(4) An individual should not be required to work more than 14 consecutive days without having 2 consecutive days off.
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However, recognizi' ; that circumstances may arise requiring deviation from the above restrictions, such deviation shall be authorized by the plant manager or his deputy, or higher levels of management in accordance with published procedures and with appropriate documentation of the cause.
If a reactor operator or senior reactor operator has been working more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during periods of extended shutdown (e.g., at duties away from.the control board), such indivicuals shall not be assigned shift duty in the control room without at least a 12-hour break preceding such an assignment.
NRC encourages the development of a staffing policy that would permit the licensed reactor operators and senior reactor operators to be periodically assigned to other duties aw3y from the control board during their normal tours of duty.
If a reactor operator is required to work in excess of 8 continuous hours, he shall be periodically relieved or primary duties at the control board, such that periods of duty at the board do not exceed about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at a time.
The guidelines on overtime do not apply to the shift technical advisor pro-vided he or she is provided sleeping accommodations and a 10-minute availability is assured.
Operating license applicants shall complete these administrative procedures before fuel loading. Development and imple. mentation of the administrative procedures at operating plants will be reviewed by the Office of Inspection and Enfercement beginning 90 days after July 31, 1980.
Evaluation In Section 22.2 of Suppleucct 4 to the Safety Evaluation Report, we evaluated the Farley 2 plans for shift manning and overtime against the requirements in Mr. D. G. Eisenhut's letter of July 31, 1980 and concluded that they were acceptable.
NUREG-0737 did not revise the shift manning requirements in our July 31, 1980 letter; however, it did change the limitations on overtime.
APC0 has agreed to modify its a'dministrative procedures to implement the overtime policy, restrictions', and administrative requirements as described in NUREG-0737. AFCo confirmed this in a letter dated February 5,1981, which reads as follow;:
"The Company will incorporate into plant administrative procedures this policy concerning the utilization of overtime. This procedure, which will establish work schedules and guidelines that control the use of overtime for the plant staff who perform safety related functions, will be approved by corporate management.
The work schedule guidelines will comply with NUREG-0737 clarification.
! For personnel required by Farley Nuclear Plant Technical Specifications, Sections 6.2.2(a) and (c), the plant manager or in his absence, the plant emergency director will approve any deviations from the overtime guidelines l
described in the plant administrative procedures.
1 For all other personnel performing safety related functions, the group supervisor or superintendent will approve any posted work schedule deviating from the overtime guidclines described in plant administrative procedures.
In those unexpected situations where the necessity exists, due to unforeseen
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shift-to-shift contengencies or emergencies to work personnel more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight, or to not provide such personnel with a break between work periods of at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in order to perform safety-related work, i
the respective group foreman may authorize such deviation. Action on the i
foreman's part in these situations will be reviewed by the respective group i
supervisor or superintendent as a part of the normal biweekly approval process for payroll tima records.
It is the opinion of Alabama Power Company that this com.itment meets the spirit of the management control process of limiting overtime in that it provides a two-tier approval and l
review for the unexpected situation described above.
This commitment will be implemented prior to fuel loading."
{
For the personnel listed in Section 6.2.2(a) and (c) of the Farley Unit 2 l
Technical Specifications, this complies with the requirements of NUREG-0737.
For all other personnel performing safety-related functions, which pill number in the hundreds, APC0's procedure will provide for on-shift approval and 1
management review of overtime. We conclude that this meets the objectives of the NUREG-0737 requirements when applied to a large group of people.
1 1.A.2.1 Immediate Upgrading of Operator and Senior Operator Training and Qualification Requirements (1) Applicants for SR0 license shall have 4 years of responsible power plant-experience, of which at least 2 years shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shall be academic or related technical training.
Certifications that operator license applicants have learned to operate the 2
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controls shall be signed by the highest level of corporate management for plant operation.
These requirements shall be met on or after May 1,1980.
See letter of March"28,1980 (Ref. 27).
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. (2)
Revise training programs to include training in beat transfer, fluid flow, thermodynamics, and plant transients.
This requirement shall be ret by August 1,1980. See letter of March 28, 1980.
(3)
An applicant for a senior reactor operator-(SRO) license will be required to have experience equivalent to one year's experience as a licensed operator.
This requirement shall be met by December 1,1980.
See letter of October 31, 1980 (NUREG-0737).
Evaluation In ' action 22.5 of Supplement 4 to the Safety Evaluation Report, the staff concluded that Alabama Power Company (licensee) has satisfied the first two requirements of this item.
By letter dated January 14, 1981, the licensee stated it will meet the third requirement for all applications for licenses for SR0 after December 1980.
We conclude that licensee has satisfactorily met the requirements of Item I.A.2.1.
I.A.2.3 Administration of Training Programs for Licensed Operators Requirements (1)
Training instructors who teach systems, integrated responses, transient and simulator courses shall successfully complete a SR0 examination.
Applications shall be submitted by August 1,1980. See letter of March 28, 1980.
(2)
Instructors shall attend appropriate retraining programs that address, as a minimum, current operating history, problems and changes to procedures and administrative limitations.
In the event an instructor is a licensed i
SRO, his retraining shall be the SR0 requalification program.
Programs shall be initiated by May 1,1980. See letter of March 28, 1980.
f (3)
Pending accreditation of training institutions, training center ar J facility i
instructors who teach systems, integrated responses, transient, ans simulator courses shall demonstrate senior reactor operator (SRO) qualifications and be enrolled in appropriate requalification programs.
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1 i Documentation should be submitted 2 months prior to the issuance of an operating license.
See letter of October 31,1080 (NUREG-0737).
Evaluation In Supplement 4 to the SER, we concluded that licensee complied with requirements (1) and (2) of this item.
By letter of January 14, 1081, licensee has stated that all current and future plant instructors will be SRO-licensed or certihel and will attend the SRO requalification program.
We conclude that licensee has satisfactorily met the requirements of Item I.A.2.3.
I.C.6 Guidance on Procedures for Verifying Correct Performance of Operating Ac ti vi tie s Requirement Procedures shall be reviewed and revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided as a means of reducing human errors and improving the cuality of normal operations.
This will reduce the frequency of occurrence of situations that could result in or contribute to accidents. Such a verification system may include automatic system status monitoring, human verification of operations and maintenance activities independent of the people performing the activity, or both.
Imple-mentation of automatic status monitoring if required will reduce the extent of human verification of operations and maintenance activities but will not eliminate the need for such verification in all instances. The procedures adopted by the licensees may consist of two phases--one before and one after installation of autorstic status monitoring equipment, if required, in accordance with Item I.D.3 of NUREG-0660.
This requirement shall be met by January 1,1981 or prior to fuel load.
See NUREG-0578, Recommendation 5 and letter of October 31,1980 (NUREG-0737).
Evaluation Modified procedures, policies, and directives will be used at the Farlay Nuclear Plant to assure operating activities have been adequately veriflad.
A description of these procedures has been reviewed by the staff _and comn.ents given to the licensee.
The description has been revised to incorporate these comments and other comments generated within the applicant's organization.
Representatives of the Office of Inspection and Enforcement will verify imple-i mentation of these procedures in accordance with NUREG-0737.
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4 We conclude that the program described by the licensee for verification of correct performance of operating activities is adeouate to support operation up to 100 percent of rated power.
Environmental Considerations We have determined that the amendment does not authorize a change in effluent types, total amounts or an increase in design power level of 2774 MWt.. The test program will not result in any environmentsl impacts other than those evaluated in the Staff's Final Environmental Statement since the test program is encompassed by the overall activity evaluated in the Final Environmental 1~c.tement.
Conclusions The augmented low power test program for Farley Unit 2 involves tests at low power levels conducted over a short period of time and with a very low fission product inventory.
Similar tests have been conducted at Sequoyah Unit 1 and North Anna Unit 2.
On the basis of the above considerations, the proposed operational safety criteria and the safety evaluations which include the effects of the exceptions to the Technical Specifications and operation under natural circulation conditions, the staff concludes that the augmented low power test program will not result in undue risk to public health and safety and is acceptable.
Therefore, we have concluded based on the considerations discussed above, that:
(1) the low power test program does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regu-lations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Also, we reaffirm our conclusions as otherwise, stated in our Safety Evaluation Report and its Supplements related to the operation of Farley Unit 2.
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a TABLE A EXCEPTIONS TO TECHNICAL SPECIFICATIONS FOR AUGMENTED LOW POWER TESTS Technical Specification Test No.
1 2a 2b 2c 3
4 6
2.1.1 Core Safety Limits X
X X
X
~ X 2.2.1 Various Reactor Trips Overtemperature AT X
X X
X X
Overpower AT X
X X
X X
Steam Generator Level X
X X
X X
3.1.1. 4 Moderator Temperature Coefficient X
3.1.1. 5 Minimum Temperature for Criticality X
3.3.1 Various Reactor Trips Overtemperature AT X
X X
X X
Overpower AT X
X X
X X
Steam Generator Level X
X X
X X
3.3.2 Safety Injection - All automatic functions X
X X
X X
X Auxiliary Feedwater Initiation 3.4.4 Pressurizer X
X X
- 3. 7.1. 2 Auxiliary Feedwater X
X f
3.8.1.1 AC Power Sources.
X X
l 3.8.2.1 AC Onsite Power Distribution l
System X
X 3.8.2.3 DC Distribution System X
X 3.10.3 Special Test Exceptions -
l Physics Tests X
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