ML20151C409

Proposed Tech Specs,Revising Administrative Controls Sections,Clarifying Sections 1.0 & 4.8.4.a,correcting Typos in Sections 4.7.7.1 & 4.8.1.1.2.e.2 & Deleting Inapplicable Footnotes

ML20151C409
Person / Time
Site:	Perry
Issue date:	04/08/1988
From:	CLEVELAND ELECTRIC ILLUMINATING CO.
To:
Shared Package
ML20151C402	List: ML20151C409 PY-CEI-NRR-0823, Application for Amend to License NPF-58,revising Tech Specs Administrative Controls Section,Clarifying Wording in Sections 1.0 & 6.8.4.a,correcting Typos in Sections 4.7.7.1 & 4.8.1.1.2.e.2 & Deleting Footnotes.Fee Paid
References
NUDOCS 8804120426
Download: ML20151C409 (20)

Text

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OEFINITIONS p The sealing mechanism associated with each primary containment f.

penetration; e.g. , welds, bellows or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM (PCP) 1.34 The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 61,10 CFR Part 71 and Federal and State regulations, burial ground requirements and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING 1.35 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that . replacement air or gas is required to purify the confinement.

RATED THERMAL POWER

. . 1.36 RATED 1HERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3579 MWT.

b REACTOR PROTECTION SYSTEM RESPONSE TIME 1.37 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval when the monitored parameter exceeds its trip setpoint at theThe channel response time sensor until da-energization of the scram pilot valve solenoids.

may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REPORTABLE EVENT 1.38 A REPORTABLE EVENT shall be any of those conditions specified in

' ' -Sections--50.72 and 50.73 to 10 CFR twt 50. t 3.

R00 OENSITY 1.39 R00 DENSITY shall be the number of controlAll rod notches rods fully inserted inserted as a fraction of the total number of control rod notches.

is equivalent to 100% R00 DENSITY.

SECONDARY CONTAINMENT INTEGRITY _

1.40 SECONDARY CONTAINMENT INTEGRITY shall exist when: .

a. All penetrations terminating in the annulus and required to be closed during accident conditions are either:

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SURVEILi.AN d REQUTREMENTS (Continued)

1. Confirms the accuracy of the test by verifying that the differ-ence between the supplemental data and the Type A test data is within 0.25 L,. The formula to be used is: .

=

[L, + L, - 0.25 L,] 1 L 1 e [L, + L, + 0.25 L,] where Le supplemental test result; L, = superimposed leakage; L, =

measured Type A leakage.

2. Has durition sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.

3. Requires the quantity of gas injected into the primary contain-ment or bled from the primary containment during the supple-mental test to be between 0.75 L, and 1.25 L,.

d. Type 8 and C tests shall be conducted with gas at P ,11.31 psig*, l at intervals no grer>ter than 24 monthsf except for fests involving. .

1. Air locks,

2. Main stea:n line isolation valves,

3. Valves pressurized with fluid from a seal systes,

4. . All containment isolation valves in hydrostatically tested lines

, per Table 3.6.4-1 wMch' penetrate the primary containment, and

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5. Purge supply and exhaust isolation valves with resilient material seals.

e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f. Main stua:n lin isolation valves shall be leak tested at least once per 18 months. "

g. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provistor.s of Appendix J of 10 CFR 50 Section III.C.3, when determining the conbined leakage rate provided the seal system and valves are pressurized to at least 1.10 P maintain,12.44psig,andthesealsystemcapacityisadequateto system pressure for at least 30 days.

h. All containment isolation valves in hydrostatically tested lines per Table 3.6.4-1 which penetrate the primary containment shall be leak tested at least once per 18 months.f "Unless a hydrostatic test is required per Table 3.6.4-1. .

• • Except for valvea : .121-f022A - " ' " '" "^ ^ ^A , w h i c h s ha l l b e -lea k-tes ted

-prior te July 12 - 1987. This uceptica expires en July-1-2,1987.

1. A Type C test interval extension to the first refueling outage is permissible for primary containment isolation valves listed in Table 3.6.4-1, which are e identified in letter PY-CEI/NRR-0714 L (dated Sept.?mber 11,1997) as needing

\ a plant outage to test. For this one time test interval, the provisions of Specificatien 4.0.2 are not applicable.

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PERRY - UNIT 1 3/4 6-5 k endment No. 5, 10

, PLANT SYSTEMS 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM 4 LIMITING CONDITION FOR OPERATION j 3.7.3 The react r core, isolation cooling (RCIC) system sna11 be OPERABLEYwitn an OPERABLE ' flow path capable of automatically taking suction from the sup-pression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3" with reactor steam dome pressure greater than 150 psig.

ACTION:

With the RCIC system inoperable, operation may continue provided the HPCS system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUT 00W within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to 150 psig

. within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILI.ANCE REQUIREMENTS .

[ 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water, c' 2. Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

3. Verifying that the pump flow controller is in the correct position.

b. When tested p msuant to Specification 4.0.5 by verifying that the RCIC pump de dops a flow of greater than er equal to 700 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1020 + 25 - 100 psig (ste n dome pressure).*

• i

• The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

l

1. C system shall be operable for the period April 10, 1987 the ugh May 31, e 4th the following provisions:

1. The RCIC injectio ve shall not be capabl f' automatic opening.

2. The operational procedures sTa 9 W dified and the operators shall be trained such that the RC yttem is%etgerated (except as necessary i

for testing) unle other high pressure systems fail to maintain an adequa* ctor water level.

l Continued on page 3/4 7-6a) -

PERRY - UNIT 1 3/4 7-6 AMENCMENT NO. 1, 4 t

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PLANT SYSTEMS

( f,.' MEACTORCOREISOLATIONCOOLINGSYSTEM S VEILLANCE REQUIREMENTS (Continued) )

3. P1 t operation subsequent to the reactor water level instrumentation test shall depend on the results as follows:

Channe Success Criteria A No anomalies over entire range of test conditions C ] , No anomalies over entire range of test conditions 0 e /. , No anomalies over entire range of test conditions B 4 No anomalies at pressure > 400 psig Successful Channels Permissible Power Length of Time With RCIC Ooeratino o Level at Power Level A+B+C 'fd' 75% until 5/31/87 8+C+0 75% until 5/31/87

. A+B+C+0 r(' 75% until 5/31/87 A + C + 0 + explanation ,

75% until 5/31/87 of B anomalies > 400 psig All other combinations 50% 2 weeks, then

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shutdown  ;

4. If reacter water level instrumentation shows significant dev!ations related to thermal hydraulic conditions onrelated to the RCIC tests, declare all reactor water level instrumentation inoperable and take whateveractionsarerequiredbyTechnicg1 Specifications.

The provisions of Specification 3.0.4 are \not applicable during this period.

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PERRY - UNIT 1 3/4 7-6a Amend 4aent No. 4

l PLANT SYSTEMS 3/4.7.7 FUEL HANDLING BUILDING FUEL HANDLING BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION -

3.7.7.1 At least three Fuel Handling Building (FHB) ventilation exhaust subsystems shall be OPERABLE.

APPLICABILITY: When irradiated fuel is being handled in the Fuel Handling Building.

ACTION:

With one FHB ventilation exhaust subsystem inoperable, restore the inoperable system to OPERABLE status within 7 days or suspend handling of irradiated fuel in the FHB. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.7.7.1 Each of the required FHB ventilation exhaust subsystem shall be .

demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operatec for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters OPERABLE.

b. At least once per 18 months or (1) after any structural mainter.ance on the HEPA filter or charcoal adsorber housing, or (2) following painting, fire or chemict1 release in any sentilation zone communicating with the subsystem by:

1. Verifying that the subsystem satisfies the in-place penetration testing acceptance eriteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.S.a. C.5.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the geh systey) flow rate is 15000 scfm 10%.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6 b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,

- by showing a methyl iodide penetration of less than 1% when tested at a temperature of 30*C and a relative humidity of 70%

in accordance with ASTM 03803; and

3. Verifying a subsystem flow rate of 15000 scfm 210% during system operation when tested in accordance with ANSI N510-1980.

PERRY - UNIT 1 3/4 7-17

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ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMEN15 (Continued)

7. Verifying the pressure in all air start receivers for each diesel generator to be greater than or equal to 210 psig.

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or =:;ual to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by checking for ind removing accumulated water from the day tank.

c. At least once per 92 days by checking for and removing accumulated water from the fuel oil storage tanks,

d. At least once per 92 days and from new fuel oil prior to its addi-tion to the storage tanks by verifying that a sample obtained in accordance with ASTM-0270-1975 meets the following minimum require-ments in accordance with the tests specified in ASTM-0975-1977:

1) A water and sediment content of less than or equal to 0.05 volume percent;

2) A saybolt universal viscosity at 100*F of greater than or equal to 32.6 sus, but less than or equal to 40.1 sus:

3) An API gravity as specified by the manufacturer at 60*F of greater C than or eaual to 26 degrees, but less than or equal to 36 degrees;

4) An impurity level of less than 2 mg of insolubles per 100 ml -

when tested in accordance with ASTH-02274-70; analysis shall be completed within 7 days after obtaining the sample but may be sampled and analyzed after the addition of new fuel oil; and

5) The other properties specified in Table 1 of ASTM-0975-1977 and Regulatory Guide 1.137, Revision 1, October 1979, Position 2.a., whto tested in accoraance with ASTM-0975-1977; analysis shall be ccmpleted within 14 days after obtaining the sample but may be sampied arit analyzed af ter the addition of new fuel

- oil.

e. At least once per 18 months *, during shutdown, by:

1. Subjecting the diesel to an inspection in accordance with instructions prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load of greater than or equal to 1400 kw (LPCS pump) for diesel generator Div 1, greater than or equal to kw (RHR B pump or RHR C pump)

• For any start of a diesel, the diesel must be loaded in accordance with the h manufacturer's recommendations.

PERRY - UNIT 1 3/4 8-5

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

(24n)n for diesel generator Div 2, and greater than or equal toGWO-kw (HPCS pump) for diesel generator Div 3 while maintaining speed less than nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint or 15% above nominal, whichever is less.

3. Verifying the diesel generator capability to reject a load of 5800 kw for diesel generators Div 1 and Div 2 and 2600 kw for diesel generator Div 3 without tripping. The generator voltage shall not exceed 4784 volts for Div 1 and Div 2 and 5000 volts for Div 3 during and following the load rejection.

4. Simulating a loss of offsite power by itself, and:

a) For divisions 1 and 2:

1) Verifying de-energization of the emergency busses and load shedding from the emergency busses.

2) Verifying the diesel generator starts

• on the auto-start signal, energizes the emergency busses with per-manently connected loads within 10 seconds, energizes the auto-connected loads through the load sequence (individual load tiners) and operates for greater (-

than or squal to 5 minutes while its generator is so e loaded. After energization, the steady state voltage and frequency of the emergency busses shall be main- ,

tained at 4160 420 volts and 60 1.2 Hz during this test.

b) For division 3: ,

1) Verifying de-energization of the emergency bus.

2) Verifying the diesel generator starts" on the auto-start signal, energizes the emergency bus with the per-manently connected loads within 13 seconds and operates for greater than or equal to 5 minutes while its gen-erator is so loaded. After energization, the steady

• All diesel generator starts for the purpose of this Surveillance Requirement may be preceded by an engine prelube period. The diesel generator start (10 sec)/ load (60 see) from ambient conditions shall be performed at least once per 184 days in these surveillance tests. All other engine starts for the purpose of this surveillance testing may be preceded by other warmup pro-cedures recommended by the manufacturer so that the mecnanical stress and wear on the diesel engine is minimized.

l PERRY - UNIT 1 3/4 8-6

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6.0 ADMINISTRATIVE CONTROLS

. fr 3 T## ' 6.1 RESPONSIBILITY 6.1.1 T er, Perry Plant Operations Department, shall be responsible for overall unit operation and shall delegate in writing the succession to this

.' responsibility during his absence.

6.1.2 The Shift Supervisor or, during his absence from the control room, a designated individual shall be responsible for the control room command function.

, A management directive to this effect, signed by the Vice President - Nuclear Group shall be reissued to all station parsonnel on an annual basis.

6.2 ORGANIZATION

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@ sert p ..2.1 The- cerperste organi::sti;n fcr erf t ?: nag;;;nt :nd technic 1 upport hell be as she-c. en Tiguce 0.2.1-1.

UNIT STAFF 6.2.2 The unit Orgea42"4^a shal' be ?? 5 5" nn F4gure 5.2.2-1 :nd;

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1; I b. At least one licent.ed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONCITION 1, 2 or 3, at least one licensed Senior-Operator shall be in tae control room;

c. A Health Physics Technician

• shall be on site when fuel is in the r2 actor; d, ALL CORE ALTERATIONS shall be observed and directly supervised by either a '.fr.ensed Senior Operator or licensed Senior Operator Limited to Fuel Hand 1fng who has nc other concurrent responsibilities during this operation; and "The Health Physics Technician may be less than the minimum requirements for a period of tide not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

PERRY.- UNIT 1 6-1 d

Insart 1 9

3 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety ot the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through inter =ediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and

• job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the USAR and updated in accordance with 10 CFR 50.71(e).

b. The General Manager, Perry Plant Operations Department (PPOD), shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The Vice President, Nuclear Group shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those vho carry

' out health physics and quality assurance functions may report to the appropriat9 casite manager; however, they shall have sufficient organizatie -1 freedom to enstre their independance from operating pressures. .

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JEMY - UMIT 1 6-5 Amendment No. 9

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ADMINISTRATIVE CONTROLS "r

6. b 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of simi-lar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications,

, maintenance activities, operations activities, or other means of freproving unit safety to the ";.x;;r_, Nuclear Engineering Department.

COMPOSITION

• 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers or technically oriented individuals located onsite. Each shall have either (1) a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field, or (2) equivalent work experience as described in Section 4.1 of ANSI /ANS 3.1, December 1981.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification

• that these activities are per-formed correctly and that human errors are reduced as much as practical.

(- RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to theimogee, Nuclear Engineeri.ng Department. CDi rec t oP; 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of tht unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineer-ing discipline and shall have received specific training in the response and anal-

~

ysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI N18.1-1971 for comparable positions, except for 0; 2ri:r ^-^r-S r: 2:rcint;r %: th ; " t .,, '. _.d the Plant Health Physicist who shall*

meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licansed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

• Not rasponsible for sign-off function. .

PERRY - UNIT 1 ,

6-7

ADMINISTRATIVE CONTROLS P fh 1

7' 6.4 TRAINING L

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6.4.1 A retraini'ng and. replacement training program for the unit staff shal be maintained under the direction of the Perry Training Section -if aOrr:'

n a ge0E;:b aisas, and shall meet or exceed the requirements and recommendations of Sec-tion 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemen-tal requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW CCMMITTEE (PORC)

FUNCTION 6.5.1.1 The PORC shall function to advise the danage

%ned Cooe r a e i_ons)

Perry Plantwepar =en on all matters related to nuclear safety, ~

and'the Director, Perry Plant Techni h Department, s(PP70), [

COMPOSITION 6.5.1.2 The PORC shall be composed of the:

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(, 49ee-Chai rman.'":t :-- Technical Superintendent, Perry Plant Technical Depar+aent - E nage a _

Vice-Chairman / Member: " '- ' r' " r' r - 0; r:ti r:In-inrr]pperationssect: en] '

.;..-  : Ca;cd ~.p r .i r r, 5 ^ :t' -: I ^ * ' ~.

Member: CM Member: %anaaerf*C:::

g f---- ' E;:-:'i rL;r~':'  ; ~ruinrr, r, Ma'ntenance Technical Section Section Member: Reactor Engineer Member:

• • • " ' '"rt-"':'  ; S;'nn r, Radiation (Managep*

• Protection Section Member: Plant Health Physicist Member: (ganaggC--- ' E: -"':' ; E;inrr, Instrumentation and Control Section Member:

^~^"" "'?'""# I' ; ~ ;' r r, Licensing and s

(Manager [ Compliance Section Member: Cr: :' E;r icir; *. ;' rr, St:;: r ' ;

e.. 4 .

ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve en a temporary basis; however, no more than wee +. alternates shall participate as voting memcers in PORC activities at any one ti MEETING FREOUENCY 6.5.1.4 The PORC shall meet at least once per calendar monta and as convened i - by the PORC Chairman or his designated alternate.

PERRY - UNIT 1 6-3 .

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l ADMINISTRATIVE CONTROLS

$1 QUORUM 6.5.1.5 The quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least 44*

members including alternates, four RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for:

a. Review of all Administrative Procedures;

b. Review of the safety evaluations for (1) proposed procedures /

instructions, (2) changes to procedures / instructions, equipment, systems or facilities, and (3) tests or experiments performed under the provisions of 10 CFR 50.59 to verify that such actions do not constitute an unreviewed safety question;

,@$ c. Review of proposed procedures / instructions and changes to procedures /

tt instructions, equipment, systems or facilities which involve an

- unreviewed safety question as defined in 10 CFR 50.59;

$$ d. Review of proposed tests or experiments which involve an unreviewed 55 safety question as defined in 10 CFR 50.59;

(( e. Review of proposed changes to Technical Specifications or the 3y Operating License; I g; f. Investigation of all violations of the Technical Specifications e ,y including the preparation and forwarding of reports covering evalua-

~c tion and recommendations to prevent recurrence to the Vice President -

• j Nuclear Group and to the Nuclear Safety Review Committee; Review of all REPORTABLE EVENTS;

&[ g.

"E h. Review of the plant Security Plan and Security Contingency Instruc-

. *E tions rd :9 'it:' :' rr:- : .d:d :hr;;;; t: tt: " d =- S '-t>

"xxw

"-"' r * :- ' t t r; .

i t i. Reviaw of the Emergency Plan and implementing instructips ame-+wb-2 ', _- -- _

Jt j. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems;

\ E" l nv

k. Review of any accidental, unplanned or uncontrolled radioactive l 8 .5 E: release including the preparation of reports covering evaluation, i

• recommendations, and disposition of the corrective action to prevent l recurrence and the forwarding of these reports to the 1- r , t cf r l"xt :: rr: .t:, the Nuclear Safety Review Committee and the eat" g? lJ Vice President - Nuclear Group;

1. Review of Unit operations to detect potential hazards to nuclear safety;

m. Investigations or analysis of special subjects as requested by the l

Chairman of the Nuclear Safety Review Committee; and l n. Review of the Fire Protection Program and implementir.g procedures ame

., _ . . . - - _ + 1, .. u. . % . e s m . ,, u % ._ , + m

_:1... +_7 PERRY - UNIT 1 6-9 l - 'L

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ADMINISTRATIVE CONTROLS .

. p ,, _

WW .e RESPONSIBILITIES (Continued)

.6.5.1.7 The PORC shall:

General Manno,er. PPOD/Di_r.e g r, PPTD

a. Recommend in writing to *.he -

_ ._, approval or disapproval of items considered under Specifications 6.5.1.6a.

through e. , h. , i. , J. , and k. , above prior to their implementation;

b. Render determinations in writing with regard to whether je not each item considered under Specifications 6.5.1.6b. through e. , above, constitutes an unreviewed safety question; and o c. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -

Nuclear Group and the Nuclear Safety Review Committee of disagreement (ceneraD between the PORC and the#anager, Perry Plant Operations Department;

@niFII( --'

however, the, Manager, Perry Plant Operations Department, shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.

RECORDS 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that, at a minimum, document the results of all PORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President - Nuclear Group and the Nuclear Safety Review Committee.

{

6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE (NSRC)

FUNCTION 6.5.2.1 The NSRC shall function to provide independent review and audit of designated activities in the areas of:

l l a. Nuclear power plant operations, L b. Nuclear engineering, M c. Chemistry and radiochemistry, .

i d. Metallurgy, .

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical engineering, l h. Quality assurance practices and administrative controls, and

f. Nondestructive testing.

l The NSRC shall report to and advise the Vice President - Nuclear Group on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

(

PERRY - UNIT 1 6-10 '

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. . . _ . ADMINISTRATIVE CONTROLS '

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t.c 6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES .

6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:

a. Procedures / instructions required by Specification 6.8 and other procedures / instructions which affect plant nuclear safety, and changes thereto, shall be prepareri, reviewed and approved. Each such c procedure / instruction or procedure / instruction change shall be re-2 viewed by a qualified individual (s) other than the individual (s) which prepared the procedure / instruction or procedure / instruction

J change, but who may be from the same section as the individual (s)

Ee which prepared the procedure / instruction or procedure / instruction t E$ f change. Instructions shall be approved by appropriate management

E' personnel as designated in writing by PORC, and approved by the J appropriate managers, Perry Plant Departments. Tr.: ";n;:::, ":r /

5C  ? t - :-t :-t:, shall approve Administrative Procedures. E ;- : /

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b. Proposed modifications to plant structures, systems and components l that affect nuclear safety shall be reviewed by individuals desig-nated by the "----- , Nuclear Engineering Department. Each such Director modification shall be reviewed by a qualified individual (s) other than i.

the individual (s) which designed the modification, but who say be from the same section as the individual (s) which designed the modifications.

Proposed modifications to plant structures, systems and components that affect nuclear safety shall be reviewed by PORC and approved prior to implementation b the "WiaI Manager. "'- ' l;;;---- ':

PPOD/ Director. P m

lga r: . not L_,.

C c. Proposed tests and experiments which affect plant nuclear safety shall be prepared, reviewed, and approved. Each such test or experi-ment shall be reviewed by a qualified individual (s) other than the individual (s) which prepared the proposed test or experiment.

Proposed tests and experiments shall be approved before implementa-tion b the , m "-llie;c=:.t.h _

kner.,,,"..__.aFManager, PPOD/ Director,~ PPT 6h

\

PERRY - UNIT 1 6-14

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ACHINI$TIATIVE CONTROLS '

i.)

j h*3 ACTIVITIES (Continued)

, d. Sections responsible for reviews, including cross-disciplinary re-o views, performed in ac:ordance with Specifications 5.5.3.la. , and g 6.5.3.lc. , shall be designated in writing by PORC and approved by the

) r:m:":n "r:;r, 'r y "rt ::;r rt. The individual (s) per-I g{., forming the review shall meet or exceed the qualification requirements yI of appropriate section(s) of ANSI N18.1-1971; 7 J le. Each review shall include a detersination pursuant to 10 CFR 50.59

%3 of whether or not the potential for an unreviewed safety question

• % 3% exists. If such a potential does exist, a safety evaluation oer

~3 a E. 10 CFR 50.59 .o determine wnether or not an unreviewed safety question t*8 is involved .nall be performed. Pursuant to 10 CFR 50.59, NRC 51 2 accroval of items involving unreviewec safety questions snall be obtained prior to implementation; and

f. The Plant Security Plan and Emergency Plan, and icolementing instruc-tions, snall be reviewed at least once per 12 aantas. Rec:menced (Directohchanges to the "r:;r, Perry inclementing Plant Tecnnicalinstructions Decaruent.shallRec:=:

be accreved.by the ended cMnges *.o the Plans shall be reviewed pursuant to the requirements of Soecifi-cationi 5.5.1.5 r : :. :. :. ' and ecoroved by the "r:;r' Tecnnical Decarment. NRC aoproval snall De obtained accrocriate.

as3. Perry Plant 6.5 REPO BEhEY[CIO b 6.5.1 The follcwing actions shall be taken for REPORTABLE EVENTS:

a. The Ccamission snall be notified

'" - ~ 2. ': a '.: :n Tm . %, and a recort sucmittad pursuant to the requirements of Section 50.73 to 10 CFR Part !O, and

b. Each RE?0RTABLE EVENT shall be reviewed by the PORC and the results of the review submitted to the NSRC and the Vice President - Nuclear Group.

5. 7 SAFETY t.IMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by talephone as soon as possible and in all cosas within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Vice President - Nuclear Group and the NSRC snall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. A Safety Limit Violation Recort shall be precared. The report shall be reviewed by the PORC. This report shall describe (1) acolicsole circumstances creceding the violation, (2) effects of the violation ucon unit comoonents, systems, or structures, and (3) corrective action taken to prevent recurrence.

c. The Safety Limit Violation Recort shall be sucaitted to the Commission, the NSRC, and the Vice President - Nuclear Group within 30 days of the violation.

d. Critical oceration of the unit shall not be resumed until authori:ed by the Commission.

PERRY - UNIT 1 5- 15 Menement No. 5

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7. *

,.Y;h. 6.8 PROCEDURES / INSTRUCTIONS MO PROGRAMS

- 6.8.1 Written procedures / instructions shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recoseended in Appendix A of Regulatory Guide 1.33, Revisinn 2, February 1978.

f b. The applicable procedures required to implement the requirements of i NUREG-0737 and supplements thereto.

[ c. Security Plan implementation.

[ d. Emergency Plan implementation.

PROCESS CONTROL PROGRAM implementation.

g e.

,- f. OFFSITE DOSE CALCULATION MANUAL implementation,

((

j ", g. Radiological Environmental Monitoring Program implementation.

~

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h. Fire Protection Program implementation, E%g  ;

6.8.2 Each administrative procedure of Specification 6.8.1, and changes EE t. hereto, shall "- -be ^--revi6wed

.__. as... 's, priorbytothe PORC and shall implementation. be approved All procedures by the %

/ instructions shall be reviewed periodically as set forth in administrative procedures.

Qnserc2h -

6.8./4 The following programs shall be established, implemented, and maintained:

~

a. Primary Coolant Sources Outside Containment ,

5 A program to reduce leakage from those portions of systems outside containment that could contain highly rWioactive fluids during a serious transient or accident to as low as practical levels. The systems include the HPCS,46, RHR, RCIC, LFC", feedwater leakage control system..and post-accident sampling _ systems. The program shall include the fo11owingi7the nyorogen analyzer portion oD

1. Preventive maintenancebom .stible Gas and periodic visu'a'l Control lii~pec (tion requirements, and h Integrated leak test requireeents for each system at refueling

- 2.

cycle intervals or less.

b. In-Plant Radiation Monitorina A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel,

2. Precedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equip 3ent.

PERRY - UNIT 1 6-16

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. Insert 2 6.8.3 TEMPORARY CHANGES Temporary changes to procedures / instructions which do not change the intent of the approved procedures / instructions shall be approved for implementation by two members of the plant management staff, at least one of whom holds a Senior Operator license. These temporary changes shall be documented. The temporary changes shall be approved by the original approval authority within 14 days. For changes to procedures / ins ructions which may involve a change in intent of the procedures / instructions, the original approval authority shall approve the change prior to implementation.

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