ML17146A736

Amends 63 & 34 to Licenses NPF-14 & NPF-22,respectively, Revising Tech Specs Re Administrative Changes Per Generic Ltr 85-19 & Correcting Errors Contained in Tech Specs

ML17146A736
Person / Time
Site:	Susquehanna
Issue date:	04/06/1987
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML17146A737	List: ML17146A736 ML17146A738
References
GL-85-19, NUDOCS 8704090070
Download: ML17146A736 (42)

Text

pit,S AEQIi "o

0o 0

I p

gO

+**++

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POWER 5 LIGHT COMPANY, ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT I AMENDMENT TO FACIL'ITY OPERATING LICENSE Amendment No. 63 License No. NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for the amendment filed by the Pennsylvania Power 5 Liaht Company (PPSL) dated December 26, 1985,-complies with the standards and requirements of the Atomic Eneray Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applicatiotI, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance w'ith 10 CFR Part 51 of the ComIIission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the enclosure to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 63, and the Environmental Protection Plan con-tained in Appendix B are hereby incorporated in the license.

PP8L shall operate the facility in accordance with the Technical Specifica<<

tions and -the Environmental Protection Plan.

8700090070

PDR,

• DOCK P

87040 b 05000387 PDR

3.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY CO~NISSION

Enclosure:

Chances to the Technical Specifications Date of Issuance:

April 6, 1987 Elinor G. Adensam, Director Bh'P Project Directorate No.

3 Division of BWR Licensing

ENCLOSURE TO LICENSE AMENDMENT N0.63 FACILITY OPERATING LICENSE NO.

NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE 3/4 4-13 3/4 4-14 3/4 6-23 3/4 6-24 3/4 6-25 3/4 6-26 3/4'-9 3/4 7-10 3/4 7-11 3/4 7-12 3/4 7-13 3/4 7-14 B 3/4 4-3 B 3/4 4-4 6-3 6-4 6-7 6-8 6-17 6-18 1MSERT 3/4 4-13 3/4 4-14 3/4 6-23 (overleaf) 3/4 6-24 3/4 6-25 3/4 6-26 3/4 7-9 (overleaf) 3/4 7-10 3/4 7-11 3/4 7-12 (overleaf) 3/4 7-13 3/4 7-14 (overl eaf)

B 3/4 4-3 B 3/4 4-4 (overleaf) 6-3 6-4 (overleaf) 6-7 (overleaf) 6-8 6-17 6-U'a 6-18 (overleaf)

REACTOR COOLANT SYSTEM 3/4.4. 5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of'he primary coolant shall be limited to:

a.

Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131, and 1

b.

Less than or equal to 100/E microcuries per gram.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

a 0 In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the primary coolant:

Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

Greater than 100/E microcuries per gram, be in at least HOT SHUT-DOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

C.

In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with,the specific activity of the primary coolant greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

In OPERATIONAL CONDITION 1 or 2, with:

1.

THERMAL POWER changed by more than 15X of RATED THERMAL POWER in one hour, or 2.

The off-gas level, at the SJAE, increased by more than 10,000 microcuries per second in one hour during steady state operation at release rates less than 75,000 microcuries per

second, or 3.

The off-gas level, at the SJAE, increased by more than 15K in one hour during steady state operation at release rates greater than 75,000 microcuries per

second, SUSQUEHANNA - UNIT 1 3/4 4"13 Amendment No.

63

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued ACTION (Continued) perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

SURVEILLANCE RE UIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.

SUSQUEHANNA - UNIT 1 3/4 4-14 Amendment No. 63

TABLE 3.6. 3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Manual Isolation Valves (Continued)

RCIC Suction HV-149F031 RCIC Turbine Exhaust HV-149F059 RCIC Vacuum Pum Oischar e

HV-149FQ60 HPCI In ection HV-155F006 1-55-038 RHR - Shutdown Coolin Return/

LPCI In ection HV-151F015 A,B RHR - Su ression Pool Suction HV-151F004 A,B,C,O RHR Heat Exchan er Vent~ )

HV-151F103 A,B CS In ection HV-152F005 A,B HV-152F037 A,B CS Suction~ )~ )

HV"152F001 A,,B Containment Instrument Gas SV-12654 A,B SUS(UEHANNA - UNIT 1 3/4 6-23 Amendment No. 36

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Manual Isolation Valves (Continued)

SLCS( )

HV"148F006 Demineralized Water 1"41-017 1-41-018 ILRT 1-57"193 1-57-194 HPCI Turbine Exhaust HV-155F066 RHR - Shutdown Coolin Return/

LPCI In ection - Pressure ualizin Valve HV-151F122 A,B c.

Other Valves Feedwater 141F010 A, 8 RHR - Shutdown Coolin Suction PS V-151F126 RHR - Shutdown Coolin Return/

HV-151F050 A, B SUSQUEHANNA - UNIT 1 3/4 6"24 Amendment No.

63

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Other Valves (Continued)

RHR-Minimum Recirculation Flow HV-151F007, A,B RHR - Relief Valve Dischar e( )

PSV-151F055 A, B PSV-15106 A, 8 PSV-151F097 CS In'ection HV-152F006 A,B CS Minimum Recirculation Flow HV-152F031 A,B Containment Instrument Gas 1-26"072 1-26-074 1-26-152 1-26-154 1-26"164 Recirculation Pum Seal Mater 143F013 A, B XV-143F017 A, B TIP Shear Valves C51"J004 A,B,C,D,E SLCS( )

148F007 HPCI Turbine Exhaust 155F049 SUSQUEHANNA - UNIT 1 3/4 6-25 Amendment No.

63

TABLE 3.6. 3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Other Valves (Continued)

HPCI Minimum Recirculation Flow HV-155F012 HV-155F046 RCIC Turbine Exhaust 149F040 RCIC Minimum Recirculation Flow FV-149F019 HV-149F021 RCIC Vacuum Pum Dischar e

149F028 d.

Excess Flow Check Valves HPCI XV-155F024 A,B, C, D

~Core S ra XV-152F018 A,B RHR XV"15109 A,B, C, D RCIC XV-149F044 A,B,C,D RWCU XV"14411 A,B,C,D XV-144F046 SUSQUEHANNA - UNIT 1 3/4 6-26 Amendment No.

63

s V 6

~

s s

~

m eel IA( w s(

l @it ~ s<<a 0 s,

% \\

~'

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Con'tinued C,

At. least once per 18 months by:

1.

Performing a system functional test which includes simulated automatic actuation and restart and verifying that each automatic valve in the flow path actuates to its correct position, but may exclude actual injection of coolant into the reactor vessel.

d.

2.

Verifying that the system will develop a flow of greater than or equal to 600 gpm in the test flow path when steam is supplied to the turbine at a pressure of 150,

+. 15, -0 psig."

3.

Verifying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.

4.

Performing a

CHANNEL CALIBRATION of the condensate transfer

'ump discharge low pressure alarm instrumentation and verifying.

',the iow pressure a1arm setpoint'to be greater than or equal to 113 pslg In the event the RCIC system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

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 tests.

SUSQUEHANNA - UNIT 1 3/4 7-9 Amendment No.

36

3/4.7.4 SNUBBERS-LIMITING CONDITION FOR OPERATION 3.7.4 All snubbers shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3 and OPERATIONAL CONDITIONS 4 and 5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS.

ACTION:

, Mith, one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.4.c on the supported component or declare the supp'orted system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a.

Visual Ins ections The first inservice visual inspection of snubbers shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all snubbers.

If all snubbers of each type are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection shall be performed at the first refueling outage.

Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:

No.. Inoperable Snubbers er Ins ection Period 0

1 2

3,4 5,6,7 8 or more Subsequent Visual Ins ection Period"¹ 12 months

+ 25K 6 months 2 25K 124 days + 25K 62 days 2 25K 31 days t 25K The snubbers may be categorized into two groups:

Those accessible and those inaccessible during reactor operation.

Each group may be inspected independently in accordance with the above schedule.

"The inspection interval shall not be lengthened more than one step at a time.

¹The provisions of Specification 4.0.2 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 7-10 Amendment No. 63

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS b.

Visual Ins ection Acce tance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) that attachments to the foundation or supporting structure are secure, and (3) in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up.

Snubbers which appear inoperable as a result of these visual inspections may be determined OPERABLE for the purpose of establishing-the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible, and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Surveillance Requirements 4.7.4.d.

c.

Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of at learnt that number of snubbers which follows the expression 35 (1 + -)

2 where c = 4, is the allowable number of snubbers not meeting the acceptance criteria 'selected by the operator, shall be functionally tested either in-place or in a bench test.

For each number of snubbers above c which does not meet the functional test accept-ance criteria of Specifications 4.7.4.d.,

an additional sample selected according to the expression 35 (1 + -) (

)

(a - c) shall c

2 2

2 c+1 be functionally tested, where a is the total number of snubbers found inoperable during the functional testing of the representative sample.

Functional testing shall continue according to the expression b [35 (1 + 2) (

>) ] where b is the number of snubbers found inoper c

2 2

able in the previous re-sample, until no additional inoperable snubbers are found within a sample or until all snubbers have been functionally tested.

The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers.

At least 25K of the snubbers in the representative sample shall include snubbers from the following three categories:

l.

The first snubber away from each reactor vessel nozzle.

2.

Each snubber within 5 feet of heavy equipment,

valve, pump, turbine, motor, etc.

3.

Each snubber within 10 feet of the discharge from a safety relief valve.

SUSQUEHANNA - UNIT 1 3/4 7-11 Amendment No. 63

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued Functional Test (Continued)

In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period.

If a spare snubber has been installed in place of a failed snubber, then both the failed snubber, if it is repaired and installed in another

position, and the spare snubber shall be retested.

Test results of these snubbers may not be included for the re-sampling.

If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same design subject to the same defect shall be functionally tested.

This testing requirement shall be independent of the require-ments stated above for snubbers not meeting the functional test acceptance criteria.

d.

For any snubber(s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber(s).

The purpose of this engineering evaluation shall be to determine if'he components supported by the snubber(s) were adversely affected by the inoperability of snubber(s) in order to ensure that the

. supported component remains capable of meeting the designed service.

Mechanical Snubbers Functional Test Acce tance Criteria The mechanical snubber functional test shall verify that:

1.

The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

Orag force shall not have increased more than 5'ince the last surveillance test.

3; Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

r Snubber release

rate, where required, is within the specified range in compression or tension.

For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

SUSQUEHANNA - UNIT 1

'3/4 7-L2 Amendment No. 36

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued e.

Snubber Service Life Monitorin A record of the service life of each

snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be main-tained as required by Specification

6. 10.2.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated ser-vice life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review.

This re-evaluation, replacement or reconditioning shall be indicated in the records.

F SUS(UEHANNA - UNIT 1 3/4 7"13 Amendment No. 63

h

)

3/4.7. 5 SEALED SOURCE CONTANINATIUh LIMITING CONDITION FOR OPERATION 3.7.5 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material eha11 be free of greater than or equa1 to 0.005 microcurie of removable contamination.

APPLICABILITY: At all times.

ACTION:

ae Mith a sealed source having removable contamination in excess of the above limit, withdraw the sealed source from use and either:

l.

Decontaminate and repair the sealed

source, or 2.

Dispose of the sealed source in accordance with Commission Regulations.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

5 SURVEILLANCE RE UIREMENTS 4.7.5.1 ~TE 1

-5 h

14 h114 df 1

kg and/or contami.nation by:

a.

The licensee, or b.

Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least

0. 005 microcurie 29 per test sample.

4.7.5.5 T~fi -5 h

kg~ f 1d,

• 1df g 5

sources and fssslon detectors previously subjected to core flux, shall be tested at the frequency described below.

Sources in use - At least once per six months for all sealed sources contasnsng radioactive material:

1.

Nth a half-life greater than 30 days, excluding Hydrogen 3

(tritium), and 2.

In any form other than gas.

SUSQUEHANNA - UNIT 1

'/4 7-14 Amendment No.

36

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued)

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.

When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within 'their acceptable limits.

With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant s specific activity greater than 0.2 microcuries per gram OOSE E(UIVALENT I-131, but less than or equal to 4.0 micro-curies per gram 00SE E(UIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to'ake corrective action.

SUSQUEHANNA - UNIT 1 B 3/4 4-3 Amendment No. 63

REACTOR COOLANT SYSTEM BASES 3/4. 4. 6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal" load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are 1imited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions, i.e.,

no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall,of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature:.

limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the 'time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RTNDT.

The results of these tests are sh'own in Table B 3/4.4. 6-1.

Reactor operation and resultant fast neutron, E greater than I MeV, irradiation will cause'n increase in the RTNDT.

Therefore, an adjusted reference temperature, based upon the fluence, phosphorus content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1. 99, Revision I, "Effects of Residual Elements on Predicted Radiation= Damage to Reactor Vessel Materials."

The pressure/

tempera-ture limit curve, Figure 3.4.6. 1-1, curves A', B'nd C', includes predicted adjustments for this shift in RTNDT for the end of life fluence.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50, Appendix M, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence.in predicting SUS(UEHANNA - UNIT 1

. B 3/4 4-4 Amendment No. 56

SR. VICE PRESIDENT-NUCLEAR HANAGER-NUCLEAR QUALITY ASSURANCE VICE PRESIDENT-NUCLEAR OPERATIONS HANAGER-NUCLEAR ADMINISTRATION HANAGER-NUCLEAR SAFETY ASSESSMENT HANAGER-.

NUCLEAR TRAINING PLANT SUPERINTENDENT-SUSQUEHANNA HANAGER-NUCLEAR SUPPORT HANAGER-NUCLEAR PLANT ENGINEERING MANAGER-NUCLEAR LICENSING ON SITE Figure 6.2.1-1 OFFSITE ORGANIZATION

SUPERINTENDENT OF PUNT ASSISTAIIT SUPERINTENDENT ICC SUPERVISOR SUPERVISOR OF OPERATIONS SRO SUPERVISOR OF NAINTENAIICE HEALTH PHYSICS/

CHENISTRY SUPERVISOR TECIINICAL SUPERVISOR PERSONNEL Ci ADN IN I STRAT ION SUPERVISOR SUPERVISOR OF SECURITY Ii>> <<

I 'I; I:

STAFF SHIfT SUPERVISOR SRO RADIOIDGICAL PROI'ECTION SUPVR/STAPF CNENISTRY STAFF UNIT I OPEIUITIONS SRO/RO REACTOR ENGINEERING PlJNT ENGINEERING SNIfT TECIUIICA-ADVISOR FIGURE 6.2.2-1 UNIT ORGANIZATION O

v ADMINISTRATIVE CONTROLS 6'.3 NUCLEAR SAFETY ASSESSMENT GROUP NSAG FUNCTION 6.2.3. 1 The NSAG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources

. of plant design and operating experience information, including plants of similar

design, which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The NSAG shall be composed of at least five dedicated, full-time engineers with at least three Tocated onsite, each with a bachelor's degree in engineering or related science and at least two years professional level experience in his field, at least one year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The NSAG shall be responsible for maintaining surveillance of unit activities to provide independent verification* that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The NSAG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to the Senior Vice President-Nuclear.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

6. 3 UNIT STAFF UALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed t'e minimum qualifi-cations of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Radiological Protection Supervisor or Health Physics/Chemistry Supervisor who shall meet or exceed the qualifica-tions of Regulatory Guide li8, September 1975, and'he shift Technical Advisor who shall meet or exceed the qualifications referred to in Section 2.2.1.b of Enclosure 1 of the October 30, 1979 NRC letter to all operating nuclear power plants.

6. 4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall b'e maintained under the direction of the Manager - Nuclear Training, shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18. 1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section 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.

Not responsible for sign-off function.

SUS(UEHANNA - UNIT I 6-7 Amendment No. 47

7 ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT

6. 5.1 PLANT OPERATIONS REVIEM COMMITTEE PORC FUNCTION 6.5. l. 1 The PORC shall function to advise the Superintendent of Plant-Susquehanna on matters related to nuclear safety as described in Specification 6.5.1.6.

COMPOSITION 6.5. 1.2 The PORC shall be composed of the:

Chairman:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Superintendent of Plant-Susquehanna Assistant Superintendent of Plant-Susquehanna Assistant Superintendent

- Outages Supervisor of Operations Technical Supervisor Supervisor of Maintenance I8C/Computer Supervisor Reactor Engineering Supervisor or Unit Reactor Engineer Health Physics/Chemistry Supervisor Shift Supervisor or Unit Supervisor ALTERNATES E

6.5. 1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting members in PORC activities at any one time.

MEETING FRE UENCY 6.5. 1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate.

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 four members including alternates.

SUSQUEHANNA " UNIT 1 6-8 Amendment No. 63 E

~

ADMINISTRATIVE CONTROLS 6.9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office, unless otherwise noted.

STARTUP REPORTS 6.9. 1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel

supplier, and (4) modifications that may have significantly altered the nuclear, thermal,'r hydraulic performance of the unit.

6.9. 1.2 The startup report shall address each of the startup tests identified in the FSAR and shall include a description of the measured values of the oper-ating conditions or characteristics obtained during the test program and a

comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license condi-tions based'n other commitments shall'be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or, commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

Zf the Startup Report does not cover all three events, i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation, supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS~

6.9.1.4 Annual reports covering the activities of the unit as described below for the prhvious calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

ao A tabulation on an annual basis of the number of station, utility, and other personnel, including contractors, receiving exposures greater than 100 mrem/yr and their associated manrem exposure accord-ing to work and job functions,"" e.g., reactor operations and sur-veillance, inservice inspection, routine maintenance, special main-tenance (describe maintenance),

waste processing, and refueling.

"A single submittal may be made for a multiple unit.station.

The submittal should combine those sections that are common to all units at the station.

• • This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.

SUS(UEHANNA " UNIT 1 6-17 Amendment No.

63

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.

In the aggregate, at least 80 per-cent of the total whole body dose received from external sources should be assigned to specific major work functions.

b.

The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5.

The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentation and one other radioiodine isotope concentation in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

SUSQUEHANNA " UNIT 1 6-17a Amendment No.63

s

~P4 ADMINISTRATIVE'ONTROLS MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam system safety/

relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U;S. Nuclear Regulatory Commission, (lashington, D.C.

20555, with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days of when the change(s) was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the PORC.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT" 6.9. 1.7 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and'an analysis of trends of the results of the radiological environmental survei11ance activities for the report period, including a comparison (as appropriate),

with preoperational

studies, operational controls and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include'he results of land use censuses required by Specification 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of all radiological environmental samples and of all environmental radiation. measurements taken during the report period pursuant to the loca-tions specified in the tables and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Branch Technical Position, Revision 1, November 1979.

In the event'that some individual results are not available for inclusion with the report, the report shall. be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following:

a summary description of the radiological environmental monitoring program; at least two legible maps""

covering all sampling locations keyed to a table giving distances and direc-tions from the centerline of one reactor plant; the results of licensee participation in the Interlaboratory Comparison

Program, required by Specifi-cation 3..12.3; discussion of all deviations from the Sampling Schedule of Table 3.12. 1-1; and discussion of all analyses in which the LLD required by Table 4.12.1-1 was not achievable.

I Amendment Mo.

29

(

"A single submittal may be made for a multiple unit station.

""One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

SUSQUEHANNA - UNIT 1 6-18

H

c<gp,ll RE0y

~o 0

I

~

/40 qO

+>>*<<+

UNITFD sTATES NUCLEAR REGULATORY COMMISSION INASHINGTON,O. C. 20555 PENNSYLVANIA POWER 8 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUSOUEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE 2.

Amendment No.

34 License No. NPF-22 The Nuclear Regulatory Comission (the Commission or the NRC) has found that:

I A.

The application for the amendment filed by the Pennsylvania Power 5

Light Company (PP8L) dated December 26, 1985, comp'1ies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility'will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of'his amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Accordingly, the license is amended by changes to the Technical Specif'ica<<

tions as indicated in the enclosure to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-22 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

34, and the Environmenta1 Protection Plan con-tained in Appendix 8 are hereby incorporated in the license.

PPSL shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.

P 3.

This amendment is effective as of the date of issuance FOR THF. NUCLEAR REGULATORY C01WISSION

Enclosure:

Changes to the Technical Specifications Date of Issuance:

April 6 l987 Elinor A. Adensam, Director 8WR Pro,iect Directorate No.

3 Division of RWR Licensing

ENCLOSURE TO LICENSE AMENDMENT NO. 34 FACILITY OPERATIN(i LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages nf the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE 3/4 4-13 3/4 4-14 3/4 6>>25 3/4 6-26 B 3/4 4-3 B 3/4 4-4 6-3 6-4 6-7 6>>8 6-17 6-18 INSERT 3/4 4-13 3/4 4-14 3/4 6-25 3/4 6-26 B 3/4 4-3 B 3/4 4-4 (overleaf) 6-3 6-4 (overleaf) 6-7 (overleaf) 6-8 6-17 (overleaf) 6-17a 6-18 (overleaf)

REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:

a.

Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131, and b.

Less than or equal to 100/E microcuries per gram.

APPLICABILITY:

OPERATIONAL CONDITION 1, 2, 3, Wd 4.

ACTION:

a 0 b.

C.

In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity of the primary coolant:

1.

Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12'hours.

2.

Greater than 100/E microcuries per

gram, be in at least HOT SHUT-DOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

In OPERATIONAL CONDITION 1 or 2, with:

1.

THERMAL POWER changed by more than 15K of RATED THERMAL POWER in one hour, or 2.

The off-gas level, at the SJAE, increased by more than 10,000 microcuries per second in one hour during steady state operation at release rates less than 75,000 microcuries per second, or 3.

The off-gas level, at the SJAE, increased by more than 15K in one hour during steady state operation at release rates greater than 75,000 microcuries per second, SUSQUEHANNA UNIT 2 3/4 4-13 Amendment No. 34

REACTOR COOLANT SYSTEM e

r a

LIMITING CONDITION FOR OPERATION Continued ACTION (Continued)

\\J perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the speqjfic activity of the primary coolant is restored to within its limit.

SURVEILLANCE RE UIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.

SUS(UEHANNA - UNIT 2 3/4 4-14 Amendment No. 34

TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Other Valves (Continued)

RHR - Relief Valve Dischar e

PSV-251F055 A, B PSV-25106 A, 8 PSV" 251F097 CS In'ection HV"252F006 A,B CS Minimum Recirculation Flow HV"252F031 A,B Containment Instrument Gas 2-26"164 2-26"072 2"26"074 2"26-152 2-26-154'ecirculation Pum Seal Mater 243F013 A) B XV-243F017 A,B TIP SHEAR VALVES C51-J004 A,B,C,D,E SLCS(b) 248F007 HPCI Turbine Exhaust 255F049 HPCI Minimum Recirculation Flow HV-255F012

'V"255F046 SUS(UEHANNA " UNIT 2 3/4 6"25 Amendment No.

34

I I

~

~

TABLE 3.6. 3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES VALVE FUNCTION AND NUMBER Other Valves (Continued)

RCIC Turbine Exhaust 249F040 RCIC Minimum Recirculation Flow FV"249F019 HV-249F021 RCIC Vacuum Pum Dischar e

249F028 d.

Excess Flow Check Valves HPCI XV-255F024 A,B,C,D

~Core S ra XV"252F018 A, B RHR XV-25109 A,B,C,D RCIC XV-249F044 A,B,C,D RMCU XV-24411 A,B,C,D XV"244F046 SUSQUEHANNA - UNIT 2 3/4 6"26 Amendment No. 34

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued)

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.

When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits.

With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in. excess of the limits will be detected in sufficient time to take corrective action.

3/4.4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant s specific activity greater than 0.2 microcurie per gram DOSE E(UIVALENT I-131, but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the param-eters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside contain-ment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

SUSQUEHANNA UNIT 2 B 3/4 4-3 Amendment No.

34

REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions, i.e.,

no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress control,ling location, each heatup,rate of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RTNDT The results of these tests are shown in Table B 3/4;4.6-1.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDT.

Therefore, an adjusted reference temperature, based upon the fluence, phosphorus content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

The pressure/

tempera-ture limit curve, Figure 3.4.6.1-1 includes predicted adjustments for this shift in RTNDT for the end of life fluence.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 1.

SUSQUEHANNA " UNIT 2 B 3/4 4-4

SR. VICE PRESIDENT-NUCLEAR MANAGER-NUCLEAR QUALITY ASSURANCE VICE PRESIDENT-NUCLEAR OPERATIONS MANAGER-NUCLEAR ADMINISTRATION MANAGER-NUCLEAR

'SAFETY ASSESSMENT MANAGER-NUCLEAR TRAINING PLANT SUPERINTENDENT-SUSQUEHANNA MANAGER-NUCLEAR SUPPORT MANAGER-NUCLEAR PLANT ENGINEERING MANAGER-NUCLEAR LICENSING ON SITE O

Figure 6.2.1-1 OFFSITE ORGANIZATION

- ~ ~

~

~ s

I

~

8

~ I tg

~ ~

SUPERItiTEttDEttT OF PLANT ASSISTANT SUPERINTENDENT ICC SUPERVISOR SUPERVISOR OP OPERATIONS SRO SUPERVISOR OF NAItiTENANCE IIEALTH PHYSICS/

CIIENISTRY SUPERVISOR TECllNICAL SUPERVISOR PERSOtiNEL 4 ADHINISTRATIOH SUPERVISOR SUPERVISOR

'P SECURITY STAFF SIIIFT SUPERVISOR SRO STAPP RADIOMGICAL PROTECTION S UP VR/STAPF CItENISTRY STAFF UNIT I OPERATIONS SRO/RO REACTOR EHGIHEERING PLANT EHGI HERR INC SllIPT TECIUIICAL ADVISOR tD Ct.

B tD FIGURE 6.2 2-1 UNIT ORGANIZATION O

ADMINISTRATIVE CONTROLS 6.2.3 NUCLEAR SAFETY ASSESSMENT GROUP NSAG FUNCTION 6.2.3.1 The NSAG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of plant design and operating experience information, including plants of similar

design, which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The NSAG shall be composed of at least five dedicated, full-time engineers with at least three located onsite, each with a bachelor's degree in engineering or related science and at least two years professional level experience in his field, at least one year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The NSAG shall be responsible for maintaining surveillance of unit activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The NSAG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to the Senior Vice President-Nuclear.

6. 2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

6. 3 UNIT STAFF VALIFICATIONS 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 and the supplemental require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Radiological Protection Supervisor or Health Physics/Chemistry Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September

1975, and the Shift Technical Advisor who shall meet or exceed the qualifications referred to in Section 2.2.1.b of Enclosure 1 of the October 30, 1979 NRC letter to all operating nuclear power plants.

6. 4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Manager - Nuclear Training, shall meet or exceed the requirements and recommendations of Section 5.5 o'f ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental 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.

"Not responsible for sign-off function.

SUSQUEHANNA - UNIT 2 6-7 Amendment No. 12

ADMINISTRATIVE CONTROLS

6. 5 REYIEM AND AUDIT

6. 5. 1 PLANT OPERATIONS REVIEM COMMITTEE PORC FUNCTION 6.5. 1. 1 The PORC shall function to advise the Superintendent of Plant-Susquehanna on matters related to nuclear safety as described in Specification 6.5.1.6.

COMPOSITION 6.5.1.2 The PORC. shall be composed of the:

Chairman:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Superintendent of Plant-Susquehanna Assistant Superintendent of Plant-Susquehanna Assistant Superintendent

- Outages Supervisor of Operations Technical Supervisor Supervisor of Maintenance I8C/Computer Supervisor Reactor Engineering Supervisor or Unit Reactor Engineer Health Physics/Chemistry Supervisor Shift Supervisor oi Unit Supervisor ALTERNATES 6.5. 1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting members in PORC activities at any one time.

MEETING FRE UENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate.

UORUM 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 four members including alternates.

SUSQUEHANNA - UNIT 2 6"8 Amendment No. 34

~,,

i

~,

~

ADMINISTRATIVE CONTROLS

6. 9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administratop of the Regional Office, unless otherwise noted.

STARTUP REPORTS 6.9. l. 1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel

supplier, and (4) modifications that may have significantly altered the nuclear,

thermal, or hydraulic performance of the unit.

6.9. 1.2 The startup report shall address each of the startup tests identified in the FSAR and shall include a description of the measured values of the operat-ing conditions or characteristics obtained during the test program and a compari-son of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events, i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation, supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS" 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9. 1.5 Reports required on an annual bas'is shall include:

a.

A tabulation on an annual basis of the number of station, utility, and other personnel,.including contractors, receiving exposures greater than 100 mrems/yr and their associated man-rem exposure according to work and job func4ions,"" e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter,

TLD, or film badge measurements.

Small exposures totalling less than 20K of the individual total dose need not be accounted for.

In the aggre-gate, at least 80X of the total whole body dose received from external sources should be assigned to specific major work functions.

"A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units, at the station.

""This tabulation supplements the requirements of f20.407 of 10 CFR Part 20.

SUSQUEHANNA - UNIT 2 6-17

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) b.

The results of specific analysis in which the primary coolant exceeded the limits of Specification 3.4.5.

The following informa-tion shall be included:, (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to ex-ceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the'-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a func-tion of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

SUS(UEHANNA " UNIT 2 Amendment No. 34

~ ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9. 1.6 Routine reports of operating statistics and shutdown experience, in-cluding documentation of all challenges to the main steam system safety/relief valves, shall. be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days of when the change(s) was made effec-tive.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the PORC.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6. 9. l. 7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a compar-ison (as appropriate), with preoperational

studies, operational controls and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also in-clude the results of land use censuses required by Specification 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of all radiological environmental samples and of all environmental radiation measurements taken during the report period pursuant to the locations specified in the tables and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of; the table in the Radiological Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results'he missing data shall be submitted as soon as possible in a supple-mentary report.

The reports shall also include the following:

a summary description of the radiological environmental monitoring program; at least two legible maps*"

covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor plant; the results of licensee participation in the Interlaboratory Comparison

Program, required by Specification 3.12.3; discussien of all deviations from the Sampling Schedule of Table 4. 12. 1-1; and discussion o'f all analyses in which the LLD required by Table 4.12. 1-1 was not achievable.

"A single submittal may be made for a multiple unit station.

""One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

SUSQUEHANNA - UNIT 2 6-18

~ '

p a