ML20235E876

Proposed Tech Specs Consisting of Administrative Changes to Enhance Clarity of Specs & Bases.Changes Delete Section 3.9 & Relocate Tech Spec 6.10.W/three Oversize Figures

ML20235E876
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1989
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20235E856	List: ML20235E854 ML20235E860 ML20235E876
References
NUDOCS 8902220292
Download: ML20235E876 (43)

Text

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. TSCR No.187 1.0 Proposed Technical Specification Change Request GPUN requests that the following pages of the TMI-1 Technical Specification be replaced as indicated below:

Replace pages: ii, v, viii, 3-13a, 3-14, 3-18g, 3-23, 3-24, Table 3.5-1 (page 3-29), 3-37, 3-37a, 3-61, 3-62a, 3-106, 3-107, 3-109, Table 4.1-1 (page 4-4), Table 4.1-2 (4-8), 4-39, 4-41, 4-42, 4-44, 4-46, 4-55, 4-55b, 4-58, 4-59, 4-61, 4-64, 4-80, 4-81, 4-82, 5-1, 5-10, Figures 5-1, 5-2, and 5-3; 6-2, 6-5, 6-19, and 6-19a.

Eliminate pages: 4-80a, 5-11 and Figure 5-4 2.0 Reason for Change This change is requested in order to make numerous administrative revisions to the document for the purpose of improved clarity and bases statement revisions. The following changes requested do not change the intent of any technical specifications but rather improve consistency of the document: TS 3.5.3.1, TS 3.15.1.2.a. and .b, Table 3.5-1, TS 3.22.1.1, TS 3.22.1.2, TS 3.22.1.3, Table 4.1-1, Table 4.1-2, TS.

4.5.1.1, TS 4.5.1.2.a, TS 4.5.2.2.a, TS 4.5.2.3.a, TS 4.5.3.1, TS 4.6.1.b, TS 4.12.1.2.a,.b,.c, TS 4.12.2.2.a,.b,.c, and .d, TS 4.15.1, TS 4.16.1, TS 4.17.1.e, and .i, TS 4.19.4.b, TS 5.1.1, TS 6.2.1, TS 6.5.2.1, and TS 6.9.4.3.3. In addition, the following Technical Specification bases have been revised to provide clarifications: TS 3.1.6, TS 3.1.13, TS 3.3, and TS 4.19.

3.0 Safety Evaluation Justifying Changes The changes proposed by Technical Specification Change Request No.187 are administrative in nature and serve only to enhance the clarity of the Technical Specifications and bases. None of the above changes have any impact on safety as discussed below in the statements of no significant hazards consideration.

For ease of review the changes needed will be discussed on a page by page basis:

Page ii Contents change only, reflecting correction of content deletion of Section 3.9 of the Technical Specifications pursuant to Amendment No. 129.

Page v Contents change only to reflect the relocation of TS 6.10 to page 6-19a.

Page viii Contents change only, reflecting deletion of Figure 5-4 from Section 5.0 of the Technical Specifications.

jf,h92 890131 p K 05000289 PDC 7442f/0171f N _ _ _ _ . __ __________ ___ ______ _-__________ _ __- _____ _ _ __ -_ - - -

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. TSCR No. 187 Pg 3-13a/14 Correction of typographical error in TS 3.1.6 bases statement with respect to referent surveillance specification and correction of reference to FSAR table.

Page 3-18g Clarification of TS 3.1.13 bases statements regarding environmental qualification of the high point vents has been provided, and the technical specification wording was not changed.

Page 3-23/24 A clarification to the bases statements concerning TS 3.3 for the BWST volume and baron concentration has been made by reorganization of existing information and reflects original design information including an appropriate UFSAR reference, and deletion of the original references.

Table 3.5-1 The RPS trip setting limit for variable low reactor Page coolant system pressure was eliminated by Amendment 3-29 No. 142 (Cycle 7 reload) but the associated instrumentation and operating conditions was inadvertently not deleted from the TS table, on line A.6.

Pages Technical Specifications 3.22.1.1, .2, and .3 refer to 3-106 Figure 5-4 for liquid effluent outfall locations. As 3-107 Figure 5-4 is being deleted by this TSCR, the appreoriate 3-109 reference is changed to Figure 5-3. l Table 4.1-1 The variable low pressure trip setpoint was elimina-Pages ted by Amendment No. 142 (Cycle 7 reload) but the associ-4-4 ated surveillance on-the RCS instrumentation was inad-vertantly not deleted from the TS table, on line 11.

Pages 3-37 Elimination of unnecessary footnotes related to previous 3-37a cycle specific operation has been made to Table 4.1-2; and 3-61 Technical Specifications: 3.5.3.1, 3.15.1.2.a, and .b, 3-62a 4.5.1.1; 4.5.2.1.a, 4.5.2.2.a, and 4.5.2.3.a.; 4.5.3.1; 4-8 4.6.1.b; 4.12.1.2.a,.b, and .c; 4.12.2.2.a,.b,.c, and .d; 4-39 4.12.2.2.a,.b,.c, and .d; 4.16.1; 4.17.1.e, and .1; and 4-41 4.19.4. Note, after deletion of cycle specific information 4-42 on page 4-80a items 7 and 8 were relocated to 4-44 page 4-80 and thus page 4-80a is eliminated.

4-46 4-55 4-55b 4-59 4-61 4-64 4-80 4-80a (deleted) 1 7442f/0171f ,

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n,. .o c.~TSCR No.'187f ,

Page'4-58 Reference to "AEC" has'been changed to "NRC" in Technical-Specification 4.15.1.

Page 4-81 .A clarification of OTSG' tube removal'from service'and; 4-82 repair. mechanisms has..been:made to the TS 4.19.4.b and the

' associated bases statement. The method of repair was not intended to be restricted and therefore the Technical' Specification and its Bases statement needed clarification.

In addition, the 6R cycle specific. paragraphs related to plugging criteria for the primary side tube freespan-c Jed..

by Amendment No. 116 have been deleted.

Page 5-1 A' revision to TS 5.1.1 has been made to reflect-the deletion of TS Figure 5-4 and provide editorial clarification of the

-remaining referenced figures.

Figures. .These figures have been revised and updated to provide 5-1 improved legibility consistent with the Updated FSAR.

5-2 The locations of liquid effluent outfalls of TS 5-3 Figure 5-4 has been added to the revised TS 5 Figure.5-3, and Figure 5-4 is thus eliminated.

y" Page 5-10.- The information concerning Liquid Effluent Outfall

'5-11 ' Descriptions has been revised to reflect the NPDES and relocated to page 5-10. .Page 5-11 is thus eliminated.

Page 6-2 Table 6.2-1 of TS 6.2.1 has been revised to reflect 10 CFR 50.54.

Page 6-5 An editorial clarification of TS 6.5.2.1 has been made to more properly describe the responsibility.that each Vice

' President has for ensuring that independent safety reviews are performed, by deletion of the word " periodic," which has '

no relevence in its present context.

Page 6-19 A change to TS 6.9.4.3.3 has been made by revising the 6-19a parenthetical reference to TS Figures 5-3 and 5-4. As ,

Figure 5-4 is being deleted by this TSCR, Figure 5-4 is thus 1 eliminated from the reference. TS 6.10, Record Retention is J relocated to page 6-19a for consistency of format.

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.TSCR No.187' 4.0 Ho Significant' Hazards Consideration-I GPUN has determined thatithis TSCR poses-no significant hazards-consideration as defined in 10CFR50.92 in that operation of TMI-l in accordance with the' proposed changes'will not:

1. Involve:a significant: increase in the probability or consequences of-any accident previously evaluated.- The.

probability of occurrence or'the consequences of previously_

evaluated accidents are not affected b'y these changes because the majority of the changes.are administrative in nature, or serves to conform to existing regulations which do not affect the plant configuration or operation.

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2. Operation of the facility in accordance with the proposed Technical Specification changes would not create the possibility of a new or different kind of accident _from any previously evaluated. As stated above, these changes are administrative in nature, conform to existing regulations.

3. Operation of the facility in accordance'with the proposed changes would not involve a.significant reduction in the margin of safety. The administrative changes do not reduce the margin of safety because'of the nature of such changes which serve to provide additional clarity or enhanced understanding of existing Technical Specifications and l

bases statements.

The propored amendment combines Examples (i), and (vii) of amendments that are c1nsidered not likely to involve significant hazards considerat wn (48 FR 14870) in that the changes are purely ,

administrative; and, serves to make the license conform to changes in the regulations.

5.0 Implementation It is requested that this amendment become effective 60 days after issuance to allow for implementation of any procedure changes that are requireo.

6.0 Amendment Fee (10 CFR 170.21)

In accordance with the provisions of 10 CFR 170.21, a check for $150.00' in herewith submitted with TSCR 187, Rev. O.

7442f/0171f

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' TABLE OF CONTENTS L

- Section' Page

'2 SAFETY' LIMITS AND LIMITING' SAFETY SYSTEM SETTINGS- 2-1 K '2.1 Safety Limits, Reactor Core 2-1

, 2.2. Safety Limits, Reactor System Pressure- 2-4

( 2.3 Limiting Safety System Settings, Protection i .

Instrumentation 2-5 3 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 General Action Requirements 3-1 3.1 ' Reactor Coolant System 3-la l '3.1.1 Operational Components .

3-la 3.1.2 Pressurization, Heatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves 3-18b 3.1.12 Pressurizer Power Operated Relief Valve (PORV) 3-18c and Block Valve 3.1.13 Reactor Coolant System Ver.ts 3-18f 3.2 Makeup and Purification & Chemical Addition Systems 3-19 3.3 Emergency Core Cooling, Reactor Building Emergency Cooling and Reactor Building Spray Systems 3-21 3.4 Decay Heat Removal Capability .

3-25 3.4.1 Reactor Coolant System Temperature Greater than 250'F 3-25 3.4.2 Reactor Coolant System Temperature 250*F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5.2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation Setpoints 3-37 3.5.4 Incore-Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.5.6 Chlorine Detection Systems 3-40f 3.6 Reactor Building 3-41 3.7 Unit Electrical Power System 3-42 3.8 fuel Loading and Refueling 3-44 3.9. Deleted 3-46 3.10 -Miscellaneous Radioactive Materials Sources 3-46 3.11 Handling of Irradiated Fuel 3-55 3.12 Reactor Building Polar Crane 3-57 3.13 Secondary System Activity 3-58 3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around TMI 3-59 3.14.2 flood Condition for Placing the Unit in Hot Standby 3-60 3.15 Air Treatment Systems 3-61 3.15.1 Emergency Control Room Air Treatment System 3-61 3.15.2 Reactor Building Purge Air Treatment System 3-62a 3.15.3 Auxiliary and Fuel Handling Building Air Treatment System 3- 62 c 3.15.4 fuel Handling Building ESF Air Treatment System 3-62e 11 Amendment No 50, 72 78. 9 ', 48 ill '2, 136

. I TABLE'0F CONTENTS Section ~ Page 5 DESIGN' FEATURES 5-1 5.1 SITE 5-1 5.2133TTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW'AND' SPENT FUEL ~ STORAGE ^ FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 ,

5.5 AIR ~ INTAKE TUNNEL ~ FIRE' PROTECTION' SYSTEMS 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT' STAFF' QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND' AUDIT 6-3

6. 5.. ] TECHNICAL REVIEW AND CONTROL 6-4 6.5.2 INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.5.4 INDEPENDENT ONSITE SAFETY REVIEW GROUP 6-8 6.6 REPORTABLE EVENT ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-10 6.8 PROCEDURE 5 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 SEMI ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6-18 6.10 RECORD RETENTION 6-19a l 6.11 RADIATION PROTECTION PROGRAM 6-21 6.12 HIGH RADIATION AREA 6-21 6.13 PROCESS CONTROL PROGRAM 6-21 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6-22 6.15 DELETED 6-22 6.16 POST ACCIDENT SAMPLING' PROGRAMS 6-22 NUREG 0737 (II.B.3, II.F.1.2) 6.17 MAJOR CHANGE 5 TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6-23 i

-v- I Amendment No. 11, 47, 72, 77, 129 i

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. LIST OF FIGURES Figure Title 3.5-2K Axial Power Imbalance Envelope for Operation from 40+10/-0 to 100+10/-0 EFPD, TMI-l 1

3.5-2L Axial Power Imbalance Envelope for Operation after 100+10/-0 EFPD, TMI-l 3.5-2M LOCA Limited Maximum Allowable Linear Heat Rate l 3.5-1 Incora Instrumentation Specification Axial Imbalance Indication, TMI-1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, TMI-1 3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from Cask Loading Pit 4.17-1 Snubber Functional Test - Sample Plan 2 5-1 Extended Plot Plan TMI 5-2 Site Topography 5 Mile Radius 5-3 Locations of Gaseous Effluent Release Points and Liquid Effluent Outfalls viii Amendment Nos. 72, 77, 126, 139, 142 i

1 8 Elases (Continued)

Although some leak rates on the order of gallons per minute may be -

tolerable from a dose point of view, it is recognized that leaks in the order of drops per minute through any of the barriers of the primary system could be indicative of materials failure such as by stress corrosion cracking. If depressurization, isolation, and/or other safety measures are not taken promptly, these small leaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature and location of the leak, as well as the magnitude of the leakage, must be considered in the safety evaluation.

When reactor coolant leakage occurs to the Reactor Building, it is ultimately conducted to the Reactor Building sump. Although the reactor coolant is safely contained, the gaseous components in it escape to the Reactor Building atmosphere. There, the gaseous components become a potential hazard to plant personnel, during inspection tours within the Reactor Building, and to the general public whenever the Reactor Building atmosphere is periodically purged to the environment.

When reactor coolant leakage occurs to the Auxiliary Building, it is collected in the Auxiliary Building sump. The gases escaping from i reactor coolant leakage within the Auxiliary Building will be collected in the Auxiliary and Fuel Handling Building exhaust ventilation system and discharged to the environment via the unit's Auxiliary and Fuel Handling Building vent. Since the majority of this leakage occurs within confined, separately ventilated cubicles within the Auxiliary Building, it incurs very little hazard to' plant personnel.

In regard to the surveillance specification 4.2.7, the isolation valves may be tested at a reduced pressure in accordance with the Franklin Research Center Report titled " Primary Coolant System Pressure Isolation Valves for TMI-1" (FRC Task 212) dated October 24, 1980, Section 2.2.2.

When reactor coolant leakage occurs to the nuclear services closed cooling water system, the leakage, both gaseous and liquid, is contained because the nuclear services closed cooling water syrtem surge tank is a closed tank that is maintained above atmospheric pressure. The leakage would be detected by the nuclear services closed cooling water system monitor and by purge tank liquid level, both of which alarm in the control room. Since the nuclear services closed cooling water system's only potential contact with reactor coolant is in the sample coolers, it is considered not to be a hazard. However, if reactor coolant leakage to this receptor occurred and the surge tank's relief valve discharged, radioactive gases could be discharged to the environment via the Unit's l auxiliary and fuel handling building vent.

Order dtd. 4/20/81 3-13a

4 Bases (Continu'ed)

When reactor coolant leakage occurs to the intermediate cooling closed cooling water system, the leakage is indicated by both the intermediate cooling water monitor (RM-L9) and the intermediate cooling closed cooling water surge tank liquid level indicator, both of which alarm in the control room. Reactor coolant leakage to this receptor ultimately could result in radioactive gas leaking to the environment via the unit's auxiliary and fuel handling building vent by way of the atmospheric vent on the surge tank.

When reactor coolant leakage occurs to either of the decay heat closed cooling water systems, the leakage is indicated by the affected system's radiation monitor (RM-L2 or Rii-L3 for system A and B, respectively) and surge tank liquid level indicator, all four of which alarm ic. the control room. Reactor coolant leakage to this receptor ultimately could result in radioactive gas leaking to the environment.via the unit's auxiliary and fuel handling building vent by way of the atmospheric vent on the surge tank of the affected

. system.

Assuming the existence 'he maximum allowable activity in the reactor coolant, a reactor coolant leakage rate of less than one gpm unidentified leakage within the reactor or auxiliary building or any of the closed cooling water systems indicated above, is a conserva-tive limit on what is allowable before the guide lines of 10 CFR 20 would be exceeded. This is shown as follows: if the specific activity of the ceactor coolant is 130/E uCi/ml and the gaseous portion of it (as identified by UFSAR Table 11.1-2) is discharged to l the environment via the unit's auxiliary and fuel handling building vent, the yearly whole body dose resulting from this activity at the site boundary, using an annual average X/Q = 4.5 x 10-6 sec/m3, is 0.34 rem. This may be compared with the 10 CFR 20 guideline of 0.5 rem / year whole body dose.

When the reactor coolant leaks to the secondary sides of either steam generator, all the gaseous components and a very small fraction of the ionic components are carried by the steam to the main condenser. The gaseous components exit the main condenser via the unit's vacuum pump which discharges.to the condenser vent past the condenser off-gas monitor. The condenser off-gas monitor will detect any radiation, above background, within the condenser vent.

However, buildup of radioactive solids in the secondary side of a steam generator and the presence of radioactive ions in the conden-sate can be tolerated to only a small degree. Therefore, the appear-ance of activity in the condenser off-gas, or any other possible in-dications of primary to secondary leakage such as water inventories, condensate demineralized artivity, etc., shall be considered positive indication of primary to secondary leakage and steps shall be taken to determine the source and quantity of the leakage.

i Amendment flo. 11, 22, 77, 139 3-14

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' Gases The safety function enhanced by this venting capability is core cooling. For events beyond the present design basis, this venting capability will substantially. increase the plants ability to deal with large quantitles of noncondensible gas.which could interfere with natural ~ circulation (i.e., core cooling).

The. reactor vessel head vent (RC-V42 & RC-V43 in series) provides the capability of venting noncondensible' gases from the majority.

of the reactor vessel head as well as the Reactor Coolant hot legs (to the elevation.of the top of the outlet nozzles) and cold legs-(through vessel internals leakage paths, to the elevation of the top

! of the inlet nozzles). This vent.is routed to containment atmosphere.

Venting for the pressurizer steam space (RC-V28 and RC-V44 in series) has been provided to assure that the pressurizer is available for Reactor Coolant System pressure and. volume control.

This vent is routed to the Reactor Coolant Drain Tank.,

Additional venting capability has been provided for the Reactor Coolant ~ hot leg high points (RC-V40A, B, RC-41A, B), which normally cannot be vented through the Reactor vessel head vent or pressurizer steam-space vent. These vents relieve to containment atmosphere through a rupture disk (set at low pressure).

The above. vent systems are seismically designed and environmentally qualified in accordance with the May 23, 1980 Commission Order and  !

Memorandum per.NUREG-0737, Item II.B.l. The high point vents do not fall within the scope of 10 CFR 50.49, since the vents are not relied upon during or following any design basis. event. The power operated valves (2 in series in each flow path) which are powered from emergency buses fail closed on loss of power. All vent valves for the reactor vessel head vent, pressurizer vent and loop B high point vent are powered from the class lE "B" bus. The vent valves for the loop A high point vent are powered from the class IE "A" bus. The power operated valves are controlled in the Control Room.

The individual vent path lines are sized so that an inadvertent valve opening will not constitute a LOCA as defined in 10 CFR 50.46(c)(1). These design features provide a high degree of assurance'that these vent paths will be available when needed, and that inadvertent operation or failures will not significantly hamper the safe operation of the plant.

3-18g Amendment No. 97 I

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. ' 3.3.3 Exceptions to 3.3.2 shall be as follows:

a. Both core flood tanks shall be operable at all times.

b. Both the motor operated valves associated with the core flood tanks shall be fully opened at all times.

c. One reactor building cooling fan and associated cooling unit shall be permitted to be out-of-service for seven days.

3.3.4 Prior to initiating maintenance on any of the components, the duplicate (redundant) component shall be tested to assure operability.

Bases The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate engineered safety features are operable. Two engineered safeguards makeup pumps, two decay heat removal pumps and two decay heat removal coolers (along with their respective cooling water systems components) are specified. However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required because a single core flooding tank has insufficient inventory to reflood the core for hot and cold line breaks.

The operability of the borated water storage tank (BWST) as part of i the ECCS ensures that a sufficient supply of borated water is  ;

available for injection by the ECCS in the event of a LOCA l (Reference 2). The limits on BWST minimum volume and boron concentration ensure that 1) sufficient water is available within contain*nent to permit recirculation cooling flow to the core, and 2) the reactor will remain at least one percent suberitical at 70 F without any control rods in the core following mixing of the BWST and RCS water volumes (Reference 3). j J

The contained water veiume limit of 350,000 gallons includes an ]

allowance for water not usable because of tank discharge location. l The limits on contained water volume, Na0H concentration and baron l concentration ensure a pf. value of between 8.5 and 11.0 of the l solution sprayed within containment after a design basis accident.

The minimum pH of 8.5 assures that iodine will remain in solution while the maximum pH of 11.0 minimizes the potential for caustic damage to mechanical systems and components. Redundant heaters i maintain the borated water supply at a temperature greater than 40 F.

3-23

• The post-accident reactor building emergency cooling may be accomplished by three emergency cooling units, by two spray systems, or by a combination of one emergency cooling unit and one spray system. The specified requirements assure that the required post-accident components are available.

The iodine removal function of the reactor building spray system requires one spray pump and sodium hydroxide tank contents.

The spray system utilizes common suction lines with the decay heat removal system. If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.

When the reactor is critical, maintenance is allowed per Specification 3.3.2 and 3.3.3 provided requirements in Specification 3.3.4 are met which assure operability of the duplicate components.

The specified maintenance times are a maximum. Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5.

An allowable maintenance period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be utilized if the operability of equipment redundant to that removed from service is demonstrated immediately prior to removal.

In the event that the need for emergency core cooling should occur, operation of one makeup pump, one decay heat removal pump, and both core flood tanks will protect the core. In the event of a reactor coolant system rupture their operation will limit the peak clad temperature to less than 2,300*F and the metal-water reaction to that representing less than 1 percent of the clad.

Two nuclear service river water pumps and two nuclear service closed cycle cooling pumps are required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant.

REFERENCES (1) Updated FSAR, Section 6.1 - Emergency Core Cooling System (2) Updated FSAR, Section 14.2.2.3 - Large Break LOCA (3) Updated FSAR, Section 14.2.2.1 - Fuel Handling Accident ,

l 3-24 Amendment No. 80

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3 5.3 ENGINEERED' SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS Applicability:

This specification applies to the engineered safeguards protection system. actuation setpoints.

Objective:

To provide for automatic initiation of the engineered. safeguards protection system in'the event of a breach of Reactor Coolant System integrity.

. Specification:

3.5.3.1 The engineered safeguards protection system actuation' setpoints and permissible bypasses shall be as follows:

Initiating Signal Function Setpoint-High Reactor Building Reactor Building Spray 1 30 psig Pressure (1) Reactor Building Isolation 1 30 psig High-Pressure Injection 1 4 psig Low-Pressure Injection 1 4 psig-Start Reactor Building Cooling & Reactor Building Isolation _1 4 psig Low Reactor Coolant High Pressure Injection > 1600(2) and System Pressure 2 500(3) psig Low Pressure Injection 2 1600(2) and 2 500(3) psig Reactor Building Isolation 2 1600 psig(2) 4.16 kv.E.S. Buses Undervoltage Relays Degraded Voltage (5) Switch to Onsite Power Source and load shedding 3595 volts (4)

Degraded grid timer 10 sec (5)

Loss of voltage Switch to Onsite Power Source and load shedding 2400 Volts (6)

Loss of voltage timer 1.5 sec (7)

(1) May be bypassed for reactor building leak rate test.

(2) May be bypassed below 1775 psig on decreasing pressure and is automatically reinstated before 1800 psig en increasing pressure.

(3) May be bypassed below 925 psig on decreasing pressure and is automatically reinstated before exceeding 950 psig on increasing pressure.

Amendment No. 70, 73, 78, 89 3-37

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(4)' Minimum allowed setting is 3560'9.~ Maximum' allowed' setting is

'3650 v.

-(5) Minimum allowed time is 8Lse'c. maximum. allowed time is 12 sec.

.(6) Minimum allowed setting is 2200 volts, maximum allowed setting' is 2860 volts

'~(7) Minimum' allowed time is (1.0) second, maximum allowed time is 1

(2.0) seconds.

Bases-High Reactor Building Pressure The basis for the 30 psig and 4 psig setpoints for.the high pressure-signal is to establish a setting which would be reached in adequate time in the event of a LOCA, cover a spectrum of break sizes and yet be far enough.above normal operation maximum internal' pressures to-prevent spurious initiation.

Low Reactor Coolant System Pressure The basis for'the 1600 and 500 psig low reactor coolant pressure setpoint for high and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum to break sizes and is far enough below normal operating pressure to prevent spurious initiation.

Bypass of HPI below 1775 psig and LPI below 925 psig, prevents ECCS actuation during normal system cooldown.

4.16 KV ES Bus Undervolt. age Relays The basis for N a degraded grid voltage relay setpoint is to protect the safety related electrical equipment from loss of function in the event of a sustained degraded voltage condition on the offsite power system. The timer setting prevents spurious transfer to the onsite source for transient conditions.

The loss of voltage relay and timers detect loss of offsite power condition and initiate transfer to the onsite source with minimal -!

time delay. j 1

)

i 3-37a Amendment No. 70, 73, 78, 89

~ __ - _ ___- - _ _ - _ _ _ _ - - _ _ _ _ - _ _ - _ _ _ _ _ _ - - _ - - _ - _ _ _ _ - .

' 3.15 AIR TREATMENT SYSTEMS 3.15.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM Applicability Applies to the emergency control room air treatment system and its associated filters.

L Objective To specify minimum availability and efficiency for the emergency control room air treatment system and its associated filters.

Specifications 3.15.1.1 Except as specified in Specification 3.15.1.3 below, both l emergency treatment systems, AH-E18A fan and associated filter AH-F3A and AH-E18B fan and associated filter AH-F3B shall be operable at all times, per the requirements of Specification 3.15.1.2 below; when containment integrity is required and when irradiated fuel handling operations are in progress.

3.15.1.2 a. The results of the in-place DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal absorber banks shall show <0.05% D0P per.etration and <0.05% halogenated hydrocarbon penetration, except that the 00P test will be conducted with prefilters installed.

b. The results of laboratory carbon sample analysis shall l show 90% radioactive methyl iodide decontamination efficiency when tested at 125 F, 95% R.H.

c. The fans AH-E18A and B shall each be shown to operate within 14000 CFM of design flew (40,000 CFM).

3.15.1.3 From and after the date that one control room air treatment system is made or found to b.e inoperable for any reason, reactor operation or irradiated fuel handling operations are permissible only during the succeeding 7 days provided the redundant system is demonstrated to be operable per 4.12.1.1 and 4.12.1.3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and daily thereafter.

3.15.1.4 From the date that botn control room air treatment systems are macle or found to be inoperable or if the inoperable system of 3.15.1.3 cannot be made operable in 7 days, irradiated fuel handling operations shall be terminated in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reactor shutdown shall be initiated and the reactor shall be in cold shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l

3-61 Amendment No. 55, 67, 76

L 3.15.2 REACTOR BUILDING PURGE AIR TREATHENT SYSTEM 1

' Appl'icabili ty .

Applies to the reactor building purge air, treatment system and its assoc'iated filters.

' Objective-To'specify minimum availability and' efficiency for the reactor building purge air treatment system and its associated filters.

. Specification 3.15.2'.1 Except as specified in Specification 3.15.2.3 below, the Reactor Building Purge Air Treate.ent System filter AH-Fi shall be operable as defined by the' Specification below at all times when containment integrity is required unless the' Reactor Building purge isolation valves are closed.

3.15.2.2 a. The results of the in-place DOP and halogenate'd .

l

-hydrocarbon tests at maximum available flows on HEPA filters and charcoal adsorber banks for AH-Fi shall show less than 0.05% DOP penetration and less than 0.05% halogenated hydrocarbon penetration; except that the DOP test will be conducted with prefilters installed.

b. The results of laboratory carbon sample analysis for l the reactor building purge system filter carbon shall show greater.than or equal to 90% radioactive methyl iodide decontamination efficiency when tested at 250*F,95% R.H.

3.15.2.3 From and after the date that the filter AH-F1 in the reactor building purge system is made or found to be inoperable as defined by Specification 3.15.2.2 above, the Reactor Building purge isolation valves shall be closed until the filter is made operable.

Bases The Reactor Building Purge Exhaust System filter AH-F1 while normally used to filter all reactor building exhaust air. It is necessary to demonstrate operability of these filters to assure readiness for service if required to mitigate a fuel handling accident in the Reactor Building and to assure that 10CFR50 Appendix I limits are met. Reactor Building purging is required to be terminated if the filter is not operable.

3-62a Amendment No. 55, 67, 76, 108

- _- _ _ - _ _ . .- _ - ___ _ __ D

. . 3.22 ' RADI0 ACTIVE EFFLUENTS 3.22.1 LIQUID EFFLUENTS 3.22.1.1' CONCENTRATION I LIMITING COMDITION'FOR OPERATION 3.22.1.1 The concentration of radioactive material released at anytime from the unit to unrestricted areas (see Figure 5-3) shall l be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-3 uC1/cc total activity.

APPLICABILITY: At all times ACTION:

a. With the concentration of radioactive material released from the unit to unrestricted areas exceeding the above limits, immediately restore concentration within the above limits.

b. If action "a" cannot be met, then be in:

1. At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,

2. At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

3. At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

BASES This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the unit to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures with (1) the Section II. A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106 (e) to the population. The concentration limit for noble gases is baced upon the assumption the Xe-135 is the controlling radioisotope? and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

3-105 i Amendment No. 72,137 i

a -

RADI0 ACTIVE EFFLUENTS e - DOSE -

LIMITING' CONDITION FOR OPERATION 3.22.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to the site boundary (see Figure 5-3) shall be limited: l

a. During any calendar quarter to < 1.5 mrem to the total body and to i 5 mrem to aiiy organ.

b. During any calendar year to < 3 mrem to the total body and to i 10 mrem to any- organ.

APPLICABILITY: At all times ACTION:

a. With the calculated dose from the release'of radioactive materials in . liquid effluents exceeding any of the above

' limits, prepare and submit to the NRC Region I Administrator within 30~ days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the.

corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters so that the cumulative dose or dose commitment to any individual- from such releases during these four calendar quarters.is within 3 mrem to the total-body and 10 mrem to any organ. This Special Report shall also include (1) the result of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

BASES This specification is provided to implement the requirements of Sections II. A, III. A, and IV. A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in l Section II.A of Appendix I. The ACTION statements provide the J required operating flexi'oility and at the same time implement the j guides set forth in Section IV.A of Appendix I to assure that the j releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by )

plant operations, there is reasonable assurance that the operation  !

of the facility will not result in radionuclides concentrations in )

the finished drinking water that are in excess of the requirements of 10 CFR 20. The dose calculations in the ODCM implement 3-107 )

Amendment No. 72, 129, 137 i

--_ _.m__... .--_ -_ _ _ _ .._._..____.-___._______.m

RADI0 ACTIVE EFFLUENTS

. LIQUID'RADWASTE' TREATMENT' SYSTEM LIMITING ~ CONDITION'FOR'0PERATION 3.22.1.3 The appropriate' portions of the liquid radwaste treatment system shall be used'to reduce the radioactive materials in liquid wastes prior to their discharge when the pro-jected doses due to the liquid effluent from the unit to unrestricted areas (see Figure-5-3) would exceed 0.06 mrem l to -the total body or 0.2 mrem _ to any organ in any calendar month.

APPLICABILITY: At all times ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason forl in-operability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and,

3. A summary description of action (s) taken to prevent a recurrence.

BASES .

The requirement that. the appropriate portions of this system be used,_when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The intent of Section II.D. is to reduce effluents to as low as is reasonably achievable in a cost effective manner. This LC0 satisfies this intent by establishing a dose limit which is a small fraction (257) of Section II.A of Appendix I,10 CFR Part 50 dose requirements.

This margin, a factor of 4, constitutes a reasonable reduction, i

3-109 Amendment No. 72,129

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_ ' g -- ,. TABLE 4.1 L MINIMUM EQUIPMENT TEST FREQUENCY-Item Test Frequency

1. Control Rods' ' Rod: drop times of all Each Refueling shutdown.

full length rods

2. Control Rod Movement of each rod Every two weeks, when ~

Movement- reactor is critical l 3. Pressurizer lSetpoint* 50% each refueling

-Safety Valves' period

4. ' Main Steam Setpoint 25% each~ refueling-Safety Valves period l

5. Refueling System Functional Start of each Interlocks refueling period'

6. Main Steam (See Section 4.8)-

Isolation Valves

7. Reactor Coolant Evaluate Daily, when reactor System Leakage coolant system

' temperature is greater than 525'F

~8. Deleted

9. . Spent Fuel Functional Each refueling period Cooling System prior to fuel handling

10. Intake Pump (a) Silt Accumulation- Each refueling period House Floor Visual inspection (Elevation of Intake Puiap 262 ft. 6 in.) House Floor (b) Silt Accumulation Quarterly Measurement of Pump House Flow

11. Pressurizer Block Functional ** Quarterly Valve (RC-V2)

• The setpoint of the pressurizer code safety valves shall be in accordance with ASMF Boiler and Pressurizer Vessel Code,Section III, Article 9, Winter, 1968.

• • Function shall be demonstrated by operating the valve through one complete cycle of full travel.

4-8 Amendment No. 55, 68, 78

.l

b 4,5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM & REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Loading Sequence Applicability: Applies to periodic testing requirements for safety actuation systems.

Objective: To verify that the emergency loading sequence and automatic power transfer is operable.

Specifications:

4.5.1.1 Sequence and Power Transfer Test

a. During each refueling interval, a test shall be conducted to demonstrate that the emergency loading sequence and power transfer is operable.

b. The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred to the emergency power as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.

-M. U. Pump

-D. H. Pump and D. H. Injection Valves and D. H. Supply Valves

-R. B. Cooling Pump

-R. B. Ventilators

-D. H. Closed Cycle Cooling Pump

-N. S. Closed Cycle Cooling Pump

-D. H. River Cooling Pump

-N. S. River Cooling Pump

-D. H. and N. S. Pump Area Cooling Fan

-Screen House Area Cooling Fan

-Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B.

30 psig Pressure Test Signal.)

-Motor Driven Emergency feedwater Pump

c. Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generator then reclosed to verify block load on the reclosure.

4.5.1.2 Sequence Test

a. At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred power or energency power.

h. The test will t,e considered satisfactory if the pumps and k fans listed in 4.5.1..lb have been successfully started and )

the valves listed in 4.5.1.lb have completed their travel as evidenced by the cc,ntrol board component operating lights, and either the station computer or pressure / flow indication.

4-39 Amendment No. 70, 78

_.--______________j

' 4.5.2 EMERGENCY CORE COOLING SYSTEM Applicability: Applies to periodic testing requirement for emergency core cooling systems.

,0bjective: To verify that the emergency core cooling systems are operable.

Specification 4 5.2.1 High Pressure Injection

a. During each refueling interval and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

l The M. U. Pump and its required supporting auxiliaries will be started manually be the operator and a test signal will be applied to the high pressure injection valves MU-V-16A, B, C, D to demonstrate actuation of the high pressure injection system for emergency core cooling operation.

b. The test will be considered satisfactory if the valves have completed their travel and the M. U. Pumps are running as evidenced by the control board component operatinc lights.

Minimum acceptable injection flow must be greater than or equal to 500 gpm per HPI pump when pump discharge pressure is 600 psig or greater (tha pressure between the pump and flow limiting device) and when the RC pressure is equal to or less than 600 psig.

c. Testing which requires HPI flow thru MU-V16A, B, C, D shall be conducted only under either of the following conditions:

1) T avg. shall be greater than 320*F.

2) Head of the Reactor Vessel shall be removed.

4.5.2.2 Low Pressure Injection l

a. During each refueling period and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

The auxiliaries required for low pressure injection are all included in the emergency loading sequence specified in 4.5.1.

I

b. The test will be considered satisfactory if the decay heat pumps listad in 4.b.1.lb have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.

Amendment No. 19, 57, 68 4-41

l.  !

, ' c. When-the Decay Heat System is required to be operable, the

correct position of DH-V-19A/B shall be' verified by obser-vation within four hours of each valve stroking operation or-valve maintenance,'which effects the position indicator. 1 l- i 4.5.2.3 . Core Flooding

a. During each refueling period, a system test shall'be

~

conducted to demonstrate proper operation of the system.

During depressurization of the Reactor Coolant System, l-verification shall be made that the check and isolation valves in the core cooling flooding tank discharge lines operate properly,

b. The. test will be' considered satisfactory if control board indication of core flooding tank level verifies that all valves have opened.

I 4.5.2.4 ' Component Tests

a. At intervals not to exceed 3 months, the components required for emergency core cooling will be tested,

b. The test wl.ll be considered satisfactory if the pumps and fans have been successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.

Bases The emergency core cooling systems are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the' bypass valves in the borated water storage tank fill line.

This allows-water to be pumped from the borated water storgage tank through each of the injection lines and back to the tank.

The minimum acceptable HPI/LPI flow assures proper flow and flow split'between injection legs.

With the. reactor shutdown, the valves in each core flooding lines are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.

4-42 Amendment No. 57, 68

I

( . .

. b. Reactor Building Cooling and Isolation Systems

1. During each refueling period, a system test shall be conducted to demonstrate proper operation'of the system. A test signal l will actuate the R.B. emergency cooling system valves to demonstrate operability of the coolers.

! 2. The test will be considered satisfactory if the valves have completed their expected travel as evidenced by the control board component operating lights, and either the station computer or local verification.

I l 4.5.3.2 Component Tests

a. At intervals not to exceed three months, the components required for reactor building cooling and isolation will be tested.

i

b. The test will be considered satisfactory if the valves have completed their expected travel as evidenced by the control board component operating lights, and either the station computer or local verification.

Bases The reactor building cooling and isolation systems and reactor building spray system are designed to remove the heat in the containment atmosphere to prevent the building pressure from exceeding the design pressure.

The delivery capability of one reactor building spray pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump.

With the pumps shut down an0 the borated water storage tank outlet closed, the reactor building spray injection valves can each be opened and closed by the operator action. With the reactor building spray inlet valves closed, low pressure air can be blown through the test connections of the reactor building spray nozzles to demonstrate that the flow paths are open.

The equipment, piping, valves and instrumentation of the reactor building cooling system are arranged so that they can be visually inspected. TM cooling units and associated piping are located outside the secondary concrete shield. Personnel can enter the reactor building during power operations to inspect and maintain this equipment.

The reactor building fans are normally operating periodically, constituting the test that these fans are operable.

Reference (1) FSAR, Section 6 4-44 Amendment No. 68

. 4 *. 6 EMERGENCY POWER SYSTEM PERIODIC TESTS-L

! Applicability: Applies to periodic testing and survelliance l- requirement _of-the emergency' power system.

Objective: To verify that the emergency power system will respond promptly and properly when required.

Specification:

The following tests and survelliance shall be performed as stated:

4.6.1 Diesel Generators

a. Manually-initiate start of the-diesel generator, followed by manual synchronization with other. power-sources and assumption of load by the diesel generator up to.the name-plate rating (3000 kW). This test will be. conducted every month on each diesel' generator. Normal plant. operation will not be effected.

b. Automatically start and leading the emergency diesel l generator in accordance with Specification 4.5.1.1.b/c including the following. This test will be conducted every refueling interval on each diesel generator.

(1) Verify that the-diesel generator starts from ambient condition upon receipt of the'ES signal and is ready.

to load in 1 01 seconds.

(2) Verfiy that the diesel block loads upon simulated loss of offsite power in 130 seconds.

(3) The diesel operates with the permanently connected and auto connected load for 15 minutes.

(4) The diesel engine does not trip when the generator.

breaker is opened while carrying emergency loads.

(5) The diesel generator block loads and operates for 15 minutes upon reclosure of the diesel generator breaker.

c. Each diesel generator shall be given an inspection at least annually'in accordance with'the manufacturer's recommendations for this class of stand-by service.

4.6.2 Station Batteries ,

a. The voltage, specific gravity, and liquid level of each cell will be measured and recorded monthly.

b. The voltage and specific gravity of a pilot cell will be measured and recorded weekly,

c. Each time data is recorded, new data shall be compared with old to detect signs of abure or deterioration.

Amendment No. 70 4-46

= ,

"4 ,12 ' AIR TREATMENT-SYSTEM. 1 9

.4.12.1 EMERGENCY CONTROL-. ROOM AIR TREATMENT SYSTEM

' Applicability-Applies to the emergency control room air treatment. system and associated components.

Objective i a., . l.

To. verify that this system and associated components will be able to perform its, design functions.

Specification-4.12.1.1- At least every refueling interval or once every 18 months, whichever comes first, the pressure drop across the combined HEPA filters and charcoal adsorber. banks of AH-F3A and 3B shall be demonstrated to be less than 6inchesofwateratsystemdesignflowrate(i10%).

4.12.1.2 .a. The tests ~and sample analysis required by l' Specification 3.15.1.2 shall be performed initially and at least once por year for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting,.: team, fire or chemical release in any ventilation zone communicating with the system that could contaminate the 1: EPA filters or charcoal adsorbers.

b. ' DOP testing shall be performed af ter each complete or l ~'

partial replacement of'the HEPA filter bank or after any structural maintenance on the system housing which could affect the HEPA filter bank bypass leakage.

c. Halogenated hydrocarbon testing si.all be performed l after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing which could effect the charcoal adsorber bank bypass leakage.

d. Each AH-E18A and B (AH-F3A and B) fan / filter circuit shall be operating at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

4.12.1.3 At least once per refueling interval or once every 18 months, whichever comes first, automatic initiation of the Control Building isolation and recirculation Dampers AH-D28, 37, 39, and 36 shall be demonstrated as operable.

4.12.1.4 An air distribution test shall be performed on the HEPA l filter bank initially, and after any maintenance or testing that could affect the air distribution within the system. The air distribution across the HEPA filter bank shall be uniform within 120%. The test shall be performed at 40,000 cfm (1107.) flow rate.

Amendment No. 55, 68 4-55

4012.2 ' REACTOR BUILDING PURGE AIR' TREATMENT SYSTEM Applicability: Applies'to the reactor building purge air treatment

, system and associated components.

Objective: To verify that this Tystem and associated components will be able to perform its design functions.

4 Specification 4.12.2.l' At least once per refueling interval or once per' ._

18 months, whichever comes first it shall be demonstrated that the' pressure drop across the. combined HEPA filters and charcoal ~adsorber banks is less than 6 inches of water at system design flow rate (1101.) .

4.12.2.2 a. The tests and sample analysis required by Specification l 3.15.2.2, shall be performed initially, once per re-fueling interval or 2 years, whichever comes first, or

> within 30 days prior to the movement of irradiated fuel in containment and following;sigr.ificant paint-ing, steam, fire, or chemical release in any ventila-tion zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers,

b. DOP testing shall be performed after each tom'plete or. l partial replacement of a HEPA filter bank or after any structural maintenance on the system housing which could-affect HEPA frame bypass leakage.

c. Halogenated hydrocarbon testing shall be performed l-after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system "ousing which could affect the charcoal adsorber bank bypass. leakage.

d. The DOP and halogenated hydrocarbon testing shall be l performed at the maximum available flow considering physical restrictions, i.e., purge valve position, and gaseous radioactive release criteria,

e. Each refueling, AH-E7A&B shall be shown to operate within i 5000 cfm of design flow (50,000 cfm) with purge valves fully open.

4.12.2.3 An air distribution test shall be performed on the HEPA filter bank initially and after any maintenance or testing that could affect the air distribution within the system.

The air distribution across the HEPA filter bank shall be uniform within 1207.. The test shall be performed at 50,000 cfm (1107.) flow rate with purge valves fully open.

I Bases Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once every refueling interval to show system

, performance capability.

4 5Sb Amendment No. 55, 68, 108 L________________

\

~

4.15- MAIN STEAM SYSTEM INSERVICE INSPECTION Applicability This. tschnical specification applies to the inservice ' inspection of those welds in the main steam system identified as Numbers.3, 4, and 5 of Figure 6, Supplement 2, Part IX and Number 3 of Figure 9, Supplement 2, Part IX.

Objective The objective of.this inservice inspection program is to provide assurance of the continuing integrity of that portion of the main steam system in which a postulated failure would produce pressures in excess of the compartment wall and/or capacities.

Specification 4.15.1. The four weld joints identified above shall be 100 percent ultrasonically inspected in accordance with the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems dated January 1, 1970 as modified by the Winter 1970 Addenda and the provisions of Appendix IX-3400 of Section III of the ASME Boiler and Pressure Vessel Code. Inspections are to be performed prior to.

startup and subsequently at 3-1/2 year intervals (or nearest refueling outage).

Prior to initial plant operation a preoperational inspection of the identified weld joints will be performed and any data acquired will be recorded to form a baseline on which to compare results of subsequent inspections.

Upon completion of several inspection cycles, the technical benefit of the inspection program frequency will be reviewed. The conclusions of this review shall be submitted to the NRC for evaluation. l Bases Calculations reveal that postulated breaks in the main steam lines at the containment penetrations in small compartments No. 2 and No.

5 could produce pressures in excess of wall and/or slab capacities.

l-l 4-58

/

.4.16,fREACTOR INTERNALS VENT VALVES SURVEILLANCE Applicability.

Applies to Reactor Internals Vent.. Valves.

Objective-

.To verify that no reactor internals vent valve is stuck in the open position and *. hat each valve continues to exhibit freedom of-movement.

Specification Item- . Test Frequency 4.16.1 Reactor Internals Demonstrate Operability ' Each Refueling Vent Valves By: Shutdown l

-a. Conducting a remote visual inspection of.

visually accessible sur-faces of the-valve body and disc sealing faces and evaluating any observed surface irregu-larities.

b. Verifying that the valve is not stuck in an open position, and

c. Verifying through manual actuation that the valve is fully open with a force of < 400 lbs. (applied vertically upward).

Bases Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operation _and therefore insures the conservatism of Core Protection Safety limits as delineated in Figures 2.1-1 and 2.1-3, and the flux / flow trip setpoint.

l 4-59 Amendment No. 65

SHOCK SUPPRESSORS (SNUBBERS) _

SURVEILLANCE REQUIREMENTS (Continued)

c. Refueling Outage Inspections At least once each refueling cycle during shutdown, a visual inspection shall be performed of all safety related snubbers attached to sections of safety systems piping that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems,

d. Visual Inspection Acceptance Criteria Visual inspections shall verify: (1) that there are no visible indications of damage or impaired operability and (2) attachments to the foundation or supporting strJCture are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection in-terval, provided 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 Specification 4.17-lf. When the reservoir outlet port of a snubber is found to be uncovered by fluid, the snubber shall only be declared operable if functional testing in both extension and retraction directions is satisfactory and an engineering evaluation concludes that this snubber is operable.

e. Functional Tests

• At least once each refueling interval during shutdown, a representative sample of snubbers shall be tasted using one of the following sample plans. The sample plan shall be selected prior to the test period Ord cannot be changed during the test period. The NRC Regicaal Administrator shall be notified in writing of the sample plan selected prior to the test period, or the sample plan used in the prior test period shall be used:

1) At least 107. of the total each type of snubber in use -

in the plant shall be functionally tested either in- I place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.17.lf, an additional 107.

of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or The four 550,000 lb reactor coolant pump snubbers are not included. The functional test program for reactor coolant pump I snubbers is implemented in accordance with the schedule and other requirements of the snubber testing program.

4-61 l Amendment No. 30, 106, 110 l

q

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SU0CK SUPPRESSORS (SNUBBERS) u

.' , SURVEILLANCE REQUIREMENTS-(Continued)

F- 1. Snubber Seal Service Program-p A snubber seal service. life program shall'be developed

-whereby'the seal service life of. hydraulic snobbers-is- ,

monitored to. ensure that the service life is not exceeded 1

between' surveillance inspections. The-designated.sarvice H

life for the various seals shall be-established baseo on engineering information. The seals shall be. replaced so

! that.the indicated service life will not be exceeded during a period when the_ snubber is required to be

. OPERABLE. .The seal replacements.shal l be documented and L the documentation shall=be retained in accordance with-Specification 6.10.2.m.

1 i

~

1

[

4-64 Amendment No. 106, 110

2. A seismic occurrence greater than the Operating Basis Earthquake.

3. A loss of coolant accident requirinatio the engineering safeguards, or

4. A major main steam line or feedwater line break.

4.19.4 Acceptance Criteria

a. As used in'this Specification:

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawing or specifications. Eddy current testing indications below 20% of the nominal tube wal thickness, if detectable, may be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of.a tube.

3. Degraded Tube means a tube containing imperfections 20% of the nominal wall thickness caused by degradation.

4.  % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

5. Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

6. Repair Limit means the extent of degradation at or beyond which the tube shall be repaired or removed from service because it may become unserviceable prior to the next inspection.

This limit is equal to 40% of the nominal tube wall thickness.

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a steam line or feedwater line break as specified in i

4.19.3.c., above,

8. Tube Inspection means an inspection of the steam generator tube from the bottom of the upper tubesheet completely to the top of the lower tubesheet, except as permitted by 4.19.2.b.2, above.

4-80 Amendment No.116

, ;F.

4 L,.

'4.19.4 Acceptance Criteria (Continued)-

b. The steam generator shall be determined'0PERABL E af ter completing the corresponding actions 1(removal from service by plugging, or repair by kinetic expansion, sleeving, or

.other methods, of all ~ tubes exceeding the repair limit and p .all tubes containing 'throughwall cracks) required' by Table 4.19.2.

4~.19.5 Reports

a. ~ Following'the completion of each inservice inspection of1 steam generator tubes, the number of tubes repaired-or removed from service in each steam generator shall

-be reported to the NRC within 15 days.

b. The complete results of the steam generator tube inservice inspection shall be reported'to the NRC L

'within 3 months following completion of the inspection.-

This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfect' ion.

'3. Identification of tubes repaired or removed from service.

c. .Results of steam generator tube inspections which fall into Category C-3 require notification in accordance with 10 CFR 50.72 prior to resumption of plant operation. .The written followup of this. report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence in accordance with 10 CFR 50.73.

Bases' The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on modification of Regulatory Guide 1.83, Revision 1. In-service inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event'that there is evidence of mechanical damage or progressive .

I degradation due to design, manufacturing errors, er inservice conditions. Inservice inspection of steam generator tubing also pr7vides a means of characterizing the nature and cause of any tube i l

degradation to that corrective measures can be taken. l Amendment No. 47, 83, 91, 103, 129 l 4-81

r l .

Bases (Continued)

The Unit is expected to be operated in a manner such that the primary and secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the primary or secondary coolant chemistry is not maintained within these chemistry limits, localized corrosion may likely result.

The extent of steam generator tube leakage.due to cracking would be limited by the secondary coolant activity, Specification 3.1.6.3.

The extent of cracking during plant operation would be limited by the limitatior, of total steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 gpm). Leakage in excess of this limit will require plant shutdown and an unschedu'ed inspection, during which the leaking tubes will be located and y aired or removed from service.

Wastage-type defects are unlikely with proper chemistry treatment of the primary or the secondary coolant. However, even if a defect would develop in service, it will be found during scheduled in-service steam generator tube examinations. Steam generator tube irispections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Reaoval from service by plugging, or repair by kinetic expansion, sleeving, or other methods, will be required for degradation equal to or in excess of 40% of the tube nominal wall thickness.

I Where experience in similar plants with similar water chemistry, as documented by USNRC Bulletins / Notices, indicate critical areas to be inspected, at least 50% of the tubes inspected should be from these critical. areas. First sample inspections sample size may be modified subject to NRC review and approval.

Amendment No. 47, 86, 116 4-82 I

I i

' 5.0 DESIGN FEATURES 5.1 SITE 1

Applicability Applies to the location and extent of the exclusion boundary, restricted area, and low population zone.

Objective To define the cbove by location and distance description.  !

Specification 5.1.1 The Three Mile Island Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, PA. It is in Londonderry Township of Dauphin County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties. The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven Dam. Figure 5-1 is an extended plot plan of the site showing l the plant orientation and immediate surroundings. The Exclusion Area as defined in 10 CFR 100.3, is a 2,000 ft.

radius, including portions of Three Mile Island, the river surface around it, and a portion of Shelley Island, which is owned by Met Ed. The minimum distance of 2,000 ft. occurs on the shore of the mainland in a due easterly direction from the plant as shown on Figure 5-1 for the Exclusion Area.

Figure 5-3 showing the physical location of the fence definos the " Restricted Area" surrounding the plant. The minimum distance of the " Restricted Area" is approximately 560 feet and is from the centerline of the TMI Unit 2 Reactor Building to a point on the westerly shoreline of Three Mile Island.

The minimum distance to the outer boundary of the low population zone is two miles as shown on T.S. Figure 5-2, which also depicts the site topography for a radius of five miles. T.S. Figure 5-3 depicts the locations of gaseous effluent release points and liquid effluent outfalls (as tabularized on page 5-10), and the meteorological tower location (designated as ' weather tower' on the figure).

5-1 Amendment No. 72, 137

. _ ,y;;;. - - - - - -

q . [. :p "

+c 9 47. g r: : . .

-ELEVATIONS'FOR-GASE0VS' EFFLUENT RELEASE POINTS (See. Figure 5-3)- Jl 1 ,

i Unit 1" Stack 483'-7"'

. Unit 1 Turbine Buildings 425' 4"

.p g

i Unit 1: Fuel. Handling ~ Building- 348' l- ESF Vent Stack-L LOCATIONS OF LIQUID EFFLUENT OUTFALLS PURSUANT TO NPDES-(See Figure 5-3).

Outfall No. Description DSN 001 Main Station Discharge

~

DSN 002- Emergency Discharge'from' Unit 2 (if DSN 001 is blocked)

DSN 003 Emergency Discharge from Unit 1; (if DSN 001 is blocked)

DSN 004 Emergency. Discharge from Unit 1

.(if Unit.1 NDCT blocked)

DSN 005 Stormater and yard drainage and dewatering'of natural draft cooling towers, maintenance dredging desiltation and basin dewatering, fire r brigade training facility runoff, fire service water runoff.

5-10  !

' Amendment No. 72, 122

1 b-6 . A s -[ s l

TABLE 6.2-1 MINIMUM SHIFT CREW COMP 0SITIONU) I LICENSE CATEGORY

. QUALIFICATIONS Tave > 200* Tave < 200*

SR00") 2 10) l 0

R0 0 ") 2 1 l Non-Licensed Auxiliary Operator 2 1 Shift Technical Advisor 10 ' ) None Required l' O

~ (i)u Does not include the Licensed Senior Reactor Operator or l Senior Reactor Operator Limited to Fuel Handling, supervising (a) irradiated fuel handling and transfer activities onsite,.

and (b) all unirradiated fuel handling and transfer activities to and from the Reactor Vessel.

(ii) May be on a different shift rotation than licensed personnel. l (iii) Except-for the Shift Supervisor, shift crew composition may be l one less than the minimum requirements for a period of time 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 of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew 5.osition to be unmanned upon shift change due to an incoming shift crewman being late or absent.

(iv) Pursuant to the requirements of 10 CFR 50.54(m).

l 6-2 Amendment No. 11, 32, 77

I o m" .

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a. <.

b 6.5.1.9 The Emergency P1an'and implementing procedures shall.be ~

reviewed by' a--knowledgeable ' individual (s)/ group other than -

<the individual (s)/ group which prepared them. ,

6.5.1.10 A knowledgeable individual (s)/ group shallI review every; unplanned onsite release of radioactive, material to the

-environs including the preparation.and forwarding of 4 reports to the Vice President THI-1. covering evaluations, recommendations-and disposition of.the corrective action to prevent' recurrence.

6.5.1.11 Major changes to' radwaste systems shall be' reviewed. by a .

knowledgeable individual (s)/ group other.than the individuals (s)/ group which prepared them.

6.5.l.12 .. Individuals responsible for reviews performed in accordance with 6.5.1.1 through 6.5.1.4 shall include a determination of whether or not additional cross-disciplinary review is necessary. If-deemed necessary, such review shall be performed by the appropriate personnel. Individuals responsible-for reviews considered under 6.5.1.1 through 6.5.1.5 shall render determinations in writing with regard to whether or not 6.5.1.1 through 6.5.1.5 constitute an unreviewed safety question.

. RECORDS 6.5.1.13 Written records of activities performed under Specifi-cations 6.5.1.1 through 6.5.1.11 shall be maintained.

QUALIFICATIONS' 6.5.1.14 Responsible Technical Reviewers shall meet or exceed the

. qualifications of ANSI /ANS 3.1 of 1978 Section 4.6, or 4.4 for applicable disciplines, or have 7 years of appropriate experience in the field of his specialty. Credit toward experience will be given for advanced degrees on a one-to-one basis up to a maximum of two years. Responsible Technical Reviewers sha te designated in writing.

6.5.2 IIIDEPENDENT SAFETY REVidW FUNCTION

6. 5. 2.1 The Vice President of each division within GPU Nuclear Corporation shall be responsible for ensuring the l independent safety review of the subjects described in 6.5.2.5 within his assigned area of safety review responsibility, as assigned in the GPUN Review and Approval Matrix.

6.5.2.2 Independent safety review shall be completed by an individual / group not having direct responsibility for the performance of the activities under review, but who may be from the same functionally cognizant organization as the individual / group performing the original work.

6.5.2.3 GPU Nuclear Corporation shall collectively have or have access to the experience and competence required to independently review subjects in the following areas:

Amendment No. 11, 22, 77,.139 6-5 ]

-_L_--__-_-___-___--_-_--__________________}

1. . s..-

' 6.9.4.2.5 The Radioactive Effluent-Release Reports shall include the l- instrumentation not returned to OPERABLE status within 30 3 days.per TS 3.21.1.b and TS 3.21.2.b.

.rc .

6.9.4.3 The following information shall be included in the Radioactive Effluent Release Report to be submitted 60 days after_ January 1 of each year.

6.9.4.3.1 The Radioactive Effluent Release Report to be' submitted 60 I days after January 1 of each year shall include an' annual I summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmosphere stability, and. precipitation (if measured)'on magnetic tape, or.in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability.

6.9.4.3.2 The Radioactive Effluent Release Report to be submitted 60=

days after January 1 of each year shall: include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

6.9.4.3.3 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall include an assessment of the radiation doses-from radioactive liquid and gaseous effluents to MEMBERS 0F THE PUBLIC due to their activities inside the site boundary (Figure 5-3) l during the report period. All. assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM).

6.9.4.3.4 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and othe'r nearby uranium fuel cycle sources including doses from primary effluent pathways and direct radiation for the previous 12 consecutive months to show conformance with 40 CFR 190

" Environmental Radiation Protection Standards for Nuclear Power Operation". Acceptable methods for calculating the dose contributions from Liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1. l 6-19 Amendment No. 72, 77, 129, 137 1

h?$h,

~

?'" ~ 6L10 RECORD RETENTION (

-; 1 6.10.1' The-following records.shall.be' retained for-at least.five years:  ;

u La. '

Records of normal station.operationTincluding power'

. levcis_and periods offoperation at eachipower level.

b.- Records of principal maintenanceJactivities, including' inspection, repairs,; substitution,for: ..

replacement of principal items lof. equipment related

'? to nuclear, safety,

c. All' REPORTABLE-EVENTS

d. Records of periodic checks, tests-and' calibrations.

Records of reactor physics tests and other'special-

~

e.

tests related to nuclear safety,

f. Changes to. procedures required by_ Specification' 6.8.1.

g. Records of solid radioactive shipments.

6-19a l Amendment No. 72, 77, 129, 137, 141

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SEE APERTURE CARDS l NUMBER OF OVERSIZE PAGES FILMED ON APERTURE CARDS

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APERTURE CARD /HARD COPY AVAILABLE FROM DECORDS AND REPORTS MANAGEMENT BRANCH

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