ML20066B415

Amend 133 to License DPR-35,modifying Spec Having cycle- Specific Parameter Limits by Replacing Value of Units W/Ref to Core Operating Limits Rept for Values of Limits

ML20066B415
Person / Time
Site:	Pilgrim
Issue date:	01/02/1991
From:	Wessman R Office of Nuclear Reactor Regulation
To:
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ML20066B417	List: ML20066B415 ML20066B419
References
NUDOCS 9101070125
Download: ML20066B415 (42)

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((,-.., k UNITED STATES r

g NUCLE AR REGULATORY COMMISSION l

W ASHINGTON, O C. 20665

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i BOSTON EDISON COMPAtlY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMEfDMENT TO FACILITY OPERATING LICENSE Amendment No. 133 License No. DPR-35 1.

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

A.

The application for amendment filed by the Boston Edison Company (thelicensee)datedAugust 21, 1990, as supplemented on November 8, 1990 and December 3, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter It B.

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

There is reasonable assurance: (1) 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 Comission's regulations set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is e,n, ended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No OPR-35 is hereby amended to read as follows:

Technical Specificatio_n_s The Technical Specifications contained in Appendix A, as revised 4

through Amendment No.133, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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P

2 3.

This license amendment is effective as of its date of issuance and shall be innplemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION D

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Richard Wessne, Director Project Directorate T-3 Division of Reactor Projects - 1/I!

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 2,1991

)

tjiACfNENT___TO LICENSE AMENDMENT NO.133 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET _NO. 50 293 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

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

Remove Insert i

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9 10 10 11 12 13 13a 13b 14 15 16 17 18 19 20 21 22 23 24 25 26 26 27 27 29 29 36 36 37 37 38 38 39 39 40 40 40a*

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Remove Insert 126 126 127 127 127A 127a 145 145 146 143 205A 205a 205A-1 205B 205b 205B-1 2058-2 205C 205c 205C-1 205C-2 205C-3 2050-4 205C-5 205C-6 205D 205d 205E 205e 205E-1 205E-2 205E-3 205E-4 205E-5 205E-6 205F 205f 205G 205H 206m 206m 216 216 217 217 217a*

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l TABLE OF CONTENTS Pace No.

ifdlNITIONS 1

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.0

'AFETY LIMITS l

h 2.1 h fety L!M t!

6

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2.2 Safety Limit Violation 6

Limitina Conditions For Ooeration Surveillance 'neauirement 3.1 REACTOR PROTECTION SYSTEM 4.1 26 3.2 PROTECTIVE INSTRUMENTATION 4.2 42 3.3 REACTIVITY CONTROL 4.3 80 A. Reactivity Limitations A

80 B. Control Rods B

81 C. Scram Insertion Times C

83 D. Control Rod Accumulators D

84 E. Reactivity Anomalies E

85 F. Alternate Requirements 85 G.-Scram Discharge Volume G

85 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 95 A. Normal System Availability A

95 B. Operation with Inoperable Components B

96 C. Sodium Pentaborate Solution C

97 D. Alternate Requirements 97 l

_3.5 CORE AND CONTAINHENT COOLING SYSTEMS 4.5 103

.A. Core _ Spray and LPCI Subsystems A

103-B. Containment Cooling Subsystem B

106 C. HPCI Subsystem C

107

(

D. RCIC Subsystem D

108 l

E. Automatic Depressurization System E

109.

l F.~Hinimum Low Pressure Cooling System F

110 L

and Diesel Generator Availability s

G. (Deleted)

G 111 H. Maintenance of Filled Discharge Pipe H

112 3.6 PRIMARY SYSTEM BOUNDARY 4.6

-123

'A' Thermal and Pressurization Limitations A

123 B. Coolant Chemistry-B 124 C. Coolant Leakage C

125

'D.' Safety-and Relief Valves D

126 E. Jet Pumps E

_127 F. Jet Pump Flow Mismatch'

_ F_

127 G. Structural Integrity G

127a H. Deleted H

127a I. Shock Suppressors-(Snubbers)

I 137a Amendment No. - 75,.45, 65, 133, 1

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Surveillante Pace No.

l 3.7 CONTAINHENT SYSTEMS 4.7 152 l

A.

Primary Containment A

152 B.

Standby Gas Treatment System and B

138 Control Room High Efficiency l

Air Filtration System j

C.

Secondary Containment C

159 3.B RADIOACTIVE EFFLUENTS 4.8 177

)

A.

Liquid Effluents Concentration A

177 i

B.

Radioactive Liquid Effluent B

177 4

Instrumentation i

C.

Liquid Radwaste Treatment C

178 D.

Gaseous Effluents Dose Rate D

179 4

E.

Radioactive Gaseous Effluent E

180 Instrumentation l

F.

Gaseous Effluent Treatment F

181 G.

Main Condenser G

182 H.

Mechanical Vacuum Pump H

183 i

3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 194 A.

Auxiliary Electrical Equipment A

194 B.

Operation with inoperable Equipment 196 l

3.10 CORE ALTERATIONS 4.10 202 l

A.

Refuelin Interlocks A

202 B.

Core Mon toring 8

.202 C.

Spent Fuel Pool Hater Level C

203 D.

Multiple Control Rod Removal D

203 3.11 REACTOR FUEL ASSEMBLY 4.11 205a 1

A.

Average Planar Linear Heat A

205a Generation Rate (APLHGR)

B.

Linear Heat Generation Rate (LHGR)-

B 205b-C.

Minimum Critical Power Ratio (MCPR)

C-205b 0.

Power / Flow Relationship D

205d 3.12 FIRE PROTECTION 4.12 206 A.

Fire Detection Instrumentation A

206 l

B.

Fire Water Supply-System B

206a l

L C.

Spray and/or Sprinkler Systems C

206c l

D.

Halon System D

206d E.

Fire Hose Stations E-206e F.

Fire Barrier-System F

206e-1 l

G.

Alternate Shutdown Panels G

-206e-1 1

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Amendment No.

15, 27, 45,'84, 89, 113, 174, 133

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4.0 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES 206k i

4.1 Sealed Source Conamination 206k i

4.2 Surveillance Requirements 206k 4.3 Reports 2061 j

4.4 Records Retention 2061 5.0 MAJOR DESIGN FEATURES 206m i

L 5.1 Site Features 206m 5.2 Reactor 206m 5.3 Reactor Vessel 206m 5.4 Containment 206m 5.5 Fuel Storage 207 i

5.6 Seismic Design 207 6.0 ADMINISTRATIVE CONTROLS 208 i

6.1 Responsibility 208 L

6.2 Organization 200 6.3 Unit Staff Qualifications 209 t

6.4 Training 209 3

6.5 Review and Audit 211 6.6 Reportable Event Action 215 6.7 Deleted 216 l

6.8 Procedures 216 6.9 Reporting Requirements 217 6.10 Record Retention 219 4

6.11 Radiation Protection Program 221 6.12 (Deleted) 6.13 High Radiation Area 221 6.14 Fire Protection Program 222 i

Ocetational Obiettives IVntillance

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7.0 RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM 8.0 229 7.1 Honitoring Program-8.1

'229 7;2-Dose - Liquids 8.2 232 7.3 Dose - Noble Gases 8.3 233 7.4 Dose - Iodine-131. Iodine-133, 8.4 234 Radioactive Material in L

Particulate Form, and Tritium 7.5 Total Dose 8.5 234 Amendment No. 35, 45, 88, 89, 95, 122, 132, 133 tii er"*"-pi---

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i 1.0 DEFINITIONS j

k The succeeding frequently used terms are explicitly defined so that a I

uniform interpretation of the specifications may be achieved.

l A. Safety Limit - The safety limits are limits below which the reasonable I

maintenance of the cladding and primary systems are assured.

Exceeding such a limit is cause for unit shutdown and review by the i

Nuclear Regulatory Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to e

regulatory review.

B. Limitino safety System Settino (LSSS)

.The limiting safety system 1

i settings are settings on instrumentation which initiate the automatic i

protective action at a level such that the safety limits will not be exceeded.

The region between the safety limit and these settings a

represent margin with normal operation lying below these settings.

The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.

C. Limitina Conditions for Ooeration (LCO) - The limiting _ conditions for

_ operation specify the minimum acceptable levels of system performance L

necessary to assure safe startup and operation of the facility.

When these conditions are met, the plant can be c,,,erated safely and

. abnormal situations can be safely controlled.

D. CORE OPERATING LIMITS REPOSI The CORE OPERATING LIMITS REPORT is a reload-cycle specific document.

o its supplements and revisions, that provides core operating l'.mits for the current operating reload cycle.

These cycle specific core

^

operating limits shall be determined for each reload cycle in accordance with Specification 6.9. A.4.

Plant operation within these operating limits is addressed in individual specifications.

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1 Amendment No. 133 1

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2.0 SAFETY LIMITS l

l 2.1 SAFETY-LIMITS 3

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I 2.1.1 Hith the reactor steam dome pressure < 785 osig or core flow < 10% of rated core flow:

l THERMAL POWER shall be 1 25% of RATED THERMAL POWER.

l 2.1.2 Hith the reactor steam dome pressure 1 785 psig and core flow 2 10% of rated core flow:

i HINIMUM CRITICAL PONER RATIO shall be 2 1.04.

l 2.1.3 Whenever the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of j

the normal active fuel zone.

I 2.1.4 Reacter steam dome pressure shall be 1 1325 psig at any time when irradiated fuel is present in the reactor vessel.

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2.2 SAFETY LIMIT VIOLATION With any Safety Limit not met the following actions shall be met:

2.2.1 Hithin one hour notify the NRC Operations Center in accordance with 10CFR50.72.

2.2.2 Hithin two hours:

A.

Restore compliance with all Safety Limits, and B.

Insert all insertable control. rods, 2.2.3 Th( Station Director and Senior Vice President'- Nuclear c-and the Nuclear Safety Review and Audit Committee (NSRAC) shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.2.4 A Licensee Event Report shall'be prepared pursuant to-10CFR50.73.

The Licensee Event Report shall be submitted to the Commission, the Operations Review Committee (ORC),

the NSRAC and the Station Director and Senior Vice President Nuclear within 30 days of the violation.

2.2.5 --Critical operation of the unit shall.not be resumed until authorized by the Commission.

1 Amendment No. 15, 27, 42, 72, 133, 6

i' 44 B 2.0 SAFETY LIMITS BASES e

INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients.

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a stepback approach is used to establish a Safety Limit such that the Minimum Critical Power Ratio (HOPR) is not less than the limit specified in Specification 2.1.2.

HCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some

- corrosion or use-related cracking may occur during the life of.

the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross a.

rather than incremental cladding deterioration.

Therefore,.

the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling (i.e., MCPR of 1.0).

These conditions represent a significant departure from the condition intended by design for planned operation.

The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling.

FUEL CLADDING GE critical power correlations are applicable for all INTEGRITY critical power calculations at pressures at or above 785 psig (2.1.1) or core flows at or above 10% of rated-flow.

For operation at low pressures and low flows another basis is used as follows:

(continued)

Amendment No. 15, 42, 72, 10b, 129, 133-7 m-m.

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o BASES (continued)

FUEL CLAODING Since the pressure drop in the bypass region is essentially INTEGRITY all elevation head, the core pressure drop at low power and (2.1.1) flows will always be greater than that with a bundle flow of 28 x 10j.5 psi. Analyses show (continued) lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bundle f willbegreaterthan28x10jowwitha4.5psidrivinghead lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MHt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

Thus, a THERMAL' POWER 1imit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (1), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.

The GEXL correlation is valid over the range of condi*. tons used in the tests of the data used to develop the correlation.

These conditions are:

Pressure:

0.1to1.25x10gg 800 to 1400 ps lb/hr-ft2 Max Flux:

Inlet Subcooling:

0 to 100 Btu /lb local Peaking:

1.61 at a corner rod to 1.47 at an interior rod Axial Peaking:

Shape Max / Avg.

Uni form 1.0 Outlet Peaked 1.60 Inlet Peaked 1.60 Double Peak 1.46 and 1.38 Cosine 1.39 Rod Array 16,64 Rods in an 8x8 array 49 Rods in an 7x7 array HINIHUM The fuel cladding integrity Safety Limit is set such that no CRITICAL POWER fuel damage is calculated to occur if the limit is not RATIO violated. Since the parameters which result in fuel damage (2.1.2) are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.

Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at which (continued)

Amendment No. 15, 42, 72, 105, 129, 133, 8

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boiling transition is calculated to occur has been adopted as I

a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering l

the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations.

Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1.

Reference 1 includes a tabulation of the uncertainties used N the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis.

The statistical analysis used to determine the MCPR safety limit is based on a model of the BWR core which simulates the process computer function.

The reactor core selected for these analyses was a large 764 assembly, 251 inch reload core.

Results from the large reload core analysis apply for all operating reactors for all relcad cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical representation of the core and the resulting bundle critical power ratios.

Details of this statistical analysis are presented in Reference 2.

REACTOR Hith fuel in the reactor vessel during periods when the HATER LEVEL reactor is shutdown, consideration must be given to water (Shutdown During periods when the reactor is shutdown, consideration l

Condition) must also be given to water level requirements due to the (2.1.3) effect of decay heat.

If reactor-water level should drop.

below the top of the active fuel'during this time, the ability to cool the core is reduced.

This reduction in. core cooling L

capability could lead to elevated cladding temperatures and clad perforation.

The core can be cooled sufficiently should the water level be reduced to two-thirds the core height.

Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin.

This level will be continuously monitored.

-Amendment No. 75, 42, 72, 133, 9

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BASES (continued)

REACTOR The Safety Limit for the reactor steam dome pressure has been STEAH DOME selected such that it is at a pressure below which it can be PRESSURE shown that the integrity of the system is not endangered.

(2.1.4)

The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressdre Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.

The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system.

The reactor coolant system is designed to the USAS Nuclear Power Piping Code, Section B31.1.0 for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig at 562'F for suction piping and 1241 psig at 562'F for discharge piping. The pressure _ Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.

REFERENCES 1.

" General Electric Standard Application for Reactor Fuel,"

NEDE-240ll-P-A (Applicable Amendment specified in the CORE OPERATING LIMITS REPORT).

2.

General Electric Thermal Analysis Basis (GETAB):

Data Correlation and Design Application, General Electric Co.

BWR Systems Department, November 1973 (NEDO-10958).

3.

Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June 1974

.(NED0-20340).

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- Amendment No. ]f,133 (Next page is 26) 10

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM 3

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Aeolicability:

Acelicability:

i l

I Applies to the instrumentation Applies to the surveillance of l

and associated devices which the instrumentation and initiate a reactor scram.

associated devices which initiate reactor scram.

}

Obiettive:

1 Obiettive:

To assure the operability of the reactor protection system.

To*specify the type and frequency I

of surveillance to be applied to Seecification:

the protection instrumentation.

s l --

A.

The setpoints, minimum number of Seecification:

j trip systems, and minimum number of instrument channels that must A.

Instrumentation systems shall be j

be operable for each position of functionally tested and the reactor mode switch shall be calibrated as indicated in Tables as given in Table 3.1.1, The 4.1.1 and 4.1.2 respectively.

system response times from the opening of the sensor contact up B.

Verify the maximum fraction of to and including the opening of limiting power density is-less 1

the trip actuator-contacts shall than or equal to the fraction of not exceed 50 milli-seconds, rated power once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after trermal power is greater B.

The maximum fraction of limiting than or equal to 25% of rated l

power density (HFLPD) shall be thermal power and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> less than or equal to the thereafter, fraction of rated power (FRP) when thermal power is greater than or equal to 25% of rated thermal power.

1.

If HFLPD is greater than FRP, l

adjust the APRM high flux scram and rod block trip i

l setpoints to the relationships:specifled in the. CORE OPERATING LIMITS REPORT within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

.If the required actions and associated completion times of Specification 3.1.B.1, Labove cannot be met, reduce thermal power to less-than 25% of' rated thermal power within 4 f.ours.

Amendment No. 42,129. I33 26

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TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum Number Modes in Hhich Function Operable Inst.

Trip Function Trip Level Setting Must Bg_Qugrable Actien(I)

Channels per Refuel (7) Startup/Hct Run Trio (1) System StandDY 1

Mode Swi.ch in Shutdown X

X X

A 1

Manual Scram X

X X

A IRM 1 20/125 cf full scale X

X (5)

A 3

High Flux 1

3 Inoperative X

X (5)

A APRM 2

High Flux (15)

(17)

(17)

X A or B l 2

Inoperative (13)

X X(9)

X A or B 2

High Flux (151) 1151 of Design Power X

X (16)

A or 8

$ 085 psig X(10)

X X

A 1

2 High Reactor Pressure 2

High Drywell Pressure 12.5 psig X(8)

X(8)

X A

2 Reactor Low Hater Level 29 In. Indicated Level X

X X

A 2

High Hater Level in Scram Discharge Instr. Volume 139 Gallons X(2)

X X

A 1

2 Main Condenser Low Vacuum 223 In. Hg Vacuum X(3)

X(3)

X A or C 2

Main Steam Line High 17X Normal Full Power Radiation Background (18)

X X

X(18)

A or C 4

Main Steam Line Isolation 1 01 Valve Closure X(3)(6)

X(3)(6)

X(6)

A or C Valve Closure 1

1 50 psig Control Oil 2

Turbine Control Valve 1

Fast Closure Pressure at Accelera'. ion Relay X(4)

X(4)

X(4)

A or D 5 01 Valve Closure X(4)

X(4)

X(4)

A or D 4

Turbine Stop Valve Closure 1

Amendment No. J5, 42, E6, 92, II7.133 27

'3 NOTES FOR TABLE 3.1.1 (C "0) 10.

Not rmired +- be operable when the reactor pressure vessel head is not boltM to the vessel.

11.

Selet :d 12.

OtW ed 13.

An APRH will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of LPRH's to an APRK.

14.

0.e'eted 15.

The APRM high flux trip level setting shall be as specified in the CORE OPERATING LlHITS REPORT, but shall in no case exceed 120% of rated thermal power.

16.

The APRM (15%) high flux scram is bypassed when in the run mode.

17.

The APRH flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.

18.

Hithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen.

The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection.

The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.

Amendment No. 15, 27, 42, 86, !!7, IIB,133 29

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3.1 Mi[S (Cont'd) 4.i Mi[i (Cont'd) l been provided to allow for To facilitate the implementation bypassing of one such channel, of this technique, Figure 4.1.1 is provided to indicate an APRM appropriate trend in test J

interval.

The procedure is as The average power range follows:

monitoring (APRM) system, which i s calibrated using heat balance 1.

Like sensors are pooled into data taken during steady-state one group for the purpose of i

conditions, reads in percent of data acquisition.

design power (1998 MHt).

Because fission chambers provide the 2.

The factor M is the exposure basic input signals, the APRM bcurs and is equal to the-system responds directly to number of sensors in a average neutron flux.

During group, n, times the elapsed transients, the instantaneous time T (M = nT).

rate of heat transfer from the fuel (reactor thermal power) is 3.

The accumulated number of less than the instantaneous unsafe failures is plotted neutron flux due to the time as an ordinate against M as constant of the fuel.

Therefore, an abscissa on Figure 4.1.1.

during abnormal operational transients, the thermal power of 4.

After a trend is the fuel'will be less than that established, the appropriate indicated by the neutron flux at monthly test interval to

-the scram setting. Analyses satisfy the goal _will-be the demonstrated that with a 120 test interval to the left of percent scram trip setting. none the plotted points, of the abnormal operational transients analyzed violate the 5.

A test interval of one month fuel safety limit and there is a will-be used initially until substantial margin from fuel a trend is established.

damage.

Therefore, the use of flow referenced scram trip Gr'>up (B) devices utilize an provides even adottional margin.

analog sensor followed by an amplifier and a bi-stable trio An increase in the APRM scram circuit.

The sensor and setting.would decrease the margin amplifier are active components present before the fuel cladding and a failure is almest always integrity safety limit is accompanied by an alarm and an reached.

The APRM scram setting indication of the source of was determined by an analysis of trouble.

In the event of margins required to provide a failure, repair or substitution reasonable range f or maneuvering can start immediately.

An

- during operation.. Reducirg this

• as-is" failure is one that operating margin would increase

" sticks" mid-scale and is not-the frequency of sp'urious scrams, capable-of going either up or which have an adverse effect on down in response to an reactor safety because of the out-of-limits input.

This type resulting thermal stresses, of failure for analog devices is l

i Amendment No. 79.133 36 e

s

+..-r

,%,,,.,...,e s-,-

.ve,.

w.-

-,4w.,y c

,w_-.,

.-m.,,,-.--..,,._,-w.m,,%m-wwn,-

,e~.e.,

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1 2

}

l 3.1' E (Cont'd) 4.1 E (Cont'd) 1 Thus, the APRM setting was a rare occurrence and is j

selected because it provides detectable by an operator who adequate margin for the fuel observes that one signal does not cladding Integrity-safety limit-track the other three. -For i.

yet allows operating margin that purpose of analysis, it is reduces the possibility of assumed that this rare failure t

i unnecessary scrams, will be detected within-two hours, j

Analyses of the limiting The bi-stable trip circuit which J

transients show.that no scram is a part of the Group (B) adjustment is required'to assure devices can sustain unsafe

]

MCPR greater than the-Safety

' failures which are revealed only Limit MCPR when the transient is on test.

Therefore, it is 4

l-initiated from HCPR above the necessary to test them j

operating limit MCPR.

periodically.

For operation in the startup mode A study was conducted of the while the reactor.is at low instrumentation channels included pressure, the APRH scram setting in the Group (B) devices to of 15 percent of rated power

-calculate their " unsafe" failure o

provides adequate thermal margin rates.

The analog devices L

between the setpoint and the (sensors and amplifiers) are i

safety limit, 25 percent of predicted to have an unsafe rated.

The margin is adequate to falgure rate of less than 20 X accommodate anticipated maneuvers 10- failure / hour. The bi-stable associated with power plant trip circuits are predicted to i

startup.

Effects of increasing haveanunsafefa{lurerateof-less than 2 X 10-p pressure at zero or low void content are minor, cold water failures / hour. Considering the from sources available during two hour monitoring interval for l

startup is not much colder than the analog devices _as assumed that already in the system, above, and a weekly test interval temperature coefficients are for the bi-stable trip circuits, small, and control rod patterns the design reliability goal of L

are constrained to be uniform by 0.99999 is attained with ample operating procedures backed up by

margin, the rod worth minimizer.

The bi-stable devices are

'Horth of individual rods is very monitored during plant operation low in a uniform rod pattern, to record their failure history Thus, of all possible sources of and establish a test interval reactivity input, Uniform control using the curve of Figure 4.1.1.

rod withdrawil is the most There are numerous identical-probable case of significant bi-stable devices used throughout power rise.

Because the flux the plant's instrumentation distribution associated with system.

Therefore, significant uniform rod withdrawals does not data on the failure rates for the involve high local peaks, and-bi-stable devices should be because several rods must be accumulated rapidly, moved to change power by a significant percentage of rated l

Amendment No. 79,133 37

i 4g 3

3.1 LAjiS (Cont'd) 4.1 Mif,3 (Cont'd) power, thu rate of power rise is The frequency of calibration of

}

very slow.

Generally ths heat the APRH flow Biasing Network has j

flux is in the near equilibrium been established as each with the fission rate.

In an refueling outage.

The flow 4

l assumed uniform rod withdrawal biasing network is functionally j

approach to the scram level, the tested at least once per month rate of power rise is no more and, in addition, cross 4

that five percent of rated power calibration checks of the flow per minute, and the APRH system input to the flow biasing network 4

j would be more than adequate to can be made during the functional assure a scram before power could test by direct meter reading.

exceed the safety limit.

The 157.

' There are several instruments APRM scram remains active until which must be calibrated and it l

the mode switch is placed in the will take several days to perform RUN position.

This switch occurs the calibration of the entire when reactor pressure is greater network. While the calibration than 880 psig, is being performed, a zero flow i

signal will be sent to half of The analysis to support operation the APRH's resulting in a half at various power and flow scram and rod block condition.

relationships has considered Thus, if the calibration were operation with two recircuiation performed during operation, flux

pumps, shaping would not be possible.

Based on experience at other l!LM generating stations, drif t of instruments, such as those in the i

The IRM system consists of-8 Flow Biasing Network, is not chambers, 4 in each of the significant and therefore, to reactor protection system logic avoid spurious scrams, a channels.

The IRH is a 5-decade calibration frequency of each instrument which covers the range refueling outage is established, of power level between that covered by the SRM and the APRM.

Group (C) devices are active only The 5 decades are covered by the during a given portion of the IRM by means of a range switch operational cycle, for example, and the 5 decades are broken down the IRH is active during startup into 10 ranges, each being and inactive during full-power one-half of a decade in size, operation.

Thus, the only test that is meaningful is the one The IRM scram setting of 120/125 performed just prior to shutdown of full scale is active in each or startupl 1.e., the tests that range of the IRM.

For example, are performed just prior to use if the-instrument were on range of the instrument.

-1, the scram setting would be a 120/125 of full scale for that Group (D) devices, while similar rangel likewise, if the in description to those in Group instrument were on range 5, the (B), are different in use because scram would be 120/125 of full they (the analog transmitter / trip scale on that range.

Thus, as unit devices) provide alarms, the IRH is raged up to trips or scram functions.

An i

L accommodate the increase in power availability analysis is detailed in NED0-21617A (12/78).

l Amendment No. 79,133 38 l

N-

-9 o

3.1 M S (Cont'd) 4.1 fMSES (Cont'd) level, the scram setting is also Surveillance frequencies for the ranged up.

The most significant SOV system instrumentation is sources of reactivity change detailed in Amendment Number 65.

during the power increase are due NRC concurrence with this to control rod withdrawal.

For surveillance program is contained in-sequence control rod in the Safety Evaluation Report withdrawal, the rate of change of and its associated Technical power is slow enough due to the Evaluation Report (TER-C-5506-66) physical _ limitation of dated 11/10/82.

withdrawine control rods that heat flux is in equilibrium with Calibration frequenry of the the neutron flux, and an IRM

' instrument channel ii divided scram would result in a reactor into two groups.

Thise are as shutdown well before any safety follows:

limit is exceeded.

1.

Passive type indicating In order to ensure that the IRM devices that can be compared provided adequate protection with like units on a against the single rod withdrawal continuous basis, error, a range of rod withdrawal accidents was analyzed.

This.

2.

Vacuum tube or semiconductor analysis included starting the devices and detectors that accident at various power drift or lose sensitivity, levels.

The most severe case involves an initial condition in Experience with passive type which the reactor is just instruments in generating subcritical and the IRM system is stations and substations not yet on scale.

This condition indicates that the,pecified exists at quarter rod density..

calibrations are adequate.

For Additional conservatism was taken those devices which employ in this analysis by assuming that amplifiers, drift specifications the IRM channel closest to the call for drift to be less than withdrawn rod is bypassed.

The 0.4%/ month; i.e., in the period results of this analysis show of a month a drif t of.4% would that the reactor is scrammed and occur and thus providing for peak core power limited to one adequate margin.

For the APRM percent of rated power, thus:

system, drift of electronic maintaining MCPR above the Safety apparatus is not the only Limit MCPR.

Based on the above consideration in determining a analysis, the IRM provides calibration frequency. Change in protection against local control power distribution and loss of rod withdrawal errors and chamber sensitivity dictate a continuous withdrawal of control chlibration every seven days, rods.in sequence and provides Calibration on this frequency backup protection for the APRM.

assures plant operation at or below thermal limits.

Reactor low Hater Lev.ej, The set point for low level scram is above the bottom of the separator skirt.

This level has 39 Amendment No. 79,133 l

- ~ _ - _ _. - -

.,. _ _. - _ _ _ - _. - - - -. _ -,. -,, _ _ _., _, =,

y o

3.1 E (Cont'd) 4.1 E (Cont'd) been used in transient analyses A comparison of Tables 4.1.1 and dealing with coolant inventory 4.1.2 indicates that two l

decrease.

The results show that instrument channels have not been scram at this level adequately included in the latter Table.

protects the fuel and the These are:

node switch in pressure barrier, because MCPR shutdown and manual scram.

All

)

remains wall above the safety of the devices or sensors i

limit MCPR in all cases, and associated with these scram system pressure does not reach functions are simple on-off the safety valve settings.

The switches and, heitee, calibration scram setting is approximately 25 during operation is not in, below the normal operating applicable, i.e., the switch is range and is thus adequate to either on or off.

avoid spurious scrams.

B.

The Maximum Fraction of Limiting lurbine Stoo Valve Closure Power Density (MFLPD) shall be checked once per day to determine The turbine stop valve closure if the APRM scram requires scram anticipates the pressure, adjustment.

This will normally neutron flux and heat flux be done by checking the LPRM increase that could result from readings. Only a small number of rapid closure of the turbine stop control rods are moved daily and valves.

With a scram trip thus the MFLPD is not expected to setting of 1 10 percent of valve change significantly and thus a closure from full open, the daily check of the MFLPD is resultant increase in surface adequate.

heat flux is limited such that MCPR remains above the safety The sensitivity of LPRM detectors limit MCPR even during the worst decreases with exposure to case-transient that assumes the neutron flux at a slow and turbine bypass is closed, approximately constant rate.

This is compensated for in the

-Turbine Control Valve Fast Closur_t APRM system by calibrating every three days using heat balance

.The turbine control valve fast data and by calibrating L

closure scram anticipates the individual LPRH's every 1000 l

pressure, neutron flux, and heat effective full power hours using flux increase that could result TIP traverse data.

j from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves.

The reactor protection system initiates a scram when fast closure of the control valves is initiated by the acceleration relay.

This. setting and the fact that control valve closure time' is approximately twice as long as 40 Amendment No. 42,133

q o

i 3.1 MSES (Cont'd) that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR.

l Main Condenser low Vacuum To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves.

To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low condenser vacuum initiates a scram.

The low vacuum scram set point is selected to initiate a scram before the closure of the turbine stop valves is initiated.

Main Steam line Isolation Valve Closure The low pressure isolation of the main steam lines at 880 psig (as specified in Table 3.2.A) was provided to protect against rapid reactor depressurization and the resu!U Q rapid cooldown of the vessel.

Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur; thus providing protection for the fuel cladding integrity safety limit.

Operation of the reactor at pressures lower than 785 psig requires that the reactor mode switch be in the STARTUP position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram and APRM 15% scram.

Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.

In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure.

With the scrams set at 10 percent of valve closure, neutron flux does not increase, Hiah React 6r Pressure The high reactor pressure scram setting is chosen slightly above the maximum normal operating pressure to permit normal operation without spurious scram, yet provide a wide margin to the ASME Section III allowable reactor coolant system pressure (1250 psig, see Basis Section 3.6.D).

Hich'Drvwe11 Pressure Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to crMicality. This instrumentation is a backup to the reactor vessel water level instrumentation.

40a l

Amendment No.133 i

~

~

i l

1-l 3.1 BASES (Cont'd)

Main Steam Line Hiah Radiation l

l High radiation levels in the main steam line tunnel abovt that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven j

times normal background.

The purpose of this scram is t; reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination.

Discharge of excessive amounts of radioactivity to i

the site environs is prevented by the air ejector off-gas monitors which cause an isolation of the main condenser off-gas line.

4 Reactor Mode Switch A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

Ref. Section 7.2.3.7 FSAR.

Manual Scram The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation, i

_ Scram Discharae Instrument Volume The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.

The two scram discharge volumes accommodate in excess of 39 gallons of water each and are at the low points of the scram 4

discharge piping. No credit was taken for these volumes in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.

During normal operation the scram discharge volume system is empty; however, should it fill with water, the water discharged to the piping could not be accommodated, which would result in slow scram times or partial control rod insertion.

To preclude this occurrence, redundant and diverse level detection devices in the scram discharge instrument volumes have been provided which will alarm when water level reaches 4.6 gallons, initiate a control rod block-at 18 gallons, and scram the reactor when the water level reaches 39 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods.

This function l

shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.

A source range monitor (SRH) system is also provided to supply additional neutron level information-during_ start-up but has no scram functions.

t l.

Ref. Section 7.5.4 FSAR.

The APRH's cover the " Refuel" and "Startup/ Hot Standby" modes with the APRH 15% scram, and the power range with the flow Amendment No.133 40b l

3.1 Mil 5 (Cont ' d) biased rod block 6nd scram.

The IRH's provide additional protection in the " Refuel" and "Startup/ Hot Standby" modes.

Thus, the IRH and APRM 15%

scram are required in the " Refuel" and "Startup/ Hot Standby" modes.

In the power range the APRM system provides the required protection.

Ref, Section 7.5.7 FSAR.

Thus, the IRH system is not required in the "Run" mode.

The high reactor pressure, high drymell pressure, reactor low water level and scram discharge volume high level scrams are required for Startup/ Hot Standby and Run modes of plant operation.

They are, therefore, required to be operational for these modes of reactor operation.

The requirement to have the scram functions, as indicated in Table 3.1.1, operable in the Refuel mode is to assure that shifting to the Refuel mode during reactor power operation does not diminish the need for the reactor protection system.

The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit.

Below 305 psig turbine first stage pressure (45% of rated), the scram signal due to turbine stop valve closure is bypassed because flux and pressure scram are adequate to protect the reactor.

The requirement that the IRH's be inserted in the core when the APRH's read 2.5 indicated on the scale assures tnat there is proper overlap in the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation.

The provision of an APRM scram at 115% design power in the " Refuel" and "Startup/ Hot Standby" modes and the backup IRP scram at 1120/125 of full scale assures that there is proper overlap in the neutron monitoring systems and, thus, that adequate coverage is provided for all ranges of reactor operation.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (HFLPD) and reactor core thermal power.

The scram setting is adjusted in accordance with the formula in the CORE OPERATING LIMITS

]

REPORT when the HFLPD is greater than the fraction of rated power (FRP).

In a similar manner, the APRM rod block trip setting is adjusted downward if HFLPD exceeds FRP, thus preserving the APRH rod block safety margin.

Amendment No.133 40c l

tY l

PNPS TABLE 3.2.C-2 CONTROL ROD BLOCK INSTRUMENTAT10N SETPOINTS i

Trio FunctioD Trio setootnt l

APRM Upscale (1) (2) i APRM Inoperative Not Applicable APRM Downstale 1 2.5 Indicated on Scale Rod Block Monitor (Flow Biased)

(1) l l

Rod Block Monitor Inoperative Not Applicable Rod Block Honitor Downscale 15/125 of full Scale IRH Downscale 1 5/125 of Full Scale

{

IRM Detector not in Startup Position Not Applicable IRH Upscale 1 108/125 of full Scale IRM Inoperative Not Applicable e

j SRM Detector not in Startup Position Not Applicable SRM Downscale 1 3 counts /second 5 counts /second SRH Upscale 1 10 SRM Inoperative Not Applicable Scram Discharge Instrument Volume 1 18 gallons Hater Level - High Scram Discharge Instrument Volume -

Not Applicable

' Scram Trip Bypassed Recirculation Flow Converter - Upscale 1 120/125 of Full Scale Recirculation Flow Converter -

Not Applicable Inoperative Recirculatior Flow Converter -

1 10% Flow Deviation for > 80%

i Comparatof Hismatch Rated Power, and l

1 15% Flow Deviation for 1 80%

Rated Power i

(1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, L

i (2) When the reactor mode switch is in the refuel or startup e

positions, the APRH rod block' trip setpoint shall be less than or equal to 13% of rated thermal power, but always less than the APRH flux scram trip setting..

Amendment No. 42, 110, 129, 133 55a

Y l

3.2 SMIS (Cont'd)

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the Safety j!

Limit MCPR.

The trip logic for this function is 1 out of n:

e.g., any trip on one of six APRH's, eight ILH's, or four SRH's will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBH may be reduced by one for maintenance, '.esting, or calibration.

This time period is only 3% of the operating * 'me in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.

The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a i

HCPR less than the Safety Limit MCPR.

This rod block set point, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal.

The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range.

The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 107% of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRH rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin.

In the startup and refuel modes, the APRM rod block function is setdown below the APRH flux scram trip.

The RBH rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern.

The IRH tod block function provides local as well as gross core The scaling arran ement is such that trip setting is less protection.

than a factor of 10 above the fndicated level.

I A downscale indication on an APRM or IRH is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The downscale trips are as shown in Table 3.2.C-2.

i 71 Amendment No. 15, 42, 110, 129, 133 l

"3 3.2 1A1[1 (Cont'd)

The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety.

3 The refueling interlocks also operate one logic channel, and are l

required for safety only when the mode switch is in the refueling

]

position.

for effective emergency core cooling for small pipe breaks, the HPCI g

system must function since reactor pressure does.not decrease rapidly l

enough to allow either core spray or t.PCI to operate in time.

The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate.

The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.

The trip settings given in the specification are adequate to assure the above criteria are met.

The specification preserves the effectiveness of the system during periods of maintenance, t

testing or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Four radiation monitors are provided which initiate the Reactor Building Isolation and Control System and operation of the standby gas treatment system.

The instrument channels monitor the radiation from the 4

refueling area ventilation exhaust ducts.

four instrument channels are arranged in a 1 out of 2 twice trip logic.

4 Trip settings of < 100 mr/hr for the monitors in the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment-system.

Flow integrators are used to record the integrated flow of liquid from the drywell sumps.

The alarm unit in each integrator is set to annunciate before the values specified in Specification 3.6.C are exceeded. A system whereby the time interval to fill a known volume-will be utilized to provide a back r to the ficw integrators. An air 4

sampling system is also provided 9 J, <ct leakage inside the primary containment.

l I

l

- Amendment No. 89,133 72 i

.q 3

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C (Rolant leakug (Cont'd) 4.6 power operation is permissible only during the succeeding seven days.

3.

If the conditions in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Safety and Relief Valves D.

Safety and Relief _ Valves 1.

During reactor power operating 1.

At least one safuty valve and two conditions and prior to reactor relief / safety valves shall be startup from a Cold Condition, or checked or replaced with bench whenever reactor coolant pressure checked valves once per operating is greater than 104 psig and cygte.

All valves will be tested temperature greater than 340*F every two cycles, both safety valves and the safety l

modes of all relief valvn shall 2.

At least one of tne relief / safety be operable, valves shall be disassembled and inspected each refueling outage.

The nominal setpoint for the relief / safety valves shall be 3.

Whenever the safety relief valves selected between 1095 and 1115 are required to be operable, the psig. All relief / safety valves discharge pipe temperature of each shall be set at this nominal safety relief valve shall be logged setpoint 11 psi.

The safety

daily, valves shall be set at 1240 psig 2 13 psi.

4.

Instrumentation shall be calibrated and checked as indicated in Table l

2.

If Specification 3.6,0.1 is not 4.2.F.

met, an orderly shutdown shall be initiated and the reactor coolant 5.

Notwithstanding the above, as a pressure shall be below 104 psig minimum, safety relief valves that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Note: Technical have been in service shall be Specifications 3,6.0.2 - 3.6.D.5 tested in the as-found condition apply only when two Stage S rget during both Cycle 6 and Cycle 7.

Rock SRVs are installed.

3.

'f the temperature of any safety relief discharge pipe exceeds 212*F during normal reactor power operation for a period of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an engineering evaluation shall be performed justifying continued operation for the corresponding temperature increases.

Amendment No. 42, 56, 88, 133 126 I

,. 9 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.D Safety Relief Valves (Con't) 4.

Any safety relief valve whose discharge pipe temperature exceeds-212'F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more shall be removed at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more,-

tested in the as-found condition,

-and recalibrated as necessary prior-to reinstallation.

Power operation shall not continue beyond 90 days from the initial discovery of discharge pipe temperatures in excess of 212*F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC approval of the engineering evaluation delineated in 3.6,0.3.

5.

The limiting conditions of-operation for the instrumentation that monitors tail pipe temperature are given in Table 3.2.F.

E.

Jet Pomos E. -Jet Pumos 1.

Whenever the reactor is in the Wheaever there is recirculation

.startup or run modes, all jet flow with the reactor in the l

pumps shall be operable.

If it is startup or run-modes, jet. pump i

determined that a jet pump is operability shall be checked daily inoperable, an orderly shutdown by verifying that the following I

L shall be initiated and-the reactor conditions do not occur

'~

shall be in a Cold. Shutdown simultaneously.

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

1.

The two recirculation loops have a flow imbalance of 15% or more when the pumps are operated-at the same speed.

2.

The indicated value of core flow ~ rate varies from the value' derived from loop flow measurements by more than-10%.

3.

The diffuser to lower ~ plenum differential pressure reading on an individual jet pump varies-from established jet pump delta P characteristics by l

more than 10%.

Amendment.No. 42, 56, 77, 93,133 127

49 4 LidITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.:.F Jet Pumo Flow Hismatch 4.6.F Jet Pumo Flow Hismatch 1.

Whenever both recirculation pumps Recirculation pump speeds shall be are in operation, pump speeds checked and logged at least once shall be maintained within 10% of per day.

each other when power level is greater than 80% and within 15% of each other when power level is less than:or equal to 80%.

2.

If Specification 3.6.F.1 is exceeded immediate corrective action shall-be taken.

If recirculation pump soeed mismatch is not corrected within 30 minutes, an orderly shutdown shall be initiated and the reactor shall l

be in the Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the e

recirculation pump speed mismatch is brought within limits sooner.

G.

Structural Intearity G.

Structural Inteority 1.

The structural integrity of the Inservice inspection of components L

primary system boundary shall be shall be performed in accordance maintained at the level required with the PNPS Inservice Inspection f

by the ASHE Boiler and Pressure Program.- The results obtained from l

Vessel Code,.Section XI " Rules for compliance with this program will j

Inservice Inspection of Nuclear be evaluated at the coinpletion of Power Plant Components," Articles each ten year interval.

The r

IWA, IHB,- IHC, IHD and IWF and conclusions of this evaluation will mandatory appendices as required be reviewed with the NRC.

by 10CFR50 Section 50.55a(g),

l except where specific relief has been granted 'oy the NRC pursuant to 10CFR50, Section 50.55a(g)(6)(i).

l l

l 1:

1 l

Amendrient h_

19, 93, 133 127a i

l

@l 1

MS.ES:

13.6.0 and 4.6.Q Safety and Relief Valvn The valve sizing analysis considered four, 10% capacity relief / safety valves and two 8% capacity safety valves.

These-are sized and set pressures are established in accordance with the following three_ requirements of Section III of the ASME-Code:

1.

The lowest safety valve must be set to open at or below vessel design pressure and the highest safety valve be set at or below 105% of design pressure.

2.

The valves must limit the reactor pressure to no more than 110% of design

pressure, 3.

Protection systems directly related to the valve sizing transient must not

-be credited with action (i.e., an indirect scram must be assumed).

A main steam line isolation with flux scram has been selected to be used as the safety valve sizing transient since this transient results in the highest peak vessel-pressure of any transient when analyzed with an indirect scram.

The original FSAR analysis concluded that the peak pressure transient with indirect-scram would be caused by a loss of condenser vacuum (turbine trip with failure of the bypass valves to open). However, later observa ins have shown that the long lengths of steam lines to the turbine buffer the f aster stop valve closure isolation and thereby reduce the peak pressure caused by L

.this transient to a value below that produced by a_ main steam line isolation l

with _ flux scram.

Item 3 above indicates that no credit be taken for the primary scram signal generated by closure of the main steam isolation valves.

Two other scram initiation signals would be generated, one due to high neutron flux and one due to high reactor-pressure.

Thus item 3 will be satisfied by-assuming a scram due-to high neutron flux.

Relieving capacity of 40% (4 relief / safety valves) results in a peak pressure during the transient conditions used in the_ safety valve sizing analysis which is well below the pressure safety limit, The relief / safety valve settings satisfy the Code requirements that the lowest safety valve set point be at or below the vessel design pressure range to prevent unnecessary cycling caused by minor transients.

The results of postulated transients.where inherent-relief / safety valve actuation is required are given in Appendices R and Q of the Final Safety Analysis Report.

Experience in safety valve operation shows that a testing of-at:lecst 50% of the safety valves per refueling outage is adequate to detect failures or deterioration.

The tolerance value of 1% is in accordance with Section III of the ASME Boiler and Pressure Vessel Code. An analysis has been performed which shows that with all safety valves set 11 higher, the reactor coolant pressure safety limit of 1375 psig is not exceeded.

Amendment No. 75, 56,133 145

- -~ -

?

BASES:t 3.6.0 and 4.6.0

' Safety and Relief Valves The. relief / safety valves have two functions; i.e., power relief or self-actuated by high pressure.

Power relief is a solenoid ~ actuated function

-(Automatic Pressure Relief) in which external instrumentation signals _of coincident high drywell-pressure and low-low water level. initiate the.uives-to~open.

This function is discussed in Specification 3.5.0.

In addition, the valves can be operated manually.

Pilgrim's experience with 2 stage safety / relief' valves has demonstrated that minimum _ leakage exists when the tailpipe temperature is 215' Fahrenheit.

Therefore, a reporting requirement triggered by a temperature of 212'F is conservative,.and assures timely reporting before leakage reaches significant proportions.

l Amendment No. 42.133 146

1 4

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

- 3.11 REACTOR FUEL ASSEMBlv 4.11 REACTOR FUEL ASSE@j,y A3011 cabili ty:

Aeolicability:

The Limiting Conditions for The surveillance requirements Operation associated with fuel apply to the parameters which rods apply to those parameters monitor the fuel rod operating.

which monitor the fuel rod conditions.

-operating conditions.

Obiective:

Obiettive:

The-Objective of Limiting The Objective of the Surveillance Conditions for Operation is; to Requirements is to specify the assure the performance of the type and frequency of fuel rods.

surveillance to be applied to the fuel rods.

Soecifications:

Soecifications:

A.

Averace Planar Linear Heat A.

Averace Planar Linear Heat Generation Rate (APLHGR)

Generation Rate (APLHGR).

During power. operation with both The APLHGR for each type of fuel recirculation pumps operating, as a function of average planar the-APLHGR for each type of fuel exposure shall be determined as a function of average planar daily during reactor operation at exposure shall not exceed the 1257, rated thermal power.

-applicable limiting value specified in the CORE OPER$ TING LIMITS REPORT.-

If at any time during operation i

it is determined by normal surveillance.that the limiting value for APLHGR is being exceeded, action s L11 be

. initiated within 15 minutes to.

restore operation-to within the prescribed limits..If the APLHGR is.not returned to within the prescribed limits-within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding 1

action.shall continue until reactor operation is within the' prescribed limits.

l Amendment No. 15, 24, 27, 42, 59, 100,133 205a l

y; 4:

t LIMITING CONDITIONS FOR OPERATION SURVEILLANE REOUIREMENTS B.~ Linear Heat Generation Rate (LHGR)

B.

Linear Heat Generation Rate (LHGR)

During reactor powc.* operation, The LHGR as a function of core the LHGR shall not exceed the height shall be checked daily limits specified in the CORE during reactor operation at 125%

OPERATING LIMITS REPORT.

rated thermal power.

If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

C.

Minimum Critical Power Ratio (MCPR)

C.

Hinimum Critical Power Ratio (MCPR) 1.

During power operation HCPR shall

.1.

HCPR shall be determined daily be'l the HCPR operating limit during reactor power operation at specified in the Core Operating

> 25% rated thermal power and Limits Report.

If at any time following any change in power during operatinn it is determined level-or distribution that would by normal surveillance-that the cause operation with a limiting limiting value for HCPR is being control rod pattern as described exceeded,-action shall be in the bases for Specification initiated within 15 minutes to 3.3.B.S.

restore operation to within the prescribed limits.

If the steady 2.

The value of x in Specification state _HCPR is.not returned to 3.11.C.2 shall-_be equal to 1.0 within.the prescribed _ limits unless determined from the result-L within two (2) hours, the reactor

- of surveillance testing of shall be brought to the Cold Specification 4.3.C as follows:

Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and a) x is defined as corresponding acti_on shall continue until reactor operation tave - Te is within the prescribed limits.

t =

l 1.275 - te Amendment No. Ji, !7, 39, 42, 54,105,133 205b

. ~ _ _

~

ql LIMITING CONDITIONS FOR OPERATIDS

. SURVEILLANCE REOUIRMENTS C.

Minimum Critical Power Ratio MCPR

.C.

Minimum Critical Power Ratio MCPR (Cont'd)

(Cont'd) for core flows other than rated the b)

The average scram time to the-HCPR limits shall be.the. limits 30% insertion position is identified above times Kr where Kr is determined as follows:

as-shown in Figure 3.3.l'of the Core n

Operating Limits Report.

E Hj tj tave " i"I As an alternative method providing n

equivalent thermal-hydraulic E Ng

. protection at core flows other than il

. rated, the calculated MCPR may be divided by Kr. where Kr is as shown in Where: an n - number of surveillance Figure 3.3.1 of the Core Operating tests performed to date in the cycle.

Limits Report.

Ni - number of active control rods 2.

The operating limit MCPR values as measured in the Ith surveillance a function of the t are given in test.

Table 3.3.1 of the Core Operating ti - average scram time to the 30%

Limits Report where t is given by insertion position of all rods speci fication 4.11.C.'2.

measured in the lith surveillance test.

c) - The adjusted analysis mean scram time (t follows:g) is calculated as

~

Nj te - p + 1.65 o

n 1-1 Where:

p - mean of the distribution for average scram insertion time to the 30% position 0.945 sec.

Ni = total number of active control rod o - standard deviation of the distribution for average scram insertion time to the 30%

posi tion, 0.064. sec.

Amendment No.

57, 133 205c

,NPQ o

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0fJIREMENTS i

D.

Power / Flow Relationshio Durina-D.

Power / Flow Reit.tionshio Durino Power Ooeration Power Ooeration

-The power / flow relationship shall Compliance with the power / flow not exceed the limiting values relationship in Section 3.11.0 specified in the CORE OPERATING shall.be determined daily during LIMITS REPORT.

reactor operation.

If at any time during power operation it is determined by normal surveillance that the limiting value for the power / flow relationship is being exceeded, action shall be initiated within

'15 minutes to restore operation-to within the prescribed limits.

If the power / flow relationship is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and

orresponding action shall continue until reactor operation is within the prescribed limits.

i L

l Amendment No. 15, 54, 133 205d l

9%,

j BASES:

3.ll.A Averace Planar Linear Heat Generation Rate (APLHGR)

.This specification assures that the peak cladding temperature 4

following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10CFR50, Appendix K.

The analytical method used_to determine the APLHGR limiting values is described in the topical reports listed -in Specification 6.9 A.4.

3.11.8 Linear Heat Generation Rate (LHGR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation rate.

The analytical method used to determine the LHGR limiting value is described in the topical reports listed in Specification 6.9. A.4.

3.11.C Minimum Critical Power Ratio (MCPR)

Ooeratino limit MCPR for any abnormal operating transient analysis-with -the initial condition of the reactor at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming the instrument trip settings given in Tables 3.1.1, 3.2.A and 3.2.B.

The analytical method used to determine the Operating Limit MCPR values in the CORE OPERATING LIMITS REPORT is described in the topical reports listed in Specification 6,9.A 4.

By maintaining MCPR greater than or equal to the Operating Limit MCPR, the Safety Limit MCPR specified in Specification 2.1.2 is maintained in the event of the most limiting abnormal operating transient.

Amendment No. 15, 24, 42, 133 205e

.~

1%j T

BASES:

3.11.0 Power / Flow Relationshio Durina Power Ooeration The power / flow curve is the locus of core thermal. power as a function of flow from which the occurrence of abnormal operating transients b

will yield results within defined plant safety limits.

Each transient and postulated accident applicable to operation of the plant was analyzed along the power / flow line..The analysis-justifies the operating envelope bounded by the power / flow curve as long as other operating limits are satisfied. Operation under the power / flow line is designed to enable the direct ascension to full power witnin the design basis for.the plant.

4.11.C Hinimum Critical-Power Ratio (MCPR)

]

At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and tne moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant-experience indicated-that the resulting HCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent -core flow increase would only place operation in a more conservative mode relative to MCPR.

The daily requirement for I

calculating HCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating HCPR when a limiting control rod pattern is approached ensures that HCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

l

-Amendment No, 205f l 42, 100, 70E. 133

~ _.,..

q?!

5.0 MA30R'0ESIGN FEATURES 5.1 SITE FEATURES Pilgrim Nuclear Power Station is located on the Western Shore of Cape Cod Bay in-the Town of Plymouth, Plymouth County, Massachusetts.

The site is located at approximately 41'51' north latitude and 70'35' west longitude on the Manomet Quadrangle, Massachusetts, Plymouth County 7.5 Minute Series (topographic) map issued by U.S. Geological Survey.

UTM coordinates are 19-46446N-3692E.

The reactor (center line) is located approximately 1800 feet from the nearest property boundary.

5.2 REACTOR CORE The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC-approved codes and methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR II.

5.3 REACTOR VESSEL The reactor vessel.shall be as described in Table 4.2.2 of the FSAR.

The applicable design codes shall be as described in Table 4.2.1 of the FSAR.

5.4 CONTAINHENT A.

The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR.

The applicable design codes shall be as described in Section 12.2.2.8 of the FSAR.

B.

The secondary containment shall be as described in Section 5.3.2 of the FSAR, C.

Penetrations to the primary containment and piping passing through such penetrations shall be designed in accordance with standards set forth in Section 5.2.3.4 of the FSAR.

Amendment No.

42, 78, 98, 105,133 206m

~

1. (

6.7 Deleted 6.8 PROCEDURES A.

Hritten procedures and administrative policies shall be established, implemented and maintained that meet or-exceed the requirements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7 - 1972 and Appendix "A" of USNRC Regulatory Guide 1.33, except as provided in 6.8.8 and 6.8.C below.

B.

Each procedure of 6.8.A above, and changes thereto, shall be reviewed by the ORC and approved by the responsible department manager prior to implementation.

These procedures shall be reviewed periodically as set forth in administrative procedures, NQII:

ORC review and approval of procedures for vendors / contractors, who have a QA Program approved by Boston Edison Company, is not required for work performed at the vendor / contractor facility.

C.

Temporary changes to procedures of 6.8.A above may be made provided:

1.

The intent of the original procedure is not altered.

2.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license on the unit affected.

3.

The change is documented, subsequently reviewed by the ORC within 14 days of implementation, and approved by the responsible department manager.

D.

Written procedures to implement the Fire Protection Program shall be established, implemented and maintained.

I-l I

Amendment No. 29, 30, 46, 74, 88, 122, 132, 133 216

g 1

6.9 REPORTING REOUIREMENTS In addition to the applicable reporting requirements of Title 10 Code of Federal Regulations, the following identified reports shall be submitted to the Commission.

A.

Routine Reports 1.

Startuo Recort 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 plant.

The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating 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 conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion 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 power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2.

Monthlv Ooeratina Reoort Routine reports of operating statistics, shutdown experience and forced reductions in power shall be submitted on a monthly basis to the Commission to arrive no later than the 15th of each month following the calendar month covered by-the report.

The Monthly Operating Report shall. include a narrative summary of operating experience that describes the operation of the facility, including safety-related maintenance, for the monthly report period.

I I

I N** 30, 68, 88, 103, ygg, 333 217 L

_m._.

i y

O 6.9.A Routine Recorts (Continued) 3.

Orcuoational Exoosure Tabulation A tabulation of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g. reactor operations and surveillance inservice inspection, routine maintenance, special maintenance (including a description), waste processing, and refueling shall be submitted on an annual basis.

This tabulation supplements the requirements of 20.407 of 10 CFR 20.

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

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

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

4.

Core Ooeratina Limits Reoort a) CoreoperatinilimitsshallbeestablishedanddocumentedintheCORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle, b) The analytical. methods used to determine the core opersting limits shall be those previously-reviewed and approved by the NRC in NEDE-24011-P-A, " General Electric Standard Application for Reactor fuel," (the approved version at the time the reload analyses are performed shall be identified in the-CORE OPERATING LIMITS REPORT) and in NE00-21696, " Loss of Coolant Analysis Report for Pilgrim Nuclear Power Station," dated August 1977,-(the approved version at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT),

c) The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS' limits -nuclear limits such as shutaown margin, and transient and accident analysis limits) are met.

d) The_ CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle. to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.9.8 Deleted Amendment No.

217a l

_ __