ML17285B153

Amend 80 to License NPF-21,revising Tech Spec 3.4.2, Safety/Relief Valves, by Changing Applicability Requirement

ML17285B153
Person / Time
Site:	Columbia
Issue date:	04/04/1990
From:	Trammell C Office of Nuclear Reactor Regulation
To:
Shared Package
ML17285B154	List: ML17285B153 ML17285B156
References
NUDOCS 9004120019
Download: ML17285B153 (15)

Text

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UNITED STATES NUCLEAR R EG ULATORY COMMISSION WASHINGTON, D. C. 20555 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 NUCLEAR PROJECT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 8o License No.

NPF-21 1.

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

A.

The application for amendment filed by the Washington Public Power Supply System (the licensee),

dated January 9,

1990 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

9004120019 900404 PDR ADOCK 05000397 P

PDC

~II i

y, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.

NPF-21 is hereby amended to read as follows:

(2)

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

80, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of the date of issuance.

Attachment:

Changes to the Technical Specifications Date of Issuance:

">~~1 4

FOR THE NUCLEAR REGULATORY COMMISSION Cp4~ /'11~rusuu'. //

Charles M. Trammell, Acting Director Project Directorate V

Division of Reactor Projects - III, IV, V and Special Projects

II II i(i t

ENCLOSURE TO LICENSE AMENDMENT NO. 80 FACILITY OPERATING LICENSE NO.

NPF-21 DOCKET NO. 50-397 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

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

Also to be replaced are the following overleaf pages.

AMENDMENT PAGE 3/4 4-7 3/4 4-7a B 3/4 4-la OVERLEAF PAGE 3/4 4-8 B 3/4 4-2

REACTOR COOLANT SYSTEM 5/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 a) The safety valve function of at least 12 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code saf ety val ve functi on l ift settings: ~

2 safety/relief valves 8 4

safety/relief valves 8

4 safety/relief valves 8 4

safety/relief valves 8 4

safety/relief valves 8 1150 psig +1%/-3X 1175 psi g +lX/-3X 1185 psig +3%/-3X 1195 psig +1%/-3X 1205 psig +1K/-3X APPLICABILITY:

OPERATIONAL CONDITIONS 1, and 2, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

b) The safety valve function of at least 4 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:"

2 safety/relief valves 8 1150 psig +1%/-3X 4

safety/relief valves 8 1175 psig +1%/-3X 4

safety/relief valves 8 1185 psig +3%/-3X 4

safety/relief valves 8 1195 psig +3%/-3X 4

safety/relief valves 8 1205 psig +1%/-3X APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3, when THERMAL POWER is less E

ALP ACTION:

a.

With the safety valve function of one or more of the above required safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With one or more safety/relief valves stuck open, provided that suppression pool average water temperature is less than 90'F, close the stuck open safety/relief valve(s); if unable to close the open "The lift setting pressure shall correspond. to ambient conditions of the valves at nominal operating temperatures and pressures.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-7 Amendment No.

80

0 L

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REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION ACTION: (Continued)

C.

valve(s) within 2 minutes or if suppression pool average water tempera-ture is llO~F or greater, place the reactor mode switch in the Shut-down position.:

With one or more safety/relief valve acoustic monitors inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4 ~ 4.2 The acoustic monitor for each safety/relief valve shall be demonstrated OPERABLE by performance of a:

a.

CHANNEL CHECK at least once per 31 days, and a

b.

CHANNEL CALIBRATION at least once per 18 months.""

""The provisions of Specification 4. 0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-7a Amendment No. 8O

REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTBlS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

a.

The primary containment atmosphere gaseous radioactivity monitoring

system, b.

The primary containment sump flow monitoring system, and c.

The primary containment atmosphere particulate radioactivity monitoring system.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACT10N:

With only two of the above required leakage detection systems. OPERABLE, operation may continue for up to 30 days provided grab samples of the con-tainment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREHENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:

a.

Primary containment atmosphere particulate and gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a

CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b.

Primary containment sump flow monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least once pe<

31 days and a CHANNEL CALIBRATION TEST at least once per l8 months.

WASHINGTON NUCLEAR - UNIT 2 3'-8

REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:

a.

The primary containment atmosphere gaseous radioactivity monitoring

system, b.

The primary containment sump flow monitoring system, and c.

The primary containment atmosphere particulate radioactivity monitoring system.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With only two of the above required leakage detection systems. OPERABLE, operation may continue for up to 30 days provided grab samples of the con-tainment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> s and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:

a.

Primary containment atmosphere particulate and gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b.

Primary containment sump flow monitoring system-performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-8

h II

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES (Continued) the dual purpose safety/relief valves in their ASME Code qualified mode (spring lift) of safety operation.

The overpressure protection system must accommodate the most severe pres-surization transient.

There are two major transients that represent the most severe abnormal operational transient resulting in a nuclear system pressure rise.

The evaluation of these events with the final plant configuration has shown that the MSIV closure is slightly more severe when credit is taken only for indirect derived scrams; i.e.,

a flux scram.

Utilizing this worse case transient as the design basis

event, a minimum of 12 safety/relief valves are required to assure peak reactor pressure remains within the Code limit of llOX of design pressure.

Testing of safety/relief valves is normally performed at lower power.

It is desirable to allow an increased number of valves to be out of service during testing.

Therefore, an evaluation of the MSIV closure without direct scram was performed at 25K of RATED THERMAL POMER assuming only 4 safety/relief valves were operable.

The results of this evaluation demonstrate that any 4 safety/relief valves have sufficient flow capacity to assure that the peak reactor pressure remains well below the code limit of 110K of design pressure.

Demonstration of the safety/relief valve lift settings will be performed in accordance with the provisions of Specification 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3. 1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"

May 1973.

MASHINGTON NUCLEAR - UNIT 2 B 3/4 4-la Amendment No. 80

REACTOR COOLANT SYSTEM BASES 3/4.4. 3. 2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reacto~

will be shut down to allow further investigation and corrective action.

Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are subject to high stress or that certain relatively

stagnant, intermittent, or low flow fluids, requires additional surveillance and leakage limits.

The surveillance requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve fai lure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0. 5 ppm concentration of chlorides is not considered harmful during these periods.

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

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

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

The surveillance requirements provide adequate assurance that concentrations in excess of the limits wi 11 be detected in sufficient time to take corrective action.

WASHINGTON NUCLEAR - UNIT 2 8 3/4 4-2

~

~

REACTOR COOLANT SYSTEM BASES 3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was ~iso con-sidered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action.

Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are subject to high stress or that certain relatively,

stagnant, intermittent, or low flow fluids, requires additional surveillance and leakage limits.

The surveillance requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4. 4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

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

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

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

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

WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-2