ML17332A497
| ML17332A497 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/05/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17332A496 | List: |
| References | |
| NUDOCS 9501100151 | |
| Download: ML17332A497 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
188 TO FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO.
DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NOS.
1 AND 2 DOCKET NOS.
50-315 AND 50-316
- 1. 0 INTRODUCTION By letter dated November 15,, 1993, the Indiana Michigan Power Company (the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License Nos.
DPR-58 and DPR-74 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2, respectively.
The letter dated November 15,
- 1993, proposed to delete the list of Event V reactor coolant system pressure isolation valves (PIVs) and the associated surveillance requirements from the technical specifications and test the valves in accordance with the plant's inservice testing program (IST) prescribed by Section 50.55a, "Codes and Standards,"
of Part 10 of the Code of Federal Re ulations (10 CFR 50.55a).
In conversations with the staff on September 26, 1994, the licensee agreed to revise the, requested change such that rather than deleting the list of valves and the associated surveillance requirements from the technical specifications, a change to the acceptable leakage rates would be proposed.
Subsequently, the licensee submitted a revised request by letter dated October 7, 1994.
The October 7, 1994, revision clarified the original amendment request by explicitly stating the proposed new requirements as opposed to incorporating them as part of the IST Program.
Since the same new requirements would have been implemented by both the original and supplemental submittal, the staff's initial proposed no significant hazards consideration determination did not change.
Therefore, renoticing was not warranted.
2.0 BACKGROUND
The Reactor Safety Study (RSS),
WASH-1400, identified in a pressurized-water reactor an intersystem loss-of-coolant accident (LOCA) which is a significant contributor to risk of core melt accidents, referred to as "Event V" in the study.
The design examined in the RSS contained in-series check valves isolating the high pressure primary coolant from the low pressure injection system piping.
The scenario which leads to the Event V accident is initiated 950ii00i5i 950i05 PDR ADOCK 050003i5 P
by the failure of these check valves to function as a pressure isolation barrier.
This causes an overpressurization and rupture of the low pressure piping which results in a LOCA that bypasses containment.
To better de'fine the Event V concern, all light-water reactor licensees were requested by NRC's letter dated February 23, 1980, to review the configurations for their high/low pressure interfaces and determine if any Event V configurations existed at their plant(s).
By letter dated March 24,
The NRC issued a safety evaluation for an Event V configuration at the D. C.
Cook Nuclear Plant, Units I and 2, and issued an Order for Modification of License to add the list of PIVs and the associated surveillance requirements to the plant's technical specifications.
The specific valve configurations are as follows:
Units I a d
2 Valve Numbers Low-Head Safety Injection Loop 2, cold leg SI-170L2 and RH133 Loop 3, cold leg SI-170L3 and RH-134 Each pair of check valves is in series with a normally-closed, motor-operated valve.
The allowed leakage through any reactor coolant system (RCS)
PIV is currently limited to I gallon per minute (gpm) in Technical Specification 3.4.6.2.f.
However, per Technical Specification 3.4.6.2 ACTION c, leakage is allowed to be up to 5 gpm if the most recent measured leakage does not exceed the previous measured leakage by an amount that reduces the margin between the most recent measured leakage and the maximum limit of 5 gpm by 50X or more.
The Technical Evaluation Report attached to the NRC's safety evaluation stipu1ated that when leakage tests are made in cases using pressures lower than function maximum pressure differential, the observed leakage shall be adjusted to "function maximum" pressure differential value.
The adjustment is to be made by calculation appropriate to the test media and the ratio between test and function pressure differential, assuming leakage to be directly proportional to the pressure differential to the one-half power.
Subsequently, the NRC commissioned a study by the Idaho National Engineering Laboratory to survey a number of plants and make recommendations to improve valve leak testing requirements for the Event V valves.
The results of the study were compiled in EGG-NTAP-6175, "Inservice Leak Testing of Primary Pressure Isolation Valves," February 1983.
One recommendation in the report was that the current technical specification allowances of I gpm to 5 gpm leakage should be changed to make the leakage allowance proportional to valve
- size, based on an allowance of 0.5 gpm per inch nominal valve diameter at the functional differential pressure, with a maximum allowed leakage of 5 gpm, unless it can be shown that the low pressure piping has installed relief capacity suc'h that it wou1d not overpr essurize with greater allowed leakage.
This recommended allowance has been incorporated into NUREG-1431, "Standard Technical Specifications for Westinghouse Plants,"
September 1992.
3.0 LICENSEE'S PROPOSED CHANGES AND EVALUATION OF CHANGES The licensee proposes to change Technical Specification 3.4.6.2(f), which currently specifies the leakage limit of 1
gpm for the reactor coolant system pressure isolation valves listed in Technical Specification Table 3.4-0, to state the following:
The leakage from each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-0 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi
[pounds per square inch] of the nominal full pressure value.
The resulting allowed leakage as identified in revised Table 3.4-0 is as follows:
SI-170L2 SI-170L3 RH-133 RH-134 10-inch diameter 10-inch diameter 8-inch diameter 8-inch diameter 5 gpm 5 gpm 4 gpm 4 gpm The Technical Specification action statement will be changed to delete provisions to analyze the leakage of any one of the valves if it is measured between 1
gpm and 5 gpm, and comparing it to the previous leakage, prior to determining the need to declare the valve inoperable.
With the proposed
- change, any valve that exceeds the leakage limit of Technical Specification 3.4.6.2(f) will be declared inoperable and isolated by a closed de-energized motor-operated valve.
In addition, the licensee proposes to delete the specific schedule for performing the leakage surveillance test requirements from the Technical Specification and instead refer to the specification that incorporates the regulatory requirements for performing inservice testing.
The current inservice testing program for the D.
C.
Cook Nuclear Plant, Units 1 and 2, was developed to the requirements of the 1983 Edition, including addenda through Summer 1983 Addenda, oF the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (the Code),
Section XI.
The inservice testing program is due to be updated to the requirements of the 1989 Edition of the ASNE Code in July 1996.
The leakage testing requirements of the later Code edition are essentially the same as the 1983 edition.
In accordance with the licensee's inservice testing
- program, and the requirements of the ASNE Code, the leakage testing will be performed once every 2 years and following repair, replacement, or maintenance that could affect the leakage.
The Code requirements of paragraph IWV-3420, "Valve Leak Rate Test," include the test frequency, adjustments to the function pressure differential, the methods for measuring the seat
- leakage, test medium, analysis of leakage rates, and corrective actions.
The corrective action requirements of the ASNE Code include provisions for valves 6 inch nominal pipe size and larger such that the previously measured leakage rate is trended and if a higher leakage reduces the margin between the measured leakage rate and the maximum permissible rate by 50% or greater, the
0
test frequency is doubled.
Repair or replacement is required if tests show a
leakage rate increasing with time, and a projection based on three or more tests indicates that the leakage rate of the next scheduled test will exceed the maximum permissible leakage rate by greater than 10X.
Therefore, the proposed Technical Specification change will allow the licensee to monitor a
valve with increasing leakage rates rather than declare the valve inoperable.
Past experience has shown that higher leakage rates are not always indicative of degradation, and maintenance has been performed based on higher leakage rates unnecessarily.
The Code requirements provide an adequate level of quality and safety for monitoring degradation of the valve seats.
The change will eliminate unnecessary maintenance and reduce personnel exposure without compromising safety.
The change between the current test schedule and the proposed test schedule is minimal.
Because the test can be performed only during plant shutdown conditions, the test will be performed on a refueling outage schedule similar to the current requirement to leak test after each refueling outage.
The Code requires leak testing following any repair, replacement, or maintenance that could affect the leakage rate, similar to the current Technical Specification requirement.
Finally, the deletion of the TS requirement to perform a leak test during startup whenever the plant has been in cold shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9
months, will actually eliminate very few tests over the life of the plant unless the plant experiences frequent cold shutdowns.
The Code requirement to double the test frequency if degrading conditions are indicated will have the same effect as testing every 9 to 12 months during cold shutdown conditions; however, if the valve has not exhibited increasing seat
- leakage, additional testing will not be required.
For as low as is reasonably achievable
- purposes, increasing the number of tests only when degrading conditions are indicated will eliminate unnecessary personnel dose without impacting the ability to adequately monitor the valves.
4.0 INDIVIDUAL PLANT EXAMINATION RISK ASSESSMENT The licensee's change request included an assessment of the change in risk based on the Individual Plant Examination study.
While the assessment results indicate that the change represents essentially no impact on the risk of operating the plants, the staff did not review the validity or applicability of the licensee's probabilistic risk assessment and did not base its conclusion on the assessment.
5.0
SUMMARY
The proposed changes are essentially equivalent to the current technical specification requirements in that (I) the valves are maintained in the specifications and therefore remain subject to the surveillance requirements of inservice testing and (2) the proposed allowed leakage rates're 5 gpm and 4 gpm, depending on valve size, which are the same or more conservative than the allowed maximum leakage of 5 gpm in the current specifications.
The changes also achieve the intent of monitoring the valves for degrading conditions and precluding catastrophic failure by (I) allowing a set limit based on the size of the valve rather than a limit of I gpm to 5 gpm, thereby
decreasing the potential for valve repairs when there is no substantial degradation, (2) ensuring that radiological exposure to personnel will not be increased, and could be decreased by eliminating maintenance, but requiring additional testing, if only minor increases in leakage are measured, and (3) following the requirements of inservice testing of valves in accord with the regulations and technical specifications.
Based on these factors, the proposed technical specification changes are acceptable.
6.0 STATE'CONSULTATION In accordance with the Commission's regulations, the Michigan State offici,al was notified of the proposed issuance of the amendments.
The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released
- offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a
proposed finding that the amendments involve no significant hazards consideration and there has been no public'omment on such finding (59 FR 623).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
8.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonahle assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
Patricia Campbell, NRR Date:
january 5, 1995