ML20245J975

Safety Evaluation Supporting Amend 29 to License NPF-39

ML20245J975
Person / Time
Site:	Limerick
Issue date:	06/22/1989
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8907030237
Download: ML20245J975 (11)

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UNITED STATES y

7, NUCLEAR REGULATORY COMMISSION 5

j WASHING TON, D. C. 20555

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t SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE NO. NPF-39 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION, UNIT 1 DOCKET NO. 50-352

1.0 INTRODUCTION

By letter dated February 14, 1986, Philadelphia Electric Company _(the licensee) requested an amendment to facility Operating License No. NPF-39 for the Limerick Generating Station, Unit 1.

The proposed amendment would make administrative changes to the Technical Specifications (TSs) to achieve consistency, remove outdated material, make minor text changes and correct errors.

2.0 EVALUATION The proposed changes to the TSs related to 28 items. Because of the diversity of the changes, each item identified in the application is discussed and evaluated separately.

i Item 1, Page 3/4 7-22 Section 3.7.6.2, page 3/4 7-22, specifies the fire protection spray and sprinkler systems that shall be operable. After two of the five zones (zones 41 and 42A), there is an asterisk to a note at the bottom of the page. The note states that the two zones were not required to be i

operable until prior to exceeding 5% of rated thermel power. The note is no longer applicable, since Unit I received a full power license on August 8, 1985. The proposed change is to delete the asterisk and note.

This is a purely administrative change to delete something that is no longer valid and has no safety significance.

Item 2. Page 6-14 Section 6.8.4 of the Administrative Controls Section of the TSs (page 6-14) specifies certain programs that shall be established, implemented and maintained. Subsection C, which describes the post-accident sampling program, has an asterisk referring to a note at the bottom of the page which states that the program is not required prior to exceeding 5% of rated thermal power. The proposed change is to delete the asterisk and note, since the note was only applicable when Unit I had a low power license. This is a purely administrative change that has no safety i

significance.

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i Item 3, Page 3/4 3-85 Table 3.3.7.5-1 (page 3/4 3-85) of the TSs lists the accident monitoring instrumentation that is required to be operable.

Under applicable operational conditions, four of.the instruments have a pound sign (#) in 3

the column referring to a note at the bottom of the page.

The note i

states that these four instruments were not required to be operable until initial criticality.

Since Unit 1 has operated for over four years, the note is no longer valid.

The proposed changes are to delete the

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references to the note and the note itself.

This is a purely administrative change with no safety significance.

Item 4, Page 3/4 3-21 and Item 5, Page 3/4 3-22 Table 3.3.2-2 (pages 3/4 3-21 and 3/4 3-22) lists the trip setpoints for instrumentation that actuates primary and secondary containment 1

isolation.

For the instrumentation that measures differential pressure between the outside atmosphere and the reactor enclosure, the TSs now list the trip setpoint as being equal or greater than 0.1 inch.

The proposed revision is to change " inch" to " inches of water", since this is the required unit of measurement (as opposed, for example, to inches of mercury).

This is a purely clarifying administrative change that has no safety significance, since the allowable trip setpoints are not being modified.

Item 6, Page 3/4 3-23 l

Table 3.3.2-3 (page 3/4 3-23) lists the isolation system response time of f

each isolation trip function.

The 26" main steam lines are isolated by the main steam isolation valves (MSIVs) which are spring and/or pneumatic closing, piston-operated valves.

They close on loss of pneumatic pressure to the valve operator.

The main steam lines are isolated upon a signal of low, low (Level 2) or low, low, low (Level 3) water level in the reactor 4

vessel or upon a signal of high radiation, high steam flow or low pressure in the steam line.

The isolation signals not only initiate

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closure of the MSIVs but also the two small (3") motor operated drain valves.

The MSIVs are required to close in a short period of time to i

limit the radiological consequences and loss of coolant from a steam line break outside containment but not so fast as to cause a significant pressurization transient.

This quick closure is assured even upon loss of offsite power by having air operated valves.

It is not necessary for the small drain valves to close as quickly as the MSIVs.

The motor operated drain valves are normally powered from the safeguards buses.

Upon an isolation signal, the power to these drain valves is part of the load that is automatically shed, to be subsequently picked up by the diesel generators.

Each diesel generator is capable of attaining rated voltage and frequency within 10 seconds after receiving a starting signal.

After rated voltage and frequency are obtained, the diesel generators are connected automatically to their respective emergency i

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buses.. Under accident conditions, the required Class 1E loads of the divisions will be connected in a predetermined sequence to their respective diesel generator.

Power to the motor operators on the drain valves is. restored in less than 3 seconds after the diesels are up to l

speed when the 480V load center breaker is closed.

In the present Table j

3.3.2-3,thelistedresponsetime(inseconds)forhighraglionorlow pressure isolation of the main steam lines is "5 1.0*/5 13 Note

"(a)" clarifies that the '.' isolation system ' instrumentation response time specified includes 10 seconds diesel generator starting and 3 seconds for sequences loading' delays." In the present TSs, the response time listed

-for the Level 1 low water level isolation is 51.0 seconds with reference i

to a note that states that this value applies only to the MSIVs.

To

"/513g,thedrainvalveisolationtimes,thelicenseeproposestoadd encompa

" after.the present time to be consistent with the. response

. times cited for.other isolation functions and to the designation in the BWR' standard TSs.

The NRC. staff has evaluated the proposed change and considers this a clarifying administrative addition with no safety considerations that were not implicit in our original evaluation of the isolation requirements and the associated instrumentation, controls and electrical power systems.

Item 7, Page 3/4 4-19 The requirements on removal of flux wire specimens on page 3/4 4-19 refers to figures B 3/4 4.6-1 and B 3/4 4.6-2.

The latter was not included in the TSs issued by the Commission (NUREG-1149).

The reference to the second figure is an error which the licensee proposes to correct by the requested changes.

The proposed change is acceptable since it corrects an error in the Bases (which are not part of the TSs).

Item 8, Page 3/4 4-23 Section 3/4 4.7 of the TSs, page 3/4 4-23, describes the operability requirements for the MSIVs, specifically, that the valves achieve full closure in no less than 3 and no more than 5 seconds.

The action statement provides 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore operability prior to isolating the line.

The MSIV, are isolation valves and are included with other isolation valves in Table 3.6.3-1 (page 3/4 6-17).

The action statement for Table 3.6.3-1 allows four hours to restore operability prior to isolating the penetration.

Generally, when there is an apparent conflict in the TS requirements, the more restrictive requirement is controlling.

The licensee had proposed to change the'8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> requirement in section 3/4 4/7 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to be consistent.

However, there is a basis for the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and it is in all the BWR standard TSs (NUREG-0123 revision 1, 2 and

3) and in the TSs issued for other plants (e.g. NUREG-1142 for River Bend, NUREG-1202 for Hope Creek, etc.) as well as Limerick's TSs.

The 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> restoration time applies (vs 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) as long as the MSIVs close within 5 seconds.

With agreement of the licensee, thir, proposed change is not being processed as part of the subject amendment.

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4 Item 9, Page 3/4 5-4, Item 10, Page 3/4 5-5, item 21, Page 3/4 7-9, Item 22, Page 3/4 7-10 The High Pressure Coolant Injection (HPCI) System is required to be operable whenever reactor steam dome pressure is more than 200 psig.

Likewise, the Automatic Depressurization System (ADS) and Reactor Core Isolation Cooling (RCIC) System are required to be operable whenever the rer.ctor pressure is more than 100 psig and 150 psig, respectively. The HPCI and RCIC systems use turbine-driven pumps and thus require sufficient steam pressure to operate. Testing the ADS valves also requires sufficient steam pressure to verify operability. To allow plants to get out of the shutdown mode, there is a note to the surveillance test requirements that allows the operability tests to be 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". No directions were given, however, for the condition i

whereby the surveillance fails and/or HPCI/RCIC/ ADS operability cannot be demonstrated during the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period following the availability of sufficient steam supply.

The licensee has proposed to add a clarifying sentence to the existing notes. The added sentence would specify that in the event the system (i.e, HPCI, RCIC or ADS) is not successfully.

demonstrated to be operable during the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period, then the reactor l

dome pressure would be reduced to below that pressure at which the system is required to be operable (i.e., below 200 psig if the HPCI system is found inoperable, below 150 psig if the RCIC is involved and below 100-psig for the ADS system). The staff noted that there was no time limit specified to take this action. As proposed, the wording could be interpreted as requiring an imm d ate reduction in reactor pressure if the system has not been determined to be operable.

The intent of the requirement is that during startup, the operability of the ECCS systems be demonstrated as soon as possible in the startup mode prior to reaching any significant power level. The operability of the systems can be determined within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> but if a problem is detected, it is not always possible to diagnose the cause without steam to operate the turbines.

If the HPCI, RCIC or an ADS valve is found to be inoperable while the plant is at power, the LCOs allow operation for 14 days before shutdown is required.

If these systems are found to be inoperable during startup, the intent is that the inoperable equipment be fixed before proceeding with power ascension. To provide a reasonable time to troubleshoot potential problems that may be detected in these ECCS systems during startups, the staff and license agreed to a time limit of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before reactor pressure has to be reduced below the point where the systems are not required to be operable (e.g., below 200 psig for the HPCI system).

This clarification only permits the plant to hold pressure at a point where sufficient steam is available to operate the HPCI and RCIC pumps for testing; it does not permit power ascension to continue. The staff concludes that the sentence that is proposed to be added to the note associated with the surveillance requirements is a clarification of the action to be taken. The proposed action is more restrictive than what would be permitted if the plant were at power and is acceptable.

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Item 11, Page 3/4 6-9 and B 3/4 6-2 The Bases for Specification 3/4 6.1.6 page B 3/4 6-2, (which discusses the Drywell and Suppression Chamber internal pressure) limits the initial containment pressure prior to a LOCA in order to ensure that the containment peak pressure would not exceed 44.02 psig.

The Limiting Condition for Operation on page 3/4 6-9, however, is not in accordance with the Bases and incorrectly restricts the internal pressure to be between 0.0 and +2.0 psig, while the Bases calls for -1.0 to +2.0 psig.

The FSAR Section 6.2.1.1.4 evaluated negative pressure in the containment and determined that the primary containment was designed for a negative pressure of -5 psig.

In order to achieve consistency and correct these errors, a change is proposed to page 3/4 6-9 to have it agree with the BASES by changing the limitations of the Limiting Condition for Operation to -1.0 to +2.0 psig.

Further, the Bases for the drywell and suppression chamber specify that "...-l.0 to +2.0 psig.(is) for initial positive containment pressure...." As the allowable containment pressure range is actually both negative and positive, the deletion of the word " positive" on page B 3/4 6-2 is proposed as it is incorrect.

The -1.0 psig limit, when viewed from the Bases, is more conservative in that a higher additive pressure would now be required in order to exceed the 44.02 psig peak pressure than was previously required.

These proposed changes correct errors in tne present TSs and are acceptable.

Item 12, Page 3/4 6-10 The drywell average air temperature is the calculated volumetric average of the temperature readings at four drywell elevations.

At elevation 330' there are three installed temperature sensors, and there are also three sensors installed at rlevation 320'; three at 260' and six at 248'.

The volumetric calculation requires only that one sensor at each elevation plane be read, without regard to the sensor (compass) location, i.e., only the elevation location is of interest in the calculation and not the azimuth location.

The azimuth has been listed, however, in the surveillance requirements on page 3/4 6-10 as an informational guide so that the exact location of each sensor may be readily ascertained even though the azimuth of each sensor is not a factor in the actual calculations.

Because the azimuth of each sensor is listed, the relocation of any of these sensors would require an amendment to the TSs, even though the intent of the Technical Specifications clearly does not require this information as a limit.

Further, some readings taken from these sensors would be erroneous if the sensor azimuth was not changed due to the close proximity of pipes to some of the sensors.

Some of these pipes are substantially different in temperature than the ambient drywell air.

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The licensee proposes to revise the nomenclature of the drywell temperature sensors from " azimuth" to " quantity" to allow sensor azimuth relocation and preclude erroneous drywell readings and also to eliminate J

the need for amendments to the TSs each time a drywell modification calls

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for the relocation of a temperature sensor.

In order to allow for

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physical limitations in the installation of equipment, the licensee also

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proposes to add the word " Approximate" before elevation.

1 The drywell average air temperature is required to be maintained below 135*F.

The limitation on drywell average air temperature ensures that containment peak air temperature does not exceed the design temperature of 340 F during steam line break conditions. The average temperature is determined by using one reading from each of four elevations spaced over a height of 82 feet.

The purpose of the temperature sensors is to determine the average drywell temperature to ensure that the average temperature does not exceed 135'F. The precise location of a sensor is not important. What j

is more important is that the sensors are reading ambient drywell air temperatures. The proposed changes to the TSs will permit the licensee to locate the sensors so they are not near steam pipes, coolers or other equipment that would distort the temperatures readings.

The proposed changes are to assure that the sensors will more accurately achieve the intended function.

The proposed changes to the TSs will support the intended purpose of obtaining representative measurements of ambient air temperatures in the drywell and are acceptable.

Item 13, Page 3/4 6-12 i

i Section 3.6.2.1 of the TSs (page 3/4 6-12) specifies limits on thc average I

temperature of water in the suppression pool. To measure the average temperature, the TSs require that at least eight indicators be operable.

There are eight locations in the suppression pool where temperature instruments are located. More than one temperature instrument is located at each location. The present TS's call for the operability of at least eight indicators without a specification as to their location.

The licensee has proposed to modify the TSs by adding an a'dditional limitation that one of the indicators in each of the eight locations be operational. This additional limitation or control will help to ensure that the average temperature is being measured during surveillance testing and is acceptable.

Item 14, Page 3/4 6-47, Item 15, Page 3/4 6-52 Section 3.6.5.1.2 of the TSs (page 3/4 6-47) specifies the operability requirements for the refueling area secondary containment integrity.

Section 3.6.5.3 of the TSs specifies the operability requirements for the standby gas treatment system (SGTS). There is a note at the bottom of -

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p each page which states when the systems must be operable..The note states that-refueling area secondary containment must be maintained "When irradiated fuel is being handled in the refueling area secondary containment and during core alterations and operations with a potential for draining the reactor." The note on operability for the SGTS is similar.

The use of the "and" type logic is incorrect.

The present wording could be interpreted as only requiring secondary containment integrity.when all three conditions are present.

The licensee proposes to replace the word "and" with "or" to correctly state that the systems must be operable when any one or more of the three conditions exist.

The proposed changes are clarifications and are acceptable.

Item 16, Page 3/4 6-23, Item 17, Page 3/4 6-42 Table 3.6.3-1 lists the primary containment isolation valves.

On page 3/4 6-23 and in note 16 to this table, there is a typographical error in that the penetration number for the tip purge check valve is listed as "035A" when it should be "035B".

The proposed change corrects an error and is acceptable.

Item 18, Page 3/4 6-42 There are three pages of notes to Table 3.6.3-1 (mentioned above).

Note 17 applies to two excess flow check valves on the recirculation pump l

cooler flow instrument lines.

The instrument lines are connected to a closed cooling water system inside containment and can leak only if tne line or instrument should rupture.

The note presently states that "leaktightness of the line is verified during the integrated leak rate test (Type A test).

To clarify that these excess flow check valves are only tested during the'ILRT, the proposed change is to add a sentence stating.that "the excess flow check valves are subject to operability L

testing, but no Type C test is performed or required".

The proposed change is acceptable in that it is a clarification and does not modify any of the existing requirements.

Item 19, Page 3/4 6-17, Item 20, Page 3/4 6-18 The instrumentation lines inside containment include lines for Reactor Water Cleanup, Reactor Core Isolation Cooling, High Pressure Coolant Injection, and Drywell Sump level.

Although all of these instrument l

lines might be indirectly described as " reactor" instrument lines, they are generally described by their own system name such as RWCU instrument, or RCIC instrument line.

The present TSs on pages 3/4 6-17 and 3/4 6-18

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refer to " reactor" instrument lines listed in Table 3.6.3-1 and do not l

reference the other system instrument lines such as RCIC, HPCI and RWCU that are also listed in Table 3.6.3-1.

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The proposed change is to delete " reactor" and reference only

" instrumentation" line excess flow check valves.

This proposed change would then allow all of the instrument lines to be referenced, rather than only the reactor instrument lines. The proposed change constitutes an additional control and is acceptable.

Item 23, Page 3/4 8-1 Section 3.8.1.1 of the TSs specifies the requirements for A.C. electrical power sources that have to be operable. Action item b. describes the requirements for verifying the operability of the remaining A.C. power sources if two of the four diesel generators become inoperable. Action item e describes the requirements for verifying the operability of all required systems, subsystems, trains, components and devices that depend on the remaining A.C. power sources if two of the four diesel generators become inoperable. Both action items pertain to the same plant condition. Action item e now has a statement referring the operators to also see action item b.

The proposed change is to add a similar statement to action item b to "Also see action item e of 3.8.1.1".

This is a purely administrative change and is acceptable. However, page 3/4 8-1 is not being revised by this amendment. The clarifying sentence will be included in the more extensive revisions to page 3/4 8-1 which the ifcensee has proposed in an application dated September 9, 1988 on Diesel Generator Testing. Staff review of the latter has essentially been completed.

Item 24. Page 3/4 6-16 The proposed change is to correct the address and the addressee for the monthly ope *ating reports required to be submitted to the USNRC by Section 6.9.1.6.

The reports will be sent to " ATTN: Document Control Desk".

This is a purely administrative change with no safety significance and is acceptable.

Item 25, Page 3/4 1-2 In Section 4.1.2, the surveillance requirements on reactivity anomalies require that the reactivity equivalence of the difference between the actual Rod Density and the predicted Rod Density shall be verified to be less than or equal to 1% delta K/K. However, the reactor must first be in operational condition 1 or 2 in order to carry out the surveillance (i.e., in power operation or startup).

It seems rather obvious that the reactor has to be critical in order to perform the surveillance and that the reactor has to be in operable condition 2 (startup) in order to go critical. To make this clear, the licensee proposes to add " note C" to Section 4.1.2 to state that "the provisions of Specification 4.0.4 are not applicable". Specification 4.0.4 states:

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r Entry into an operational condition or other specified applicable condition shall not be made unless the surveillance requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified."

The proposed change would allow entry into conditions 1 or 2 prior to the completion of the surveillance requirements. The proposed change is a clarification and does not change the requirement. The proposed change is acceptable.

Item 26, Page 3/4 3-8 Table 4.3.1.1-1 on " Reactor Protection System Instrumentation Terveillance Requirements" has 10 notes associated with the Table. Note g calls for verification of core flow using baseline data obtained during the startup testing program. All of the startup test programs were completed prior to the commercial operation date (February 1,1986) so this note is no longer applicable. Deletion of Note g is acceptable.

Item 27, Page 3/4 3-100 Section 3.11.1.1 of the TSs requires that the concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II. Column 2 for radionuclides other than dissolved or entrained noble gases. To ensure compliance with this limit, there is a surveillance requirement that specifies that all radioactive liquid wastes shall be sampled and analyzed according to the program in Table 4.11.1.1.1-1 (page 3/4 11-2). The Table specifies the sampling frequency, the minimum analysis frequency, the t measured and the lower limit of detection (LLD) ype of activity to be for each type of a ctivi ty. There is a whole page associated with this Table describing how the LLD is defined and how it is determined. For both batch and x10guousreleases,theLLDforprincipalgammaemittersislistedas5 conti microcuries/ml. The Table also specifies the principal gamma emitters for which the LLD specification applies.

Section 3.3.7.11 of the TSs specifies the operability requirements on the radioactive liquid effluent monitoring instrumentation. This section requires that the instrumentation listed in the associated Table (Table 3.3.11-1) shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 (discussed above) are not exceeded. The notes for this table specify what actions are to be taken if the minimum number of instrument channels are not available. One action specified (when the minimum number of instrument channels is not available) requires sampling and analysis in accordance with Specification 4.11.1.1.1 discussed in the previous paragraph. Another action specified 1

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i is to collect and analyse grab sampig for gross radioactivity (beta or gamma) at a limit of detection of 10 microcuries/ml. To be consistent withSpecification 4.11.1.1.1, the licensee proposes to changes this LLD

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to 5 x 10~ microcuries/ml. The proposed change is acceptable.

Item 28, Page 3/4 3-94 j

Limerick Units 1 and 2 have a common control room on the 269' elevation.

Likewise,- the auxiliary equipment room located on the 289' elevation above the control room is common to both units.

The auxiliary equipment room is designated as fire zone 25 in the Limerick Fire Protection Evaluation Report (FPER, page 5-40). Table 3.3.7.9-1 of the TSs lists the fire protection instrumentation that must be operable in each fire zone. For Fire Zone 25, the present TSs list fifteen heat detectors as being installed in the raised non-Power Generation Control Complex for Unit 1.

However, as shown in Figure B-21 of the FPER, two of the fifteen heat detectors are on the Unit 2 side and will be included in future TSs for Unit 2, but should not have been included in the Zone 25 listing for Unit 1.

The proposed change is to revise the number of heat detectors from 15 to 13.

The proposed change corrects an error and is acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that the amendment involves 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 signifi-cant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets 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 nor environmental assessment need be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The Comission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (54 FR 18176) on April 27, 1989 and consulted with the State of Pennsylvania. No public comments were received and the State of Pennsylvania did not have any comments.

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and

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(2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and the security nor to the health and safety of the public.

Principal Contributor:

Dick Clark, F. Witt, F. Litton, T.M. Su, N.

Trehan, M. LaMastra and E. Trottier Dated: June 22, 1989 i

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