ML20133C311

Safety Evaluation Supporting Amend 169 to License DPR-35

ML20133C311
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Site:	Pilgrim
Issue date:	12/27/1996
From:	NRC (Affiliation Not Assigned)
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NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D.C. 30006 4001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.169 TO FACILITY OPERATING LICENSE NO. DPR-35 BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION l

DOCKET NO. 50-293 l

1.0 INTRODUCTION

I By application dated May 1,1996, as supplemented November 26, 1996, Boston l

Edison Company (the licensee) requested changes to the Technical Specifications (TSs) ;for Pilgrim Nuclear Power Station. The proposed changes will modify Table 3.1.1, " Reactor Protection System (SCRAM) Instrumentation l

Requirement," Table 3.2.C.1, " Instrumentation That Initiates Rod Blocks," and TS 3/4.4, " Standby Liquid Control." The November 26, 1996, letter provided clarifying information and additional changes that did not change the initial proposed no significant hazards consideration determination.

l 2.0 TABLE 3.1.1. " REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT" The licensee's May 1,1996, letter to the NRC proposes to modify Table 3.1.1,

" Reactor Protection System (SCRAM) Instrumentation Requirement." The amendment will modify Note 7 to Table 3.1.1.

Currently, the note specifies that when the mode switch is in the refuel position and the reactor coolant system (RCS) temperature

< 212" F and the reactor is subcritical with fuel in the vessel, the following SCRAM functions: Mode Switch in Shutdown, Manual SCRAM,. High Flux (intermediate-range monitor (IRM), SCRAM Discharge Volume High Level, average power range monitor (APRM) (15%) High Flux SCRAM are required to be operable.

For the above SCRAMS, the proposed TS change now makes an exception to the SCRAM function operability requirement when the mode switch is in the refuel position and the reactor vessel head is removed and all control rods are fully inserted in cells containing one or more fuel assemblies.

In addition, to be consistent with the Standard Technical Specification (STS), the licensee has applied Note 7 to the IRM Inoperative and APRM Inoperative SCRAMS. The refueling definition deviates from the improved STS as refueling is described as head-off while the improved STS would only require a bolt to be detensioned. The Pilgrim FSAR describes limitations on station operations necessary to satisfy operational criteria.

The final safety analysis report (FSAR) states that refueling operations are i

accomplished with the head-off so that the primary system cannot be repressurized. This definition was used in the development of the station TSs and is consistent with the FSAR.

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When in the refueling mode, the refueling interlocks are engaged including the j

one-rod-out interlock. The one-rod-out interlock precludes moving more than a l

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single control rod at a time and requires all other rods be fully inserted.

l This condition is allowable because TS 3.3.A.1, " Reactivity Margin - core l

loading," assures that the core can be made subcritical with one-rod-out. The one-rod-out requirement is analyzed for the most reactive condition during the operating cycle with the strongest operable control rod in its full put position and all other rods fully inserted and, therefore, bounds the-refueling condition. The shutdown margin is determined as part of the reload analysis and confirmed by insequence criticality testing during startup and j

monthly monitoring of critical rod configuration. When a rod is being moved, 1

as allowed by the refueling interlocks, the SCRAM function will be operable.

3 In addition, during fuel movement all control rods are required to be fully 1

inserted. This assures that the reactor is maintained in a shutdown condition I

during refue' ling for all possible conditions such that no event requiring a j

SCRAM will occur.

Based on the above, the NRC staff has determined that the subcriticality is maintained during all possible refueling conditions and, l

therefore, the SCRAM function is y aeded and this TS change is acceptable.

l In addition, the NRC staff has reviewed these proposed changes and concluded that the proposed changes are consistent with the STS.

3.0 TABLE 3.2.C.1. " INSTRUMENTATION THAT INITIATES ROD BLOCKS" The licensee's May 1,1996, letter to the NRC proposes to modify Table 3.2.C.1, " Instrumentation that Initiates Rod Blocks." The proposed amendment will add Note 6 to Table 3.2.C.1.

The current TSs specify that all rod block functions associated with the applicable functions of Table 3.1.1 are required to be operable at all times.

The proposed change makes an exception to the rod block function operability requirement when the mode switch is in the refuel position and the reactor vessel head is removed and all control rods are fully inserted in cells containing one or more fuel assemblies, for the following rod blocks; APRM Upscale, APRM Inoperative, IRM Downscale, IRM Detector not in Startup Position, IRM Upscale, IRM Inoperative, SRM Detector not in Startup Position, SRM Downscale, SRM Upscale, SRM Inoperative, SCRAM Discharge Instrument Volume Water Level-High, and SCRAM Discharge Instrument Volume-SCRAM Trip Bypassed. When a control rod is being moved as allowed by the refueling interlocks, all rod block functions in the TS will be operable.

The control rods provide the primary means for control of reactivity changes.

The control rod block instruentation is designed to ensure that specified fuel design limits are not exceeded. When placed in the shutdown mode, the reactor receives an automatic SCRAM concurrent with a rod block. When in the refueling mode, the refueling interlocks are engaged including the one-rod-out interlock. The one-rod-out interlock is the rod block function during refueling which allows moving a single control rod if all other rods are fully inserted. The one-rod-out interlock also precludes moving more than a single control rod at a time. Specification 3.3.A.1, " Reactivity Margin - core loading," requires that the core can be made subcritical in the most reactive condition during the operating cycle with the strongest operable control rod in its full out position and all other rods fully inserted. Therefore, TS 3.3.A.1 and the one-rod-out interlock assures that the reactor is maintained in a suberitical condition during refueling for all possible conditions such that no other rod block is required and the staff concludes that this TS

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l change is acceptable In addition, the staff has reviewed these proposed i

changes and concluded that the proposed changes are consistent with the STS.

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The STS require a shutdown rod block function to be operable in the shutdown mode and a refueling mode rod block.

The current TSs did not include the i

shutdown rod block function. To establish requirements consistent w'It,h the i

STF, the licensee proposed to add the shutdown rod block to Tables 3.2.C.1, l

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3.2.C-2 and 4.2.C.

The staff has reviewed these changes and concluded they provide an additional limitation or restriction on the plant operation and are l

acceptable.

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4.0 STANDBY LIOUID CONTROL The licensee's May 1, 1996, letter to the NRC proposes to reformat and modify TS 3.4, " Standby Liquid Control System," to make them consistent with the STS.

The staff review comp'ared the current requirements with the proposed requirements to ensure that either current TS limits are maintained or that a i

reasonablesafetybas;esexistforadoptingtheSTSrequirements.

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4.1 Requirements Maintained or Changed by the Proposed TS Change J

I The staff compared the current TS requirements with the proposed TS to ensure i

all the original requirements are maintained or a basis for modification or deletion is provided. An evaluation of the proposed TS changes are provided by the notes that follow.

If a note does not appear, the staff has concluded i

that the original TS requirements have been maintained.

Below is a table of the current requirements and their equivalent in the proposed TS:

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4.2 Current Reauirements Location In Proposed TS 3.4 Applicability 3.4 Applicability (Note 1) 3.4.A Normal System Availability 3.4.C, 3.4.D (Note 2) 3.4.B Operation with Inoperable Components 3.4.B 3.4.C.1 The net volume 3.4.A, 3.4.C. 4.4.4 (Note 2) 3.4.C.2 Temperature 3.4.C 4.4.5 (Note 2) 3.4.C.3 Enrichment 4.4.7 3.4.D.1 Cold Shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.4.C, 3.4.D (Note 2) 3.4.D.2 Enrichment Requirements 4.4.6, 4.4.7, BASES (Note 3) 3.4.D.3 Enrichment Requirements Original 4.4.A, 4.4.7, BASES (Note 3)

NOTES 1.

Currently, the Standby Liquid Control (SLC) may be inoperable in the cold shutdown condition when all operable control rods are fully inserted (that is the SLC is required to be operable in the run, startup, and hot shutdown). The proposed TS changes the applicability statement such that the SLC is required to be operable in the run and startup modes only. The safety objective of the system SLC is to provide a backup method, independent of the control rods, to maintain the reactor suberitical as the nuclear system ccols in the event that not enough control rods can be inserted to counteract the positive reactivity effects of a

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. colder moderator.

It also provides a method to mitigate the effects of Anticipated Transients Without Scram (ATWS). With the reactor in the hot or cold shutdown condition, criticality is prohibited because the mode switch position " Shutdown" prevents' rod withdrawal.

In the refuel mode, only a single control rod can be withdrawn from a core cell containing fuel assemblies.

Demonstration of adequate shutdown margin, required by 3.3.A.1,

" Reactivity Margin - core loading," ensures that the reactor will not become critical. Therefore, it is not possible to make the reactor core critical in the hot or cold shutdown conditions and the SLC system is not required to be operable for the hot or cold shutdown conditions. Based on the above, the staff concludes that the 'cManges to the applicability statement are acceptable.

In addition, the staff has reviewed these proposed changes and concluded that the proposed changes are consistent with the improved STS.

2.

The current SLC TSs specify five limiting conditions that would make both trains of the SLC inoperable and require the plant to be placed in cold shutdown (outside mode of applicability) with all rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The staff has determined that the five limiting conditions relating to the SLC operability have been maintained by the proposed TS. However, the proposed TS allows 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the plant to restore one train to operable status, and if a train cannot be restored, the plant must be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The licensee justified these time limits based on the following:

a.

The 8-hour completion time is considered acceptable given the low probability of a design-basis accident (DBA) or transient occurring concurrent with the failure of the control rods to shut down the reactor, and b.

The 12-hour completion time to bring the plant to hot shutdown is reasonable based on operating experience to reach hot shutdown from full power in an orderly manner and without challenging plant systems.

The intent of the action statements is to place the plant in a mode outside of its applicability statement in a safe and reasonable time frame. Based on the above, the staff concludes that the proposed TS is more conservative, by allowing less time to place the plant in a mode outside of the applicability statement and represents an enhancement to current TS. The staff l

has reviewed these proposed changes and found them acceptable and consistent with the STS.

3.

The requirement to verify S-10 enrichment by test any time boron is added to the solution and, during each refueling outage, is replaced with verifying the enrichment of the B-10 prior to its i

addition to the tanks. The B-10 cnrichment will be independently j

verified upon receipt to ensure that the enrichment is as 4

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certified by the vendor. Since enrichment of the solution in the

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tank cannot change by any other means but chemical addition, the TS requirement to ensure the enrichment at the required level is i

maintained because only properly enriched material is available-for addition.

The current TS requires that any time boron is added to the F

solution, and during each refueling outage, the enrichment level shall be verified by analysis.

If the results of that analysis i

are not received within 30 days or the analysis determines the enrichment does not meet the specifications, the licensee must report to the NRC their plan of action to restore the required enrichment. TS Table 6.9.1 provides the reporting requirements.

j The reporting requirements of Table 6.9.1 for boron enrichment l

will be deleted because they are no longer applicable. The verification of enrichment will be made prior to the need to use i

l the material and verification of the solution is no longer necessary. The bases for this change is discussed above.

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addition, the staff has reviewed these proposed changes and j

concluded that the proposed changes are consistent with the STS.

l Surveillance Reouirements Location In Procosed TS i;

4.4.A.1 Pump Test 4.4.1 i

4.4.A.2 Relief Valve Setpoint Deleted (Note 4)

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i 4.4.A.b Manual Initiation Test 4.4.2 (Note 5) 1 4.4.A.c 4.4.A.b Test interval 4.4.2 (Note 5) j 4.4.B.1 Testing of redundant component Deleted (Note 6) f 4.4.C.1 Daily volume check 4.4.4 4.4.C.2 Daily temperature check 4.4.5, 4.4.9 (Note 7) 4.4.C.3 Concentration check 4.4.4, 4.4.6 4.4.C.4 Enrichment check 4.4.7, BASES (Note 3) 1 Table 6.9-1 SLC enrichment out of Deleted (Note 3) j.

specification report

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4.

The requirements to verify the proper operation and setpoint of the relief valves in the SLC system are redundant to the Inservice Test Program as required by 10 CFR 50.55a(g). The program is required by TS 3/4.13 and is in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

Therefore, this requirement can be deleted from the current TS.

5.

These TSs were reworded and combined to make the TS more readable, and, therefore, more understandable, by plant operators as well as other users. During the rewording, no technical changes (either actual or interpretational) to the TSs were made.

6.

The requirement for demonstrating operability of the redundant subsystems was originally specified because there was a lack of plant operating history and insufficient equipment failure data to justify not testing redundant equipment. Since that time, plant operating experience has demonstrated that testing of the

,. redundant subsystem when one subsystem is inoperable is not necessary to provide adequate assurance of operability of the remaining subsystem. Deletion of this requirement is based on the acceptability of taking credit for normal periodic surveillance as a demonstration of operability and availability of the remaining components.

This change is consistent with changes granted undgr Amendment No. 135.

7.

The current TS requires that if the sodium pentaborate solution temperature falls below 48 'F, the system will be flow tested to verify flow path. The TS does not state when the flow test must be performed or how often. The proposed TS requires that the sodium pentaborate solution and pump suction piping be maintained abovs 48 'F or place the plant in TS 3.4.C, which would require the plant to return the temperature to greater than 48 'F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or shut down the plant in the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In addition, when the temperature is restored to > 48 'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the concentration must be verified and all heat traced piping between the storage tank and pump must be verified to be unblocked.

Even though the flow test requirement has been deleted, the staff has concluded that the revised TS more clearly specify what actions to take if the temperature is out of limit and provides time frames for completing the actions. The adequacy of the 8-hour action time was discussed in Note 2.

The staff has concluded that the proposed changes are consistent with the STS.

New Reauirementi 3.4.A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Boron concentration action time (Note 8) i 3.4.A and 8 10 days from discovery (Note 9) 4.4.3 Continuity of Explosive Charge (Note 10) 4.4.8 Heat traced piping unblocked (Note 11) 4.4.9 Pump suction temperature (Note 12) 8.

As a result of ATWS rule, the shutdown criteria for the SLC system changed.

The current TS allows the sodium pentaborate solution enrichment to be less than the required enrichment for up to 7 days if the enrichment, concentration, and volume meet the original pre-ATWS design criteria. The original concentration requirement had been determined to be a 4.46% sodium pentaborate concentration. The proposed TS limiting condition of operation allows continued operation for 3 days with a minimum concentration of 8% sodium pentaborate solution. The 8% concentration was selected based on the original pre-ATWS shutdown criteria with margin. The enrichment limit set by the TSs was discussed previously. The volume limits have not changed from the current TS. This proposed TS is consistent with the STS.

9.

The proposed LCO for concentration of boron in solution and subsystem inoperability include a limit on the maximum time 1

allowed for SLC subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. This new 1

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! restriction is intended to prevent exceeding the assumptions regarding out-of-service time for a SLC subsystem as a result of sequential inoperabilities of SLC subsystems due to boron concentration not within limits and an SLC subsystem inoperable due to other reasons. This proposed TS is consistent with the recommendations in the STS.

10.

A new monthly surveillance was added to verify continuity of the explosive charge associated with the explosive valves.

11.

A new surveillance was added to verify all heat traced piping between the storage tank and pump suction is unblocked every 24 months and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> whenever the solution temperature must be restored to >48 'F.

I 12.

A new surveillance was added to verify temperature of the pump suction piping. This surveillance was added to ensure the temperature remains at least 10 'F above the boron precipitation value.

The staff has reviewed these new requirements and conclude they result in an enhancement to the existing TSs. Therefore, these new requirements are acceptable.

The staff has cencluded that the revised TS specify more clearly when SLC is required, what actions to take if SLC is inoperable, and time frames for completing the actions. Based on the above, the staff concludes that the revisions enhance the current TS by making them more definitive and supplementing them with action statements and required completion times and, therefore, the changes are acceptable.

In addition, the staff has reviewed these proposed changes and concluded that the proposed changes are consistent with the STS.

5.0 1 TATE CONSULTATION In accordance with the Commission's regulations, the Massachusetts State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC 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 significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards i

consideration, and there has been no public comment on such finding (61 FR 28606). Accordingly, the amendment meets the eligibility criteria for i

categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR i

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4," 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

5 TheCommissionhasconcluded,basedontheconsiderationsdiscussedepove, 4

i that (1) there is reasonable assurance that the health and safety of the l

i public will not be endangered by operation in the proposed manner, (2) such j

activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal: Contributor:

A. Wang Date: December 27, 1996 l

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