ML18152A329

Proposed Tech Specs Re Intermediate Range High Flux Reactor Trip Setpoint

ML18152A329
Person / Time
Site:	Surry
Issue date:	10/26/1992
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
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ML18152A330	List: ML18152A329 ML18153D161
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NUDOCS 9211020272
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ATTACHMENT 1 SURRY POWER STATION PROPOSED TECHNICAL SPECIFICATION CHANGES INTERMEDIATE RANGE HIGH FLUX SETPOINT

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PDR ADOCK 05000280 P PDR

TS 2.3-1 2.3 LIMITING SAFETY SYSTEM SETIINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to trip and permissive settings for instruments monitoring reactor power; and reactor coolant pressure, temperature, and flow; and pressurizer level.

Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit.

Specification A. Protective instrumentation settings for reactor trip shall be as follows:

1. Startup Protection (a) High flux, power range (low set point) - < 25% of rated power.

(b) High flux, intermediate range (high set point) - current equivalent to 5. 40% of full power.

(c) High flux, source range (high set point) - Neutron flux 5. 106 counts/sec.

2. Core Protection (a) High flux, power range (high set point) - < 109% of rated power.

0 Amendment Nos.

" . TS 2.3-4 B. Protective instrumentation settings for reactor trip interlocks shall be as follows:

1. The reactor trip on low pressurizer pressure, high pressurizer level, turbine trip, and low reactor coolant flow for two or more loops shall be unblocked when power ~ 10% of rated power.

2. The single loop loss of flow reactor trip shall be unblocked when the power range nuclear flux ~ 50% of rated power.

• During two loop operation with the loop stop valves in the inactive loop open, this blocking setpoint, established by Permissive 8, may be increased to 60% of rated power only after the overtemperature L\T setpoint is adjusted to the mandatory two loop value. For two loop operation with the loop stop valves of the inactive loop closed, Permissive 8 may be increased to 65% of rated power only after the overtemperature L\T setpoint is adjusted to the mandatory value for this condition.

3. The power range high flux, low setpoint trip and the intermediate range high flux, high setpoint trip shall be unblocked when power s; 10% of rated power.

4. The source range high flux, high setpoint trip shall be unblocked when the intermediate range nuclear flux is s; 5 x 10-11 amperes.

Basis The power range reactor trip low setpoint provides protection in the power range for a power excursion beginning from low power. This trip value was used in the safety analysis. ( 1) The Source Range High Flux Trip provides reactor core protection during shutdown (COLD SHUTDOWN, INTERMEDIATE SHUTDOWN, and HOT SHUTDOWN) when the reactor trip breakers are closed and reactor power is below the permissive P-6. The Source and Intermediate Range trips in addition to the Power Range trips provide core protection during Amendment Nos.

TS 2.3-5 reactor startup when the reactor is critical. The Source Range channels will initiate a reactor trip at about 106 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a reactor trip at a current level proportional to ::; 40% of RATED POWER unless manually blocked when P-1 O becomes active. In the accident analyses, bounding transient analysis results are based on reactivity excursions from an initially critical condition, where the Source Range trip is assumed to be blocked. Accidents initiated form a subcritical condition would produce less severe results, since the Source Range trip would provide core protection at a lower power level. No credit is taken for operation of the Intermediate Range High Flux trip. However, its functional capability is required by this specification to enhance the overall reliability of the Reactor Protection System.

The high and low pressurizer pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The low pressurizer pressure reactor trip also trips the reactor in the unlikely event of a loss-of-coolant accident. (3)

The overtemperature Lff reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 3 seconds), and pressure is within the range between high and low pressure reactor trips. With normal axial power distribution, the reactor trip limit, with allowance for errors, (2) is always below the core safety limit as shown on TS Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor limit is automatically reduced. (4)( 5 )

The overpower and overtemperature protection system setpoints have been revised to include effects of fuel densification on core safety limits and to apply to 100% of design flow. The revised setpoints in the Technical Specifications will ensure that the combination of power, temperature, and pressure will not exceed the revised Amendment Nos.

TS 3.7-13b TABLE 3.7-1 (Continued)

ACTION 4. With the number of channels OPERABLE one less than required by the Minimum OPERABLE Channels requirement and with the THERMAL POWER level:

a. Below P-6, (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour. Two Source Range channels must be OPERABLE prior to increasing THERMAL POWER above the P-6 setpoint.

b. Above P-6, operation may continue.

ACTION 5. With the number of channels OPERABLE one less than required by the Minimum OPERABLE Channels requirement, verify compliance with the SHUTDOWN MARGIN requirements within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

ACTION 6.A. With the number of OPERABLE Channels equal to the Minimum Operable Channels requirement, REACTOR CRITICAL and POWER OPERATION may proceed provided the following conditions are satisfied:

1. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2. The Minimum OPERABLE Channels requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.1 .

6. B. With the number of OPERABLE Channels one less than required by the Minimum Operable Channels requirement, be in Hot Shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Amendment Nos.

e TS 3.12-2 culations and physics data obtained during Unit Startup and subsequent operation, will be permitted.

c. The shutdown margin with allowance for a stuck control rod assembly shall be greater than or equal to 1.77% reactivity under all steady-state operation conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (T avg ~

547°F) if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron.

4. Insertion limits do not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin indicated above must be maintained except for the low power physics test to measure control rod worth and shutdown margin.

For this test the reactor may be critical with all but one full length control rod, expected to have the highest worth, inserted.

Amendment Nos.

e e ATTACHMENT 2 SURRY POWER STATION DISCUSSION AND SIGNIFICANT HAZARDS CONSIDERATION

. PROPOSED TECHNICAL SPECIFICATION CHANGES INTERMEDIATE RANGE HIGH FLUX SETPOINT

DISCUSSION OF PROPOSED CHANGES INTRODUCTION Virginia Electric and Power Company proposes a revision to the Technical Specification (TS) Intermediate Range (IR) High Flux trip setpoint, and the deletion of TS 3.12.A.4 which governs the predicted critical rod position at subcritical conditions. An additional change to the Table 3.7-1, Operator Action 4, is proposed to confirm the availability of the Source Range channel and, hence, to ensure that startup protection is provided by the Source Range channel. A discussion of these proposed changes is provided in the follow1ng:

• TS 2.3.a.l(b) is being revised to increase the Intermediate Range High Flux trip setpoint from the current equivalent of~ 25% of full power to the current equivalent of~ 40% of full power. This setpoint revision will increase the operational margin between the permissive P-10 and the IR High Flux trip setpoint, thereby allowing more latitude in trip settings and reducing the potential for unneeded reactor trips.

• TS 2.3 Basis is being revised to more fully describe the startup accident protection provided by the Source Range and Intermediate Range channels.

• TS 3.12.A.4 is being deleted to eliminate the excessive RCS boration required to ensure that the specification, as written, is met.

Above permissive P-6, TS 3.12.A.2 and startup procedures ensure that the plant operating conditions remain within the current TS 3.12.A.4.

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Below P-6, reactor protection for reactivity insertion transients during approach to critical continues to be provided by the Source Range trip, which is automatically activated below permissive

• TS Table 3.7-1, Operator Action 4, is revised to limit the duration one Source Range channel can be inoperable at a thermal power level below P-6. This revision establishes an allowed outage time for the Source Range channel when the power level is below P-6.

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BACKGROUND The Surry Power Station Intermediate Range High Flux trip provides backup protection to the Power Range High Flux trip (low setpoint) during reactor startup. Both IR channels include an independent compensated (gamma) ion chamber which is located external to the core along a major core axis.

The IR circuitry provides monitoring of the flux level over an eight decade range ( 1. OE-11 to 1. OE-3 amperes). A reactor trip is generated based on one out of two channels exceeding a current equivalent of 25%

rated thermal power. The channels can be manually bypassed whe~

permissive P-10 (2 of 4 power range channels> 10% of rated thermal power) is active.

The Intermediate Range High Flux level trip provides protection for reactivity (power excursion) accidents during reactor startup. However, the Surry Updated Final Safety Analysis Report (UFSAR) accident analyses do not take credit for this protection in accident analysis models. The UFSAR assumes that a reactor trip is provided by either the Source Range or the Power Range High Flux level trips. A summary of the startup protection required by Surry Technical Specification 2.3.A.1 follows:

• High Flux, Power Range, low setpoint (2 out of 4 channels~

25% rated thermal power).

• High Flux, Intermediate Range (1 out of 2 channels~ current equivalent to 25% rated thermal power).

• High Flux, Source Range (1 out of 2 channels~ l.OE+6 counts/sec).

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e The Surry Power Station Technical Specification Amendment No. 117 (1) for both units revised the Source Range protection to bring the Surry 11 Technical Specifications in closer agreement with NUREG-0452, Standard Technical Specification for Westinghouse Pressurized Water Reactors, 11 Revision 4 (10). The Nuclear Flux Source Range trip protection was expanded to provide separate entries for modes with reactor trip breakers closed and modes with the breakers open. Notably, the Source Range High Flux level trip must be operable whenever the reactor trip breakers are closed, the control rod drive system is capable of rod withdrawal, and the Thermal Power level is below P-6 (i.e., the modes of Cold Shutdown, Intermediate Shutdown, and Hot Shutdown). Conversely, the Power Range High Flux level trips must be operable above Hot Shutdown. With these Technical Specification requirements, there is no operating condition in which the Intermediate Range High Flux level trip provides the sole overpower protection. It can be concluded that the Intermediate Range High Flux level trip is needed only to provide backup trip protection, since it is not necessary to assume that this trip function initiates a reactor trip for the UFSAR safety analyses.

Because the IR neutron detectors are located outside the reactor vessel, the IR channel response has been observed to be sensitive to reactor core loading patterns. For example, the long-lived, low leakage patterns currently in use at Surry Power Station have a different IR detector response than the more traditional patterns used for initial core loadings. In addition, the IR detectors do not cover as much of the axial core length as the Power Range channels, so the IR detector response is also sensitive to the core axial flux distribution. As a result, control I ~ .

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e rod position effects and within-cycle power redistribution can have a significant impact on the IR channel response. These variations in IR channel response make it difficult to maintain the IR channel protective action setpoints.

Miscalibration of excore detectors was first reported in 1982 by Zion 1 for Power Range detectors in INPO SER 72-82 (2). All four units at North Anna and Surry discovered a similar problem in June 1983 involving improper settings of Intermediate Range High Flux trip setpoints, as reported in Licensee Event Reports (LER) 280-83027, 338-83048, 339-83038, and USNRC Inspection* and Enforcement Information Notice 83-43 (3),(4),(5),(6). Additional Intermediate Range detector events occurred in May and June 1985 at both Surry units, as reported in LER 1 s 280-85010 and 280-85010-01 (7), (8). Subsequent to the 1983 events, North Anna pursued a license amendment to revise their Technical Specification nominal Intermediate Range High Flux trip setpoint from 25% to 35% rated thermal power (9).

TS 3.12.A.4 requires the predicted critical rod position at subcritical conditions to be above the zero-power rod insertion limit. Following reactor trip or controlled shutdown, this specification effectively requires boration of the reactor coolant system (RCS) to ensure compliance. As is demonstrated in the evaluation of the proposed Technical Specifications changes, TS 3.12.A.2 and startup procedures ensure that plant operating conditions remain within those defined by TS 3.12.A.4 above permissive P-6. Below P-6, reactor protection for reactivity insertion transients during approach to critical continues to

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be provided by the Source Range trip, which is automatically activated below permissive P-6. TS 3.12.A.4 is therefore unnecessary and may be deleted without consequence to the allowable operating conditions assumed in supporting safety analyses.

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SPECIFIC CHANGES INCREASED INTERMEDIATE RANGE TRIP SETPOINT (TS 2.3.A.l.b)

The revision of the Intermediate Range High Flux trip setpoint from~ 25%

to~ 40% power was evaluated by determining a maximum ~ttainable IR trip setpoint, and then evaluating that value for the various startup accidents that could require a backup trip for the Power Range High Flux trip (low setpoint). The IR High Flux trip setpoint will be set nominally at the current equivalent of~ 35% of Rated Power.

The maximum attainable IR trip setpoint was established by adding a combination of instrument and process measurement uncertainties to the Technical Specification allowable trip setpoint of~ 40% power. Using a methodology similar to the Westinghouse setpoint methodology (12),

Virginia Electric and Power Company evaluated the total channel error for the Intermediate Range High Flux trip. This evaluation included a statistical combination of the instrumentation and random process measurement uncertainties, along with an arithmetic combination of nonrandom process measurement uncertainties, which were treated as biases. For the purposes of this conservative evaluation, both the contra l rod insertion effects and the reactor vessel down comer temperature effects were treated as biases. Within-cycle power redi stri buti on effects are accommodated by periodic adjustment of the actual plant IR trip setpoint equivalent current.

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The control rod insertion and within-cycle power redistribution effects were evaluated for a recent Surry cycle by modelling the IR detector signal (neutron flux) as a function of rod position and burnup. A similar calculation is performed for each reload core to support adjustment of the IR trip setpoint equivalent current for within-cycle power redistribution, and to verify that the calculated IR channel uncertainty does not violate the assumptions of the safety analysis. The decreasing downcomer temperature effect accounts for the temperature difference between the maximum cold leg temperature during power operation (which is the maximum downcomer temperature at IR channel calibration) and the minimum temperature for criticality given in Technical Specification 3.1.E.4.

The total IR channel error for a recent Surry cycle was determined to be 27% power, which includes the conservative application of a 12.5% power allowance for the nonrandom process measurement uncertainties (control rod position and downcomer temperature effects), and which assumes a nominal setpoint of~ 35% of full power. Combining this total error with the proposed IR Technical Specification allowable trip setpoint of 40%

power yields a maximum attainable IR trip setpoint of 67% power.

Previous submittals (13),(14),(20) have demonstrated that an increase in the effective flux trip setpoint has an insignificant impact on reactivity accidents (power excursions) initiated from low power conditions. These analyses considered an increase to the effective flux trip setpoint (i.e.,

the accident analysis trip setpoint) from 35% power to 118% power, and addressed both the rod withdrawal from subcritical and rod ejection 8

e accidents. Since the conservative application of errors to the proposed allowable IR Technical Specification trip setpoint yields a maximum attainable IR trip setpoint at a power (67%) which is well below the 118%

maximum value evaluated, it is concluded that the proposed revision to the IR trip setpoint will not impair the ability of the trip to provide backup protection to the Power Range High Flux trip (low setpoint) for reactivity accidents initiated from low power conditions.

MODIFICATION OF TS 2.3 BASIS The Basis of TS 2. 3 has been revised to reflect the startup accident protection philosophy that is supported by the present UFSAR accident analysis. The revised Basis specifies that no credit is taken for operation of the Intermediate Range High Flux trip for the UFSAR startup accident analysis. Bounding transient results assume reactor trip is provided by the Power Range High Flux trip (low set point); however, if the transient is initiated below P-6, the Source Range High Flux trip is not blocked, and less severe results are expected because of the Source Range trip protection. The functional capability of .the Intermediate Range High Flux trip at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

ALLOWABLE SOURCE RANGE OUTAGE TIME (TS TABLE 3.7-1)

The revision to Operator Action 4 of TS Table 3.7-1 provides an allowed outage time for conditions below a power level of P-6 with one Intermediate Range channel inoperable. The revised specification states:

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With the number of channels OPERABLE one less than required by the minimum OPERABLE Channels requirement and the THERMAL POWER 1eve 1 :

a. Below P-6, (Block of Source Range Reactor Trip) setpoint, restore inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour. Two Source Range channels must be OPERABLE prior to increasing THERMAL POWER above the P-6 setpoint.

This Operator Action, when combined with the other Source Range requirements for power above P-6 in Table 3.7-1, yields the same requirements as those for Operator Action 13 in NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors, 11 Revision 4 (10). The Surry Technical Specifications do not currently limit the duration for which one Source Range channel can be inoperable at a thermal power level below P-6. Thus, the intent of this proposed change is to confirm the availability of the Source Range channel, and hence, to ensure the startup protection provided by the Source Range channe 1 .

EVALUATION OF PROPOSED DELETION OF TS 3.12.A.4

Background

TS 3.12.A.4 requires the predicted critical rod position at subcritical conditions to be above the zero-power rod insertion limit. The purpose 10

of this specification is to ensure that the control rods and, hence, the core power distributions, are within the limits assumed in the reload core design analysis. Following reactor trip or controlled shutdown, boration of the Reactor Coolant System (RCS) is often required to ensure compliance with TS 3.12.A.4. For example, the critical rod position following reactor trip may be below the rod insertion limits due to the reactivity defect between 0% and 100% power.

TS 3.12.A.2 governs control rod position at criticality. For normal operation ascension to criticality or for inadvertent rod withdrawal events initiated from a subcritical condition above permissive P-6, it may be demonstrated that TS 3.12.A.2 effectively restricts operation to ensure that criticality is achieved above the rod insertion limits without the operational burden of unnecessary boration following reactor trip or controlled shutdown associated with TS 3.12.A.4. For rod withdrawal events initiated from a subcritical condition below P-6, reactor protection is provided by the Source Range High Flux reactor trip.

Normal Operation Evaluation TS 3.12.A.2 requires the control rods to be above the zero-power critical rod position at criticality. Plant startup procedures ensure that the requirements of TS 3.12.A.2 will be met at criticality during a controlled startup or in the event of inadvertent control rod withdrawal from a subcritical condition above permissive P-6. To reach permissive P-6, the rod withdrawal sequence must have already been initiated. During a normal operation approach to critical, the calculation of estimated critical rod 11

position and required boron concentration is completed prior to initiating the rod withdrawal sequence. In effect, the actions which must be taken to ensure that the requirements of TS 3.12.A.2 are met (i.e.,

when critical, the predicted critical rod position is above the zero power rod insertion limit) also ensure that the requirements of TS 3.12.A.4 are met (i.e., when subcritical, the predicted critical rod position is above the zero power rod insertion limit).

Accident Conditions Evaluation The original intention of TS 3.12.A.4 was to ensure a critical rod position and, hence, a core power distribution consistent with that assumed in the Rod Wi.thdrawal from Subcritical (RWSC) UFSAR accident analysis. Reactor protection for subcritical and low power rod withdrawal events was assumed to be provided by the Power Range High Flux trip (low setpoint) both above and below permissive P-6. Source Range protection was not assumed to be available, since the Source Range channel lacked the redundancy required to assume trip availability in UFSAR accident analyses. More recently, a Technical Specification change was sought to increase the required number of available Source Range Channels below permissive P-6 from 1 to 2 (15)(16). This change, along with Source Range trip bistable operability testing to verify Source Range channel response characteristics, validates the assumption of Source Range High Flux trip availability in accident analyses. In addition, the availability of the Source Range channel is further enhanced by imposing an allowable Source Range channel outage time. This pro vi des greater confidence in the overall reliability of the reactor protection system.

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• Rod withdrawal from subcritical events may be initiated from above or below permissive P-6. Below P-6, less than three reactor coolant pumps may be operating, or RCS cooling may be provided by the Residual Heat Removal (RHR) System. For any operating condition below P-6, Source Range protection or open trip breakers provide reactor protection against a rod withdrawal from subcritical event. Additional protection is provided by the other operable reactor protection system circuitry, including the Intermediate Range and Power Range (low setpoint) reactor trips. As was demonstrated in the North Anna Core Uprating submittal (17), and subsequent responses to NRC questions (18)(19), events initiated from allowable operating conditions below P-6 will not result in significant power generation or core heat flux when a reactor trip is actuated on the Source Range channel. Therefore, reactor protection is provided for all operating conditions below P-6, including one-RCP, two-RCP, and RHR operation.

Consideration must also be given to the effect of the proposed change on accidents initiated above permissive P-6. Operating Procedures require three reactor coolant pumps to be running above P-6. In safety analysis, reactor protection for the RWSC above P-6 is assumed to be provided by the Power Range High Flux (low setpoint) reactor trip. Because TS 3.12.A.2 and the associated plant startup procedures ensure that the reactor will not go critical with the control rods below the zero-power rod insertion limit, the current UFSAR analysis remains valid. The analysis demonstrates that the duration of power generation and the peak heat flux achi ev.ed during a RWSC event initiated above P-6 with three 13

e reactor coolant pumps running does not approach a condition of departure from nucleate boiling (DNB).

Other potential reactivity insertion events at subcritical, such as a boron dilution event at hot shutdown, are bounded in severity by the analyzed cases of the RWSC. Specifically, the potential reactivity insertion rates considered in the RWSC analysis result in more severe DNB ratios than the reactivity insertion rates associated with the maximum achievable boron dilution rate. The proposed deletion of TS 3.12.A.4 has no impact on the analysis of the boron dilution event, in terms of avoiding inadvertent criticality, since minimum shutdown margin requirements continue to be maintained.

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SAFETY SIGNIFICANCE The conservative application of errors to the proposed allowable IR Technical Specification trip setpoint yields a maximum attainable IR trip setpoint below 100% of Rated Thermal Power, and well below the Power Range High Flux reactor trip. Therefore, the IR channel continues to fulfill its role as an enhancement to the overall reliability of the reactor protection system. Furthermore, the IR reactor trip remains a viable backup trip to the Power Range High Flux trip (low setpoint), since it continues to provide protection for reactivity accidents (power excursions) from low power initial conditions.

TS 3.12.A.2 effectively ensures equivalent operating conditions to those ensured by TS 3.12.A.4 above permissive P-6. Below permissive P-6, reactor protection for reactivity insertion transients continues to be provided by the Source Range trip (which is automatically activated below permissive P-6). TS 3.12.A.4 may be deleted with no impact on the allowable operating conditions assumed in supporting reload core design and safety analyses and, therefore, with no impact on the results of safety analyses supporting plant operation.

The proposed a 11 owed outage time for the Source Range channel improves the startup protection provided by the Source Range channel.

These proposed revisions have been evaluated for their impact on the UFSAR safety analysis, with the conclusion that there are no changes in the

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UFSAR accident probabilities or consequences and that there are no changes in Technical Specification margins.

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REFERENCES

1. Letter from C. P. Patel (NRC) to W. L. Stewart (VP), 11 Surry Units 1

& 2 - Issuance of Amendments RE: Changes to Sections 3.7 and 4.1 (TAC Nos. 55380, 55381, 60307, 60308), 11 Serial No.88-096, February 17, 1988.

2. Institute of Nuclear Power Operations Significant Event Report 72-82, "Miscalibration of Nuclear Power Range Detector Channels Resulting from Core Design Changes," December 2, 1982.

3. Letter from J. L. Wilson (VP) to J. P. 0 1 Reilly (NRC), Serial No.83-051, transmitting Lincensee Event Report 83-027 for Surry Unit 1, July 15, 1983.

4. Letter from E. W. Harrell (VP) to J. P. 0 1 Reilly (NRC), Serial No.

N-83-094, transmitting Lincensee Event Report 83-048 for North Anna Unit 1, July 6, 1983.

5. Letter from E. W. Harrell (VP) to J. P. 0 1 Reilly (NRC), Serial No.

N-83-095, transmitting Lincensee Event Report 83-038 for North Anna Unit 2, July 6, 1983.

6. USNRC Office of Inspection and Enforcement, IE Information Notice No. 83-43, "Improper Settings of Intermediate Range (IR) High Flux Trip Setpoints, 11 June 24, 1983.

7. Letter from R. F. Saunders (VP) to USNRC, Serial No.85-016, transmitting Licensee Event Report 85-010-00 for Surry Unit 1, June 14, 1985.

8. Letter from R. F. Saunders (VP) to USNRC, Serial No. 85-016A, transmitting Lincensee Event Report 85-010-01 for Surry Units 1 and 2, July 30, 1985.

9. Letter from W. L. Stewart (VP) to USNRC, 11 Proposed Technical Specification Changes," Serial No.86-740, November 9, 1987.

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10. USNRC, Standard Technical Specifications for Westinghouse Pressurized Water Reactors," NUREG-0452, Revision 4, Fall 1981.

11. USNRC, 11 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, 11 NUREG-0800, Revision 2, July 1981.

12. Letter from C. M. Stallings (VP) to H. R. Denton (NRC), Serial No.78-541, September 29, 1978, containing the Westinghouse report, "Westinghouse Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology. 11

13. Letter from W. L. Stewart (VP) to H. R. Denton (NRC), Serial No.85-457, "Safety Evaluation of Rod Withdrawal from Subcritical at Low Flow Conditions, 11 April 1, 1985.

14. Letter from W. L. Stewart (VP) to H. R. Denton (NRC), Serial No.

85-523A, February 14, 1985.

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15. Letter from W. L. Stewart to USNRC, "Virginia Electric and Power Company; Surry Power Station Units 1 and 2; Proposed Technical Specifications Changes, 11 Serial No.87-212, dated May 22, 1987.

16. Letter from C. P. Patel (USNRC) to W. L. Stewart, "Surry Units 1 and 2 - Issuance of Amendments Re: Changes to Sections 3.7 and 4.1 (TAC Nos. 55380, 55381, 60307, and 60308), 11 Serial No.88-096, dated February 17, 1988.

17. Letter from W. L. Stewart to USNRC, Amendment to Operating Licenses NPF-4 and NPF-7; North Anna Power Station Units 1 and 2; Proposed Technical Specifications Changes, 11 NRC Letter Serial No.85-077, dated May 2, 1985 (North Anna Core Uprating Project).

18. Letter from W. L. Stewart to H. R. Denton, "Virginia Electric and Power Company; Response to NRC Request for Additional Information; Core Uprate Program; North Anna Power Station Units 1 and 2, 11 Serial No. 85-772A, dated February 6, 1986.

19. Letter from L. B Engle (NRC) to W. L. Stewart, Amendments 84 and 71 to Facility Operating Licenses NPF-4 and NPF-7, NRC Letter Serial No.86-575, dated August 25, 1986 (North Anna Core Uprating Project).

20. Letter from W. L. Stewart to H. R. Denton (USNRC), 11 VEPCO Reactor System Transient Analyses," Serial No. 376A, dated August 24, 1984.

"Westinghouse Reactor Protection System/Engineered Safety Features Actuation System Setpoint Methodology."

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e e BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATIONS (10 CFR 50.92)

Virginia Electric and Power Company has proposed changes to the Surry Power Station Technical Speci fi cations which revise the Intermediate Range High Flux trip setpoint. It has been determined that the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination may be stated as follows.

1. The proposed changes will not increase either the probability of occurrence of any of the UFSAR accidents or their potential consequences. The Intermediate Range High Flux trip provides backup trip protection for the Power Range High Flux trip (low setpoi nt) during reactivity accidents (power excursions) from low power initial conditions. The proposed revision to the Intermediate Range High Flux trip setpoint will not increase the probability of any UFSAR accident because the actuation of a reactor trip is not considered an accident initiator. Further, the proposed revisions should decrease the probability of unnecessary Intermediate Range trips. The proposed revision does not alter the consequences of any UFSAR accident because the Intermediate Range trip functions as a backup trip and is not credited for mitigation within the UFSAR analysis for any accident.

Hence, the revision of a backup trip setpoint does not impact the sequence of events considered for an accident, and will not alter the consequences of the accident as analyzed.

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e The establishment of an allowed outage time for the Source Range channel at power levels below P-6 also does not increase the probability of any UFSAR accident or its consequences, because the required action of this new requirement (i.e., open the reactor trip breakers at a power level below P-6) has been previously evaluated in the safety analysis.

The deletion of TS 3.12.A.4, which governs the predicted critical rod position at subcritical conditions, does not increase either the probability of occurrence or the potential consequences of any of the UFSAR accidents. TS 3.12.A.2 and startup procedures ensure equivalent conditions to TS 3.12.A.4 above permissive P-6 (i.e.,

ensure that the control rods are above the zero power insertion limit at criticality). Below permissive P-6, reactor protection against reactivity insertion transients is provided by the Source Range channel, which is automatically activated below permissive P-6.

2. The proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The Intermediate Range trip setpoint revision simply alters the time during an accident at which reactor trip actuation would occur, if assumed to be actuated by the Intermediate Range protective action.

Such a change in reactor trip actuation time has previously been evaluated and found not to create new or different kinds of accidents.

The es tab 1 i shment of an a 11 owed outage time for the Source Range channel at power levels below P-6 also does not create the possibility 28

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e of new or different kinds of accidents, because the required action of this new r~quirement (i.e .* open the reactor trip breakers at a power level below P-6) has been previously evaluated.

The deletion of TS 3.12.A.4, which governs the predicted critical rod position at subcritical conditions, does not create the possibility of a new or different kind of accident from any accident previously evaluated. TS 3.12.A.2 and startup procedures ensure equivalent conditions to TS 3.12.A.4 above permissive P-6. Below permissive P-6, allowable normal operation operating conditions are effectively restricted by other governing Technical Specifications, and accident operating conditions are restricted by the protective action of the Source Range channel. The deletion of TS 3.12.A.4 does not create an allowable operating mode or core operating condition which could create a new or different kind of accident from those previously evaluated.

3. The proposed changes will not involve a significant reduction in a margin of safety. As stated above, the Intermediate Range High Flux trip provides backup protection for the Power Range High Flux trip (low setpoint). Because a backup trip function is not assumed within the UFSAR accident analysis, no explicit margin of safety is dependent upon the actuation of the backup trip protective action. Accordingly, a revision in a backup trip setpoint will not alter any explicit margin of safety. The proposed changes do alter the Technical Specification basis to explicitly state that the functional capability of the Intermediate Range High Flux trip at the specified 21

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trip setting is required to enhance the overall reliability of the Reactor Protection System. The proposed Intermediate Range High Flux trip setpoint revision has been evaluated and determined to maintain the functional capability of the trip at the proposed revised setpoint. Thus, any margins inferred by this proposed basis change are maintained by the proposed Intermediate .Range trip setpoint revision.

The establishment of an allowed outage time for the Source Range channel at power levels below P~6 also does not involve a significant reduction in a margin of safety, because the required action of this new requirement (i.e., open the reactor trip breakers at a power level below P-6) has been previously evaluated.

The deletion of TS 3.12.A.4, which governs the predicted critical rod position at subcritical conditions, does not involve a significant reduction in margin of safety. TS 3.12.A.2 and startup procedures ensure equivalent conditions to TS 3.12.A.4 above permissive P-6 (i.e., ensure that the control rods are above the zero power insertion limit at criticality). Below permissive P-6, reactor protection against reactivity insertion transients is provided by the Source Range channel. The deletion of TS 3.12.A.4 does not cause the results of any existing accident analyses to become more limiting. The margin of safety as defined in the bases of applicable Technical Specifications is not reduced.

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