ML18144A044

Vepco Surry Units 1 & 2 Safety Balance Assessment for Elimination of RCS Main Loop Pipe Break Protective Activities.

ML18144A044
Person / Time
Site:	Surry
Issue date:	11/30/1985
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18144A041	List: ML18144A040 ML18144A042 ML18144A043 ML18144A044
References
TAC-59821, TAC-59822, NUDOCS 8512110277
Download: ML18144A044 (35)

Text

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• VIRGINIA ELECTRIC AND POWER COMPANY SURRY UNITS 1 AND 2 SAFETY BALANCE ASSESSMENT FOR ELIMINATION OF REACTOR COOLANT SYSTEM MAIN LOOP PIPE BREAK PROTECTIVE DEVICES ------------

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8512110277 851203 PDR ADOCK 05000280 P PDR NOVEMBER 1985 ATTACHMENT 3

• I. I I. III. IV. ---------------------------

TABLE OF CONTENTS Introduction Safety Balance Assessment Summary and Conclusions Development of Safety Balance 1 3 5 A. Public Health Risk Avoidance Attributable to Protection 5 from Dynamic Effects Associated with Pipe Breaks B. Reduction in Occupational Radiation Exposure 9 Resulting in a Decision Not to Use Protection Against Dynamic Effects Associated with Pipe Breaks 1. Occupational Exposure -Accidental

2. Occupational Exposure -Routine 9 11 a. Avoided Occupational Dose Associated with Not 11 References Installing Backfit Modifications for Asymmetric Blowdown Loads b. Avoided Occupational Dose Associated with Eliminating Removals and Reinstallations for Maintenance and Functional Testing of the 18 Large Bore Snubbers Per Unit c. Avoided Occupational Exposure due to Eliminating Visual Inspections of 18 Large Bore Snubbers Per Unit d. Avoided Occupational Exposure due to Eliminating Impact on Other Items by Removal of 18 Large Bore Snubbers Per Unit 14 16 18 21 Appendix -Radiation Exposure Associated with Removal and Reinstallation of Large Bore Snubbers Table A-1: Exposure Records -Unit 1, 1984 (B-P) A-9 Table A-2: Exposure Records -Unit 2, 1985 (B-P) A-10 Table A-3: Exposure Records -Unit 2, 1985 (Pathon) A-11 Table A-4: Man-hour Records --Unit 2, 1985 A-12 60-MSW-3303B-2

• L Introduction This report is submitted in support of Virginia Electric and Power Company's request for exemption to General Design Criterion 4 (GDC-4) applicable to Surry Power Station Unit Nos. 1 and 2, to the extent that protection against the dynamic effects of postulated pipe rupture on primary system components and piping may be eliminated.

The scope of the request would allow elimination of any requirement to backfit those modifications for asymmetric blowdown loads addressed in NRC Generic Letter 84-04 (Reference 1). In addition, the scope of the request would allow elimination of 18 large bore snubbers on the reactor coolant system (RCS) which are required only for pipe rupture loadings.

This report presents a safety balance assessment of the consequences of eliminating these pipe rupture protective devices, weighing the increased risk to the public vs. the benefits of avoided occupational exposure due to their elimination.

Generic Letter 84-04 provided the NRC safety evaluation concluding that an acceptable technical basis exists so that the blowdown loads resulting from double-ended pipe breaks in the RCS primary loop need not be considered as a design basis, provided that certain conditions can be met for the reviewed plants, including Surry Unit Nos. 1 and 2. The following value impact analysis utilizes methodology consistent with that utilized in Generic Letter 84-04. Data contained in Generic Letter 84-04 is used as the basis for obtaining estimates of increased risk to the public health and of occupational exposure due to a pipe rupture accident, 60-MSW-33038 Page 1

• and for the exposure associated with the proposed modifications for asymmetric blowdown loads. Actual man-Rem exposure records from the last two years form the basis for the estimates of occupational exposure associated with removal and reinstallation of the large bore snubbers over the remaining life of the units. The results confirm that the positive safety benefits of avoided occupational exposure of eliminating the pipe rupture devices and large bore snubbers dramatically outweigh the negative impacts of risk to the public safety . 60-MSW-33038 Page 2

• *--~

• II. Safety Balance Assessment Summary and Conclusions A summary of the results of the safety balance is shown below. All ranges of estimates support the request to eliminate consideration of the dynamic effects of RCS main loop pipe breaks in the Surry design basis, and to permit removal of the 18 large bore snubbers per unit required only for RCS pipe rupture loadings.

TABLE 1 SAFETY BALANCE VALUE

SUMMARY

Value (man-Rem)*

Lower Nominal Upper Estimate Estimate Estimate Public Health 0.0 -0.56 -5.94 Occupational Exposure (Accidental) 0.0 -0.11 -3.67 Occupational Exposure (Routine) (a) Due to Eliminating Back-Fit Modifications for Asymmetric Pressure {Units 1 and 2) 122 366 1098 (b) Due to Eliminating Removals and Reinstallations of 18 Snubbers (Unit 1) 728 1520 3036 (Unit 2) 242 507 1518 (c) Due to Eliminating Visual Inspections of 18 Snubbers (Units 1 and 2) 19 27 54 (d) Due to Eliminating Impact on Other 15 20 30 Items By Removal of 18 Snubbers (Units 1 and 2) TOTAL QUANTIFIED VALUE 1126 2439 5726 *Positive numbers indicate increased benefit in safety; Negative numbers indicate decreased benefit in safety . 60-MSW-3303B Page 3

• Value Impact Conclusions These estimates clearly demonstrate that implementation of "leak-before-break" on the RCS primary loop piping and removal of the 18 large bore snubbers per unit at Surry is justifiable when considering public and occupational exposure (both operational and accidental) over the remainder of the 40-year life of these units. In addition, these results indicate that elimination of the 18 large bore snubbers per unit will result in 4 to 8 times the savings in occupational exposure as that avoided (at Surry) by not installing the asymmetric pressure modifications previously addressed in Generic Letter 84-04. 60-MSW-33038 Page 4

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• III.Development of Safety Balance A. Public Health Risk Avoidance Attributable to Protection from Dynamic Effects Associated with Pipe Breaks The Value-Impact Analysis enclosed in Generic Letter 84-04 has been used as a basis for.calculating the potential increase in offsite dose to the general public resulting from employment of 11 leak-before-break 11 and elimination of special design considerations for pipe breaks in the RCS primary coolant loops. However, several changes were necessary to calculate dose estimates applicable to our Surry Units 1 and 2. The following bases were used for the calculation:

1. The Surry Units 1 and 2 are three loop Westinghouse Pressurized Water Reactors (PWRs). Because Generic Letter 84-04 estimates are given in terms of one unit with two loops, adjustments have been made to consider the increased number of potential breaks locations in a three-loop unit. 2. The most significant site specific effect is population density. 60-MSW-3303B Generic Letter 84-04 assumed the U.S. average of 340 people per square mile with a uniform population density. As stated in the Surry FSAR and Environmental Report, in the year 2020 Surry 1 and 2 are predicted to have a corresponding population density of 367 people per square mile in the Oto 50 mile radius zone. As a consequence, the resulting offsite doses are non-conservative for Surry by a factor of approximately 8%. As a further Page 5

• conservatism, a uniform population density in the 50-mile radius release model was used for the evaluation.

3. Generic Letter 84-04 used the site meteorology of a typical Midwestern Westinghouse PWR site and is assumed as comparable to the Surry site for this evaluation.

4. Where required in this report, the remaining plant life is based on assuming an end of plant life 40 years from the dates of commercial operation (December, 1972 for Unit 1 and May, 1973 for Unit 2), rather than from date of issuance of construction permit. Therefore, the remaining lives are assumed as 27 years. This difference in plant lives does not significantly affect the results, and certainly does not affect the validity of the conclusions.

Generic Letter 84-04 calculated the increase in risks associated with a large LOCA using the guidelines in WASH-1400 assuming no protection from the dynamic effects of pipe rupture. As used in Generic Letter 84-04 and throughout this submittal, the term 11 plant 11 means a single unit. For a two-loop plant, the nominal risk increase from a double-ended guillotine (DEG) large LOCA within the reactor cavity from Generic Letter 84-04 was estimated to be 6E-3 man-Rem/py (man-Rem per plant-year).

Adjusting for the number of loops (3 vs. 2) and the difference in population density (367 vs. 340), the Surry nominal risk estimate is: 3/2 x 367/340 x 6E-3 = lE-2 man-Rem/py 60-MSW-33038 Page 6

• Similarly from Generic Letter 84-04, the nominal risk from a DEG large LOCA outside the reactor cavity was estimated to be 2E-4 man-Rem/py.

The Surry nominal risk estimate is: 3/2 x 367/340 x 2E-4 = 3.2 E-4 man-Rem/py Therefore, the combined Surry nominal increase in risk estimate from DEG large LOCAs both within and outside of the reactor cavity is: lE-2 + 3.2 E-4 = 1.03 E-2 man-Rem/py The upper increase in risk estimate is calculated in Generic Letter 84-04 with a procedure similar to that utilized for the nominal risk estimate, but based on a more conservative estimate of the probability of a large LOCA and core melt. The upper estimate in Generic Letter 84-04 is 0.1 man-Rem/py with no adjustments required for the number of loops per plant because this frequency is per plant-year.

Adjusting for the difference in population density, the Surry upper increase in risk estimate is: 367/340 x 0.1 = 0.11 man-Rem/py 60-MS~J-3303B Page 7

• The lower increase in risk estimate is assumed to be 0. The increases in risk to the public as determined above are tabulated as follows: Nominal Estimate Upper Estimate Lower Estimate Risk Increase (man-Rem/py)

1. 03 E-2 0.11 0 Applying these estimates of increased risk to the remaining 27 year life for both units, yields the following estimates for the combined increase in risk of exposure attributable to both units: 60-MSW-3303B Lower a.man-Rem Nominal .56 man-Rem Page 8 Upper 5.94 man-Rem

• B

• Reduction in Occupational Radiation Exposure Resulting From a Decision Not to Use Protection Against Dynamic Effects Associated With Pipe Breaks 1. Occupational Exposure (Accidental)

An increase in occupational exposure can be calculated as the increase in core melt frequency multiplied by the occupational exposure expected to occur as a result of a major accident.

Generic Letter 84-04 calculated an increase in core melt frequency by summing the contribution from the breaks inside the reactor cavity and the breaks outside the reactor cavity, and then adjusting for the number of loops. For Surry Units 1 and 2, the nominal core melt frequency increase would be: 3/2 x (9E-8+0.2x(3E-6/250))

= lE-7 events/py The upper estimate of 2E-6 events/py from Generic Letter 84-04 is directly applicable because it is per plant-year and is not dependent upon the number of loops. A lower bound estimate of 0 is assumed. These estimates are then used with cleanup and decommissioning dose estimates from NUREG/CR-2601 (Reference

2) to give an estimate of occupational exposure increase per plant-year.

The dose estimates are given in two parts. The first is immediate occupational exposure (o 10) during the span of the event and its short-term control. The second is the long-term occupational 60-MSW-33038 Page 9 J .J *

• exposure (DLTO) associated with the cleanup and recovery from the accident.

The increase in occupational exposure per plant-year (D 0 A) is calculated as follows: DOA= P(DIO + D LTO) where: DOA = Increase in occupational exposure per plant-year p = Increase in core melt frequency DIO = Immediate occupational exposure DLTO = Long-term occupational exposure The results of the are conservatively upper bound DLTO" Nominal Estimate Upper Estimate Lower Estimate p (events/ pl ant-yr) lE-7 2E-6 0 calculations are shown below. propagated by the use of upper

• DIO (man-rem/

event) 1E+3 4E+3 0

• DLTO (man-rem/

event) 2E+4 3E+4 1E+4 Uncertainties bound Dro and DOA (man-rem/

pl ant-yr) 2.lE-3 6.8E-2 0 Based on cleanup and decommissioning estimates contained in NUREG/CR-2601.

Applying these estimated increases in occupational exposure per plant-year to the remaining 27 year life of the two units yields the following estimated exposures:

60-MSW-33038 Page 10

• Lower Nominal Upper 0. man-Rem 0.11 man-Rem 3.67 man-Rem 2. Occupational Exposure -Routine Two sources of avoided operational exposure are addressed in this safety balance assessment.

The first is the operational occupational exposure avoided due to not installing the modifications for asymmetric blowdown loads addressed in the safety-balance study of Generic Letter 84-04. The second source of avoided operational occupational exposure is due to eliminating future maintenance and functional testing of the 18 large bore snubbers per unit which we propose to eliminate based on this exemption request. (a) 60-MSW-33038 Avoided Occupational Dose Associated with Not Installing Backfit Modifications for Asymmetric Blowdown Loads The operational occupational exposure avoided due to not installing and maintaining the backfit modifications for asymmetric blowdown loads is calculated in a manner consistent with that utilized in the safety-balance assessment included with Generic Letter 84-04. Surry is not one of the three plants requiring extensive modifications; WCAP-9628 (Reference

3) indicates that modifications to the Steam Generator supports are not required.

Therefore, only modifications to install primary shield wall restraints, inspection port modifications, and reactor coolant pump support modifications were considered to be required.

The man-hour estimates to implement these Page 11

• J

• 1. modifications were estimated by prorating the man-hours estimated in Generic Letter 84-04 as follows: For installation of primary shield wall restraints and inspection port modifi cati ans: (56,000 total man-hours/13 plants) x 2 Surry units= 8615 manhours 2. For Modifications to the Reactor Coolant Pump Supports:

(6000 total manhours/21 loops) x 6 Surry loops= 1714 manhours The safety-balance for Generic Letter 84-04 calculated implementation dose based on 25 mR/hr inside containment.

The corresponding areas for these modifications were reviewed by Surry Health Physics and dose rates were estimated as follows: for the area around the primary shield wall, the general dose rate was 100-250 mR/hr; and in the area of the reactor coolant pump modification was 200-400 mR/hr. In addition, it was observed that it would be difficult to provide effective shielding in these areas because of the high generalized sources. However, while the 25 mR/hr appears to be low as an integrated dose rate, it will be used here. Therefore, the total calculated exposure for implementation of these modification is estimated as: Assumed Calculated Manhours Dose Rate Exposure Shield Wall & Inspection 8615 25 mR/hr 215 man-Rem Port Modifications Reactor Coolant Pump Modifications 1715 25 mR/hr 43 man-Rem TOTAL 258 man-Rem 60-MSW-3303B Page 12 In addition, occupational dose to maintain the modifications is also avoided. This is calculated in conformance with Generic Letter 84-04, by assuming that two additional man-weeks per plant-year would be spent inside containment if the modifications were made: Operational dose avoided= 27 yrs x 80 man-hr/pyx 2 units x 25 mR/hr = 108 man-Rem Total avoided occupational dose associated with not implementing these modifications is therefore estimated to be 366 man-Rem for both units. Using the model for upper and lower estimates assumed in Generic Letter 84-04 (NUREG/CR-2800, Reference 4): Dose upper= 3 dose expected Dose lower= l/3 dose expected yields the following range: 60-MSW-3303B Lower 122 man-Rem Nominal 366 man-Rem Page 13 Upper 1098 man-Rem (b) e ._)

• 60-MSW-33038 Avoided Occupational Dose Associated with Eliminating Removals and Reinstallations for Maintenance and Functional Testing of the 18 Large Bore Snubbers Per Unit In accordance with the Surry Technical Specifications, 10% of the snubbers selected at random must be functionally tested each refueling outage, with additional snubbers tested if there are results for the testing which are outside the acceptance limits. In addition, the seal life monitoring program requires removal of the snubbers for seal replacement prior to exceeding the seal life. The occupational exposure associated with removal and reinstallation of the 18 large bore snubbers per unit to be eliminated upon approval of the exemption request is estimated based on actual exposure records for the Unit 1 Fall 1984 and Unit 2 Spring 1985 refueling outages, as discussed in the Appendix.

For the 18 snubbers to be eliminated, the exposure for a single removal and reinstallation of all 18 snubbers is estimated in the Appendix as: Lower 182 man-Rem Nominal 304 man-Rem Upper 506 man-Rem For Unit 1, it is assumed that the number of removals and reinstallations for maintenance and functional testing over the remaining life of the unit will be: Page 14

• 60-MSW-3303B Lower 4 Nominal 5 Upper 6 Therefore, for Unit 1, the conservatively estimated total exposure attributable to removals and reinstallations over the life of Unit 1 are: Lower Nominal Upper 728 man-Rem 1520 man-Rem 3036 man-Rem For Unit 2, the large bore snubbers were refurbished during the Spring 1985 outage. This refurbishment included new Tefzel seals (with a projected seal life of 40 years), new self-flushing control valves, test-in-place valve connections, individual remote reservoirs to replace the common reservoirs, and general snubber upgrading.

While these upgrades were made too recently for their effectiveness to have been proven, it is assumed in the Appendix that the total exposure attributable to maintenance and functional testing over the remaining life of Unit 2 will be: Lower 242 man-Rem Nominal 507 man-Rem Page 15 Upper 1518 man-Rem (c) Avoided Occupational Exposure due to Eliminating Visual Inspections of 18 Large Bore Snubbers Per Unit 60-MSW-3303B In addition to removals and reinstallations for seal replacement and functional testing, routine visual inspections of the snubbers are required.

These 18 large bore snubbers are included in the 11 Inaccessible Snubbers" group, and the frequency of inspections depends on total visual failures within the entire Inaccessible Group, not just the large bore snubbers.

The large bore snubbers have been a significant contributor to visual failures.

One reason for this is that the large bore snubbers were on common reservoirs and all snubbers on a reservoir found to be inoperable are considered inoperable by Surry Technical Specifications.

An additional reason is the possibility of visual inspections resulting in an incorrect conclusion that the large snubber hydraulic systems are not functional.

While this type of incorrect conclusion for small bore snubbers can easily be resolved by functional testing, such on-site functional testing is not possible for large bore snubbers due both to physical size and test load capacity.

Therefore, in accordance with the Technical Specifications, more frequent visual inspections are required resulting in more frequent plant outages, with resulting additional exposure.

While these inspections could be required only once every 18 months, it will be assumed that these visual inspections take place once a year (counting both "as-found" and 11 as-left 11 inspections as a single inspection).

The estimated number of inspections assumed over the remaining 27 years of operation are: Page 16

• 60-MSW-33038 Lower 19 Nominal 27 Upper 54 It will be assumed that the visual inspections of the large bore snubbers require 20 manhours per inspection (this accounts for both 11 as-found 11 and "as-1 eft" i nspecti ans). The integrated dose rate is assumed as 25 mR/hr. Therefore, the occupational exposure per visual inspection is 0.5 man-Rem. For the estimated number of inspections above for each unit, this yields for both units (combined):

Lower 19 man-Rem Nominal 27 man-Rem Page 17 Upper 54 man-Rem (d) Avoided Occupational Exposure due to Eliminating Impact on Other Items By Removal of 18 Large Bore Snubbers Per Unit 60-MSW-3303B In addition, the snubbers limit access to other components which must be inspected.

While this is easy to demonstrate qualitatively by photographs, it is difficult to measure the impact quantitatively.

For purposes of this safety-balance assessment, it will be assumed that the existence of the snubbers adds 5 man-Rem per unit to inspection of other items every 10 year ISI cycle. This totals 20 man-Rem for each of the two units. The assumed range is: Lower 15 man-Rem Nominal 20 man-Rem Page 18 Upper 30 man-Rem

• Therefore, the total avoided occupational exposure per this exemption request is: Avoided Exposure (man-Rem)

Lower Nominal Upper (a) Due to Eliminating Back-Fit Modifications for Asymmetric Pressure (Units 1 and 2) (b) Due to Eliminating Removals and Reinstallations of 18 Large Bore Snubbers Per Unit Unit 1 Unit 2 (c) Due to Eliminating Visual Inspections of 18 Large Bore Snubbers Per Unit (Units 1 and 2) (d) Due to Eliminating Impact on Other Items by Removal of 18 Large Bore Snubbers Per Unit (Units 1 and 2) TOTAL 60-MSW-3303B Page '19 15 1126 122 728 242 19 366 1520 507 27 20 2440 1098 3036 1518 54 30 5736

• A review of these avoided occupational exposures indicates that elimination of the 18 large bore snubbers will result in 4 to 8 times the savings in occupational exposure as that avoided (at Surry) by not installing the modifications proposed for asymmetric pressure which were previously addressed in Generic Letter 84-04 . 60-MSW-3303B Page 20

• L_ IV References

1. Generic Letter 84-04, February 1, 1984, 11 Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops. 11 2. Murphy, E.S., and G.M. Holter. 1981. Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents.

NUREG CR-2601, Pacific Northwest Laboratory, Richland, Washington.

3. Andrews, W.B., et. al. 1983. Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development.

NUREG/CR-2800 (PNL-4297), Pacific Northwest Laboratory, Richland, Washington.

4. "Structural Evaluation of Steam Generator and Reactor Coolant Pump Supports for Hypothetical Reactor Coolant Pipe Breaks Outside of the Reactor Cavity --(Westinghouse Owners Group Asymmetric LOCA Loads Evaluation

-Phase B)" WCAP, 9628, November, 1979. 60-MSW-3303B Page 21

• APPENDIX RADIATION EXPOSURE ASSOCIATED WITH REMOVAL AND REINSTALLATION OF LARGE BORE SNUBBERS The purpose of this Appendix is to provide background on the difficulties of removal and reinstallation of the large bore snubbers, the exposure-related problems and our ALARA efforts already implemented, and to document the basis for our estimates for occupational exposure associated with the removal and reinstallation of the 18 large bore snubbers which we propose to eliminate.

Background The Surry Technical Specifications require periodic removal of hydraulic snubbers, which are considered active components, for functional tests and implementation of a seal service life program. The Surry large bore snubbers were not designed for quick removal, ease of inspection, or for minimizing radiation exposure during removal/inspection activities.

The man-Rem burden associated with maintenance of these large bore snubbers is extremely high due to their location near the highest radiation areas of the plant and their relatively massive size. Because of plant equipment congestion, removal of other permanent plant equipment is often required in the process of removing these snubbers.

Currently, our experience indicates that removal and reinstallation of a single pair of large bore snubbers on the reactor coolant pump support can result in as much as 100 man-Rem of exposure during a single refueling outage, although a more common 60-MSW-3303B A-1

• exposure level is about 35 man-Rem per pair. When a number of these large bore snubbers are removed for functional testing and mainten'ance, they can account for as much as one-fifth of the total radiation exposure received during a refueling outage, as well as a maintenance expense in the millions of dollars. The highest exposures are associated with the large bore snubbers on the Reactor Coolant Pump (12 11-bore Bergen-Paterson Snubbers) which are located approximately 12 to 14 feet above the floor in each loop room. Installation and removal of these snubbers must be performed on scaffolding which is, by necessity, small and often crowded making personnel entry and exit difficult and time consuming.

These snubbers are located within a network of piping, valves, and support fixtures making mechanical work difficult and historically time consuming.

A majority of the rigging for each snubber must be done from above in high dose areas located in the 18 1 4 11 level RCP Cubicles, all of which contain RCS loop stop valves, RCS piping, and Safety Injection piping, and other difficult to shield sources. This set up work is again time consuming . 60-MSW-3303B A-2

' /

• ALARA Aspects of Removal and Reinstallation of Large Bore Snubbers Our Health Physics department has been aggressive in implementing additional safeguards to minimize worker exposure.

Prior to the snubber tasks starting, the ALARA pre-planning techniques are formulated, ensuring their incorporation into the work effort and the Radiation Work Permit (RWP) requirements.

Below is a standard set of techniques already established and used for snubber work. These techniques may be found attached to the snubber RWP work packages.

1. The use of mock-ups and hands on training by workers prior to starting 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. task, when available.

Constant Health Physics coverage at each loop room. ALARA Review of Historical Data. Shielding of loop room overhead areas. Decontamination of snubber area when required.

The use of tool lay down areas. The use of temporary lighting as required.

The use of scaffolding in lieu of ladders. Minimizing the number of workers at the work area at all times. Requiring the use of ALARA low dose waiting areas. Requiring the procurement of all known required tools and rigging apparatus prior to starting job. 12. Requirement of pre-shielding and post-shielding surveys to denote hot spot and general work area readings.

13. Requirements for ALARA pre-job briefing with the foremen and workers. 14. Requirement for special dosimetry and respiratory protection when necessary . 15. Requirement to maintain traffic control within the work area. 16. Requirement to critique task when applicable.

60-MSW-3303B A-3

• Shielding efforts for dose reduction on the loop room overhead piping difficult and almost impossible at times. Numerous source terms exist in overhead which contribute to the overall whole body exposure dose rates. average amount of lead shielding placed in these loop rooms equals out to approximately 2000 pounds (+) of shielding per loop room. Even with the shielding efforts the dose rates remain high. Supporting Data for Exposure Estimates for Removals and Reinstallations is the The The man-Rem totals used in this safety-balance assessment are based on actual exposure records measured by Health Physics during removals of numbers* of these snubbers in the Unit 1 1984 Winter and Unit 2 1985 Spring outages. These are therefore 11 hard 11 numbers. Table A-1 provides the cumulative exposure associated with maintenance and functional testing of the 12 11-bore Bergen-Paterson snubbers on the Reactor Coolant Pumps, during the 1984 fall outage. Only the pair of snubbers of the 11 A 11-loop were removed for functional testing and seal-replacement; total exposure directly related to these two snubbers totaled 123 man-Rem. Only in-place maintenance work was performed on the other snubbers.

Table A-2 provides the total cumulative exposure associated with the removal and reinstallation of all six of the 12 11-bore reactor coolant pump snubbers during the 1985 spring outage. As discussed later, the Unit 2 effort is used as the primary basis for the estimates in this report. Table A-3 provides the cumulative exposure associated with the removal and reinstallation of the 12 11-bore Pathon snubbers on the steam generator support rings. During the 1984 Unit 1 effort, only half of the snubbers were removed 60-MSW-3303B A-4

\

• and reinstalled.

During the 1985 Unit 2 effort, all snubbers were removed and reinstalled.

Again, the Unit 2 effort is used as the primary basis for the exposure estimates in this report. Table A-4 provides planned and recorded manhours for the portion of the Unit 2 effort involving removal and reinstallation of the large bore snubbers.

The actual manhours could be.somewhat higher than the recorded numbers, because a considerable amount of the normal reinstallation work associated with tubing fitup, filling, etc. could have been charged to the work order number for installation of the new tubing to the remote reservoirs.

The 11 apparent dose rate" for each activity was obtained by dividing the actual man-Rem totals by the recorded man-hours.

It is important to note, however, that it is the man-Rem exposure records that are the most accurate data, and these are used as the basis for the estimates in this report. The 1985 Unit 2 work performed under Design Change Package (DCP) 85-05 is used as the primary basis for the estimates used in this report. That effort had the benefit of separate procedures for removal and reinstallation of each snubber, so that they could be tailored to address specific local interferences, special rigging requirements, installation sequences, and the like. In addition, new attachment hardware (snubber bracket pins, tubing connections, bolts, etc.) were used to facilitate reinstallation.

In addition, because most of the Unit 2 large bore snubbers were removed simultaneously, significant savings in man-hours and exposure resulted since a lot of the support work common to several snubbers (scaffolding, etc.) would have been required even for one snubber. In addition, the scheduled outage of 89 days allowed sufficient time to work at a deliberate, rather than a frenzied pace . Therefore, the Unit 2 results are approaching the limits achievable under ALARA. A lower estimate for future routine snubber removals and 60-MSW-33038 A-5

/

• reinstallations is assumed to be 10% lower than these results; a nominal estimate is assumed to be 50% higher than the Unit 2 results. An upper estimate is assumed to be 2.5 times the Unit 2 results. These are tabulated below: Estimated Man-Rem for Routine Unit 2 Removal/Reinstallation of Snubbers (1 Cycle) 1985 Lower Nominal Upper S/G Uppers 15 14 23 38 S/G Lowers 69 62 104 173 RCPs 133 120 200 333 The 18 snubbers per unit to be removed due to the elimination of LOCA pipe rupture loads are the Steam Generator lower support ring snubbers and the RCP snubbers.

Therefore, the avoided exposure per 1 cycle of removal and reinstallation of just these 18 large bore snubbers is taken to be: Lower S/G Lowers & RCPs 182 Nominal 304 Upper 506 Estimated Number of Removals over Remaining Life of the Plant The number of removals and reinstallations for maintenance and functional testing over the remaining life of the plant is estimated separately for Surry Unit 1 and Unit 2, as a result of improvements incorporated during the snubber refurbishment performed during the Unit 2 Spring 1985 refueling outage. Surry Unit 1 In accordance with the Surry Technical Specifications, 10% of the snubbers selected at random are functionally tested each refueling outage, with 60-MSW-3303B A-6

) , __

• additional snubbers tested if there are snubbers which test outside the acceptance limits. In addition, the Unit 1 snubbers have Ethylene-Propylene (EP) seals which require seal replacement every 5-7 years in accordance with the Surry seal life monitoring program. Therefore, it is estimated that seal replacement and functional testing will require removal of the snubbers a minimum of 27 yrs/6yrs = 4 times over the remaining life of the plant. It could seem this is a somewhat high estimate because the seal life of certain snubbers would be longer in relatively low radiation areas (i.e. upper Steam Generator support snubbers).

However, the removal/reinstallation of the snubbers in high-radiation areas drive the man-Rem totals, and extending the seal life of those snubbers in lower-radiation areas will not have a major effect on total man-Rem. In addition, the reliability of these snubbers has been somewhat less than desired. Therefore, some snubbers have been removed at times in addition to those dictated by the 5-7 year seal life. Therefore, we believe that 4 removals over the remaining plant life is the lower estimate.

A nominal estimate of 5 removals, and an upper estimate of 6 removals and reinstallations are assumed. Therefore, the estimated total man-Rem attributable to removals and reinstallations over the life of the plant (Unit 1) are: Lower 182 man-Rem/Removal x 4 times 728 man-Rem 60-MSW-3303B Nominal 304 man-Rem/Removal x 5 times 1520 man-Rem A-7 Upper 506 man-Rem/Removal x 6 times 3036 man-Rem J

• These values apply to Surry Unit 1. Surry Unit 2 For Surry Unit 2, the large bore snubbers were refurbished during the spring 1985 outage. This refurbishment included new Tefzel seals (with a projected seal life of 40 years), new self-flushing control valves, test-in-place connections, separate remote reservoirs to replace the common reservoirs, and general snubber upgrading.

While these upgrades were made to*o recently for their effectiveness to have been proven, it is assumed that the effort would be judged effective if the man-hours and exposure over the life of the plant were reduced to only one-third of that without the upgrade. This assumption is used to obtain nominal and lower estimates; however, the upper estimate is taken as 50% of the Unit 1 upper estimate.

Therefore, for Unit 2, the estimated doses over life of the plant are: 60-MSW-33038 Lower 242 man-Rem Nominal 507 man-Rem A-8 Upper 1518 man-Rem

)::, I \.0

• Area Task "A" Loop Rm B.P. Snubbers "B" Loop Rm B.P. Snubbers "C" Loop Rm B.P. Snubbers (*) All readings in mR/hr RWP # Area 84-1893 "A,B,C" Loop Rm SWP "A,B,C" Loop Rm SWP "A,B,C" Loop Rm 84-1879 "A,B,C" Loop Rm DIRECT EXPOSURE FOR BERGEN-PATERSON SNUBBERS Shielding Pkg Radiation Levels(*)

Hot-Spots

(*) Total RWP # TSR # Shielded I Unshielded (Up To) Man-Rem Used 84-2001 TSR-84-26 200-700 500-2000 10,000 123. 256 84-2024 TSR-84-27 200-1200 1000-3000 15,000 26.036 84-2120 TSR-84-28 500-1000 700-2000 10,000 20.694 (Note: Only the two snubbers of Loop "A" were Removed) SUPPORT WORK FOR BERGEN-PATERSON SNUBBERS Task Scaffolding Support Health Physics Support Actual Man-Rem Received 17.862 N/A Estimated Man-Rem Received N/A 4.500 H.P. Decon Support N/A 3.000 H.P. Shielding Support _1_8_.8~7~3~~~~-N~/A~~

TOTAL 44.24 Man-Rem TOTAL 169.99 man-Rem Estimate 1.5 Man-Rem Per Loop Room For Health Physics Coverage Estimate 1.0 Man-Rem Per Loop Room For Decon Support Combined Total for Bergen-Paterson Snubbers (Unit 1, 1984) = 214.23 Man-Rem 60-RKM-3344B-2 F;:; > I ...

( DIRECT EXPOSURE FOR BERGEN-PATERSON SNUBBERS Shielding Pkg Radiation Levels(*)

Area Task RWP # TSR # Shielded I Unshielded "A" Loop RM B.P. Snubber 85-1208 TSR-85-5 100-700 450-3000 "B" Loop RM B.P. Snubber 85-1214 TSR-85-6 200-700 300-3000 "C" Loop RM B.P. Snubber 85-1230 TSR-85-7 200-700 200-700 (*) All readings in mR/hr SUPPORT WORK FOR BERGEN-PATERSON SNUBBERS Rwp # Area 85-1191 "A,B,C" Loop RM SWP "A,B,C" Loop RM SWP "A,B,C" Loop RM 85-1200 "A,B,C" Loop RM Task Scaffolding Support Health Physics Support H.P. Decon Support Actual Man-Rem Received 16. 351 N/A N/A Estimated Man-Rem Received N/A 4.500 3.000 H.P. Shielding Support ~1~3~.9~9~0:...._~~~_;.;N~/Ac:...._

TOTAL 37.84 Man-Rem Combined Total for Bergen-Paterson Snubbers (Unit 2, 1985) 133.41 Man-Rem 60-RKM-3344B-3 Hot-Spots

(*) Total (Up To) Man-Rem Used 5,000 31.420 6,000 20.176 3,000 43.968 TOTAL 95.56 Man-Rem Estimate 1.5 Man-Rem Per Loop Room For Health Physics Coverage Estimate 1.0 Man-Rem Per Loop Room For Decon Support * (I) C: ::c rr, ::c ... > s;! N u, = (I) q10 0 ! !:? ::c ::c > rr, -t a= 9 0 rr, ::IE §~-.. ~:i:! i I "O "ti ::c F;; (I) al ;:s > I I Or" N z> !~elj -tC: Ni~ .. rr, -::c z ... (I) (I) -t CCIO> u,zr-s 0 z 0 "Tl

'

• AREA TASK "A" Steam Gen Upper S/G Support Ring "B" Steam Gen Upper S/G Support Ring "C" Steam Gen Upper S/G Support Ring "A" Steam Gen Lower S/G Support Ring "B" Steam Gen Lower S/G Support Ring "C" Steam Gen Lower S/G Support Ring (Note: Only Upper Steam Generator

i:,, I "A" Steam Gen Upper S/G Support Ring ,__. ,__. "B" Steam Gen Upper S/G Support Ring "C" Steam Gen Upper S/G Support Ring "A" Steam Gen Lower S/G Support Ring "B" Steam Gen Lower S/G Support Ring "C" Steam Gen Lower S/G Support Ring "C" Steam Gen Lower S/G Support Ring "C" Steam Gen Lower S/G Support Ring (*) All readings in 60-RKM-3344B-4 RWP # 84-1978 84-1978 84-1978 84-1978 84-1978 84-1978 TOTAL Support 85-1207 85-1207 85-1207 TOTAL 85-1266 85-1266 85-1266 85-1285 85-1599 TOTAL mR/hr UNIT 1 (1984) TOTAL RADIATION LEVELS (*) MAN-REM USED GEN AREA 79.573 20-40 Same As Above 20-40 Same As Above 20-40 Same As Above 22-200 Same As Above 22-200 Same As Above 22-200 79.573 Man-Rem Ring Snubbers Were Removed & Reinstalled)

UNIT 2 (1985) 15.139 5-35 Same As Above 5-35 Same As Above 5-35 15.139 Man-Rem 54.867 50-750 Same As Above 15-1000 Same As Above 7-1000 5.745 150-2000 8.125 200-700 68.737

• REMARKS "A,B,C," Upper & Lower Support Rings Listed On One RWP "A,B,C" Upper Support Ring Listed On One RWP "A,B,C" Lower Support Ring Listed on One RWP 2-RC-HSS-161 2-RC-HSS-159

• TABLE A-4 MAN-HOURS ASSOCIATED WITH REMOVAL AND REINSTALLATION OF S/G Uppers S/G Lowers RC Pumps LARGE BORE SNUBBERS (UNIT 2) Planned Recorded Exposure (man-hours) (man-hours) (man-Rem) 8904 4834 15.14 8328 5150 68.74 4884 3625 133. 41 *Apparent Dose Rate= Total Exposure/Recorded Manhours Apparent Dose Rate* (mR/hr) 3.1 13.3 36.8 It is estimated that removal accounts for approximately 20-25% of manhours, and reinstallation accounts for 75-80%. 60-RKM-33448-1 A-:-12 7