ML20059G573
| ML20059G573 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/10/1994 |
| From: | Pelletier J VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 BVY-94-02, BVY-94-2, NUDOCS 9401240277 | |
| Download: ML20059G573 (9) | |
Text
.
VEh1MONT.' YANKEE NUCLEAR POWER CORPORATION Ferry Road, Brattleboro, VT 05301-7002 mPty?o
[y')--
580 MAIN STREET ENGINEERING OFFICE BoLToN. M A 91740
y (508) 779 4 711 January 10,1994 BVY 94-02 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
References:
(a)
License No. DPR-28 (Docket No. 50-271)
(b)
Letter, USNRC to VYNPC, NVY 93-167, dated October 4,1993 (c)
Letter, USNRC to VYNPC, dated May 7,1980 (d)
Letter, VYNPC to USNRC, FVY 81-8, dated January 12,1981 (e)
NUREG-1372 dated June 1990 (f)
Letter, VYNPC to USNRC, dated January 15,1991
Dear Sir:
Subject:
Response to USNRC Request for Additional Information Regarding NUREG-0737 Habitability Requirements for the Vermont Yankee Nuclear Power Station in Reference (b), your staff indicated that a compliance review is being performed with regard to habitability requirements for the Control Room (CR) and the Technical Support Center (TSC) at Vermont Yankee. Reference (b) transmitted a request for additional information which your staff deemed necessary in order to verify Vermont Yankee compliance with the current licensing basis of the plant. A 90 day response from date of receipt was requested. Reference (b) was received by Vermont Yankee on October 12,1993.
The purpose of this letter is to provide our response to your staff's request for additional information. The two specific staff requests for additional information from the enclosure to Reference (b),
and our responses to these concerns, are as follows.
NRC Reauest No.1 Explain how the requirements of Section lil, " Design Control," and Section XI, " Test Control," of the Yankee Atomic Electric Company Operational Quality Assurance Program (YOQAP-1), as implemented pursuant to 10 CFR 50.54(a)(1), have been met with regard to the design and testing of the Control Room envelope and the Control Room Ventilation System. The licensee is requested to provide docume:itation of calculations, preoperational tests, or special tests where available, in support of the explanation. The following items are of specific concern:
Verification that the actual rate of infiltration into the Control Room envelope satisfies the assumptions of the Control Room dose analysis.
Confirmation that the Control Room receives adequate cooling in the isolation mode of ventilation system operation to maintain a suitable long-term environment for critical equipment and personnel.
000 71 Poa y
u
U.S. Nuclear Regulatory Commission vt nuoNT.YANxtr NUG LAR POwm CORPORATION January 10,1994 Page 2 Vermont Yankee Response:
WNPC's compliance with the requirements imposed on licensees by 10 CFR 50.54(a)(1) is controlled by_ the Yankee Operational Quality. Assurance Program (YOOAP-1). The requirements contained in the YOOAP are implemented by plant specific procedures. Section lit, " Design Control,"
specifies the criterion necessary to implement these controls for the design of and changes to structures, systems and components. At the plant level, these requirements are contained in various work controlling and documentation controlling procedures. This is also the case for Section XI, # Test Control." During the implementation of evaluations and modifications undertaken to satisfy W's commitments to the NRC, these procedures were invoked to provide the requisite control of the specific activities. Some of the major design controlling procedures are mentioned in the remainder of this letter. The plant, including the i
Control Room, was designed and constructed prior to the requirement for a Quality Assurance Program.
Preoperational testing was performed to show that system components functioned as designed.
Following W's notification by the NRC of the post TMI-2 requirements, W performed evaluations to determine the radiological dose to Control Room personnel and chemical toxicity effects on Control i
Room personnel from identified chemical hazards. These calculations are contained in calculation WC-39
" Tech Support Center 30 Day LOCA Doses" and WC-69 " Control Room Habitability From To'ic x
Chemicals." These calculations were performed, reviewed and approved in accordance with YAEC procedure WE-103, " Engineering Calculations and Analysis," which controls calculations in accordance with the requirements of the YOQAP.
Concurrent with these evaluations, detailed physical inspections of the Control Room envelope were conducted to determine the condition of critical components and systems.. These inspections indicated that some degradation of equipment due to aging had occurred which could potentially increase the Control Room in-leakage. A safety-related Maintenance Request was generated in accordance with.
plant procedure AP 0021, " Work Orders," which identified the scope of work. Additional implementing procedures were developed by the maintenance contractor. Both the Maintenance Request and the l
contractor's procedures (SPN 49994-700 and SPN 49994-701) received W review and approval in accordance with the procedural requirements. The contractor's procedures also received the Plant Operctions Review Committee (PORC) review and approval. Upon completion of the work activities, the work was documented in Job Order File 82-05 in accordance with the requirements of the procedure AP 6022, " Job Order Files".
Following the repairs and upgrades to the system, a test was devised to measure the Control Room in-leakage rate. The original design basis rate was verified with the Architect / Engineer, Ebasco
.i Services, Inc. A consultant was hired and Special Test Procedure, STP 82-01 was written, reviewed and approved in accordance with the requirements of AP 6001, " Installation, Test and Special Test Procedure."
The test was devised using a tracer gas technique in accordance with ASTM 741-80. The results of this 7
test indicated a near zero infiltration rate over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. These test results were significantly less than the design basis rate used in the dose calculation for control room inhabitants..
The results of the analysis conducted in calculations WC-69 indicated that additional protection was required for Control Room inhabitants due to postulated releases of toxic chemicals. Yankee Nuclear Services Division (YNSD) developed two Engineering Design Change Requests, EDCR 82-33," Toxic Gas Monitoring System" and EDCR 82-38," Control Room Pressurization Subsystems and Bio Packs." These EDCR's were written and reviewed in accordance with the requirements of Administrative Procedure AP 6004, " Engineering Design Change Request". The post implementation documentation of these EDCR's and the related implementation documentation are preserved in Job Order File 83-42.
In 1985, modifications were made to correct a deficiency in the logic for dampers which did not automatically close the Control Room kitchen and bathroom exhaust dampers or the computer room supply damper when the Control Room HVAC was switched to the recirculation mode. These modifications are i
U f clear Regulatory Commission VERMONT YANKu: NUCLLAR POWER CORPORATION Page 3 described and contro!!ed in Plant Alteration Request PAR 85-05, which was generated in accordance with i
Administrative Procedure AP 6003," Plant Alteration Requests". The PAR and its associated documents t
are permanently captured in Job Order File 86-04.
A revision to the Technical Specifications was approved by the Commission Technical Specification Amendment 96 which incorporated requirements for the operability of the Toxic Gas Monitoring System.
This reqdrement was later modified in Amendment 132 to remove these conditions based on our analysis and submitted to you in our letter Reference (f).
l The cooling capability of the Control Room HVAC System contains two (2) 100% capacity units to provide the assurance that the Control Room air temperature will not exceed 105"F under the most adverse design basis conditions. This system was installed during the original construction of the plant and no preoperational testing was performed that would simulate the most adverw design basis conditions. System functionally was tested during preoperational testing. Vermont Yanuse is currently assessing if any additional testing is appropriate under YOQAP,Section XI, " Test Control."
NRC Request No. 2 In VYNPC's reply to Generic Letter 82-33 dated April 19,1983, VYNPC committed to establish a Technical Support Center (TSC) radiologically habitable to th-c ae degree as the Control Room for postulated accident conditions. Our understanding is that the st ' ts not reviewed your actions to satisfy this commitment. Therefore, VYNPC is requested to demonstn. sat the radiological protection features of the TSC are adequate to satisfy this commitment. The license es aisc requested to provide supporting i
assumptions, tests, and calculations.
j Vermont Yankee Response:
)
Radiation exposures to Control Room (CR) operators are limited by General Design Criterion 19 to 5 rem whole body, or its equivalent to any part of the body, for the duration of postulated accidents.
l During March 1083, calculations were performed to compare Vermont Yankee's Technical Support Center (TSC) to GDC 19 in erder to ensure that the TSC is radiologically habitable to the same degree as the CR.
A summary of the calculations is provided as Enclosure 1.
The calculations demonstrate that, assuming a design basis LOCA and the Technical Specification primary containment leak rate of 0.8% per day for the 30 day duration of the accident and auunting for direct shine from the reactor building and the overhead cloud components, deses to TSC personnel remain below 5 Rem whole body for the duration of the accident.
The dose rate to occupants of the Technical Support Center during a Desiga Basis Accident was evaluated in calculation VYC-39. 'This evaluation used the same' source terms as were used for the Control Room evaluation and applied the same accident meteorology conditions for containment i
atmosphere leakage which was assumed to be filtered by the Stanc'by Gas Treatment System. Also analyzed as part of the contributing dose to TSC occupants was the direct gamma dose and skyshine from the top sector of the reacter building and the whole body gamma dose from tha overhead cloud. The results of this analysis indicated the need for additional shielding. An additional two inches of concrete shielding was added to the existing six inches of concrete. This shielding was added by Vermont Yankee Work Request 81-0344.
i lt should be noted that we have not included a TSC dose component due to leakage from main steam isolation valves (MSIV), since this is not considered as part of our licensing basis. In response to an NRC request for information [ Reference (c)], Vermont Yankee previously provided thyroid and whole body doses to CR personnel from MSIV leakage in Reference (d). However, this was provided to NRC to be responsive to your request and not as an expansion of our licensing basis. The absencti,of
U.S. Nuclear Regulatory Commission Vf RMONT YANKt t Nuct LAR l'OWER CORPORATION January 10,'T994 Page 4 consideration of MSIV leakage in the TSC dose component from our licensing basis is further supported by Regulatory Gaide 1.96 which does not recommend MSIV leakage control systems for boiling water reactor plants forwhich construction permits were issued prior to March 1,1970. The construction permit for Vermont Yankee was issued in December,1967. This staff recommendation was subsequently confirmed in the resolution of Generic issue C-8, " Main Steam isolation Valve Leakage and LCS Failure"
[ Reference (e)). In addition, although item lli.D.3.4 of NUREG-0737 requests MSIV leakage be considered in CR habitability evaluations, no similar guidance is provided relative to TSC habitability evaluations.
Further, item Ill.D.3,4 informs licensees that inclusion of MSIV leakage in CR habitability evaluations "should not be construed as altering the staff recommendations in Section D of Regulatory Guide 1.96 (Rev. 2) regarding MSIV leakage control systems." inclusion of a TSC dose component due to leakage from MSIVs is not part of our present licensing basis.
We trust that this information is responsive to your request and fully answers any remaining concerns that your staff may have surrounding this issue? However, should more information be required, please do not hesitate to contact this office.
Sincerely, Vermont Yankee Nuclear Power Corporation
~
wed James P. Pelletier Vice President - Enginee -
cc:
USNRC Region i Administrator U3NRC Resident inspector -VYNPS USNRC Project Manager - VYNPS i
.. ~
ENCLOSURE 1 f
Basis for Vermont Yankee Technical Support Center (TSC)
Habitability Calculation, VYC-39 i
The following assumptions ~were used to assess _the Technical Support Center Radiological Habitability for Vermont Yankee:
- 1. A design basis Loss of Coolant Accident (LOCA) was assumed resulting in 100% of the Noble Gases and 50% of the Halogens being immediately released to the primary containment atmosphere.
- 2. The Technical Specification Primary Containment leak rate of 0.8% per day was assumed to exist for the duration of the accident (30 days).
i
- 3. The CAD system was assumed to be used to vent the containment through the Standby Gas Treatment system at a flow rate of 20 scfm beginning at 192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br /> post LOCA and continuing for the~
duration of the accident. This assumption was based on the old containment air dilution system analysis. Venting for the present nitrogen system would occur at a later time.
- 4. The containment atmosphere leakage is assumed to be filtered by the Standby Gas Treatment System (SBGTS), removing 95% of the radiciodine before being enhausted to the stack.
i
- 5. Accident meteorology was used including a fumigation condition 5
for the first 30 minutes of the accident. The BETA / THYROID values were used to evaluate the thyroid dose and the whole body dose from the airborne activity in the TSC. The GAMMA values were used' to evaluate the contribution to whole body dose from the overhead cloud.
The following values were used for the dispersion l
coefficients; TIME POST LOCA BETA / THYROID X/Q GAMMA X/Q (HOURS)
(SEC/M )
(SEC/M )
{
3 3
0.0 - 0.5 1.979E-04
(
1.979E-04
[
0.5 - 1.0 3.074E-07 3.867E-05 1.0 - 2.0 2.740E-07 2.528E-05 2.0 - 8.0 2.275E-07 1.409E-05 8.0 - 24 1.685E-07 1.045E-05 l
24 - 96 6.820E-08 6.656E-06 96 - 720 4.879E-08 4.301E-06 I
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- 6. Occupancy factors from SRP 6.4; 100% from 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, GO%
from 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, 40% from 96 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.
l
- 7. It was a3sumed that there was uncontrolled, unfiltered inleakage' into the Tech Support Center. Complete mixing with outside air was assumed for the calculation of thyroid dose and.
the whole body dose from airborne activity in the TSC.
j
- 8. The direct gamma dose and skyshine from the top section of the Reactor Building were calculated at five locations within the TSC. This was necessary due to.the changes in the " visible" i
volume of the unshielded top section of the. reactor building.
- 9. The contribution to whole body gamma dose from direct and scattered radiation from the Reactor Building as well as the whole body gamma dose from the over head cloud were calculated i
assuming a total of 8" of concrete shielding. An additional 2" of concrete shielding was added to the existing 6" concrete floor above the TSC as a result of this analysis. The shielding was added via Vermont Yankee Work Request 81-0344, " Installation of Concrete for TSC Shielding."
The results of the analysis are shown in Table I.
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TABIZ I VY TSC HABITABILITY (VYC-39)
- I I WHOLE BODY GAMMA INTEGRATED DOSE - REM 1*' 74 Hrs.
30 Days A.
Contribution from Reactor Building
+
- 1. Location 1
- a. Direct 0.43 0.6~
- b. Skyshine 0.18 0.77
- 2. Location 2
- a. Direct 0.72 1.2
- b. Skyshine 0.22 0.96
- 3. Location 3
- j a.
Direct 1.7 3.3
- b. Skyshine 0.20 0.91
- 4. Location 4
- a. Direct 1.5 3.0
- b. Skyshine 0.19 0.85 i
- 5. Location 5 a.
Direct 1.5 2.7 l
- b. Skyshine 0.21 0.94 B. Contribution From Overhead Cloud (Same for all locations) 0.35 0,42 I
C. Contribution From Airborne Activity in TSC (Same for all locations) 0.00034 0.028
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D. Total Whole Body Gamma
- l
- 1. Location 1 0.96 1.9
- i
- 2. Location 2 1.3 2.6 i
- 3. Location 3 2.3
~ 4.7
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4 Location 4 2.1 4.3 i
- 5. Location 5 2.1 4.1 I
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